ML20030A495

Semiannual Operating Rept,Jan-June 1973

ML20030A495
Person / Time
Site:	Big Rock Point File:Consumers Energy icon.png
Issue date:	06/30/1973
From:	CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:
References
NUDOCS 8101090762
Download: ML20030A495 (69)

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( 3*MERS POWER COMPANY J ket No 50-135 License No DPR-6 4 161;i SEMIANNUAL REPC.?' OF OPERATIONS OF BIG ROCK POINT NUCLEAR PLANT Janu try 1,1973 - June 30,1973 I. INTPODUCTION'- S1!NIANNUAL OPERATING REPORT The plant was. base loaded at several differing power levels during this report period. - The off-gas release rate on January 1 was approximately 15,000 pCi/s. f l I 1 -l i 1 . (. - .) i i e w~ j .-., ~.. -. , =..... -,, .~,

yL i ri 1 ,3 II. OPERATIONS

SUMMARY

A. Changes in Plant Design There were three changes in the design of the facility which were incorporated as facility changes. They are as follows: 1. Facility Change C-169 This change involved the addition of a re=ote alarm to the particulate channel of the containment sphere exhaust continuous air monitor (CAM). To accomplish this, a spare annunciator drop (#15) on the containment sphere ventilation panel C-20 was used, and is labeled " SPHERE EXHAUST C.A.M. HIGH ACTIVITY." Any alarm occurring on C-20 panel is annunciated in the control room on the station service annunciators. This change gives the control room operators ic=ediate status on increases in the airborne activity in the containment sphere and is deemed to improve the overall safety of the facility. 2. Facility Change C-182 This change was completed on the new fuel storage area. This involved the addition of a security alarm system to meet 10CFR73 require-ments for storage of plutonium fuel. 3. Facility Changes C-215 to C-225 To ccuply with the one gpm unidentified leakage requirements, modifications were made in the containment sphere clean and dirty su=p piping. The modifications are covered by Facility Changes C-215 to C-225 which are still not fully implemented although Big Rock Point is currently in compliance with the one gpm limit. When completed they will be re-ported in the applicable semiannual report. B. Performance & aracteristics At the start of the report period (January 1,1973), the unit was on _line at 53 Weg in a coastdown mode, with one reactor feed pump in service. On January 16, control rod withdrawal was resumed in an attempt to maintain electric output at 50 Weg for the remainder of the cycle. On January 20, the plant was forced out of service due to primary coolant leakage at the packing of the reactor clean-up system discharge valve to the No 1 Teactor recirculating pump discharge piping. The unit ( returned on line on January 21 after a 25-hour outage. L

2 In anticipation of receiving six mixed oxide fuel assemblies, the new fuel storage intrusion alarm system was placed in service on January 12. Two G type (Exxon Nuclear) mixed oxide assemblies were sub-sequently received and inspected on January 15 and 16. Four mixed oxide recycle bundles were later received (February 7) from Nuclear Fuel Ser-vices. On February 9 and 26, spent fuel shipments Numbers 18 and 19 (each consisting of ten fuel assemblies) left the Big Rock Point Plant for eventual reprocessing at Nuclear Fuel Services. On February 10, the control rod pattern was changed slightly as the off-gas release rate reached the off-gas alarm point of % 25,000 uCi/s. A coastdown mode (with h8 notches remaining) was resumed QI maintained until the unit was taken off line on March 3 for the tenth annual refuel-ing outa6e. When plant output was decreased for the scheduled outage, elec-tric output had decreased to N g, and the off-gas release rate was averaging approximately 21,000 pCi/s. During the routine shutdown oper-ation, main steam isolation valve M0-7050 failed to close following mmentary manual initiation. A second manual closing initiation was attempted a few minutes later and was successful. The valve was subse-quently retested successfully by simulating a reactor protection system automatic closure signal. The failure to close was attributed to the valve packing binding the valve stem, thus causing the closing torque switch to trip. (See our letter of April 5, 1973 to the AEC.) Dry sipping of all 84 fuel assemblies commenced on March 6 and was empleted on March 11. Based on the sipping results and visual ob-servations, 21 assemblies were classified as definite leakers 'and one as " questionable." Failed bundles averaged % 14,000 mwd /t for the EG types and % 7,000 mwd /t for the F type assemblies. Fuel crud levels appeared to be lower than previous cycles and it must be noted that the crud seemed to be less tenacious. (Please r-*arence Special Report SR-lh to the AEC, July 5,1973.) During the inspection of four Exxon Nuclear, Inc assemblies (by -( EL_an's fuel team), several cobalt targets were noted to be unlocked. The cause was attributed to insufficient -forces of the springs that maintain + r + +

3 ( locking of the rods. Since the bundles were scheduled to be returned to the core, the spring forces were effectirely increased by addition of auxiliary springs. (See letter of April 10, 1973 to the AEC.) Dismantling of core internals consisted mainly of removal of 29 fuel channels with removable flow orifices and insertion of 29 new channels with fixed flow orifices. The' annual inspection of control blades D-6 and C-6 (known to have one bottom roller missing) resulted in the discovery that C-6 was missing another bottcn roller which was later retrieved from the core support plate. A subsequent inspection of all peripheral blades revealed that control rod A-2 also had one lower roller missing. (See letter of May 1, 19T3 to the AEC.) Control rod blade C-6 was subsequently replaced with a spare blade. Southwest Research Institute was also on site to assist in the annual in-service inspection. Preliminary findings indicated that all inspected welds and components were within specifications. Special Re-port SR-16 on the in-service inspection program as required by Technical Specifications Change No 37, was submitted to the AEC on August 20, 1973. After completion of Exxon's irradiated fuel inspection, the General ~ Electric fuels team inspected some of the General Ilectric fuel that had been removed from the reactor. -Only one assembly (EP-3) was reconstituted for return to the reactor. On March 28, fuel loading started and was completed on April 2. Core loading for Cycle 11 included 23 new fuel assemblies.- Six new assemblies were of the mixed oxide type (11 x 11 array), ie, G-01 and G-02 (Exxon Nuclear, Inc) and DA-01 through DA-04 (Nuclear Fuel Services, Inc). The G assemblies both contain one auxiliary neutron source each, giving a total of -four auxiliary neutron sources in the reactor core. (See Technical Specifications Change No 35.) 'lhe remaining 17 new assemblies were of a modified F type (GE 9 x 9 array). Two in-cores were also replaced in locations #11 and #18 at this time. During and following the core loading, core physics tests (re-quired by Technical Specifications) were performed, ie,.suberitical and shutdown margin checks.and the temperature coefficient test (see Section D g of this report and Special Report SR-lh submitted to the AEC July 5,1973).

b As required by Technical Specifications, Section h.l.l(i), the Annual Nil Ductility Transition Temperature calculation was per-formed (see Section D of this report). As a result, the primary system was hydroed at > 1h5 F prior to the plant's return to power. During the approach, however, control rod blade F-3 would not withdraw. This led subsequently to removal of the reactor vessel head and unloading of four fuel bundles and channels comprising the blade's cell. A control blade roller was found vedged in the control rod drive's " Dixie cup." After its retrieval, proper ope. ration of the drive was achieved and verified. The unit was returned on line April 16 and power was initially escalated to % 50 MWeg. At this power level and at equilibrium conditions, flux vires were irradiated to verify experimentally the computer predicted core power distributions. On April 21, plant power was escalated to 4 69 MWeg operation with an initial off-gas release rate of % 600 pCi/s. On May 3 and 12, power output was reduced temporarily to S 30 MWeg, while system substation work was in progress. On May 11, the off, gas release rate temporarily increased to % lb,000 pCi/s and stabilized to N 5,000 pCi/s, thus giving the first indication that some fuel rod clad integrity may have been breached. On May 21, power was reduced to N 200 MWt for a flux tilting test to deter-mine the location of the leaking bundle (s). Although the data were in-conclusive, no off-gas increases were noted when the flux was tilted by 4 10% in the quadrants containing the mixed oxide recycle fuel assem-blies. Also on May 21, spent fuel shipment No 20 (10 fuel bundles) left the site. for Nuclear Fuel Services, Inc. On June 1, noise was noted in the air ejector system and a sub-sequent inspection of the condenser and pipe tunnel areas resulted in packing adjustments on the SJAE steam supply, bypass valve and the Annin main steam bypass valve. -The latter valve was subsequently test-operated satis factorily. On June 20, the NFS-100 cask was loaded with 10 spent fuel assemblies for shipment No 21 tc' Nuclear Fuel Services for eventual -reprocessing. I h__

5 On June 29, alarming in-cores No 12 and No lh were cleared by the insertion of control rods resulting in a power reduction to N 208 !Gt. A subsequent re-evaluation of calibration values for these in-cores, based on . flux vire data and a core physics analysis, resulted in re-calibration of in-cores No 12 and No Ik. Verification was further nade that no ther:a1 limits had been exceeded. On June 30, plant output was returned to % 69 MWeg operation. At the end of this report period, the off-gas release rate, which had gradually decreased to % 2,h00 pCi/s at 69 Mweg operation, began a slightly upward trend. C. Changes in Procedures Which Were Necessitated by A and B or Which Otherwise Were Required To Improve the Safety of Facility Operations The following procedural changes were made with respect to plant operations: A3.1.3 - Technical Engineer duties and responsibilities des-ignated. A3.6.1, A3.6.2 - This section has been revised to include new fuel storage intrusion alar =s and key controls. A3.8.2.1 - Section revised to designate responsibility for check-off sheet approval and review. Ah.2.1.1 - Entire section pertaining to routine reports nee-essary to the facility. revised. Bl.3.1.6 - Designates NDT + 60 before primary system pressuri-zation. Ek.3.h - Designates records necessary to record explosive valve temperatures. B6.1.k - Technical Specifications ' change pertaining to emer-gency condenser operation with one tube bundle out of service. 38.3.13.1h - Revised for ' test operation of core spray syste= with closing of M)-7072 125v d-c breaker. B8.k.10 - Determination of severed line for core spray or back-up core spray line. B10.1.5 - Revision to include weekly testing of containment (' isolation valves CV kO27, kl03 and h102.. B10.2.2, 10.2.3, 10.2.k - Addition to include testing _of sphere ventilation valves for high friction.

6 B10.k.3 - Addition to check backup gas supply for containment building ventilation valves. B11.h.2 - Determination of high dP for radvaste demineralizer and how to correct same. B12.0 - Radvaste System - Solid - Entire section revised to include curie content and methods of handling and surveying solid radwaste. B28.5 - Emergency diesel generator cooling water pu=p suction line heater monitoring system and testing addition. B29.1.h - Primary system leakage limits specified as per Technical Specifications changes. B29.h - Addition to procedures to define periods of testing for primary system leakage and data to be collected and stored. B30.3.1.h - Procedures change reflecting the automatic start of a contrcl rod drive pump upon a scram signal. C8.2 - Operator action re, quired when any abnormal occu=rences happen while emergency condenser is in operation. D1.1 - #12 - New fuel storage intrusion alans and corrective actions. D1.3 - #35 - Emergency Generator Engine - Auto start failure and corrective actions. D15.0 - Addition to define fuel move from new fuel storage, logging of fuel moves, site security with mixed oxide fuels stored and l unirradiated. l E2.1.5 - Defines conditions which allow entry of the recircula-ting pump room as pertains to survey instruments. I E3.1.h -~ Personnel Decontamination - describes proper procedures l for personnel decontamination. E3.2.1.1 and E3.2.1.2 - Describes the alert and high-level alarm points and responsive action.- E3.3.1 - Description and limits for Radiation and Respiratory Protection Program. Eh.0 - Description of personnel survey instruments, their use and general surveying. l-(. b

~ Z-J 7 D. Results of Surveillance Test and Inspection Required by Technical Specifications The following listing shows the systems tested, the required test frequency, the dates tested during this report period and the results of the test (s): Containment Isolation-System: Containment isolation valve controls and instrumentation. Required Frequency: Quarterly. T_est Date: April 6, 1973. Results: The automatic controls and instrumentation for eight of the nine isolation valves were checked and found to function prop-erly. One valve (main steam drain, NO-7065) is maintained in the closed position, de-energized and not used. Therefore, testing the automatic controls of this valve' is not required. System: Isolation valve leak and operability test. Required Frequency: Twelve months or less. Test Date: April 7, 1973. Results : Satisfactory. System: Containment sphere penetration inspection (visual). Required Frequency: Twelve months or less. Test Date: March 23, 1973. 'Results: Satisfactory. System: Containment sphere integrated leak rate test. Required Frequency: Every two years. Test Date: No.-test required during this report period. Results: None. System: - Containment sphere component leakage rate test. _ Required Frequency: -Six months.or less. Test Dates: ' April lh through April 25, 1973. Results: The containment sphere component leakage rate test was performed ~ ~ using air 'at ? 20 psig., The results 'of this test showed a total leakage.of 715'of the allowable limit. 77% of this total' leakage, however has been attributed to one component,. the supply vent _. valve, which required repairs during this and- - previous -leak rate tests. - A new valve is now on site for sub-sequentLinstallation.

8 T Control Red Drives and Associated Ecuictent System: Reactor safety syste= scras circuits (not requiring plant shutdown to test). Required Frequenev: One nonth or less. Test. Dates: January 20,'/ebruary 20, P.ay 11, June 12, 1973. Results: The reactor safety syste= vas tested using the switches pro-vided to si=ulate sensor trips. All channel trips cecured as designed. In addition, the neutron nonitcring power range and inter =ediate range channels vere tested for trip setting. All of these tests showed the trip settings to be within the 1120 t 2% of power and the 10-second period setting. Syste=: Control rod perfor:ance - run. Required Frequency: Each major refueling and at least once every six nonths during period of per operation. Test Date: Mrrch 26,1973. Results: The control rod drive continuous withdrawal and insertion test, including withdraval timing, was perforced for each drive. This test is perforned during reactor shutdevn fci-Icving ce=pletion of other drive performance tests and ad-justments and represents the results of the final ti=ing of each drive. The results of this test shoved all drives to be operating satisfactorily with most withdrawal times at 36 seconds. No withdrawal time vas less than 23 seconds. Syster: Control rod perfor=ance - jeg. t Required Frequency: Each major refueling and at least once every six months during period of power operatien. Test Date: March 28, 1973. Results: The centrol rod drive latching test was perfor=ed for each l drive. This test showed satisfactory latching of all drives. i-t System: Control rod performance - scram. -Required Frequency: Each major refueling and at least once every six months during period of pcVer operation. I Test Date: March 28, 1973. (

9 i Results: The control rod scram time test was performed for each drive. The test included time from system trip to 100% of insertion at a reactor water temperature of about 150 F. The results 0 of this test were satisfactory for all drives. System: Reactor safety system scram circaits (requiring plant shut-down to test). Required Frequency: During each major refueling shutdown but not less frequently than once every 12 months. Test Dates: April 3 through April 5, 1973. Results: All safety sensors were checked satisfactorily and were within Technical Specifications limits with the folleving exceptions: a. Reactor enclosure high-pressure sensor PS/636. It oper-ated at 2.3 psig, which is 0.1 psig above setpoint licits. Pivot points of switch had high friction. Switch was thoroughly cleaned and exercised and operated properly, b. Reactor high-pressure scram and emergency cooling sensors (PS/RE07C and D). RE07C scram was found at kl psi, emergency condenser operation at 90 p?i and RE07D scra= found at 60 psi with emergency condenser operation at 88 psi. Technical Specifications limits for these switches are scram 50 5 psi above reactor pressure and 100 1 10 psi above reactor pressure for emergency con-denser operation. RE07C and D vere reset within toler-ances of 50 5 psi and 100 10 psi. Condenser vacuum scram bypass pressure switch (P215B). c. The setpoint for this pressure switch is 1 350 psig. The "as found" setting was 359 psig. It was recalibrated te 350 psig. System: Reactor safety system response time (requiring plant shutdown to test). Required Frequency: During each major refueling shutdown, but not less frequently.than once every 12 months. Test Date: April 6, 1973. l Results: Satisfactory.

10 System: Control rod withdrawal per=issive interlocks function. Required Frequency: 12 months or less; the refueling interlocks will be tested prior to each major refueling. Test Date: March 28, 1973. Results: Satisfactory. Syste=: Control rod drive friction tests. Required Frequency: During each major refueling, but not less fre-quently than once a year. Test Date: March 27, 1973. Results: Satisfactory. Emergency Cooling System: Core spray system check valves. Required Frequency: 12 months or less. Test Date: March 13, 1973. Results: Check valves function ;roperly. System: Post-incident spray system automatic control operation. Required Frequency: During each major refueling shutdown, but not less frequently than once every 12 months. Test Dates: Redundant core spray system, April 6, 1973; all other post-incident spray systems, March 27, 1973. Results: Test results on core spray system were normal. The redundant enclosure spray system test was satisfactory. During the tests on the enclosure spray system, M3-7068 did not open until five minutes,10 seconds had elapsed, or 10 seconds too long (ti=er setting 5/1/72 had been h minutes, 59 seconds). The timer was reset to open MD-7068 after a time delay of k minutes, 33 seconds. This is considered in the conserva-i tive direction. No instrument problem was evident. System:. Reactor emergency cooling system trip circuits. Required Frequency: 12 months or less. Test Dates: April 5 and 6, 1973.. Results: During testing of the emergency _ cooling system, 10-7053 would not open when given an automatic signal. Upon inspection, 10-7053 was found to have a cocked packing gland which was k-repacked and adjusted. Twelve subsequent operations showed no abnormalties. All other components tested satisfactorily.

11 System: Containment sphere isolation trip circuits. Required Frequency: During each major refueling shutdown, but not less frequently than once every 12 months. Test Dates: March 28 through April 3, 1973. Results: All trip circuits checked satisfactorily except the auxil-iary pressure switch for reactor building vacuum relief pressure switches dps-9051 and 9052. The setpoint for dps-9051 is -20.8" H 0. The switch oper-2 ated at -25.5" H 0' 2 The setpoint for dps-9052 is -27.7" H 0. The switch oper-2 ated at h3.0" E 0. 2 The switches were recalibrated to operate at their proper settings. These switches, as they are marginal in opera-tion, ve*e replaced with Barksdale snap action switches. Miscellaneous Systems System: Reactor shutdown margin test. Required Frequenev: After each refueling, after certain core ec=ponent 4 changes if system is cooled to atmosphe:1c condi-tions and after 35,000 IGdt have been generated. Test Dates : April 2 and 3, 1973. Results: Following the core loading for Cycle 11, the shutdown mar-gin verification test was performed as required by Techni-cal Specifications Section 5.5.2(b).- The test was conducted first by withdrawal of a control rod adjacent to a strong control rod to a position known to contribute at least 0.003 &/, then by withdrawing the strong control rod co=- k pletely. Additional shutdown margin tests comprised of complete withdrawal of two adjacent strong control rods after withdrawals of six notches (s 0.00h R/ ) n a neighboring k control rod. As pred.icted by computer calculations and verified by data obtained from extra nuclear instrumentation, - the core remained sub::ritical during each step of the test. Fina1' calculations indicated a total shutdown margin of 6% R/ with the most valuable control rod (3% R / ) #"117 k k withdrawn from the core. (See our Special Report SR-lh to the AEC dated 7/5/73.) Y

12 System: Nil ductility transition temperature calculation. Required Frequency: At least once each year. Test Date: March 26, 1973. Results : The calculation was based on an accrued reactor vessel 19 vall fluency of 1.5 x 10 n/cm. There is no indication 2 that any adverse NDT type problems are occurring. The most restrictive increase is displayed by the veld metal (NDTT % 850F). Therefore, reactor vessel pressurization in excess of 20% of normal operating pressure will not be allowed to occur at temperatures below 1k5 F (Max NDIT + 600F). System: Moderator temperature coefficient. Required Frequency: Following each najor refueling outage. Test Date: April 6, 1973. Results: Results of the test indicate a maximum reactivity addition of 12 cents while heating up the moderator coolant from ambient (% 70 F) to the turnover (reactivity decreasing) point of 137 F. The Technical Specifications limit is one dollar or approximately nine times greater than the experi-mentally determined value using 1* = 0.309231 x 10 and ~ 8,ff = 0.00613315. (See our Special Report SR-14 dated 7/5/73.) System: Suberiticality checks. Required Frequency: -During core alterations which increase reactivity. Test Dates: March'28 through April 2, 1973. Results: Incorporated into the core loading procedure, BRP-RE7, sub-criticality checks were performed throughout the reconstitu-tion of the core. 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 Position 06, the adjacent rod fully withdrawn). By monitoring instrumentation responses, verification was made that no critical condition was approached during the core loading. ( ye e-+ w a

13 System: In-service primary system inspection. Required Frequency: A continuing program being conducted during sone major refueling outages. Test Date(s): Conducted during March 1973 refueling outage. Results: Satisfactory (see separate Report No 16 dated 8/20/73). System: Refueling operation controls. Required Frequency: Each major refueling. Test Date: March 6,1973. Results : Satisfactory. S ptem: Reactor refueling safety system sensors and trip devices. Z Required Frequency: Each major refueling. Test,_Dat es : March 27 and April 6, 1973. Results: Satisfactory. March 27 date includes sensors for refueling operation. April 6 includes all safety system sensors and trip devices. Poison Systems System: Liquid poison system firing circuit test. Required Frequency: Two months or less.. Test Dates:. February 5, April 6 and June 6, 1973. Results : The liquid poison system circuits monitoring the poison in-jection valves circuits "A" and "B" were checked by opening the isolation breaker for each circuit. The monitors fune-tioned properly on each test. System: Explosive valve from equalizing line. Required Frequency: Twelve months or less. Test Date: March 26,1973. Results: No abnormalities. System: Explosive valve from nonequalizing lines. Required Frequency: Twelve months or less. Test Date: - March 26,1973. Results: No abnormalities. . Radiation M nitoring o System: Air ejector off-gas monitor system. ( Required Frequency: One month or less.

1h i Test Dates : January 31, February 28, March 25, April 25, May 24, June 28, 1973. Results: The air ejector off-gas monitoring system logarithmic moni-toring instrument was checked using the instrument calibra-5 tion test points at 1, 10 and 1 x 10 units. The checks showed the celibration to be satisfactory (within 20%). The automatic closure function of the isolation valve timer was checked. The test showed the timer calibration to be satisfactory (within 3% of the maximum timer setting) and the isolation valve to close as specified. System: Calibration and functional test of the stack-gas monitoring system. Required Frequency: One month or less. Test Dates: January 31, February 28, March 25, April 25, May 24 and June 28, 1973. Results,: The stack-gas monitoring system was checked using the built-in calibration source (Cs-137). The instrument :: heck showed the calibration to be satisfactory, result $ng in the alarm point occurring within the specified 0.1 curie per second release rate. 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 the iodine filter are reported in terms of curies re-leased in Appendix A of this report. - System: Calibration of emergency condenser vent monitors. Required Frequency: One month or less. Test Dates: January 29, February 28, March 25, April 25, May 24 and June 28,.1973. Results: The emergency condenser vent monitors are checked by com-paring with a calibrated portable instrument. The checks showed the vent monitor calibration to be satisfactory with' f 'all monitor checks within ! 5% (of full scale). y' 1-e

15 Syste=: Calibration of canal liquid process acnitor. Required Frequency: One month or less. Test Dates: January 31, February 28, March 25, April 25, May 2k and June 28, 1973. Results: The calibration of the " radioactive vaste syste= effluent to canal" monitor is a ec=parative calibration used to denonstrate operations of the monitor and to detect gross calibration changes. The results of these =onthly cali-brations shoved that the r.cnitor was functioning properly and that no gross monitor drift had occurred since the original calibration which utilized certified standards. Syste=: Canal liquid collection sa=ple. Required Frequency: Daily. Test Dstes: The analysis was conducted daily. Results: Satisfactory. E. The Result of Any Periodic Contain=ent Leak Rate Test Perforced During the Reporting Period No integrated contain=ent leak rate test was perfor=ed during the report period. F. During this report period, the following Technical Specifications changes were authorized by the Cec =ission. Change No 35 - This change provides for use of two new auxiliary anti =ony-beryllium neutron sources for use in 11 x 11 array fuel. Change No 36 - This change provides clarification in the opera-bility requirements of the e=ergency condenser. Change No 37 - This change includes new sections covering Plant Reporting Requirements and a Radiation and Respiratory Protection Progrs=. G. Changes in Plant Operating Organization Involving Key Supervisor Personnel There were no changes in key personnel as defined in Technical Specifications Paragraph 7.2.2.1.2a (1) (2) (vii) during the report period. { m e

t III. PO*='ER GENEPATICN Report Total Period To Date (i) Therra1 Fever Generated (Ka'h ) 605,7k2 11.385,713 (ii) Gross Electric Pcver Generated (Ka'hes) 192,25k 3,632,202 (iii) Net Electric Pcver Generated (MWhe) 162,k25 9 3,L39,782.8 (iv) Hours Critical (h) 3,278.6 6L,907.6 (v) Hours Generator On Line i (h) 3,2hk.3 63,131.3 1 i \\ l 1 t s

IV. SHUrDOWPS During this report period, two outages occurred. A description of each outage, listed in chronological order, follows halow: OUPAGE REPORT Type of Outage - Forced (1/73) Length of Outage - 25 h, 2 min Unit Off Line - 0050 h (1/20/73) Unit On Line - 0152 h (1/21/73) The plant was removed from service on a forced outage due to primary coolant leakage at the packing on the reactor clean-up system dis-charge valve (CU-1) to No 1 reactor recirculating pump discharge piping. New packing was installed in the valve and the plant was returned to service. The method of shutting down was a controlled, deliberate shutdown and the unit's status during the outage was cold shutdown. OUTAGE REPCRT Type of Outage - Scheduled (2/73) Length of Outage hk Days, 18 h, 17 min Unit Off Line - 231k h (3/2/73) Unit on Line - 1731 h (h/16/73) The plant was removed from service on a scheduled outage for the tenth refueling of the reactor. The unit's status during the outage was cold shutdown. For details, please refer to Section II - Operations Summary, Section V - Safety-Related Maintenance and Section VI - Changes. Tests and Experiments. =. - -

V. SAFETY-RELATED PAINTENANCE Note: Dates contained in this section generally refer to the weekly period when the maintenance was performed. A. Reactor Protection and Control System Instrumentation 1. Neutron Monitoring Channel No 1 h/5/73 - The high-voltage power supp3; for this unit was a. repla:ed following erratic operation of the channel. Bench testing of the faile'd power supply revealed that it intermittently lost regulation on the compensating supply. Repairs to this unit consisted of electron tube re-placement. This failure occurred while the reactor was shut down for refueling and caused downscale picoe.cmeter indication and downscale trip. This type of failure is considered to be within the design limitation of the equipment. b. h/17/73 - The picoammeter in this channel was replaced with the spare following erratic operation of the unit in service. Bench repair of the failed unit resulted in replacement of a microphonic electrometer tube in the d-c amplifier. Failures of this type are considered to be within the design limitation of the equipment. This failure occurred with the reactor in " cold shutdevn" prior to return to power following the tenth refueling outage. 2. Neutron Monitoring Channel No 2 h/5/73 - The compensated ion chamber and coaxial ce,bles from the chamber to the chamber drive were replaced following upscale indica-tion on the picommmeter. Water leakage from the fuel shuffling cask had drained into the chamber drive pit and saturated the chamber and cables. This failure occurred during refueling operations, causing a cessation of same until the unit was repaired and returned to service. The chamber drive pit vac reseded to prevent a recurrence of the failure. 3 Neutron Monitoring Channel No 3 1/k/73 - The picosameter in this channel was replaced following downscale failure of the unit in service. Bench testing of the failed unit resulted in replacement of a defective series regulator tube (and cracked socket) in the +150 volt internal power supply.

2 E Failures of this type are considered to be within the design limitations of the equipment. The Technical Specifications and plant de-sign provide for removal of one power range flux monitor from service with- - out compromising safety. 4. Neutron Monitoring Channel No h 6/7/73 - The high voltage power supply in this channel was re-placed with a spare unit following apparent loss of regulation of the posi-tive polarizing supply. Initial repairs consisted of realigning and tightening the voltage set-point knob; additional bench testing revealed no other problems. Failures of this type are considered to be within the design limitations of the equipment. The Technical Specifications and plant design do not require this instrument to be in service when reactor power is above 5% of rated power. 5 Neutron Monitoring Channel No 5 6/7/73 - The Log N-Period amplifier in this channel was replaced with a spare unit due to erratic period indication of the unit in service. Bench repair of the defective unit consisted of replacement of a defective electron tube in the period amplifier circuit. This type of failure is con-sidered to be withir. the design limitation 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. 6. Neutron Monitoring Channel No 6 2/1/73 - The log count rate meter in this channel was repaired following full scale failure. The failure was corrected by electron tube replacement in the internal +150 volt power supply.- Failures of this type are considered to be within the design limitation of the equipment. The Technical Specifications and plant design do not require this instrument to be in service when the reactor is at power. 7 Motor-Generator Sets 3/15/73 - The inboard and outboard flywheel bearings were re- - placed (due to excessive noise and heat) on the No 1 motor-generator set.. The motor was cleaned and the motor and flywheel realigned before returning the unit to service.

3 i All equipment povered through the No 1 motor-generator set was transferred to Instrument and Control Bus 1-Y (alternate source) during the period of time the No 1 motor-6enerator set was out of service. In ad-dition, the reactor was in the cold shutdown condition. B. Radioactive Effluent Monitoring Systems 1. Air Ejector Off-Gas System 3/15/73 - The intericr of the 2h" holdup pipe was inspected a. using the underwater television camera, steel tape and measurement jib. Ap-proximately 30' of the pipe underwent visual inspection with the camera, 90' ~ with the steel tape and the pipe diameter at the stack base was measured to be 2h". From all observations, it appears that the pipe is the di-ameter shown en the plant prints (2h") and is free of obstructions and/or moisture pockets. The inspection was performed during the time the *1 ant was shut down for refueling. b. 5/31/73 - The input electrometer tube in the Channel No 1 log radiation monitor unit was replaced following reports of erratic re-sponse on this channel. Control of the off-gas isolation valve var performed by the No 2 monitor while this unit was being repaired. 2. Stack Gas Radiation Monitoring System 3/8/73 - A defective alarm backset switch was replaced on a. l the stack gas radiation monitor recorder. l' Removal of this recorder for repair did not disable the l stack gas monitoring system. Count rate indication was available on the 1 l log count rate meters. b. 6/7/73 - The linear emplifier.in this system was replaced with a spare unit following calibration drift in the single isotope chan-nel. Repairs to the defective unit consisted of electron tube replacement.. I Failures of.this type are considered to be within the design limitation of the equipment. Removal of this system' from service is permitted by the .( Technical Specifications provided repairs are made promptly and the system is returned to service.. The off-gas monitors provide backup for this moni-toring system.

h ( 3 Liquid Process Monitoring a. 1/25/73 - Sphere Service Water Liquid Process Monitor - The linear count rate meter in this channel was replaced with a spare unit following full-scale failure. Repairs consisted of electron tube replace-ment and calibration check. b. 1/25/73 - Radvaste-to-Canal Liquid Process Monitor - The linear count rate r-ter. in this channel was also replaced with a spare unit following full-scale failure. Repairs consisted of electron tube replace-ment and calibration cheer.. c. 2/22/73 - Discharge Canal Liquid Trocess Monitor - The linear count rate meter in this channel was replaced with a spare unit fol-loving questionable operation of the count rate meter alarm. Bench testing of the unit indicated that it was performing properly. Precise calibration of alarm set point is difficult to achieve on these units with the low random count rates that exist. d. h/19/73 - Sphere Service Water Liquid Process Monitor - The linear count rate meter in this channel was replaced following erratic op-eration of same. Repairs to the faulty unit consisted of electron tube re-placement. e. h/19/73 - Sphere Cooling Water Liquid Process Monitor - The linear count rate meter in this channel was replaced following upscale fail-ure of same. Repairs to this unit also consisted of electron tube replace-ment. f. 5/10/73 - Radvaste-to-Canal Liquid Process Monitor - The linear count rate meter in this channel was replaced following failure of the unit in service.. Repairs' to the failed unit consisted of replacement of a shorted electron tube and overheated resistor. Failures of these types are considered to be within the 1e-sign limitations of the equipment. Removal of these systems from service are permitted by the-Technical Specifications provided repairs are promptly made and the system is returned to service. C. Containment and Associated Isolation Systems j. Solenoid Valve Replacement '1/18/73 - Solenoid valves for the instrument air dryer transfer (SV-kB67)land the reactor cooling water demineralizer water makeup valve

5 CV kO26 (SV-kBTT) were replaced with experimental valves during the re-porting period. The new solenoid valves use a zytel dise with stainless steel disc and resilient 0-ring seats. Selection of these valve (nonsafety related) locations vere based on the duty cycle of the valves (ie, the air dryer transfer occurs three times a day, resulting in regular operation of the solenoid valve while the reactor cooling water makeup valve may go for long periods of time with the solenoid valve in an energized condition - siellar to the operating status of several isolation valve solenoids which have failed in the past). D. E=ergency Power System E=ergency Diesel Generator 5/3/73 - Three panel alarm lamp sockets vere replaced with bayonet-type sockets to improve lamp reliability. The original units were of the candelabra base type and susceptible to vibration (bulk would beco=e loose in the socket and fail to 'li ht ). At the present time, three different types 6 of lamp sockets are presently in use on the panel and they vill all be re-placed with coccon type sockets. The lamp sockets provide alar = indication only for a co= mon trip and annunciator scheme on the diesel generator. Replacement of the la=p sockets involved low-voltage viring and no special precautions were re-quired to provide for raaetor safety. E. Fire Protection System Fire Pump Breaker - Panel 2B The Region Electric Lab adjusted the instantaneous trip devices on the breaker for maximum current of S 1700-1750 amps. This is in refer-ence to recent recommendations from the General Office Electric Operations Department. The lab verified that there is no thermal trip ' device associ-ated with the breaker. Adjustment of the trip setting provides maximum reliability of the a-c fire pump. The diesel fire pump provided fire system backup during the breaker testing period. F. - Liquid Poison System {- 1. Squib' Valves ' 2/1/73 - Temperature indicators were installed at the poison tank level to monitor the block' valve _ temperatures on CV bl21 and CV bl22.

6 These replaced the T/C temperature recorders previously in use, following a two-week test period. Data vill be taken from the indicator; at 0700, 1500 and 2300 hours and the highest reading obtained (usually at 1500 hours) vill be plotted on a graph in the control room. This will be the permanent record .of operating temperature of the squib valves in these two assemblies. The reactor was at operating temperature and pressure during this replacement. There were no safety implications involved. 2. Three-Inch Check Valve 3/23/73 - During the in-service inspection of this system, ex-ternal pitting in the vicinity of the 3" check valve adjacent to CV h020 was traced to slight water seepage from this check valve. The valve flange bolts were tightened to stop all seepage. During this repair the reactor was in the cold shutdown condition ~. 3. Explosive Squib Valves 3/29/73 - Two explosive squib valves from the poison system were test-fired (CV bl20 and CV-h122) and the liquid poison system circuit resistance tests were completed. The reactor was in cold shutdown during replacement of the squib valves and no additional precautions were required to provide for reactor safety. G. Emergency Condenser System 1. Emergency Condenser 3/5/73 - The emergency condenser was hydrostatically tested (1685 psig) to determine integrity. Minor leakaga was found from the in-ternal water box head seal on the south tube bundle. This was corrected by tightening.the head bolts. Retesting hydrostatically (1685 psig) indi-cated no leakage. The north tube bundle and condenser shell hydrostatic test (750 psig and 9 5 psig) indicated no leakage. 2. Emergency Condenser Valve 3/6/73 - The inlet valve (MO-7052) to the No 1 loop was disas-sembled for in-service inspection. The inspection revealed no significant defects. 1-

3. -Limitorque-operators h/9/73 - The limitorque operators on the emergency condenser outlet valves MO.7053 and M0-7063 were inspected and adjusted to provide-

7 proper closing and opening torque and limit switch operation. These valve limitorques have been placed in a two-year preventive maintenance schedule. The reactor was maintained in the cold shutdown condition while performing the above repairs and inspections on the emergency condenser system. h. Inlet Valve MO-7052 6/1h/73 - The emergency condenser inlet valve M0-7052 was found to be leaking steam very slightly at the packing. Tightening of the valve stem packing corrected this problem. During this repair the reactor was at operating pressure and temperature. This valve is maintained in the open condition and is required to be open to perform its safety function. H. Post-Incident System 1. Fost-Incident System Valves 3/29/73 - The 25 gpm relief valves (RV-5077 and RV-5078) a. were replaced with 7.6 gpm relief valves set to relieve at 1h5 psig. b. 4/12/73 - The limitorque operators on the post-incident system valves (MO-7051, MO-7061, MO-706h and MO-7068) vere inspected and adjusted to provide proper closing and opening torque and limit switch operation. During these repairs on the post-incident system, the re-actor was in the cold shutdown condition. I. Reactor Clean-Up System 1. 1/21/73 - High airborne activity in the' containment was traced to primary coolant leakage at the packing on the reactor clean-up system discharge valve (CU-1) to the No 1 reactor recirculating pump discharge piping. The plant was removed from service and placed in the cold shutdown condition while installing new packing in the valve. 2. 3/12/73 - The clean-up demineralizer drain valve (CU-118) was replaced due to excessive seat leakage. The replacement involved three. carbon steel butt velds. A nev 3" valve and short spool piece were in-stalled. All completed welds were inspected to the requirements of the ASME Boiler and Pressure Vessel Code, Section III, 1971. During this re-pair, the reactor was maintained in the cold shutdown condition. ' J. Reactor Vessel 1. 3/5/73 - The reactor vessel head and associated piping vere removed in-preparation for reactor reineling. b M N v v v

f 8 1' 2. h/3/73 - Installed reactor vessel head and associated piping in preparation for power operations. 3 h/12/73 - Removed reactor vessel head and associated piping to remove a control rod drive roller from control rod blade F3 " Dixie cup." h. h/14/73 - Installed reactor vessel head and associated piping in preparation for power operation. The reactor vessel was hydroed at 1000 psig after final instal-lation of the head prior to returning to power operations. K. Primary Coolant System 1. Reactor Recirculation Water Pump No 1 3/lk/73 - The No 1 pu=p and motor were removed fro = the pri=ary system, disassembled and inspected as a preventive maintenance require =ent. The pump shaft bearing proved to be in excellent condition with no measur-able voer (graphite bearing). The p1mp shaft thermal barrier area evidenced minor wear of an acceptable nature as did the impeller in the areas of the case and cover wear rings. The pump motor proved to be in excellent condf-tion with no indications of wear or deterioration. The reassembled motor 3-displayed an end float in the Kingsbury thrust bearings of 0.010 to 0.012 inch which is on the high side of the allowable 0.006 to 0.010 inch float but is well within the running clearances required in the pump-motor unit. Prior to reasse=bly of the pu=p, Southwest Research Institute completed a visual inspection of the pu=p case, an ultrasonic inspection of the pump case mating flange and magnetic particle and ultrasonic inspections of the pump cover-to-case flange bc3ts. All NDT results were satisfactory. The pump was then reinserted in the pump case cavity in the primary syste=. The pump motor was coupled to the pump through the pu=p-motor spacer. All mating surfaces were cleaned thoroughly prior to installation. The total end float on the pump auxiliary impeller after pump installation was 0.165 inch. ' With the pump shaft raised 0.090 inch (leaving 0.075 inch clearance between the upper surface of the. auxiliary impeller and the pump cover), there remained 0.035 inch clearance between the thrust disc of the pump-motor. coupling and the motor shaft. A 0.330 inch shim was added on top of the thrust disc, let7tng 0.005 inch for final compression of the car- .tridge seal springs (which properly seat tna carbon ring seals) when the motor and pump were coupled through_the. coupling spacer. (This final LJ

9 0.005 inch compression upon coupling raised the pump shaft a like amount, leaving 0.070 inch final clearance between the upper surface of the aux-iliary impeller and the pump cover.) Concentricity and parallelisms be-tween the pump and motor shafts were controlled by the proper assembly of the pump, pump-=otor spacer and motor mating surfaces. Upon reconnection of all auxiliary lines and leads, and the satisfactory check of free rota-tion of the reassembled pu=p-cutor unit shafts, the No 1 recirculating pump was released for return tt service. The oil reservoirs supplying the upper (thrust) end lower (guide) motor shaft bearings were drained and resupplied with fresh oil on No 2 reactor recirculating pu=p motor. The spare mechanical seal was cleaned, inspected and reassembled and installed with new rotating faces, one new stationary face (other stationary faces relapped) and all new seals and 0-rings. During the inspection of the recirculation water pu=p, the reactor was in the cold shutdown condition and the pu=p was isolated and properly tagged from the primary coolant system. 2. Reactor Recirculation Water Pu=p No 1 h/8/73 - During the hydrostatic test on April 7, 1973 at 100 psi (> 150 F)' the No 1 reactor recirculation water pu=p exhibited leakage in the flange area. The pump and motor were removed from the pu=p casing and inspection revealed an improperly sized flexitallic gasket in use for sealing the pump case and cover mating flange. The proper gasket was in-stalled and the pump reassembled and returned to service. During this repair, the reactor was placed in the cold shut-down condition and the pump was isolated and properly tagged from the primary coolant system. L. Control Rod Drive System

1. -Control Rod Drive Pumps a.

h/26/73 - Excessive leakage of the No 1 and No 2 control rod drive pumps necessitated replacer ent with spare valves set to relieve at 1950 raig. b. 6/21/73 - Leakage from the No 1 control rod drive pump ( suction inspection cover was corrected by replacement of the cylinder. head gasket.

10 The above repairs on the control rod drive pumps were made with the reactor at operating pressure. One of the two control rod drive pumps may be removed from service and still maintain normal operational status. 2. Control Rod Drives and Instrumentation a. 1/25/73 - The following repairs were made to the control rod position probes during the plant outage of January 20: D-h - Repaired the probe connector assembly pins for the drive temperature thermocouple, D Adjusted the "00" position lights (% 1/8") to pre-vent losing indication on this drive following a scram or shutdown, and F Repaired a broken solder joint in the probe connector to restore position indication to Positions 2, 12 and 22. The reactor was in the cold shutdown condition. b. 3/10/73 - The D-4, F-4 and A-5 drive mechanisms were replaced with rebuilt drives as a preventive maintenance measure. The drive mounting thimble "J" welds were ultrasonically inspected at this time as a part of the in-service inspection requirements. c. 3/16/73 - Due to excessive leakage at the flange 0-rings, control drive F-2 was dropped for replacement of the flande 0-rings. In e.ddition the E-6 drive mechanism was replaced due to high withdrawal _fric-tion. d. h/5/73 - Control rod drive D-b position indication probe was replaced with the spare unit to eliminate loss of indication on Posi-tions 2, 12 and 22. The defective probe was repaired (solder connection) and will be used as a spare. The reactor was in the cold shutdown condition while all L the above drives and instrumentation were repaired or replaced. -e. 3/15/73 - A new bladder and 0-ring seals were installed in ' correction of leakage from the E-2 accumulator. .New teflon seats and new piston rings were installed on the pistons on the rod drive insert wui withdrawal header Atkomatic sole-noid valves. '{ The leaking E-4 accumulator DECS valve 112 was reground and repacked.. i s L

11 k f. 3/15/T3 - The inlet and drain valves on both control rod drive filters were replaced due to excessive seat leakage. The replacement valves were inspected by.either radiographic or ultrasonic methods. The velds were made in accordance with the requirements of the ASME Boiler and Pressure Vessel Code, Section III, 1971. g. h/12/T3 - A leak in the D-1 accumulator was corrected by replacing the 0-rings between the accumulator halves. The above repairs (e thru g) were performed with the reactor in the cold shutdown condition, h. 6/21/73 - The E-3 drive accumulator leakage by the charging valves and check valve was corrected by replacement of the valve. The above repair was conducted with the reactor at operating pressure. Under these conditions, the primary hydraulic source for drive scram =ing comes from the reactor vessel. This design feature permits repair of an accu =ulator with-out affecting reactor safety. 3. Control Rod Drive Scram Accumulators a. 2/8/T3 - The auxiliary relays in the rod drive accu =ulator alarm and rod block circuits were disassembled and inspected. (Recently high contact resistance has caused noise interference en several monitoring systems.) All contacts were cleaned and the defective (broken rocker posts) F-2 relay and the C-6 relay (formerly improper type) were replaced. b. 2/15/73 - The D-5 and F h withdrawal rate set valve leaks were eliminated by replacing the steam seal 0-rings and backup rings. 2/22/73 - Leaks from the B-3 and E h accumulator valves c. were corrected by. replacing the valves.

d. - 2/22/73 A' leak from the-B-3 accumulator was corrected by-replacing the seals in 'the lover. accumulator half.

The above repairs'(a thru d) were conducted with the reactor at operating pressure. Under these' conditions, the primary hydraulic source for drive scraming comes from the-reactor vessel. This design feature permits repair of an accumulator or accumulator instrumentation. without.affecting reactor' safety. -M. Reactor Feed-Water System 1. Shutdown Flushing Line { 3/5/73 - Water seepage vas-determined to be coming from the blank flange on the shutdown flushing line where it joins the reactor ~

12 i feed-water line under the stea= dru=. Upen re:cyal of the 6" flange it was detemined that the surface of the flange had been stea= cut which was veld repaired. The "lange was then reinstalled. During this repair, the reactor was =aintained in the cold shutdown condition and the syste= isolated and properly tagged. 2. Reactor Feed Pu=p No 2 Discharge Check Valve 3/lh/73 - This valve had exhibited indicaticns of " sticking open" during operations when the No 2 feed pu=p vas re=oved frc= service. The No 2 reactor feed pu=p discharge check valve was re=cved and replaced with a spare valve. No velding was involved. No direct cause for the sticking of the valve was found. With the replace =ent valve installed the proble of " sticking" has not occurred. During this replace =ent the reactor vas' in the cold shutdevn condition and the valve was properly isolated and tagged. 3 Peactor Feed Pu=cs 3/26/73 - As routine maintenance the No 2 feed pu=p =otor a. - vindings were cleaned and inspected and the =otor lubrication syste: ec=- ponents were inspected and adjuste:1 as necessary. The pu=p packing was adjusted and the =otor-pu=p coupling was repacked and realigned. - The No 1 feed pu=p motor vindings were cleaned and in-spected and the motor lubrication syste= co=ponents were inspected and ad- -justed. During their. inspection / repairs the reactor was maintained in the cold shutdown condition. b. 5/3/73 - The outboard pu=p packing on the' No 1 pump-and inboard pump packing.on the No 2 pump were replaced due to packing failure. The Lpacking failure was caused by initial overtightening and insufficient break-in. During these repairs the reactor was at operating te=pera-ture and pressures. Lone reactor feed pump at a ti=e was removed frc= ser-vice, isolated and tagged prior to perfoming work. One reactor feed pu=p was always available and in use. . N. Main Steam System - .] . 1.~ Main' Steam Isolation valve 3/5/73'- During the. shutdown frce power operation to cold shut-down, the main steam isolation. valve' failed to close on the initial signal.

13 (- The valve closed on the second and subsequent attempts. Disassembly and inspection of the valve revealed the packing gland to be bound to the valve stem and the follower with the stem covered with considerable amounts of hard white deposits. The valve was thoroughly cleaned and reassembled with a new type braided carbon packing, tested and returned to service. The above repairs were conducted with the reactor in the cold shutdown condition. Prior to returning to power operation the valve was test-operated near operating pressure and temperature. O. Stea= Drum System Drum Relief Valves 3/7/T3 - The No 3 steam drum relief valve was removed for inspec-tion and subsequent placement in a spare status. The replacement relief valve (previously set to relieve at 1535 psig) was installed in the No 3 position. During this removal, the reactor was in the cold shutdown condi-tion. P. PrL=ary System Leak Detection System Work continued this report period on the containment sphere dirty and clean sump piping modification to establish separate drainage for " identified" and " unidentified" leakage from the primary system for a leak rate nessuring system. When completed the modification vill be incorporated as facility changes and reported as such. L

VI. CHANGES, TESTS AND FJ.FE9IMENTS A. Facility Changes Perfor=ed Pursuant to 10 CFR 50 59fb) 1. Facility Change C-151 - This change involved =odifications to the control rod drive starting circuitry to provide auto =atic start of the second (standby) pu=p following a scrs=. This change was perfo med to limit the control rod drive outlet water te=peratures following a scra=. This vill reduce the te=perature transient on the rod drive flange as well as the dunp tank and vill be most beneficial for the hot standby condition when all rods are inserted and protection systaa testing is being perfor=ed. This change was reviewed by the Safety Audit and Review Ecard (SARB) with the reco==endation that the =odifications be perfo med. SARE concluded that the change does not require a change to the Technical Specifications nor does it involve an unreviewed safety questien. 2. Facility Change C-185 - This change involved =odifications to the emergency diesel generator. It consisted of recalibration of the low lube oil pressure switches to prevent reset.of sa=e during the crank-ing cycle. The new set po.nt is 20 psig on falling pressure. This change var perfor=ed when it was observed that extended cranking on the diesel generator could increase the oil pressure t2 a point where the lube oil trip switch would reset and initiate a trip when cranking ceased. The no mal operating lube oil pressure is 52 psig. ~ The safety analysis concluded that this change vould i= prove the syste= reliability and that it does not involve a change in the Technical Specifications or an unreviewed safety question. 3 Facility Change C-196 - This change re= oves the voltage sta-bilizing transfomer in each of the air ejector off-gas systems. The safety analysis concluded that external voltage stabilization is not re- -quired since the present off-gas instrucentation contains its own' internal voltage stabilization equipeent. Therefore, it was concluded that the change does not involve a change in the Technical Specifications nor is there an unreviewed safety question. 1 I

2 h. Facility Change C-205 - This change consisted of repositioning the emergency condenser gauge glass by raising it approximately three inches to place normal water level at an observable point and not behind a junction of the gauge glass windows. Also, the gauge was rotated 90 to allow easier observation of the level. Following completion of the facility change, the instrumentation header was leak tested at 150% of containment design pressure (40.5 psig) to verify system integrity. The safety analysis concluded that this change would improve readability, reduce a personnel safety hazard and that a change in Tech-nical Specifications or an unreviewed safety question is not involved. 5. Facility Change C-210 - This change involved the addition of a new local metering panel to the emergency diesel generator to provide indication of output voltage, current, frequency and exciter voltage. The meters are mounted in a separate panel on the south wall of the emergency diesel generator room. The output voltage and current meters will be a backup for those mounted in the generator (subject to vibration), the exciter voltmeter for monitoring exciter condition (to detect field degradation) and the-frequency meter as an operating aid. The safety analysis concluded that this addition would improve system reliability and that the change does not involve a change in the Technical Specifications nor is there an unreviewed safety question. 6. Facility Change C-211 - This change involved relocation of the power source for the five-minute time delay relay (62-1) in the reactor enclosure redundant backup spray valve (MO-7068). The original power for this relay was in an auxiliary scheme receiving its power from Instrument and Control Transformer lY. This facility change places the time delay relay _in the control circuit of the valve. This change was performed to correct a deficiency revealed in a control circuit review (described in February 21, 1973 letter to the DOL). l The safety analysis concluded that this change vill improve redundant enclosure spray reliability and that it does not involve a { change in the Technical Specifications nor is there an unreviewed safety l_ question. ~ i. ~ =,.

3 { B. Tests Performed Pursuant to 10 CFR 50.59(b) 1. Hydrostatic Test of North and South Emergency Condr aser Tube Bundles 3/1/73 - On December 16, 1972, leakage of the emergency con-denser stack vent occurred during a reactor start-up and subsequent heatup (reported to AEC January 16, 1973). On March 1, 1973, after the reactor was shut down and placed in the cold shutdown condition for refueling, a special hydrostatic test and visual inspection were conducted on the north and south emergency condenser tube bundles to determine which, if any, tube bundles were leaking. A pressure of 1,685 psig was applied to the e uth tube bundle and minor leakage was found from the internal water box head seal. This was corrected by tightening the head bolts. A retest at 1,685 psig on the south tube bundle indicated no leakage visually. The north tube bundle was hydrostatically tested to 750 psig with no leakage noted visually. The condenser shell was hydrostatically tested at 9 5 psig with no leakage indicated. The conductance of this test was controlled by written proce - dures. A prior review of this test determined that this test.vas con-sistent with Technical Specifications and did not involve an unreviewed safety question. 2. Main Steam Valve MO-7050 Special Operational _ Test h/16/73 - on March 3, 1973, the main steam isolation valve MO-7050 failed to close on initittl operation of the manual controller (reported to the AEC on April 5,19(3). The Plant Review Committee recommended that a new paaking be installed in this valve and a test be conducted at or near operating pressure and temperature to demonstrate operability. On April 16, 1973, while the reactor was heating up on re-turning to power operations, the main steam isolation valve (M0-7050) was operated at 1,000 psig in the close direction for 10 seconds and then returned to the fully opened condition. This valve when 50% closed will result in a reactor scram (23 seconds from full open toward closing position). Results of this test'were satisfactory. This test was con-I' trolled by the use of a written procedure and prior review of the-

k _1 procedure indicated that it was consistent with Technical Specifications and did not involve an unreviewed safety question. 3 Emergency condenser outlet valves MO-7053 and M3-7063 Operation Due to the adjustment of Limitorque valve torque settings and limit stops, the emergency condenser outlet valves MO-7053 and MO-7063 were tested at or near operating pressure and temperature upon recon =enda-tion of the Plant Review Committee. Operation of these valves was individually performed at an operating pressure greater than 600 psig and one loop at a time of the emergency condenser was removed from ser-vice. Operation and leak tightness of these valves was satisfactory. After completion of this test, the emergency condenser was returned to two-loop operation. This test'was performed utilizing a written procedure. Prior review of this procedure indicated that it was consistent with Technical Specifications and did not involve an unreviewed safety question. t ~ The Big Rock Plant Technical-Specifications allow temporary removal of one loop from service, or isolation, if a leak develops. ( y y v

m. VI. AVERAGED DAILY THERMAI. POWER (MWt) 1973 ^ g l,, fi . p. h ~ [u.'l 't ~ l 2-' .I a .y _ ,s. t, i m. p. . :m p. l ld .L. -4 ~ s a u -gr i + -h a r ld.il-' "~' il D [; q : { a -r a ~ i TY ~' ^~7 3.10 4 i r ~ ...t p. .. l l TT~ i .t. . j., u {,}1^ {t gg u 1 y- ~ too 11l 4 r 7 7 a.: # J s s. s I t o o,Fft) a L hy%y %p;w.. m o L,.A' 2 9 ine ,a . l - k.. 4 yl Jr-;f uRVp t i s '-f - 6 l"l l .--z r w ,44 f %w, t' y!! 7 _ y _1_.__ e. l r g, j.>., Q' ^ ;9.htq g.m j . m 2. i . n t.A_ m , u.h.m l ' I al. g i g t R -4 .r. i g g,o f.Q h[gs@d,Mf47).4[jQ.,[ . o ~-. a - +. ~*1~~ ~' # T Qtg. e %@i$ gi M J$y ~ 73-F r I-N w s' n H N $t %hY T ~ lr i i a + i a 0 tio g'di.f. y.%*YjgnAgi;g19gj ~ T. .sg i .). G: 1 3 wg: k y 4. 4,.l" 'm 4.k. An,t Q P ( % "- p <e--- .L -I-i'l - 1 } W tto t "Y .e se J -'.x .M_.,, s 1, '.2. ..l r ,4 l 7f;{Q %f)AN;(.h.Q][f,4 . h*far999 t. -L,-,. H' <R f %.fg ' '40,45, (O 3

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4 VII. RADIOACTIVE EFFLUO.T RELEASES Introduction Releases of radioactive =aterial both to the atmosphere and Lake Michigan frc= January 1 to June 30, 1973 were well within the facility-licensed limits and the AEC's regulations, particularly Title 10, Code of Federal Regulations, Part 20.

i. Gaseous Effluent Gaseous releases to the atmosphere totaled 11k,000 curies of fission and activation gases. This corresponds to 0.72% of the licensed technical specification limit of 1 Ci/s. Particulate releases totaled 0 3L curie or 5.7% of the licensed limit while halogen releases were measured to be 1.5 curies or 16% of the licensed li=it.* Gross alpha measure =ents on the particulate filter revealed that the release of alpha emitting nuclides totaled 1 x 10 curies. The tritin= releases for the

_ c: period totaled L2 curies or 3.4 x 10 '% of li=it based upon =eteorological dispersion to the point of =axi=u= ground concentration. Gaseous Effluent Calculational Methods A sa=ple of off-gas is obtained weekly during power operation and analyzed by ga-a spectro =etry for ++six noble gas radionuclides. Based upon the mixture of the six nuclides, a stack release rate, which includes a total of 22 ncble gas radionuclides, is determined. The . stack release rate is based on a 16-minute holdup time for off-gas plus a 1% contribution frc= the turbine sealing steam syste= utilizing a 2-minute holdup.*** The 1% turbine seal contribution has the same distri-i bution of nuclides as the off-gas corrected for a 2-=inute decay period. This is reflected in the monthly totals shown in Appendix A. Activation gas releases are co= posed pri=arily of N-13 The rate of release is power-level dependent and is incorporated in the total monthly releases shown in Appendix A.

• Due to uncertainties in iodine collection efficiencies for various species and potential sa=ple line plateout, the measured values will be arbitrarily tripled for reporting purposes. A detailed study is scheduled to e=piri-cally quantify (and significantly reduce) the appropriate correction' factor.

t

• • The six nuclides are:

Kr-85=, -87, -88 and Xe-133, -135 and -138. t - i I

• • • Revised radiogas release data,-reflecting the 16-minute holdup ti=e and l

upgrading of applicable nuclear data, for the years 1965 through 1972, l are presented in Appendix A. ,c I' g

2 ( Particulate and halogen releases to the atnosphere are measured by counting particulate and charcoal filters weekly. These filters col-lect stack effluent continuously at a rate of 3 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, de-pending on the half-life of.the nuclide observed. Tritium releases to the atmosphere are calculated, based upon measurements made in the primary coolant and containment air and using identical concentrations for all releases as follows: a. Off-Gas - A flow rate of 10 cfm containing 100% relative humidity and 90% radiolytic gas by volume both at primary coolant tri-tium to hydrogen ratio to determine tritium releases both in vapor and molecular form. b. Turbine Sealing Steam 'Ihe measured flow rate at 100% relative humidity and primary coolant tritium to hydrogen ratio. c. Containment Ventilation - The measured flow rate and mea-sured containment building tritium concentration. The results of these calculations are also shown in Appendix A. ii. Liquid Effluents Liquid vaste releases totaled 1.22 curies of radioactive mate-rial. This release corresponds to 3 9% of technical specification limits. Additionally, 4.4 curies of tritium were released corresponding to 0.003% of 10 CFR 20 permissible concentrations in the discharge canal. Liquid Effluent Calculational Methods ~ The release pathway to Lake Michigan for all liquid effluents is through the plant's condenser circulating water discharge canal. A flow rate'of.49,000-53,200 gpm dilution for liquid effluents is obtained through th'e use of the condenser circulating water pumps, two at 24,500 gpm each, and house service water pumps, two at 2,100 gpm each. Each collected tank of liquid is sampled, analyzed for radio-active content, and discharged at a controlled rate to assure that permissible concentrations are not exceeded in the canal prior to dilution ( m-- mm

3 I' in Iake Michigan durirg the ti=e of discharge. La s sa=ple is analyzed by gn=a spectro =etry to identify as rany of the conponent nuclides as possible. (See Appendix B for results.) Per=issible concentratiens in the canal are dete mined fro the following: Ci g, I -1 i where Ci is the concentration of the ith isotope in the canal ct the given concentration ceasured in the tant diluted by the known canal flow rate. Those isotopes not identified by ga=a spectrc=etry but ceasured by a gross beta analysis are presu=ed to be Sr-90 and released on that basis. Periodic sanples of the batches are then sent to the radiological environ = ental contractor and analyzed for Sr-90 and Sr-69 Frc= concentra-tions of Sr-90 and Sr-69 found in the batches, the total curies released of these two isotopes is calculated and used in calculatire the percent of applicable limit in Appendix 3. The re=aining unidentified iso +4 pes are assigned an F2C of 3 x 10 t:Ci/nl per 10 CFR 20. Tritiu= released' are based on average concentrations in both " clean" and " dirty" vaste ta nh. iii. Solid Wastes A total ~of 9,671,946 curies of radioactive ::sterial was shipped off site during the period covered by this repert. Out of this total, 1,291,000 curies were irradiated cobalt and 8,280,9L6 curies were solid waste. See Appendix C.

VIII. ENVIROICENTAL MONITORING

i. Environmental Survey Environmental levels of radioactivity as found in the vicinity of the plant were composed almost entirely of naturally occurring radio-active caterials. In the vicinity of the circulating water discharge canal radioactive caterial of plant origin was found. These materials occurred primarily in aquatic organisms. These levels of radioactive materials, however, were extremely lov and pose no threat to the health and safety of the public. Further, these 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 20 miles southeast of the plant, respectively, to determine increaced concentrations, if any, of radioactivity of plant origin. In addition, film badges placed at each of these locations plus six additional locations on the site property boundary measure direct dose in the environment. In order to obtain greater sensitivity of measurement, a com-parative program of film vs thermoluminescent disometers (TLD) was started in late 1971. The program consists of placing a film and TLD side by side at each monitoring. station for a one-month exposure period. Average monthly doses'at the site, inner ring and background ctations are compared and any difference, at the 95% c.fidence level, is re-ported using standard "F" and t" tests. ?- results of these dosimeter L analyses are given in Appendix D. While all %e dosimeters record doses i from natural' occurring sources, the dosimeters on site can also be ex-pected to receive doses from not only the. plume but direct radiation from the plant. The site dosimeters showed, on an average, 1.06 ! O.48 mR/mo above the background station dosimeters. During the same period of time, the inner ring of. dosimeter stations did not'show a dose rate above the background station dosimeters. k

2 Air samples gathered continuously and analyzed weekly at the t' stations shown in Appendix D showed no difference, at the 95% confidence level, in 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 radioactive materials resulting from plant operations. From the known meteorological dispersion conditions, the following =aximum concentrations can be cal-culated: -1 3 Particulates (May) (1.2 pCi/s) x (0.31) x (5 0 x 10 s/cm ) -1 3 = 1.9 x 10 pCi/cm -1 3

• Halogens (March)

(1.2 uCi/s) x (132) x (5.c x 10 s/em ) -1 3 = 7.8 x 10 pCi/cm

• + Reflects measured values multiplied by three.

These compare to the minimu= detectable activity values and normal background concentrations as follows: Maximum Calculated. Minimum Detectable Nor=al Background 3 3 Activity u i/c=3 Release Concentration uCi/cm gcg1yggy ycifcm a Particulate .1 9 x lo- 'l x 10' 7 x 10-1 -1k -15 Halogen 7.8 x 10 2 x 10 Hence, the negative data obtained in the program was expected. Also, at the Big Rock Point Plant, daily co=posite condenser circulating water inlet and canal water discharge samples are taken and analyzed for radioactive content. In addition, a monthly composite of these samples.is analyzed for radioactive content. These results are shownlin Appendix D. Additional aquatic samples are taken and analyzed during the summer growing season and these results are also tabulated in ] . Appendix D. ( r-g

3 Based upon the liquid release of 1.22 curies of radioactive ma-terial (less tritium) which results in an annual average concentration in -8 the discharge canal of 2.4 x 10 pCi/ml, the e.alysis of discharge canal water should indicate an increase of radioactive material in discharge canal water samples since the minimum detectable activity for gross beta 0 measurements is about 5 x 10 pCi/ml or about 5 times lover than the average concentration discharged. The results shown plotted in Appendix D indicate an average of about 2.6 x 10' pCi/ml for the year, which is in very close agreement with the calculated concentration. ii. Environmental Dose Calculations Levels of radioactive materials in environmental media indicate that public intake is well below 5% of that which could result from contin-uous exposure to the concentration values listed in Appendix B, Table II, Part 20. a. Atmospheric Releases In order to predict potential radiation doses resulting from gaseous releases, environmental transport and uptake factors must be known. A confirmation of these calculated doses is attempted then by measuring levels of radioactive materials in the plant's environmental surveillance program. In previcus reports for the Big Rock Point Plant, the average yearly meteorological relationship between release rate and downwind concentration of radionuclides at distances from the plant up to 50 miles has been used. Using a typical equilibrium mixture of noble gas release -6 after a 30-minute decay, a concentration of 1.4 x 10 pCi/cm3 delivers in air an exposure rate-of 1 millirem per hour using semi-infinite cloud i 3 geometry. From this and the relationship X/Q = 3.45 x 10-s/cm, the release rate required to produce a dose of 500 millirems per year at the point of maximum ground concentration was 2.4 curies per second. The licensed technical specification limit on the other hand is 1 curie per second. Furthermore, the licensed technical specification limit for particulates and halogens is (1.2 x 10 ) x (MPC) which produces a con-centration at the point of maximum ground concentration some 2400 times below maximum permissible concentracion allowed in Title 10 CFR 20 for I-131 using the previous meteorological relationship. (

b The X/Q value of 3.45 x 10 s/cm3 -1 was based on the fact that the vind conveys a plume in Sectors 1 and 2 (see Appendix D) during near neutral conditions 21.25% of the time per Section 91511 of the FHSR. Using the meteorological data in the FESR, X/Q values for each sector, have been calculated which accou-t for the amount of time the different stability classes exist and the wind is blowing in each sector. 'At the site boundary, a plume from the plant has not yet reached ground level. Therefore, any dose received from plant releases at the site boundary would be a shine dose from an overhead finite cloud. Per " Meteorology and Atomic Energy - 1968," using an equilibrium mixture, the release rate required to deliver 500 millirems per year at the site boundary is LOS Ci/s in the critical sector which is Sector 4 (see Appendix D). A computer model is now used to calculate radiation dose re-sulting from plant releases of noble gases. 'Ihe integrated population dose, out to 50 miles, for 1973 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 '; hen the summation of the individual nuclide contributions. The 1973 population is estimated from the 1970 Census of c. Population on a township basis corrected by the census-determined State of Michigan growth rate of 13% per year and includes transient popula-tion as 1/4 residents. 'Ibe total estimated 1973 populat on resides 24 i hours per day all year at the same location. d. The actual mixture found during the weekly off-gas analysis is used for that week's releases and total release is further corrected. by daily measurements of off gas, Site boundary doses are finite cloud shine doses. Se=i-e. infinite cloud geometry is utilized to calculate doses after the plume reaches ground level. y c u. 4 -y

5 f. No credit is taken for the meanderin6 of the plume before it reaches the different annuli. 'Ibe maximum calculated radiation dose at the site boundary resulting from noble gas releases was 3 5 millirems. The integrated dose to the population out to 50 miles was 2.8 person-Rems. Doses from particulate iodine and tritium releases as shown in Appendix A vere negligible compared to that received from noble gases due to the conservative limits in the plant technical specifications and the absence of any significant milk food chain in the nearby area affected by the plant. i 4 4 i e i y*- y. 4 -. + m e w

6 Calculated Radiation Doses From Gaseous Releases January 1, 1973 to June 30, 1973 (Person-Rems) Distance Sector (Miles) 1 2 3 4 5 Total 1-2 Population 13 75 o lo o 98 -3 Population Dose o.001 0.o26 0.0 6.3 x 10 o,o e,og1 2-3 Population 265 269 o 51 73 658 Population Dose o.12 0.066 0.0 0.o22 0.032 0.24 3-4 562 397 0 48 58 1,065 0.17 0.073 0.0 0.016 0.018 0.28 4-5 3,343 721- -o 103 0 4,167 0 79 0.10 0.0 0.028 0.0 0 92 5-10 2,101 24 o 534 0 2,659 0.21 1 5 x 10-3 0.0 0.065 0.0 0.28 10-20 8,983 395 747 14,109 327 24,561 0.23 6.0 x 10-3 0.024 o.44 9 0 x 10-3 o.71 20-30 9,647 3,503 1,920 4,623 327 20,o20 0.07 0.015 0.018 0.044 2.6 x 10-3 o.15 l-l- 30 22,766 4,079 4,859 4,845 0 36,549 0.07 7.7 x 10-3 ~o.012 0.020 0.0

0. n 40-50 40,774 8,884 5,870 12,096 o

67,624 i o.o66 8 9 x 10-3 0.013 0.026 0.0

0. n 0-50 Populatica 88,454 18,3h8 n,451 36,208 785 u9,038 Population Dose -

1 73 0 30 0.07 o.67 0.06 2.83 site Boundary 3 0 x 10-3 1 9 x 10-3 3 5 x 10-3 3 3 x 10. Dose (Rem). . -{

7 1 b. Liquid Releases In order to predict potential radiation doses resulting fro = the liquid releases, environ = ental transport and uptake factors =ust be known. A confirmation of these calcuhted doses is then at-tenpted by =easuring levels of radioactive naterials in the plant's environ = ental radiation surveillance progra=. "ihe nearest ~micipal drinHrg vater supply intake is located in Charlevoix, Michigan, which is generall,y upstrea= of the pre-vailing current flov in Lake Michigan at this location. However, since current patterns do occur that could, at times, carry the discharged water in the direction of Charlevoix, population dose based upon this flow is calcuhted in the next section of this repod. A conservative dilution factor of 800 is taken froc the point of discharge to the City of Charlevoix based upon the report, " Big Rock Point Hydrological Sur-vey, Great Lakes Research Division, University of Michigan, Special Report No 9," by John C. Ayers, 1961. In addition, the population dose is calculated to the entire ;opulation which receives its drinking water froc Lake Michigan, based on a unifor= concentration, resulting fro: plant releases, through-out Lake Michigan. Also, radiation dose to hu=an populations can occur as a result of phnt releases through the consu=ption of fish caught in - Iake Michigan. Utilizing the measured values of radionuclides released as shown in Appendix B, the following for=ula, and the standard can model, drinking vater doses can be calcuhted as follows: Ci I D, = I (Limiting Dose Rec /Yr) ( ij D, is the individual dose in Re=/yr, vbere: Ci is the average concentration. in Iake Michigan of the individual nuclides measured, in 9C1/=1, MPC:is the concentration of each nuclide measured required to produce the Limiting Dose at continuous intake in pCi/=1 and Limiting Dose is the dose produced at continuous exposure to MPC concentrations.- (i t s.-

8 In calculatind ingestion dose from the conausption of fish, an o. 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 2200 ml of fluid consumed daily by the standard man. This, in effect, alters the MPCi by 50/2200 or 0.0227 The calculation of individual doses, both from drinking water and consuming fish, are per the previous formula while integrated popula-tion doses in man-Rem are calculnted utilizing the following parameters: a. For drinking water, the individual doses are su=med over the entire population that receives its drinking water from Lake Miaigan with discharge canal flow appropriately mixed with the lake. This _s approximately 10 million people of which approximately 7 million reside in the Chicago metropolitan area. b. The population dose due to drinking water to Charlevoix residents is based on a p2pulation of 3500 people. C.

• For fish consumption, the average concentration in Lake Michigan water, resulting from plant releases, is used with a bicacet nula-tion factor to determine the average concentration in fish.

I d. Fish do not reside continuously in the discharge canal but migrate. This can be seen in the following table which compares the fish - consumption dose based on the discharge canal water concentration and the appropriate reconcentration factors to the fish consumption dose calcu- . lated from actual concentrations in fish caught in or near the discharge canal. Population doses based upon drinking water from the.Charlevoix municipal system were'O.07 person-Rem and total-Lake Michigan drinking water consumption population dose was 1.13 person-Rems. The consumption ' of all of the Lake Michigan. fish harvested resulted in a population dose of 0 97 person-Rem. ~* ERG Special Report No 2. " Trace Element Distributions in Lake Michigan Fish: A Baseline Study With Calculations of Concentration Factors and Equilibrium-Radioisotge Distributions," March 1973 j L E b

9 As a n.easure 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 late and total water consumed from the lake. Ikth of the dose calculations are conservative in that: ri. Equilibrium is not obtained in the human body for most isotopes released. b. No credit is taken for precipitation and deposit in sedi-ment or uptake by. life forms other than fish which are seen to occur by the data shown in Appendix D. c. No credit is taken for radioactive decay which for I-131 is significant. Results are shown in the following tables. b pisui i si-

~ ,2 IX. OCCUPATIONAL EXPOSURE BREAKDOWN - For individuals receiving over 500 mrem during the first six months of 1973 A. Routine Plant Surveillance and Inspection - 36 Men B. Routine Maintenance - 28 Men - C. Special Maintenance - 7 Men (Recirculating Pump Overhaul) 10 Men (Leak Detection System Installation) D. Routine Refueling - 18 Men E. Special Refueling - 5 Men (Reactor Vessel Weld Inspection by Outside Contractor) 14 Men (Irradiated Fuel Inspection by Fuel Vendors) Film Badge Results mrem Dose 1/1/73 - 1/28/73 1/29/73 - 2/25/73 2/26/73 - 3/25/73 0-100: Maint 6 Oper 12 Maint 5 Oper 9 Maint 3 Oper 2 Supvr 18 Tech 6 Supvr 15 Tech 7 Supvr 5 Tech 2 Visitors 8 Visitors '6 Visitors 25-101-500 Maint. 8. Oper 5 Maint 7 Oper 10 Maint 9 Oper 10 Supvr 0 Tech k' Supvr 3 Tech 3 Supvr 9 Tech 7 Visitors 0 Visitors 0 Visitors 16 501-1250 Maint 0 Oper 2 Maint 0 Oper 0 Maint 6 Oper T Supvr 0- Tech 0 Supvr 0 Tech 0 Supvr k Tech 1 Visitors 0 Visitors 0 Visitors 30 g. 1251-2500 Maint 0l Oper 0 Maint 0 Oper 0 Maint _3 oper 0 Supvr 0 Tech 0 Supvr 0 Tech -0 Supvr 0 Tech 0 Visitors G Visitors 0 Visitors 2h

2501-5000 Maint 0 Oper 0 Maint 0 Oper 0 Maint 0 Oper 0

-Supvr 0 ' Tech- 0 Supvr 0 Tech 0 Supvr 0 Tech 0 Visitors 0 Visitors- 0 Visitors 0 3/26/73 h/29/73 h/30/73 - 5/27/73 '5/28/73 - 7/1/73 ~ 0-100 Maint 1 Oper 1 Maint 6 Oper 13 Maint 5 Oper 10 Supvr 8 Tech 1.Supvr 17 Tech 9 Supvr 15 Tech 8 Visitors 20 Visitors 6' Visitors k i '101-500 Maint, 1 Oper 15 Maint 6 Oper 6 Maint 6 Oper 9 Supvr 8 Tech 6-Supvr 'O' Tech 1 Supvr 2 Tech 2 Visitors 11 Visitors 0-Visitors 0-501-1250 'Maint 4 oper 3 Maint 0 Oper 0 -Maint 0 Oper 0 Supvr 2 Tech 3 Supvr O' Tech 0 Supvr' O Tech-O -Visitors 5 Visitors O' . Visitors 1251-2500 Maint 5 oper; O Maint l0 oper O_ Maint O. Oper 0 Supvr O. Tech 0 Supvr 0 Tech -0 Supvr 0 Tech 0 a Visitors - 0 Visitors 0 Visitors 0 2501-5000 Maintf 0 Oper 0 Maint. 0 Oper 0 Maint -0 Oper 0 Supvr 0 ~ Tech 0 Supvr 0 Tech 0 Supvr 0 Tech 0 Visitors. O' 1 Visitors O Visitors- 0 {; y m-g - y e -+ g r= r.--..-,, w J-m y -. we-w, -4 mv-te-

i X. RADIOACTIVE LEVELS IN PRINCIPAL FLUID SYSTD'S Mini =u= Average Maxi =u= A. Primarv Coolant Reactor Water Filtrate ("} k.35 x 10 1.09 x 10 8 73 x 10 -2 -1 ~1 901/=1 Reactor Water Crud ( } k.36 x 10 1.16 x 10 2 32 x 10 -2 ~1 -1 pCi/ml/ Turbidity Unit ~ ~ Iodine Activity 1 x 10 5 x 10 7 x 10~ pCi/=1 B.- Reactor Cooling Water Syste= -3 -2 -2 Reactor Cooling Water " 7 26 x 10 1.h5 x 10 k.35 x 10 pCi/ml C. Spent luel Pool Fuel Storage Pool ("} h.36 x 10 1.h5 x 10 k.36 x 10 -2 -2 ~ pCi/mi Fuel Pool Iodine (b) 1 10 5 x 10 7 x 19 ~I -5 -3 (a)A counter efficiency based on a decay scheme consisting of one Ea.r.a photon per disintegration at 0.662 MeV used to convert count rate to microcuries. All count rates were taken at two hours after sampling. (b) Based on efficiency of Iodine 131 two hours after sampling.

• Based on APHA turbidity units and 500 ml of filtered sample.

CONSUMERS POWER COMPANY By Nuclear Licensing Ad=inistrator Date: August 31, 1973 Sworn and subscribed to before me this 31st day of August 1973 Lois E. Barnes, Notary Public Jackson County, Michi an 6 q'; My commission expires June 20, 1976

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BIG ROCX POI T NLICLUAR PLANT-DOSE ISOPLETHS ^ ' ~ ~ ' 50 Mi r m acens b / p ~ ~~~~ ~~qgpo, Q-k / s' /p j' i SCALE 0 0 ~ P00R ORIGINAL l

s APPENDIX D (Conta) Sampling and Analysis Sn-mary - Number of Samples Frequency of Medium Description Location Collected Type of Analysis Analysis - Air Continuous at All 166 . Gross Beta. 131 Weekly 7 Approximately 1 Cfm Lake Water 1 Gal Grab ST 12 Gross Beta, Gross Gamma Monthly 90Sr, 13h,, 137Cs, SkMn, """"t*#17 e Co, bZn, 59 ' Co, O Fe - Well Water 1 Gal Grab ST 6 Gross Beta Monthly Gaassa Dose Continuous All h51 Film Dose Monthly 108 TLD Dose Aquatic Biota' Grab St, NM,- 8 Gross Beta, Gross Gamma Semiannual .Mt McSauba Spectrum a L

n APPENDIX D (Contd) I High, Low and Average Concentrations For Highest Average Sampling Location Type Type of Analysis Locati>n High Low Average Air Gross Beta-Ga-na TC 0.05 pCi/1 0.01 pCi/1 0.03 I-131 All <0.2 <0.2 <0.2 -Lake Water Gross Beta BR ST LWo 64 3.5 26 Gross Ga-na BR ST LWO 120 22 55 Sr-90 BR ST LWO .18 1.8 2.1 Cs-137 BR ST LWO 13 <2 4 Well Water Gross Beta BR ST W k.3 <.8 1.9

• LTD Doae E

7.3 mR -0.k 3.8 mR 'In excess of control dosimeter. J J

-~ ~,. APPENDIX D (Contd) Big Rock Point Plant Aquatic Biota. May 1973 Nuclide (pCi/g) Wet cm Beta 137 58 65 -60 40 Sample Cs Co Zn Co K cpm /g pCi/g e Minnows Discharge 0.2h 0.08 0.79 0.12 1 0.01 3.5 0.h Crayfish-Discharge 0.35 0.28 0.19 0.51 0.05 2.7 1 0.3 Periphyton Discharge 0.83 0.h5 1.0 1.8 1 0.2 16 1 2 1/4 Mile West' O.92 0.31 1.6 0.50 0.05 7.2 i 0.8 1/4 Mile East 1.06 0.68 0.h6 2 9 1 0.3 11 1 1 0.2h 0.10 1.0 0.20 1 0.02 5.h i O.6 Mt McSauba. '. 9 0.12 1 0.03 15 0.2 Nine Mile Point Algae Discharge 0.29 0.18 0.06 0.26 i 0.03 2.2 i 0.3 W

10 i APPETDIX D (Contd) Difference in Average TLD Readings mR/ Month Site vs Background Inner Ring vs Month Stations Lackground Stations 1 January ND ND February ND ND March ND ND April 2 9h 2.19 ND May 3.k3 1 1.85 ND June ND ND Average' 1.06 i 0.kB ND f 1 ND - No difference at the 95% confidence level. Control TLD vere not shipped with others, statistical analysis based total dose including dose received in shipment (usually 12 to lh mR). l l y s g. y -r-

11 APPdDIX D (Contd)

• Michigan Department of Natural Resources Fisheries Division Lansing, Michigan Sports Catch. Lake Michigan and Anadromous Strea=s, 1970 Estimated Total Species Nunber Caught Veight (Lb)

Perch 1,700,000 283,333 Walleye 69,000 207,000 Bass 2h6,000 h92,000 Panfish 1,300,000 260,000 Northern Pike lh6,000 292,000 Suckers h82,000 1 hh6,000 smelt 2,800,000 280,000 Lake Trout 2k5,000 1,715,000 Rainbov Trout 285,000 1,h25,000 Brown Trout 168,000 8h0,000 Brook Trout 125,000 250,000 Coho Salmon 53k,000 5,3ho,000 Chinook Salmon 180,000 2,700,000 Other Species 368.000 368,000 Total - 8,6h8,000 15,898,333 'l\\ ' Unpublished 1970 data from postcard census program of the Michigan Department of Natural Resources, Fisheries Division. ~ w

12 ( APPENDIX D (Contd)

• Michigan Department of Natural Resources Fisheries Division Lansing, Michigan Commercial Catch. Lake Michigan - 1970 Estimated Total Species Weight (Lb)

Alevives 5,981,h15 Bullheads 610 Burbot 51,261 Carp 2,39h Chubs 4,028,340 Herring 676 Lake Trout 89,939 Menominee 161,987 Perch 22 Pike 65 Rock nass 35 Sauger 1 Sheepshead 12 Smelt 1,700,365 Suckers 521,807 Walleyes 7 White Bass 1 Whitefish 1.hl7.83h 13,956,771 -( Taken Frcat-GREAT LAKES FISHERIES 1970 Data Taken From December Issue of Michigan, Ohio and Wisconsin -Landings

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AFPENDIN A (Contd) 729ERS POWN CfMPAs_T Big rock Ructeer fewer Piant Calculated Be41sttoa Dosee Frem L1guld Effluents

• Floh Conemption Ibee 1/1/73 to 6/30/T3 Average C ev eatretton Average Concentrettoe Bicacetemalation(2) in take Maidea to Fish (CFg/MPC )MFJ Pope 41stion Dose Critical 1

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• IA Co-58 4.4 a 10'I 0.1. Trect 2A0 5 9 a 10 1.7 a 10 5.8 a 10" 0 34 4

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2. 3 a 10'"

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0. 2. Tract 60 5.4 a 10'1" 3 2 a 10'I#

5.5 a 10'O 3 3 a 10"I 4

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APPENDIX A (C .) bio ROCK PoIlrP NUCIEAR PIANT Inert Radiogme Release Curies Per Year 6 lootope_ 1962 1963 1964 1965 19(4 _1067 1968 19 9 1970 1971 1972 48,380 386,000 156,690 104,190 138,e3o 98,683 126,000 95,444 xe-138 22,292 181,180 60,280 53,229 55,580 65,153 78,585 51,404 Kr-87 14,349 144,640 31,360 51,306 32,736 34,153 43,792 29,836 Kr-88 6,109 41,809 13,474 15,911 13,997 17,484 20,913 13,572 Kr-85 23,225 120,550 36,870 47,875 49,270 78,656 76,926 54,683 xe-135 8,827 45,540 44,858 23,145 17,966 25,546 30,169 21,566 xe-133 1 7E-26 4.7E-25 2.6E-26 2 3E-25 2.oE-25 9 3E-26 7 9E-26 3.oE-26 xe-143 4.4E-25 1 3E-23 7 7E-25 6.5E-24 5 7E-24 2 7E-24 2.8E-24 9 3E-25 Kr-94 2.4E-18 6.8E-17 4.1E-18 3 2E-17 2 9E-17 2.4E-17 1.2E-17 2.2E-15 Kr-93 8 9E-13 2.2E-11 1.6E-12 1.1E-11 9 5E-12 4.TE-12 4.kE-12 1 74E 12 xe-141 1 3E-11 3 3E-10 2.4E-11 1.2E-lo 1.4E-lo 7,2E-11 6 5E-n 6.6E-lo 10 - 52 Kr-91 4.1 60 11 27 25 17 18 13 56 715 100 316 3m 215 239 125 xe-140 646 6,689 1,153 2,843 2,821 2,270 2(4 1,410 Kr-90 957 9,505 1,712 4,003 4,006 3,307 3,910 2,098 xe-139 2,770 34,150 7,934 13,563 14,189 13,842 17,176 9,918 Kr-89 10,064 77,200 18,357 3o,461 31,990 31,666 39,424 22,338 xe-137 9,450 65,550 17,804 25,061 26,326 28,748 36,046 23,893 xe-135m 3,465 23,324 7,162 8,812 8,457 10,106 12,436 7,622 Kr-83m 382 3,031 1,130 1,208 656 1,052 999 788 xe-133m 34 344 146 115 2T 86 41 845 xe-131m 179 2,089 936 904 21 383 33 437 Kr-85 vu N-13 67 2,090 2,040 1,803 3,435 4,504 4,245 3,988 3,601 3,674 3,170 c, ,, no n , r e, r~, ,, s. c, ~, ,en eo ,ne ooo

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APPENDIX A Consumere Power ConInny - L13 Bock Foint Flant, Doczet 1.3 50-155 Atmospheric Belcases of Badioactive mterial tRatte January February March April my June htal Total Ibble Gases curies 47,100 45,300 6,200 905 7,530 7,130 114,000 4 .* Total Raiogens 0.15 0.Ok5 k.2 0.0A8 0.0h2 3.0 x 10 k.5 -3 4 h tal h rticulates (3, y) 0.016 2.0 x 10'3 2.8 x 10'3 7 8 x 10' 1.1 x 10 6.4 x 10 0.024 htal Tritium 7.4 6.9

1. 2 39 7.4 7.4 42 Total hrticulate

• I 1.1 x 10'I 1 3 a 10"I 1.0 x 10
• I 1.7 x 10'I 1.2 x 10 Gross Alpha 3 9 x 10*I 1.2 x 10 leastaa Noble Gee WC1/e 2.1 x 10 2.2 x 10 1.8 x 10 1.4 x 10 1.2 x 10 k.6 x 103 3

2.2 x 10 Release mate Percent of hch Spec Limite for: Noble Games 1.8 1.7 0.24 0.03 0.29 0.27 0.72

• Nalogene 4.8 1.5 132 1.5 1.3 0.006 2h hrticulates 2.2 0.41 0 38 0.11 31 0.088 57 Isotopes Released:

Curies 4.6 x 10'g O.24,g Mrticulates -3 1,3, to-3 0.23,g 7.8 x 10'] 2 9 x 10 Bolm-140 5 3 m 10 Co-134/Co-137 4.1 x 10 1.1 x 10 0.095 0.095 h 54 4 Bologens: 1 131* 0.15 0.0k5 h.2 0.Ok8 0.Ok2 1.9 x 10 h.5 1-133 9 3 x 10 1.4 x 10 3 9 x '0 2 7 x 10'g. 4 -5 -5 I-135 Gaeon ~ 9,220 1,450 255 1,820 1,270 24,000 Ie-138 10,300 Kr-87 6,290 6,320 561 94 1,110 1,080 16,000 Kr-88 3,470 4,920 (49 41 644 629 10,000, Kr-85m ' 2,270 2,4'!O 361 18 368 398 5,800." Xe-135 3,380 9,890 1,k60 68 1,680 1,780 23,200 xe-133 h,350 5,260 813 7.3 1,160 985 12,600 Xe-143 0 0 0 0 0 0 0 h94 0 0 0 0 0 0 0 D43 0 0 0 0 0 0 0 to-141 0 0 0 0 0 0 0 Kr-92 0 .3,* 0 0 0 0 0 Kr-91 1.8 <1 0 <1 0 0 4 Re-lbo ' 24 10 <1 ~ 37 <1 0 40 Kr-90 272 120 2.0 28 72 1.1 430 Ze-139 40b 180 32 38 11 13 650 Kr-89 1,830 884 32 81 65 20 2,900 Xe-137 4,200 2,070 80 171 156 105 6,800 xe-135m 3,880 2,220 151 85 213 247 6,800 Kr-83m 1,350 1,130 139 14 152 292

3. 100 Ie-133m A09 295 61 71 56 174 350 Ie-131a 4.5 69 11 0

10 10 85 Kr-85 36 342 82 0 74 69 570 N-13 347 312 19 171 k23 L67 1,740 speflects Meneured Values hittplied by Three r

l AFn h.ulX C CFF-SITE SHIReiT OF RADICAC"TVE FATERIAL Ship:ent Transfer No Date Free Transfer To Radicactive Material 3 3o7 1/h/73 DPR-6 ECo, rz 16-NFS-1 110 ft high level vaste, 7 9 Ci 308 1/8/73 DPR-6 ECo, rr 16-NFS-1 6 fuel support tubes, h8 Ci 3o9 1/5/73 DFR-6 GE, val, CA 0017-60 Vaco filter, 1 $ 1 310 1/10/73 DPR-6 ECo, rt 16-NFS-1 6 fuel support tubes, filter socks 107 Ci 311 1/12/73 DPR-6 NECo, rf 16-NFS-1 7 fuel support tubes, 6 transitics pieces, 6h Ci 312 1/26/73-DPR-6 NICo, rf 16-NFS-1 6 fuel support tubes, mise, 76 Ci 313 2/9/73 DPR-6 NFS, FI CFS-1 10 spent fuel bundles, 1,413,120 Ci 314 2/10/73 DPR-6 NKo, 12 16-NFS-1 6 fuel support tubes, 32 Ci 315 2/16/73 DPR-6 NECo, IT 16-NFS-1 6 fuel support tubes,1 coupon rack, 89 Ci 316 2/22/73 DPR-6 ECo, rt 16-NFS-1 6 fuel support tubes,1 ccupen rack, 47 Ci 317 2/26/73 DFR-6 NFS, FI CFS-1 10 spent fuel bundles, 1,565,760 Ci 318 2/27/73 DFR-6 NECo, rf 16-NFS-1 6 fuel support tubes, filter socks 195 Ci 319 3/14/73 DPR-6 Battelle-Nortir est Fuel handling tools, 15 2 1 WA, WNL-027-1 320 3/1h/73 DPR-6 Exxcn Nuclear, WA Fuel handling tools, 1 2 1 WN-IO62-1 321 .h/9/73 DPR-6 Exxon Nuclear, WA Fuel handling tools, 1 d i WN-IO62-1 322 h/12/73 DPR-6 GE, Val, CA 0017-59 7 g=s of crud & liquid radvaste samples 323 h/13/73 DPR-6 NFI, 19-12667-01 33 irradiated cobalt rods, h47,000 Ci 32k 4/17/73 DPR-6 GE, Val, CA 0017-59 Fuel inspection equip =ent,1 di 325 h/18/73 DPR-6 GE, Val, CA SE-96o 9 irradiated fuel reds, 17,200 Ci Amend 71-20 326 h/19/73 DPR-6 NFI, 19-12667 33 irradiated ectalt rods, 532,000 Ci 327 h/30/73 DPR-6 NFI, 19-12667-01 30 irradiated cobalt rods, 412,000 Ci 328 5/21/73 DPR-6 NFS, NY CM -1 Spent fuel bundles, 2,h54,800 Ci 329 6/18/73 DPR-6 NFS, NY CFS-1 Spent fuel bundles, 2,805,600 Ci 330 6/22/73 DPR-6 ' GE, val, CA S E-96o 6 irradiated fuel rods, 23,800 Ci Amend 71-20}}