ML20030A493
| ML20030A493 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 12/31/1973 |
| From: | CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | |
| References | |
| NUDOCS 8101090755 | |
| Download: ML20030A493 (74) | |
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iv CONSUMERS POWER COMPAI;Y BIG ROCK POINT PLANT SD!IANNUAL OPERATIONS REPORT IiG 19 JULY 1, 1973 - DECEMBER 31, 1973 FILES RETUR,a n r~"ULATG,r" w"NTRAL iu 1 T ROOM 016 Operating License DPR-6 Docket 50-155
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CONSUMERS POWER COMPANY BIG ROCK POINT PIRiT Nineteenth Semiannual Report July 1, 1973 - December 31, 1973 I.
INTRODUCTION - SEMIANNUAL OPERATING REPORT The plant was base loaded at 69 MWe (gross) during this report period. The off-gas release rate on July 1, 1973 was averaging approxi-mately 2,500 pCi/sec.
The outage beginning on November 1, 1973 to conduct the six-month Technical Specifications testing requirements marked the end of 198 days of consecutive power generation - a new record for domestic operating boiling water reactors. During this 198 day span (beginning on April 16,1973), a total of 321,172 MWhe(g) were produced for an average output of 67.5 MWe(g) or 90% of rated power.
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.II.
OPERATIONS
SUMMARY
A.
CHANGES IN PLANT DESIGN Changes in the design of the plant which were incorporated as
. facility changes are as follows:
1.
Facility Change C-215 (Primary Systen Leak Rate Equipmentl -
This change involved the installation of leak trace sampling tees as listed:
a.
A loop seal and leak trace sampling tee was installed in the pipeway cooling unit drain.
b.
A leak trace sampling tee was installed in the collection pot drain from the enclosure spray system relief valves.
c.
Three leak. trace sampling tees were installed in the clean-up heat exchtLger room on' selected relief valves.
2.
Facility Change C-216 (Primary System Leak Rate Equipment)
This ichange involved the addition of a two-inch drainpipe from the reactor
- recirculation pump seal collection sink to the reactor clean sump to identify process' flows.
3.
Facility Change C-218 (Primary System Leak Rate Equipment) -
7 tis ~ change. involved the construction of a temperature and dew point temperature. sampling station for sampling air from both the supply and exhaust air ducts as they enter and leave the reactor enclosure.
An air-cooling coil was modified to use the supply air.
'(after dew point measurement) to-cool the exhaust air (prior to_its
- dew point measurement) to enable a wider range of measurement of the exhaust air.
All points-(h) were connected to read out on the dew point recorder-located in the control room.
k.
Facility Change C-221'(Primary System Leak Rate Equipment) -
This change involved. rerouting'of the collection system for the control rod drive ~ pumps and associated safety relief valve discharge to the re-actor enclosure clean sump to identify process flows.
.5 Facility Change C-222-(Primary System Leak Rate Equipment) -
- This ' change involved the constractJon of an ' angle-iron dam around the -
clean sump'to prevent water, accumulating ^or: draining on the recirculation
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pump room floor, from entering the clean sump. The dam was successfully leak tested using_ water at 7/8 inch above ficor level.
6.
Facility Change C-22h (Primary System Leak Rate Equipment) -
This change involved the addition of an integrating water meter between the demineralized water storage tank and the condensate storage tank so that the amount of water transferred can be accounted for. Previously, that amount of water used for regeneration and rinsing of the makeup de=ineralizer had to be subtracted from the demineralizer flow meter integrator.
T.
Facility Change C-225 (Primary System Leak Rate Equipment) -
This change provided for the addition of a drain in the clean-up deminer-alizer room. No floor drain exists in this room. However, to facilitate early detection of unidentified leakage, the two-inch pipe stub (drain to enclosure dirty su=p) in the clean-up demineralizer room floor was drilled and tapped for 1/4 inch pipe (two holes). These holes were made approximately 1/4 inch from the floor level and a-screen was
'placed around the pipe stub. This arrangement will limit the total leakage to approximately 18 gallons'before drainage to the dirty sump begins.
8.
Facility Change C-23h - This change involved the removal of -
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the generator and bus itstantaneous differential overcurrent relay (2503).
' Following the inadvertent tripping of the unit, a study conducted by the '
Consuners Power-Electric Engineering Department revealed this relay scheme to be-unnecessary. Adequate' fault protection is provided through other existing relays.
9 Facility-Change C-235 - This change _-involved eltsination of one and relocation of two annunciator circuits associated with the system transmission lines. These changes were the result of the -instal-lation of the 3h5 kV transmission line to this area and relocation of
' alarm systems external to the plant.
B.
PERFORMANCE CHAttACTERISTICS l
LAt the ' start of this report pr od, the unit was on-line.at L69 MWe(g) (220 MW ).
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A fuel inspection tea = frc= General Electric (GE) a-rived on site July 7,1973 for rereval of individual fuel rods fec= various fuel bundles. These fuel rods were scheduled for metallurgical testing at the GE Yallecitos Test Center as a part of GE's fuel develep:ent pregrar.
Four shi; cents (25 fuel rods) were zade this re;crt period with a total of six shipments (ho fuel rods) shipped throughout the year.
Fcur shipments of spent fuel (32 fuel bundles) were =sie this report period to Nuclear Fuel Services at West Valley, New York for reprocessing,-with the first shipment leaving c: July 12, 1973 and the last en k: gust 30, T 3 A total of eight shi; cents (72 fuel buniles) were made in 1973 On July 20, plant load was reduced to 10 We to per=it investi-gatien of a ccepenent cooling vater leak in the recirculating pu=p reon.
Cc=penent cooling water was found leaking frem.a line to the zeter thrust bearing on the No 2 recirculating ymp. Following repairs, the pu=p v' returned to service and the plant lead increased to 69 We.
. On July 25, pcver was reduced to -200 W: by flev centro dur-ing a test to determine the effect of recirculating ymp flow c= stes:
drum tilt. The recirculating flev was alternately decreased in each loop-by throttling the recirculating ymp discharge valves. Folleving the-test, operatics was resumed a 220 W.
g On August 16,- yver was again reduced to 10 We to per=it
. entry into the recirculating pump room to investigate for component cooling water. leakage. ' The flex line from the a-all heat exchanger c No 1 recirculating pump was four.d to be leaking. The syste= vas re-paired and the plant vas. returned to operatics at 69 We..
Oce irradiated cobalt rod (11,000 C1) was shipped to Neutron Products, Inc cn August 18. This rod had been left in bundle D-61 during
' the recent cobalt shipping campaign 'and was -later removed 'and stored.
On September 19,' the clean-up pump tripped off and could not be: restarted. This necessitated a power reduction to 10 We to permit entry into the recirculating ymp rom for the purpose of isolating the clean-up systes.- After this was accomplished, power was increased to
..b 69 We on September 20..
Following installation of the spare clean-up F2p on September 22, power was once again reduced to pemit entry into II the recirculating p'=p room to valve the cleari-up syste= into service.
Upon completion, power was again increased to 69 We.
The se=iannual contain=ent component leak rate test was co=-
pleted on October 8.
Results, indicated the leak rate was h5% of the maxi =u= leak rate allowed in t le Technical Specifications.
On October 18, the containment ventilation syste= vas re=oved from service for a period of 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> to effect repairs to the four (h) solenoid valves (CV-9151, CV-9152, CV-9153 and CV-915h), which con-trol the supply and exhaust ventilation valve air operators. A repair kit was installed in esch solenoid valve to correct for excessive air leakage. Following repairs, the supply and exhaust ventilation valves were operated satisfactorily and returned to service.
On October 26, during fuel pool draining operations, an ir-radiated fuel rod was discovered on the botto= of the pool. Positive identification could not be =ade at the time and it was decided to store the rod in the fuel transfer. cask. For accountability purposes, the rod is being carried in the plant records as an unirradiated E-type tie rod until such time that it can be positivel; identified.
On Nove=ber 1, the plant was taken off the line for a scheduled outage to perform the se=iannual control rod drive checks. After the reactor had been taken suberitical on Nove=ber 1, a sera = occurred on low condenser vacuum _(22.8" Hg). The turbine bypass d-c isolation valve failed to close automatically on the loss of condenser vacuu::: but was closed manually after being exercised. This was ccrrected by reset-ting the limit and torque switches and relubricating the valve gear train.
The following tests were ec=pleted successfully during the outage:
1.
CRD As Found Hot Withdrawal Timing 2..
CRD Coupling Integrity, Jog and Position Indication 4.-'CRD Cold Withdrawal Timing 5
One-and Two-Rod Shutdown Margin Checks.
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I In addition, special tests of the recirculating pump interlocks and the poison system squib valve firing circuit were performed to verify sys-tem operability. The recirculating pump interlocks were tested success-fully. Tests of the poison system squib valve firing circuit conducted
.during this outage were inconclusive. However, tests were successfully accomplished during the outage in Dece=ber 1973 The unit was returned to service folloting the outage on November 4 and reached 69 MWe on November 6.
Sampling of the e=ergency condenser shell side water following return to power revealed leakage had occurred from the primary to secondary side during start-up. The north ^,ube bundle of the emergency. condenser was isolated and repairs.
vere scheduled for the next outage.
On December 3, power was reduced to 58 MWe in order to decrease the off-gas release rate below 15,000 uCi/sec. The control rod withdrawal sequence was modified as 'vell in order to limit the off-gas response to control rod withdrawal. On December 6, the off-gas release rate was r
again approaching 15,000 uci/see and a second power reduction was ordered, this tL=e.to 53 Mwe (or 1 feed pump operation).
i On December 8, the unit was forced off the line (by.means of a controlled shutdown) due to a packing failure on the level instru-mentation lover root valve at the east end of the reactor steam drum.
Repairs were completed on December 11, but the outage was extended to permit-investigation of. the emergency condenser tube leakage. The plant remained out of service throughout the remainder of the report period while the emergency' condenser was being repaired. This consisted of repairs to.three leaking tubes at the tube-to-tube sheet velds in the north tube bundle and~ modifications to the baffle plates in the inlet' water box heads of both the north and south tube bundles. This latter.
vork was contracted to South-West Research Institute for both design and installation of a baffle plate that would meet the system thermal stresses.
Other work performed during this extended outage included the.
successful testing of the liquid poison system squib valve firing p
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.i circuit, which has previously been discussed, and an investigation into the off-gas holdup piping system. Following replacement of the off-gas isolation valve, an isolation test conducted on November 30 failed to demonstrate isolation capability. During the December outage, the valve was removed from the off-gas line and bench tested.
It was found that the torque was insufficient to close the valve fully. The moment ann was increased to provide the necessary torque and the valve was then successfully bench tested. Following reinstallation into the off-gas i'
line, the entire off-gas pipe from the flow orifice below the air ejectors to the isolation valve in the stack base was successfully pressurized to 6.5 psig. A volume test of the off-gas holdup pipe 1
yielded a volume of 367 ft3, in' good agrrement with plant as-built specifications.
At the start of the report period, tne off-gas release rate was approximately 2,600 pCi/sec..This demonstrated a gradual but steadily increasing trend until just prior to the November 1, 1973 1-.
outage when the release rra.e had reached approximately 9,500 uci/sec.
Follosing return to power, the off-gas release rose sharply and reached 23,000 pCi/sec. Power was reduced on December 3 to hold the releases
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below 15,000 pCi/sec.
C.
CHANGES IN PROCEDURES WHICH WERE NECESSITATED BY A AND B OR WHICH OTHERWISE WERE-REQUIRED TO IMPROVE THE SAFETY OF. FACILITY OPERATION
'The following procedural changes were made with respect to f
plant operations:
A3.0
- Defines additon of Operations Engineer-and further c
l-defines duties and responsibilities of plant staff.
A2.2
- Defines operator requirements for control room.
A2.6L
- Revised the application of " Switching and Tagging Orders."
. A3 7 7 -
- Defines the responsibility. for " Locked Door and :
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Valve Control" and redefines its application.
A8.0
- Incorporates a new section
" Maintenance, Test,
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-Refueling and Special Procedures."
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Bl.3 3 5
- Corrects the incore calibration calculation.
B8.2.1-
- Revises plant operating requirements to reflect changes in Technical Specifications.
Bil.3.2.8 ).- Further clarifies use of the radvaste discharge f***2 valves to the canal.
B15 2.6
- Includes the addition to survey the condenser by radiation protection personnel if opened for mainte-nance purposes.
B2h.0
- Includes a new section on " Plant Operating Require-ments," for chlorinating condenser and service water systems.
B28.6.0
- Includes a new secton on the " Emergency Diesel Cooling Water Pump Sealing System," describing its use and precautions.
B29.3.1
- Includes a section on placing the " Reactor Re?ircu-lating Pump Seal-System" in service.
l Dlk.0
- Revises the fuel shuffling vinch procedure to pro-
.hibit its use for shuffling fuel in the reactor vessel.
E2.1.1
- Defires +ne " Chemical and Rad Protection Supervisor" duties.
- E2.2. 2
- Changes title of " Chemical and Rad Protection Engineer" to " Chemical ~and' Rad Protection Supervisor."
' E4.1. h'. 5
- Includes additional' information on survey instruments for detecting beta radiation.
E4.1.8.1.
-- Incorporates a new section to describe the " Snoopy" neutron survey instrument.
E5.4
- Adds a new section to describe the limitations in the " Shipment'of Waste and-Other Radioactive
' Material."
E6.l'
- Defines the responsibilities in controlling radiation-
. protection records.
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D.
RESULTS OF SURVEILLANCE TESTS AND INSPECTIONS 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 tests.
1.
Containment Isolation a.
Syster-Containment isolation valve controls and instru-mentation.
Required Frequency: Quarterly (Conducted Monthly)
Test Date: July 10, August 7, Sepetember 5, October 2, December 3 Results: The automatic controls and instrumentation for eight of nine isolation valves were checked and found to function properly. One valve (main steam drain, MO 7065) is maintained in the closed pos. ion, de-energized and not used. Therefore,
testing the automatic controls of this valve is not required.
b.
System:
Isolation valve leak and operability test.
Required Frequency: Twelve months or less.
Test Date: Not required during this report period.
Resultsi 'None.
c.
System:. Containment sphere penetration inspection (visual).
Required irequency: Twelve months or'less.
Test Date: Not required duris3 this report period.
Results: None, d.
System: Containment sphere integrated leak rate test.
Required Frequency: Every two years.
Test Date:
No_ test required during.this report period.
Result's: None.
.e.
-System: Containment-sphere. component leak rate test.
Required Frequency: -Six months or less.
-Test Date: October.5, 1973 to ocotber 8, 1973 II-8
i-Results: The contsinment component leak rate test was performed using air at 120 psig. The results of this test showed a total leakage of 45 2%
of the allowed limit. Seventy-one percent of this leakage was from the supply ventilation valve. It is anticipated that this valve will be replaced at a convenient future plant outage.
2.
Control Rod Drive System and Associate Tests a.
System: Reactor safety system scram circuits (not re-quiring plant shutdown to test).
Required Frequency: One month or less.
Test Dates: July 10, August 7, September 5, October 2, November 3, December 3
~Results: The reactor safety system was tested using the switches provided to simulate sensor trips.
All channel trips occurred as designed.
In addition, the neutron monitoring power range and intermediate. range channels were tested for trip setting. All of these tests showed the trip settings to be within 120 2% of power and 10-second period setting.
b.
System: Control rod performance - run.
Required Frequency: Each major refueling and at least once every six months during power operation.
Test Date: November 2, 1973
'Results: The control rod drive continuous withdrawal and insertion test ~ including withdrawal timing, was.
performed for each drive. This test is performed
- during reactor shutdown following completion:of other drive performance tests and adjustments and represents the results of the final timing of each drive under cold-conditions. The results of l
'this test showed all drives to be operating sat-isfactorily'with most withdrawal times at 36 II-9 1
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seconds. No withdraval time was less than 23 i
seconds.
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Syste=: Control rod performance - jog.
Required Frequency: Each major refueling and at least every six sonths during power oper-ation.
Test Date: Nove=ber 2., 1973 Results: Satisfactory latching of all drives.
d.
S:ste=: Control rod performance - seras.
Required Frequency: F,ach major refueling end at least cnce every six =cnths during pover
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operation.
Test Date: Nove=ber 2, 1973 Results: The control rod scra: test was perfor=ed for each drive. The test included ti=e fro = sys-te: trip to 100'. of insertion at a reacter temperature of about 150 F.
The results of this test were satisfactory for all drives.
e.
System: Reactor safety systems scra: circuits (requiring plant shutdown).
4 Required Frequency: During each =ajor refueling outage but not less frequently than once
-every 12 =onths.
Test Date: Not required during this report period.
Results: None.
- f.. Syste=: Reactor safety syste response ti=e (requiring plant shutdown).
Required Frequency: During each =ajor refueling shutdown,
.but not less frequently than once every 12 =onths.
Test'Date: Not required during this report period.
Results: None.
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System: Control rod withdrawal permissive interlocks function.
Required Frequency: Twelve months or less - the refueling interlocks will be tested prior to each major refueling.
Test Date: Not required during this report period.
Results: None.
h.
System: Control Bod Drive Friction Test 4
Required Frequency: During each major refueling, but not less frequently than once each year.
Test Date: Not required during this report period.
Results: None.
3 Emergency Cooling a.
System: Core Spray System Check Valves Required Frequency: Twelve months or less.
l Test Date: Not required during this report period.
Results: None.
b.
System: Post incident spray system automatic control operation.
i Required Frequency: Twelse months or less.
Test Date: Not required during this report period.
Results: None.
.c.
System: Reactor Emergency Core Cooling System Trip Circuit Required Frequency:
Twelve months or less.
Test Date: Not required during this report period.
Results: None.
d.
System: Containment Sphere Isolation Trip Circuits Required Frequency: During each major refueling shutdown, but not less frequently than once every 12 months.
Test Date: Not required during this report period.
I Results: None.
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4.
Miscellaneous Systems a.
System: Reactor Shutdown Margin Test j.
Required Frequency: After each refueling, after certain core ecmponent changes, if the system is cooled to atmospheric conditions and after 35,000 MWdt have been gen-erated.
Test Date: November 3 and h,1973 Results: The shutdown margin of 0.003 ok/k with the L
strongest rod fully withdrawn from the core was verified.
In addition, the shutdown margin of 0.003 ak/k was verified with two adjacent rods fully withd swn from the core.
b.
System: Nil Ductflity Transition Temperature Calculation Required Frequency: At least once each year.
Test Date: Not required during this report period.
j Results: None.
c.
System: Moderator Temperature Coefficient Test Required Frequency: Following each major refueling outage.
Test Date: Not required during.this report period.-
Results: None.
d.
System: Suberiticality Checks -
i Required Frequency: 'During-core alterations which increase reactivity.
Test Date: Not required during this report pericd.
Results: None.
e.
System: In-Service Primary System Inspection Required Frequency: A continuing program being conducted during some major refueling outages.
Test Date: Sot performed'during this report' period.
Results: None.
f.
System: Refueling Operation Control
-Required Frequencyi Each-major refueling.
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Test Date: Not required during this report period.
Results: Nene, g.
System: Beacter Refueling Safety Syste= Sensors and Trip Devices Required Frequency: Each =ajor refueling.
Test Date: Not required during this report period.
Results: None.
5 Poison System a.
System: Liquid Poisen Syste= Firing Circuit Tes' Required Frequency: Two =cnths or less.
Test Date: September 5, 1973 and Nove=ber 3, 1973 Results: Satisfactory. However, the test perfor=ed September 5 was unti=ely.
It should have been
.perfor=ed en er about August 6.
This was re-perted on a deviation report and reviewed by the Plant Review CH ttee.
b._ Syste=: ~ Explosive Valve Frc= Equalizing Line Required Frequency: Twelve scnths or less.
Test Late: Not r quired during this report period.
Results: N0ne.
c.
Syste=: Explosive Valve _Frc= Nonequalizing-Lines Required Frequency: Twelve months or less.
Test Date: _Not required during this report period.
Results: Nene.
- 6. -Radiation Monitoring a.
Syste=: Air Ejector and Off-Gas Monitor Syste=
Required Frequency: One =onth or less.-
Test Date:.' July 26, August 23, October 1, October 31, Nove=ber 30, Dece=ber 23,.1973 Results: Checks shoved the calibration to be satisfactorv
-(vithin 20% of the 2 5 x 103 eps alar = setting).
The autc=atic closure function of the isolation
. valve timer was checked. The test showed the.
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timer calibration to be satisfactory (within 3% of the maximum timer setting) and the iso-l lation valve closed as specified.
b.
System: Calibration and Functional Test of the Stack Gas Monitoring System Required Frequency: One month or less.
Test Date: July 26, August 23, October 1, October 31, November 30, December 28, 1973 Results: The stack gas monitoring system was checked using the built-in Cs-137 calibration source. The instrument check showed the calibration to be satisfactory, resulting in the alarm point occurring within the specified 0.1 curie per second release rate.
c.- System: Analyses of Stack Gas Particulate and Iodine Filters 7
Required Frea.uency: Weekly.
4 Test Date: The' analyses-were conducted weekly.
Results: The results of analyses of the stack gas particu-t late filter and iodine filter are reported in terms of curies released in Appendix A of this report.
d.
System: Calibration of Emergency Condenser Vent Monitor
-Required Frequency: One month or.less.
Test Date: July 31, August 23, September 28, October 31, November 29, December 28, 1973-Results: The emergency condenser vent monitors are checked by comparing with a calibrated portable instru-ment. The checks showed the vent monitor calibra-tion to be satisfactory with all monitor checks within-! 5% of full scale.
- e.. System: Calibration of Canal Liquid Process. Monitor Required Frequency: One month or less.
Test Date: July 26, August 23, October 1, October 31,
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- November 30, December-28, 1973 s
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4 Results: The calibration of the canal liquid process i
monitor is a comparative calibration used to demonstrate operations of the monitor and to i
detect gross calibration changes and/or instru-ment drift. The results of these monthly l
calibrations showed that a renitor drift has occurred since the lust calloration which i
utilized certified standards. Becalibration of the monitor with liquid standard sources will be completed shortly. Also, an accept-ance criteria for a process monitor calibration vill be developed.
f.
System: Canal Liquid Collection Sample Required Frequency: Daily.
Test Date: The analysis was conducted daily.
Results: Satisfactory.
E.
THE RESULTS OF ANY PERIODIC CONTAINMENT LEAK BATE TEST PERFORED DURING THE REPORT PERIOD
-No integrated containment leak rate test was performed during the report period.
F.
TECHNICAL SFECIFICATIONS CHANGES During this report period, one Technical Specifications change was authorized by the Commission.
Change.39 - This change describes changes in plant organization and titles associateC ith the creation of the i
Operations Engineer.and Maintenance Engineer job classifications.
G.
' CHANGES IN PLANT OPERATING ORGANIZATION INVOLVING KEY SUPERVISORY PERSONNEL i
1.
On July 1,.1973, the plant organization.was changed. ~These changes were made to make the plant organization more responsive to_
present day operating requirements. The changes involved eliminating the Assistant. Plant Superintendent position and adding the positions
.of Operations Engineer and Maintenance Engineer. The Operations,-
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Maintenance and Technical Engineers all report directly to the Plant Superintendent as does the Quality Assurance Engineer.
Mr. George Tyson assumed the job of Maintenance Engineer.
Mr. Tyson had previously held the position of Assistant Plant Super-intendent at Big Rock Point since he first reported there in 1968.
Mr. Tyson has held a Reactor Operator's license at Big Rock Point since 1969 Mr. "harles R. Abel was promoted to the position of Operations Engineer on July 1, 1973.
Mr. Abel has been at the Big Rock Point Plant since 1967 except for a brief period when he served on a special assign-ment at the Pickering Power Station in Canada.
Mr. Abel has held a Senior Reactor Operator's license at Big Rock Point since June 1969 2.
On November 15, 1973, Mr. Earl F. Peltier was promoted to the position of Assistant Shift Supervisor.
Mr. Peltier has been at Big Rock Point continuously since March 1962 where he was on the original operating crew in the position of Control Operator No 1.
Mr. Peltier has held a Reactor Operator license at Big Rock Point since 1962.
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II-16
l III. POWER GENERATION Report Total Period To Date 1.
Thermal Power Generated (K4h )
808,763 12,19h,h76 t
2.
Gross Electric Power Generated (K4he(g))
253,148 3,885,350 3
Net Electric Power Generated (M'Jhe) 2h0,287.2 3,680,070 h.
Hours Critical (h) 3,757 2 68,66h.8 5
Hours Generator On-Line (h) 3,751.0 66,882.3 h
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IV.
SHUTDOWNS A.
TYPE - SCHEDULED 1.
UnitOffLine-11/1/73 0036 T
2.
Unit on Line - 11/4/73 2339 3
Length of Outage - 95 Hours, 3 Minutes 4.
Discussion - This was a scheduled outage to perform the i
necessary semiannual license requirements on the control rod drives.
_ Power de: cent was controlled and deliberate to a cold shutdown mode.
Big Rock Point established an international record for length of oper-ation of a BWR facility without power Interruption. The p ant had been in continuous operation since April 16, 1973 generating 198 consecutive calendar days at a unit capacity factor of 90%.
B.
TYPE - FORCED 1.- Unit off Line - 12/8/73 0600 2.
Unit On Line-
~
- 3.. Length of Outage (Plant Still Shut Down at End of Reporting Period) 4.
Discussion - The unit was forced out of. service due to a
. packing gland leak on an instrumentation root valve. Off-gas release rates were in the region of 10,000 pCi/see _ (unit output was 53 MWe(g))
when the' plant was shut down for repairs. The method of shutting down was a controlled deliberate shutdown to a cold shutdown mode. Valve'
. packing was replaced; _however, the outage schedule was extended to rq air the emergency condenser. For details, please nference Section 2
/V G and VI A(5) of this rep' ort. At the end of the reporting period, the' unit was off the line in the cold shutdown condition.
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SAFETY-RELATED MAINTEIA' ICE Note: Dates contained in this section generally refer to the veekly period when the maintenance was performed.
A.
REACTOR PROTECTION AND COIiTROL SYSTE4 INSTRLHEITATION 1.
Neutron Monitoring Channel No 1 - 11/29/73 - Upscale drift in the output signal of this picca==eter was traced to defective contacts in the picoa==eter range sviten.
I==ediate repairs consisted of exercising the range switch between the "125%" and " Test Trip" positions to eliminate resistance in the range switch cc-tacts.
Subsequent repairs consisted of the same action with the picca==eter re=oved frc= service so that more switch positions (h) could be "viped."
The contacts in the range switch are in the feed-back circuit of the picoa==eter and increased resistance in the feed-back loop tends to increase picoa==eter output. This svitch will be cleaned and inspected at the next refueling outage.
Failures of thf s type are considered to be within the design limitations of the equipment. The Technical Specifications and plant design provide for the te=porary removal for maintenance of one power range flux monitor frc= service without co= pro-dsing safety.
2.
Neutron Monitoring Channel No 2 9/13/73 - The picoa==eter for this channel was replaced a.
following reports of a rise in recorder trace level. The indication was still present following picon==eter replacement and was traced to the range switch for this channel. Exercising of the range switch alleviated the proble=. This switch vill be cleaned and inspected during the next refueling outage. Failures of this type are considered to be within the design limitations of the equi; cent.
b.
12/31/73
. The high-voltage power supply for this neutron monitoring channel was replaced-_vith a spare unit following erratic flux level measurement at the most sensitive positions of the picoa=:eter range switch. Bench testing of the failed unit resulted in replace =ent of three marginal electron tubes. This failure occurred while the reactor was in " cold shutdown" for plant maintenance. Failures of this
- type are considered to be within the design limitations of the equip-ment.-
V 3
Neutron Monitoring Channel No 3 - 7/12/73 - The picoa= eter in this channel was replaced with a spare unit on July 11 following nmall variations of 2%-L% on the neutron flux recorder trace. Bench testing of the unit recoved revealed no proble: and operation of the channel remained normal. The proble= is now attributea to the range switch feed-back circuit contact resistance (similar to that observed in the other two power channels) and this switch vill be inspected and cleaned at the next refueling outage.
Failures of this type are considered to be within the design limitations of the equipment. Technical Specifications and plant design provide for the temporary re= oval of one power range flux =enitor fro:
service without co= pro =ising safety.
h.
Neutron Monitoring Channel No k a.
8/9/73 - The Log N-Period amplifier in this channel was repaired following reports of erratic period =easurement while at power.
Repairs consisted of replacement of a defective (gassy) electron tube in the period amplifier circuit. This type of failure is considered to be within the design limitation of the equip =ent.
The Technical Specifi-cations and plant design do not require this instrument to be in service when reactor power is above 5% of rated power.
b.
11/6/73 - The high voltage power supply for this channel was replaced with a spare unit following syste= response failure during instrumentation checkoff on Nove=ber 3,1973 In performing the response check, it was noted that the Long N-Period indicator readings would increase when the co=pensatien voltage was increased. This was first diagnosed as a defective chamber and the chamber and coaxial cables fro = the cha=ter to the chamber drive head vere replaced. When this did not correct the proble=, it was determined that the high-voltage power supply was defective and the unit replaced with the spare.
Inspection of the defective power supIly re-vealed that the unit was connected as a positive-positive supply instead of positive-negative as required. This unit was installed during a previous failure on June h,1973, while the reactor was at power and could not be checked for chamber response.
V-2
As a result of this error, which has been discussed by the Plant Review Ccesittee and reported an abn =al occurrence ( AO 73), appropriate steps have been tahen to veri'y polarity of replace =ent power supplies when nor-*' testing methods are not possible.
5 Neutron Manitoring Channel No 5 - 8/16/T3 - The Lcg N-Feriod a=plifier in this channel was repaired folleving reports cf erratic period ceasurement. Repairs consisted of replacenent of a defective electron tube in the period a plifier circuit. Failures of this. type are considered to be within the design limitations of the equip =ent.
The Technical Specifications and plant design do not require this instrument to be in service when reactor pcver is above 5% rated pcVer.
6.
Neutron !/onitorine Channel No 6 - 12/13/73 - The coaxial cable connector at the chamber location was repaired en this syste folleving response failure after plant shutdovn en Dece=ber S.
The coaxial cable cla=p en the cha:ber had loosened, placing strain en the cable and allowing the center wire to withdraw. The connector was repaired and the cable secured to preclude a future probles of this nature. This type of failure is considered to be within the design limitations of the equip =ent.
Technical Specifications and plant design provide for re= oval of one start-up channel for taintenance during plant shutdown w nditions.
7 Neutron Monitcring Channel No 7.- 12/31/73 - The high-voltage power supply for this channel was repaired folleving loss of count rate indication.
Inspection of the supply indicated the voltage had dropped to approxi=ately 300 volts (nor-#.11y 650). Eepairs to the supply cen-sisted of electron tube replace =ent.
This type of failure is considered to be within the design limitations of the equiptent.
Technical Specifications and plant design provide for renoval of one start-up channel for =aintenance during plant shutdown conditions.
8.
g aetor Protection Systen Sensors a.
11/S/73 - Recalibrated the high reactor pressure ser=
and high condenser pressure scra: bypass sensors to a more conservative V-3
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4 l-set point. The high condenser pressure scram bypass sensors were found to operate at a less conservative set point than required and were re-ported as an abnormal occurrence (AO-12-73). The calibration of the reactor steam drum low water level scram sensors and the high conde:mer pressure scram sensors were checked at the request of the Operations Department. All sensors checked normally and were within required limits.
All testing of the sensors was perfomed with the reactor in cold shutdown condition.
b.
12/13/73 - Calibration checks were perfomed on the high condenser pressure scram bypass sensors to determine if any instrument drift had occurred since prior calibration (11/8/73). All sensors were within calibration specifications and operated normally.
Testing of the sensors was perfomed with the reactor in cold shutdown condition.
B.
RADI0 ACTIVE EFFLUENT MONITORING SYSTDiS J
1.
Air Ejector-Off-Gas System a.
Maintenance Related to Off-Gas System Integrity Testing 11/8/73 - Off-Gas Isolation Valve - The off-gas isolation
~
valve, CV 4015, was replaced with a newly prEcured valve designed to close i
tightly enough to isolate the off-gas system. This work was perfomed with the reactor in cold shutdown.
12/31/73 - Off-Gas System - The following components vere i
inspected and repaired with the reactor in the cold shutdown condition in preparation for integrity testing of the off-gas piping:
(1) After Condenser Drain Isolation Valve, CV ho30 -
Inspection revealed scale on the valve internals and imperfect seating.
The. valve was cleaned, the seat and dise were lapped, the packing was replaced and the valve was test operated and returned to service.
(2) Air Ejector Off-Gas Drain to Radwaste Isolation Valve, CV ho35 '- Inspection revealed-the valve seat and disc to be in
" fair" condition.- The' seat and disc were lapped and the valve was
. test operated and returned to service.
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11/8/73-Off-GasMonitor (1) Purge Valve Bypass - A bypass line was added around the three-way purge valve (SV RL 25) to facilitate cleaning of this valve while the plant was on the line.
(2) Purge Valve Repair - The three-way purge valve (SV RL 25) and the two-way purge valve (SV RN 25) were disassembled and inspected (be+,h valves were quite scaly and dirty). The valves were cleaned and reassembled and returned to service.
The work was performed with the reactor in the cold shut-down condition.
c.
Off-Gas Filter Changes 11/8/73-Theoff-gasfilteranddemisterwasreplaced.
12/31/73 - The filter was replaced.
Both filter changes were made with the reactor in the cold shutdown condition.
2.
Stack Gas Radiatien Monitoring System 7/19/73 - Several compenents in this system were checked a.
tal".owing erratic operation of the single isotope channel. Replacement cf the differential dit.criminator provided some improvement due to a higher output signal level. However, the major t.aurce of the problem was discriminator shift on the log count rate meter. This was corrected by recalibration of the discriminator. The coaxial cable between the differentir21 discriminator and the log count rate meter was also re--
placed with a new bable of shorter length to reduce signal attenuation,
- b. _7/26/73 - The spare differential discriminators (2) were bench tested and calibrated. Several marginal electron tubes were re-placed and minor adjustments were performed on the regulated power sup-plies and "E" ' dial span controls, 9/6/73-Thedifferentialdiscriminat6rinthissystem c.
was replaced with a spare unit following instability in the single iso-tope channel. _ The instability was evident only during the daily cali-bration procedure, at which time minor changes were required in the detector polarizing supply to maintain system calibration. Repairs to
(
the failed unit consisted of el.ectron tube replacement.
V-5
- i. I 2
9/13/73 - The linear amplifier was removed and repaired d.
i.
in this system following reports of erratic indication. Repairs consisted of replacement of defective electron tubes and alignment following fail-ure of the automatic scan feature.
The failures above are considered to be within the design s
l limitations of the equipment. Removal of this system from service is permitted by the Technical Specificat. ions provided repairs are made promptly and the system is returned to service. The off-gas monitors provide backup for this monitoring system.
3 Liquid Process Monitoring 9/13/73 - Discharge Canal Liquid Process Monitor - The a.
linear count rate meter in this channel was replaced with a spare unit following reports of full scale failure. Subsequent bench testing re-s21ted in repair of a defective internal power supply socket (broken solder connection). A failure of this type is considered to be within the design limitation of the equipment.
b.
10/4/73 - canal Sample Pump - Failure of the canal sample pu=p to pump design capacity was ' corrected by replacement of-the pu=p impeller and casing.
- c. - 12/6/73 - Discharge Canal Liquid Process Monitor - The detector high-voltage supply cable for this channel was repaired fol-lowing loss of reading on the linear count rate meter. This occurred 7
immediately after monthly detector. calibration and was the result 'of l
moving the detector during the calibration process. The high-voltage connector.was repaired and the system returned to service.
Removal of this system from service is permitted by the
. Technical Specifications provided repairs are promptly made and the sys-tem returned to service.
C. - CONTAIN!GT. SPHERE IS01ATION ~ SYSTE i-1.
10/18/73 - Sphere Ventilation System - Leaking fit';ings were tightened in the' nitrogen-lines associated with SV 9152 in the emergency operating system for the supply ventilation. valves.
Performing leak tightness adjustments on low pressure ' gas tubing fittings 1without removing the gas system from service is within
(
the ~ scope of acceptable. maintenante practices. This approach was util-7 9
h.
V-6
l i-ized in this case and the safety of the sphere ventilation system was therefore not compromised.
D.
EGRGUiCY POWER SYSTU4 1.
Energency Diesel Generator a.
10/h/73and10/11/73-Investigaticn of a low battery charger reading on the emergency diesel battery system disclosed in-sufficient voltage and current output from the charger while en " fast charge." This was corrected by replacement of the batteries and the battery charger rectifier.
Plant operating r-quirements permit removing the emer-1 gency diesel frc::. service for periods in excess of 30 minutes with the approval of the Plant Superintendent. In each of the cases noted above, the Plant Superintendent's approval was obtained for removing the diesel from service only for the time specifically required for final hookup, troubleshooting and replacement, respectively.
b.
12/13/73 - The control panel indicating lights for this unit were replaced with new sockets and lenses to improve reliability and visibility of alam indication.
The original panel lemps were of three different-varieties, some of which had screw type bases and were susceptible to vibration (and loss of indication). The new sockets have baycnet bases and are sufficiently bright to be seen in most room locations.
The la=p sockets provide local alam. indication only
~ for a co::non trip and annunciator scheme on the diesel generator.
Replace =ent of the la=p sockets involved low-voltage wiring and no
. special precautions were required-to provide for reactor safety.
4-2.
125 V D-C Power System
- a. ' 8/16/73 - A 2-1/2 amp d-c ground on the 125 V d-c motor' control center ground test station was eliminated by replacement of low accumulator pressure scram unit No D174.
b.
12/20/73-Intermittent fluctuations and flickering of l
' the control room d-c emergency lighting disclosed the HGA relay contacts in lighting panel 5L to.be burned up. The contacts were replaced returning
..the system to service.
V Continued availability of.the battery system was assured by removing the system from service only as required for replacement of failed parts.
~ V-7
E.
FIRE PROTECTION SYSTH4 1.
Diesei Fire Pump - 10/18/73 - The starting batteries 3r the diesel fire pump were relocated for safer accessibility.
i The electric fire pump provided primary fire system supply potential while the diesel fire pump batteries were relocated.
F.
LIQUID p0ISON SYSTD4 12/20/73-Ir vestigation of air leakage from oparator diaphragm joint on poison system discharge valve CV !;O20 disclosed inadequately tightened flange bolts. The diaphragm was replaced as a preventive maintenance measure a w the flange bolts were properly tightened. This maintenance activity was conducted with the reactor in cold shutdown.
G.
D4ERGENCY CONDDISER SYSTD4 1.
12/13/73 - Emergency Condenser - Hydrostatic testing of l
both tube bundles-in the emergency condenser disclosed an inability i
of the north tube bundle to achieve test pressure. Investigation disclosed three leaking tubes in this bundle. Dye-penetrant inspection indicated the tube leaks to be in the seal weld joining the tube end to
~
the tube sheet. Upon disassembly of this tube bundle, the inlet-outlet water box baffle plate: was observed to be warped or bowed. The findings of the inspection on the north tube bundle.resulted in the decision to conduct a similar inspection on the south tube bundle.
2.
12/20/73 - Further examination of both tube bundles disclosed the required repairs to be as follows:
a.
North Tube Bundle (1.) Modify _ inlet-outlet water box baffle plate.
(2) Repair three leaks and two indications at the tube-to-tube sheet velds,
~b. ' South Tube Bundle'
-(l) Modify inlet-outlet water box baffle plate.
(2) Repair eight indications at the tube-to-tube sheet
' welds.
Southwest-Research was contracted to engineer and repair the baffle plates while Consumers Power-qualified procedures and a welder i
to conduct the tube-repair.
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3 12/27/73 - The old baffle plates were removed by air are frcm the north and south water boxes. The water boxes were ground out 4
and made ready for the installatien of the new baffle plate. New baffle plates were installed in both water boxes as described in Section VI A(5) - Facility Change C-238.
4.
Tube Sheet Welds - A mockup of the tube sheet was fabricated with 1/4 inch 304' stainless steel overlay, 20 holes drilled and tubes installed and welded, to qualify the veld procedure and then the welder to that procedure.
5 12/27/73 - Emergency condenser outlet valve M3-7053 - The valve was disassembled for inspection and nondestr tetive testing. The gate and seals were cleaned and reground before reassembly. This in-spection was required because of slight leakage experienced while con-ducting the hydrostatic test on the tube bundle.
All repairs were made with the reactor in the cold shut-down condition.
H.
REACTOR CLEAN-UP SYSTai 1.
Clean-Up Pu=p a.
9/27/73 - The clean-up demineralizer pu=p was replaced during this. report period due to failed windings. The replacement pu=p had been ec=pletely rebuilt following a winding failure whicu had oc-curred approximately one year prior to this date. The change out was again performed in accordance with applicable Quality Assurance require-ments. The failed purp is scheduled for 4-diate replacement and a facility change has been approved to convert the pu=p from a. welded-in installation to a flanged arrangement to simplify pu=p maintenance.
(Facility change work was not yet initiated.)
This work was performed during a period when the reactor was in cold shutdown.
b.
11/15/73 - The failed clean-up system pump which was re-placed was:coc:pletely disassembled and deconta=inated. 'The pump casing, bearing housings and end cover plate were-salvageable and were returnel to the manufacturer for rebuilding. The re. built pu=p will be stored as
[
a spare.
V-9.
I.
PPlMARY COOLANT SYSTE4 1.
Reactor Recirculating Water Purp a.
7/26/73 - A leaking flexible cooling waterline to the No 2 reactor recirculating water ptp thrust bearing was replaced during a load reduction this report period. The leak was caused by nor=al deterioration cf the flexible hose material.
This work was perfomed during a power reduction with the No 2 recirculating loop isolated, thus assuring personnel and reactor safety.
b.
8/16/73 - A failed flexible cooling waterline to the 3/4 inch heat exchanger for the No 1 recirculating pu=p was replaced.
This work was likewise perfor=ed during a power reduc-tien with the No 1 recirculating loop isolated.
2.
Reacter Recirculation Purp No 1 - 11/8/73 - Consumers Power Company Regicn Electric Laboratory replaced a defective thermal over-current relay (149 T4C - X phase) in the No 1 reactor recirculatien purp motor control scheme. The alar = setting en the defective relay could not be properly adjusted to the required setting.
The relay was replaced while the plant was in cold shutdevn for other testing.
J.
CCNTd3L ROD DRIVE SYSTm 1.
Centrol Rod Drive pu=ps (CRD)
- a. -8/23/73 - I= proper operation of the CRD purp discharge 1cw-pressure alar = vas traced to a plugged reference line to the switch.
The line was dimntled, flushed' and returned to service.
The repair was made during reactor operation with the CRD system at cperating pressure. Ecwever, the reference line provides sensing for alarm and. indication cnly 'and had no significant'effect on plant operation as backup indication of the CRD discharge pressure is available.
b.
9/27/73 & 10/4/73 - Investigation of pressure pulsations on the No 1 CRD pu=p disclosed that one discharge valve seat and one suction valve seat were cocked in the pump casting allowing leakage
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A between the casting and valve seats. The gouged out areas of the pu=p casting were repaired with Devcon Plastic Steel and the valve reseated.
The above repairs en the CRD pu=ps were made with the reacter at power. One of the two CED pu=ps may be remov-d frcm service and still raintain normal operational status.
2.
ControlRodDrivetandInstrumentation-11/8/73-The follewing repairs were made to the CRD position probes:
C 4 - Aligned "02" switch position.
2 C Replaced a defective reed switch for the "o3" posi-tien. The reactor was in the cold shutdown condition during the above probe repairs.
3 Control Rod Drive Scram Accu =alaters a.
7/26/73 - E 4 Accumulator - A failed gas side accu =alator burst disc was replated on the E 4 CRD accu =alator. Failure was determined to be due to gas erosion, i'
b.
8/9/73 - E-3' Accu =ulator - The nitrogen 0-ring seals on i-CRD E-3 accu =alator were replaced to step leakage.
8/30/73 - C 4 Accu =ulator - A leak from the joint between c.
the' accu =ulator halves was corrected by relacement of the 0-ring and backup ring. utilized at that point.
- d. - 8/30/73 - Replaced a defective gauge en C 4 accinlator following reports of false readings during charging of this accu =alator.
e.-. 9/13/73'- D 4 Accumulater - A leak from the joint between
.the accumulator halves was corrected by replacement of the 0-ring and backup rings used as seals at that point.
f.
9/27/73 - A 4 Accu =alator -' Leakage from the A-4 I
accu =ulator was. corrected by replacement of ceals between the accu =a-lator halves.
g.
12/20/73 -~F-2 Accumulator - Leakage from the F-2 accu =u-
-lator was corrected by replacing the bladder and the C-rings and backup rings on both the water and gas sides of the acetelator.
'~
The above repairs were ccnducted with the reactor at cper-ating pressure. Under these ecmditions{ the primary hydraulic source' for
( _
CRD scramming comes frcm the reactor vessel. This design feature permits repair of an accielator without affecting reactor safety.
. V -
f In each case, only one accu =ulator was removed from service and only for the time required to perfor= the corrective maintenance.
3 Accumulator Drain Valves a.
8/30/73 - C 4 and i * \\ccu=ulator Drain valves - Leaks frcm the waterside drain valves were repaired by relapping the valve discs and seats.
b.
9/13/73 - D 4 Accu =ulator Drain Valve - Leaks frc= the waterside drain valve were repaired by relapping the valve disc and seat.
The above repairs were made with the reactor at operating pressure. Under these conditions, the primary hydraulic source for CRD scra: ming cc=es from the reacter vessel. This design feature per-mits repair of an accu =ulator without affecting reactor safety.
t In each case, only one accu =ulator was removed from ser-dce t
and only for the time required to perform the corrective =aintenance.
4.
CRD Pu=p Belief Valves 7/12/73-ControlRodDrivePu=ps-Excessiveleakage f
a.
from the No 1 CRD pu=p relief valve necessitated its replacement with a rebuilt valve. The newly installed valve was set to relieve at
' 1900 psig.
b.
10/25/73-Control Rod Drive System - The No 2 CRD puu? relief valve was replaced due to excessive leakage.
c.
11/18/73 - No 1 CRD Pu=p Relief Valve - The valve was repaired and reset due to excessive leakage.
K.
FEED-WATER SYSTDI I
1.
Peactor Feed Pu=ps -' 12/13/73 No 2 Reactor Feed Pump - Failure of the pump to start through use of the control room hand switch resulted in the following inspections: -
- a. ' Control' Circuit '- The feed pu=p breaker was tested and I..
cleaned. Inspection revealed the trip coil linkage to be binding slightly.
r preventing ~ resetting of the coil after a pu=p trip.
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Auxiliary Oil Pump - Indications of low after-filter discharge pressure (8.5 psig) on the No 2 auxiliary oil pu=p were corrected by readjusting the relief valve to relieve at 10.1 psig (approximate correct setting). Low discharge pressure prevented the associated pressure switch from closing, completing the feed pump start circuit.
Either of the conditions in a. or b. above would have pre-vented the pump frc properly starting. Corrections of these problems have returned the No 2 feed pump to service.
The above repairs were made during the forced outage of 12/8/73, during which the plant was in the cold shutdown condition.
- r.. STEAM DRUM 1.
Steam Drum Level Sensing Root Valves a.
11/8/73 - Minor leakage from the steam drum ecst end level element bottom reference line valve was corrected by ti htening 6
the packing.
b.
12/18/73-The failed packing en the east end instrument root valve was replaced, restoring the' instrumentation system to service.
The west end instrument root _ valve packing was also replaced as a pre-
' ventive maintenance measure.
The above maintenance activities were performed with the-
- reactor in cold shutdown and the steam drum drained.
i-4 V-13
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ji VI.
CHA?GES. TESTS. EXPEPITENTS A.
FACILITY CHMGES FERF0FJED PURSUANT TO 10 CE 50.59 l.
Facility Change C-21h This change added sealing water to the shaft seal on the diesel
,i generator cooling water pump. The seal water system utilizes a head tank supplied from either the service water system or the diesel gener-ator water pu=p.
The head tank will provide 1 gpm water at approxi-mately h.h psig for 2h minutes which will be adequate to provide sealing 7
water to the pump shaft seal until the diesel is started. The 4.h psig setl water pressure will be sufficient to permit early detection of pump shaf t packing problems and eliminate pump failure due to packing air leakages. The addition of the seal water system does not consti-e tute a change in the designed-function of the diesel generator as de-
-scribed in the FHSR and the Technical Specifications; nor is the designed safety of the diesel generator impaired since the seal water system will assure adequate cooling of the diocel through proper functioning of the I
-water pump, t
2.
Facility Change C-228 This change covered the plugging of the drain line from each
[
reactor feed pump base. The original design provided for draining of accu =ulated waste (oil and/or water) to either the radwaste or turbine
. sumps. Valves were provided to clore the drains off entirely or to 4
divert the flow to either of the sumps mentioned above. During normal f.
opere.tions, these drain l!nes were valved closed. When they_were used, they drained accumulated waste from the pump bases to the turbine su=ps and from there to the clean vaste receiver tanks. The use of these
[
drains thus provided the potential of introducing oil into the clean l-Lwaste~ system. Since the normal accumulation of waste on the pump bas 5s
[
_is of a small enough volume to allow-easy removal by hand, the potential of introducing oil into the clean waste system was eliminated by plugging
~
the drain lines.. ' A safety evduation determined that this change did j
.not constitute a change'in the designed function of'the feed pumps as described in the FHSR and the Technical Specifications; nor is the de-
' [
signed ' safety of the. feed pumps impaired since accumulated waste cannot' enter into the feed pump mechanisms' or_ the components of any other
- equipment in the vicinity.
. - yI_1 s
1
(
3 Facility Change C-230 This change facilitated construction of a 12-foot diameter by 20-foot high tank for storage of radioactive materials removed from the spent fuel pool while it was being relined.
Safety analyses pur-suant to 10 CFR 50.59 were conducted to examine the effects of flooding and excosure should the tank rupture. The analyses assumed complete tank rupture and concluded:
(1) there would be no additional hazards to safety related equipment if the tank contents were limited to 1217 cubic feet of water; and (2) the resultant exposures at the nearest site boundary would remain within applicable requirements if the field at the top of the tank were lisited to 10 R/hr.
Since these criteria established that failure of the tank would not present a significant change in the hazards considerations described or implicit in the FHSR, the tank was so utilized.
h.
Facility Change C-231 This change facilitated the addition of a bypass 11ne around the three-way air purge valve on the off-gas monitoring system. The bypass line will allow inspection or repair of the three-way valve during normal plant operation.
(In the past, the effectiveness of purge in removing excess moisture has been hindered through the intro-duction of dirt into the three-way valve.)
l A aafety evaluation concluded that this change does not consti-tute a change in the designed function of the off-gas monitoring system as described in the FHSR and the Technical Specifications; nor is the l
designed safety of the system impaired sinco malfunctions can now be quickly~ corrected.
5 Facility Change C-238 The change replaced the original baffle plates on the emer-l gency condenser inlet-outlet water boxes. The original baffle plates had been warped in service and were replaced with the following re-design: Each new baffle plate consists of a center section which is
- bolted to 'a narrow ledge welded horizontally around the three interior b,
sides of inlet-outlet water box; and, which when in position on the tube bundl* butts up against a fourth ledge welded horizontally to the VI-2 L
(
tube sheet. The redesign of the baffle plates considered flow induced and thermal response loadings on the plates and bypass flow around the plate (the fourth ledge replaced a flexitallic gasket en the original design). The new design concluded that the flow induced loading and the bypass flow to be relatively insignificant and the highly flexible design of the new baffle plates to adequately withstand the thermal response loadings. It was thus concluded that the baffle plate modi-fication does not constitute a change in the designed function of the emergency condenser as described in the FHSR and the Technical Specifi-cations; nor is the designe safety of the system in: paired since the new baffle plates will withstand the loadings imposed during the in-tended service. In support of these conclusions, a baseline test of the emergency condenser north tube bundle will be conducted when the system is returned to service.
B.
TESTS PERFORMED PURSUANT TO 10 CFR 50.59(b) 1.
Liquid Poisen System Explosive Valve Firing Circuits 11/8/73 - During the scheduled shutdown for semiannual i
a.
testing, the firing circuits were tested for operability utilizing 2-ampere fuses in place of the explosive valves. The test failed due to the incompatibility of the test equipment with the circuit design. Fur-ther test results are described below, b.
12/20/73-Component checks and operability tests were l.
performed following pirnt shutdown. Resistance measurements indicated j
that no problems existed in the relay contact surfaces or fuse clip resistance.
t Firing tests were performed using various size fuses in the explosive valve test devices _(simulators). Firing with the 2-ampere fuses was marginal as a time' lag is encountered upon firing-(the paral-(
lel 2-iwpere fuses approach total circuit current capability).
Firing with the 1-ampere fuses was acceptable. The total time req'uired to open all fuses was 0.19 second (including time re-L
_ quired for control relay closure).
Firing with the 0.5-ampere fuses was nearly instantaneous.
I.
The total time required for fuse opening was approximately 0.070 second, t-
'VI-3.
including the 0.050 second required for relay closure. Firing of the 1-ampere fuses ind'.<.as the circuit is reliable in meeting the 1-ampere maximum firing cuneat criteria of the explosive valves.
12/27/73 - In order to verify an adequate safety margin c.
above the 1--
,.e maximum firing current criteria of the explosive
. valves, an operability test using 15-ampere fuses was performed on this system on December 22.
All circuits worked as designed and the total firing times for Circuits A and B were 0.550 and 0.650 second, respectively.
The reactor was in cold shutdown during all testing on the poison system and no additional precautions were required to provide for reactor' safety.
All tests were performed using approved written procedures.
Prior reviews of these procedures were performed to ensure that these tests erere consistent with Technical Specifications and did not involve an unreviewed safety question per 10 CFR 50 59 2.
Steam Drum Level Tests 4
7/25/73 - A special test was conducted to determine. if a.
4 variable reactor recirculation pump flows could affect steam drum levels at either end. The test results showed that there had not been any subtle or small changes in relative pump flows which would account for l
l the change in drum tilt over the past years.-
b.
10/31/73-A special test was committed to in order
- that data may be gathered to evaluate any change in the reactor steam drum tilting phenomena as a function' of power. Steam drum elevations were measured for both ends of the drum at power levels of 10 We(g),
.30 We(g), 50 We(g) tnd 70 We(g). There was little noticeable change over the full range of power settings. The data are tabulated below:
Yarway Bailey i
I Plant West East East West Output RE06A RE20A RE06B RE20B. Recorder IndicatoI 3
+3
-0
-1 0
+4
- 69 We(g) -
+
' 2.5 o
-1 o
+4 l-50 We(g)-
+3'
+
I 4 30 W e(g)
+3-
+3 0'
-1 0
+4
~
1 10 W e(g)
+3
+2.5.
'o
-l_
.5
+3-VI-4
,, +.
,-..,s.
,,,--..-..,w..
..m.
, = - - -
r
,r-_.,,
t This test will continue to be run at regular six-month inter-vals to verify the posture of the drum.
This test was performed using approved and written procedures.
A prior review of this test determined that this test was consistent with Technical Specifications and did not involve an unreviewed safety question per 10 CFR 50.59 3
Reactor Recirculation Purr. Interlock Tests 11/3/73 - A special test was conducted to fulfill the oper-ating requirements of Technical Specifications, Section 6.1.5(q). The logic assaciated with the pump starting circuits with the various set-tings of the valving was checked. All systems logic checked out as required. One deficiency, the annunciator on No 2 recirculating " pump trip," did not alarm. The circuit was repaired and test-operated satis-factorily.
-In conjunction with.the above tests, the valves (pump suction, pump discharge and discharge bypass) were timed over their entire open-ing and closing strokes. The pump discharge valves were found outside the limits specified in the Technical Specifications as indicated in our letter to the AEC dated December 6,1973 The gear train ratio in the limitorque valve operators will be modified to bring the timing within limits.
This test was performed utilizing an approved written pro-cedure. Prior review of this procedure indicated that it was consis-tent with Technical Specifications and did not involve an unreviewed safety question per-10 CFR 50 59 h.
Off-Gas Isolation Valve Tests
'12/29/73 -~'A special test was run--following the. installation of _ a new off-gas isolation valve. The purpose.of the tsst was to de-
- termine the integrity of the off-gas. piping system and to verify the holdup line_ volume. The initial test of the system showed leakage in r
th'e region of the absolute filter and the off-gas isolation valve.
. Following. repairs' (see Shintenance Section of this report), the~ integ-
-rity of_the-system was. verified with a static holding test.
.(
The volume of the -system was also calculated to be as designed..
- At the' close' of _the report period,' isolating tests of the off-gas -system
.~
'VI-5' q
were being formulated and will be conducted with the plant in service.
This test was performed using a written and approved procedure. Prior review of th5.6 procedure indicated that it was consistent with Technical Specifications and did not involve an unreviewed safety question per 10 CFR 50.59 l
l l
1 I
i VI-6
f VII. PADICACTIVE FFFLG7;T PELEASES A.
I!;TRODUCTION Releases of radioactive naterial both to the at=osphere and Lake Michigan from January 1 to Dece=ber 31, 1973 were well within the facility-licensed limits and the AEC's regulations, particularly Title 10, Code of Federal Regulations, Part 20.
E.
GASEOUS UyLUEIT Gasecus releases to the atmosphere totaled 22L,500 curies of fission and activation gases. This corresponds to.17% of the licensed technical specification li=it of 1 Ci/s.
Particulate releases totaled 0.37 curie or o.87% of the licensed limit while halogen releases were teasured to te h.7 curies or 27% of the licensed li=dt.
+ Gross alpha measure =ents on the particulate filter reveeled that the release of f
alpha emitting nuclides tetaled 1.6 x 10 ~ curies. The tritium re-leases for the period totaled 85 curies of 7 x 10) of litit tased upon =eteorological dispersion to the point of =axitu ground concentra-tion.
1.
Gaseous Effluent Calcule.tional Methods A ss=ple of off-gas is obtained weekly during power operaticn and analyzed by gan=a spectrometry for ++six noble gas radionuclides.
Eased upon the =ixture of the six nuclides, a stack release rate, which includes a total of 22 noble gas radionuclides, is determin.d. The stack release rate is based on a 16-=inute holdup time for off-gas plus a 1% contribution from the turbine sealing steam systet utilizing a 2-minute holdup. The 1% turbine seal contribution has the sate dis-tribution of nuclides as the off-gas corrected for a 2-=inute decay period. This is reflected in the =onthly totals shavn in Appendix A.
- Due to uncertainties in iodine collection efficiencies for various species and potential sa=ple line plateout, the =easured values will be arbitrarily tripled for reporting purposes. A detailed study is currently being nade to enpirically quantify (and significantly reduce) the appropriate corree-tion factor.
++The six nuclides are:
Kr-85=, -87, -88 and Xe-133, -135 and -138.
(
VIl-1
= -
i i
{
Accivation gas releases are composed primarily of N-13 The rate of release is power-level dependent and is incorporated in the total monthly releases shown in Appendix A.
particulate and halogen releases to the atmosphere 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.
4 Determination of release rates in this canner assumes radioactivity is I
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.
Table I, Appendix A, has been revised since the last Semi-l annual to properly distinguish between total particulates released and gross beta activity on the particulate filters and to correct an error in the reported percent of Technical Specifications limit for particulate releases.- (See footnote Table I, Appendix A.) The net beta activity, i
as now reported in Appendix A, represents the unidentified portion of the total activity present on the particulate filters (ie, gross beta j
activity minus the' identified isotopic activity). Unlike the individual isotopes, the net unidentified beta activity, due to the lack of a known half-life, has n'ot been corrected.for continuous deposition and 1
decay.
Tritium releases to the atmosphere are calculated, based upon measurements.cade in the primary coolant and containment air and using identical concentrations for all releases as-follows:
Off-Gas - A flow rate of 10 com containing 90s radiolytic a.
gas' by volume at pri=ary coolant tritium to hydrogen ratio and at 10$
- relative humidity is used to determine tritium releases both in vapor and molecular form.
b.
Turbine Sealing Steam - The measured flow rate at 100%
. relative humidity and primary coolant tritium to hydrogen ratio.
- c.. Containment Ventilation - The measured flow rate and measured. containment building tritium concentration.
The results of these' calculations are'also shown in Appendix A.
].
VII-2
~
l 4
-nr-,
4w
~,
.,s
,.,. -, -., _ + -
,--%...w
.wa.. v. w w.-
m
,.--m.
,u
-,-e.-
2+,
v.
.-.r.-
m
C.
LIQUID EFFLUENTS Liquid waste rel:sses totaled 2.65 curies of radioactive caterial. This release corresponds to 31% of Technical Specifications limits. Additionally,19.7 curies of tritius were released correspond-ing to 0.006% of 10 CFE 20 permissible concentration in the discharge canal.
1.
Licuid 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 h9,000-53,200 gp dilution for liquid effluents is obtained through the use of the condenser circulating water purps, two at 2L,500 gym each and house service water purps, two at 2,100 gp each.
Each collected tank cf liquid is sarpled, analyzed for radio-active content, and discharged at a controlled rate to assure that per=issible concentrations are not exceeded in the canal prior to dilu-tien in Lake Michigan during the time of discharge. Each sample is analyzed by ga--a spectroretry to identify as nany of the cc=ponent nuclides as possible.
(See Appendix E for results.) Fermissible con-centrations in the canal are determined from the following-Ci E y3C
- I i
where Ci is the concentration of the ith isotope in the canal at the given concentration =easured in the tank diluted by the known canal flow rate.
Those isotopes not identified by garrn spectrometry but measured by a gross beta analysis are presumed to be Sr-90 and released on that basis. Periodic sarples of the batches are then sent to the radiological environmental contractor and analyzed for Sr-90 and Sr-89 From concentrations of Sr-90 and Sr-89 found in the batches, the total curies released of these-two isotopes is calculated and used in cal-culating the percent of applicable limit in Appendix B.
The remaining
-6 unidentified isotopes are assigned an MFC of 3 x 10 ci/clper 10 CFR 20.
Tritium released are based on average concentrations in
(
both " clean" and " dirty" waste tanks.
VII-3
D.
SOLID WASTES A total of 11,563,9h8 curies of radioactive r.aterial was shipped off site during the period covered by this report. Of the total, irradiated cobalt accounted for 11,000 curies, spent fuel 11,561,575 curies and solid radvaste 11,373 curies.
See Appendix C.
(
vn_h x
r
-(
VIII. EINIRC! GNTAL MONITORING A.
EIUIRCf2GUTAL SURVEY Environmental levels of radicactivity as found in the vicinity of the plant were ec= posed almost entirely of naturally occurring radio-active caterials. In the vicinity of the circulating water discharge canal, radicactive material of plant origin was found. These caterials occurred primarily in aquatic organiJms. The levels of radicactive materials, however, were extremely low and are of no significance to the health and safety of the organisms or the public. Further, the levels of radicactive caterial found in the resident biological ec== unity are consistent _with levels found in previous years and show no upward trend.
The environmental surveillance program includes continucus sa=pling of air for particulate and halogen activity at seven locations including background sarple locations at Traverse City and Boyne City, Michigan, about 50 miles south-southwest and 20 miles southeast of the T
plant, respective 7v, to determine increased concentrations, if any, of radicactivity of plant cricin.
In addition, film bad es and thernoluminescent dosimeters (TLD),
E placed at each of these locations plus six additional locations on the site prcperty boundary, measure direct dose in the environ =ent.
Average monthly doses at the site, inner ring and backgrcund stations are ecm-pared and any difference, at_ the 95% confidence level, is reported using standard "F" and "t" tests. The results of these dosimeter analyses are
- given in Appendix D.
While all the dosimeters record doses from natural occurring sources, the dosimeters on site can also be expected to re-ceive doses from not only the plume but direct radiatien from the plant.
'The site dosimeters showed,' en an average, 0.74 t 031mR/moabovethe 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..
Air. samples gathered continucusly and analyzed weekly at the stations shown in Appendix D showed no difference, at the 95% cenfidence level, in the level-of radioactivity measured at those stations close to 1the site and those remote from the site. Both particulate filters and
~
~
4.
VIII-1
I carbon cartridges are used to measure potential concentration of radio-active mater t ls resulting from plant operations. From the known meteorological dispersion conditions, the following maximum concentra-tions can be calculated:
Particulates (May)
(1.2 uCi/s) x (0.013) x (5 0 x 10-lk s/cm) 3
-1
= 7.8 x 10 pC1/cm3
- Halogens (March)
(1.2 pCi/s) x (1 32) x (5.0 x 10-14 s/cm3)
= 7 9 x 10-lb uci/cm3
- Reflects measured values multiplied by three.
These compare to the minimum detectable activity values and normal background concentrations as follows:
Maximum Calculated Minimum Detectable Nomal Berkground Release Concentration uCi/cm3 Activity uCi/cm3-Activity uCi/cm3 Particulate 7.8 x 10-16 1 x 10-1b 7 x 10-1h Halogen 7 9 x 10-14 2 x 10-13 Hence, the negative data obtained in the program was expected.
Also, at the Big. Rock Point Plant, daily composite condenser
- circulating water inlet and canal water discharge sa=ples are taken and analyzed for radioactive. content. In addition, a monthly composite of these sa=ples is analyzed for radioactive content.. These results are shown in Appendix D.. Additional aquatic samples are taken and analyzed during the summer growing season and these results are also. tabulated in Appendix D.
Based upon the liquid release of 2.06 curies of radioactive material (less tritium and noble gases) which results in an annual aver-l
' age concentration in the discharge canal of 2.0 x 10-0 pCi/ml, the analysis of discharge canal' water should indicate an increase of radio-active material.in. discharge' canal' water samples since the minimum detectableL activity for gross beta measurements is about 5 x 10-9 pCi/ml
-or about four times lower than the average concentration discharged.
The results shown plotted in Appendix D indicate an average of about
{
(1.2 + 0 94).x 10-8 uCi/ml for the year, which is' in close agreement with the calculated concentration.
VIII-2
I i
B.
ENVIROINENTAL DOSE CAICUIATIONS Levels of radioactive materials in environmental media indicate that public intake is well below 5% of that which could result from continuous exposure to the concentration values listed in Appen'lix B, Table II, 10 CFR Part 20.
1.
Atmospneric Releases In order to predict potential radiation doses resulting frem 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.
Currently, a computer model is used to calculate radiation dose resulting from plant releases of noble gases. The integrated pbpulation dose, out to 50 miles, for lW3 is shown on the following page. The computer model utilizes the following:
X/Q values for the five sectors are avcraged over both a.
stability class and wind frequency.
b.
Doses are calculated for each of the 22 noble gas radio-nuclides and daughter products based on individual decay energies. Total dose is then the summation of the individual nuclide contributions.
c.
The 1973 population is estimated' from the 1970 Census of 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. The -total. estimated 1973 population 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 the total release is further corrected by daily measurements of off gas.
e.' Site boundary doses are-finite cloud shine doses. Semi-infinite' cloud geometry is utilized to calculate doses after the plume. reaches. ground level.
f.-
No credit;is taken for the meandering of the plume before it reaches.the different annuli.
VIII-3 L
_. = - -
t i-I-
The maximum calculated radiation dose at the site boundary resulting from noble gas releases was 7.5 millirems. The integrated dose to the population out to 50 miles was 6.0 person-Rems.
Doses from particulate, iodine and tritium releases as shcwn in Appendix A were neFligible co= pared to that received from noble gases due to the conservative limits in the plant Technical Specifications.
2.
Liquid Releases In order to predict potential radiation doses resulting from the liquid releases, environmental transport and uptake factors must be known. A confirmation of these calculated doses is then f'
ctter:pted by measuring levels of radioactive materials in the plant's environmental radiation surveillance program.
4 The nearest municipal drinking water supply intake is
[
located in Charlevoix, Michigan, which is generally upstream of the prevailing current flow in' Lake Michigan at this location. However, since current patterns do occur that could, at times, carry the dis-
- charged water in the direction of Charlevoix, pcpulation dose based upon this flow is calculated in the next section of this report. A conservative dilution factor of 800 is taken from the point of. dis-charge to the city of Charlevoix based upon the report, " Big Rock -
Point Hydrological Survey, Great.Lahes Research Divisicn, University of Michigan, Special Report No 9," by John' C.' Ayers,1961.
. In addition, the population dose is calculated to the entire o
4 population which receives its drinking water from Lake Michigan,. based
~on a uniform concentration, resulting~ frem plant releases, throughout -
Lake Michigan. lAlso, radiation dose to human populations can occur as-t a result of plant -releases through the consumption of fish caught in.
Utilizifig the measured values of radionuclides released as
~
shown in Appendix B, the>following formula, and the.-standard man model,
- drinking water doses can be-calculated as fellows:
,l.--
VIII k
.u.
CALCUIATED RADIATICU DOSES FROM GASEOUS RELEASES January 1, 1973 to December 31, 1973 (Person-Rems)
Distance Secter (Miles) 1 2
3 4
___ 5 Total 1-2 Populatien 13 75 0
10 0
98 Population Dose 0.019 0.054 0.0 0.013 0.0 0.086 2-3 Population 264 270 0
51 73 658 Pqpulation Dose 0.24 0.14 0.0 0.048 0.067 0.50 34 Populatien 562 397 0
48 58 1,065 Pcpulation Dose 0 37 0.15 0.0 0.034 0.039 0.59 4-5 Pcpulation 722 3,344 o
103 o
4,169 Population Dose 0.21 1.7 0.0 0.057 0.0 2.0 5-10 Pcpulation 2,102 24 0
534 0
2,660 Populatien Dose 0.44 0.003 0.0 0.14 0.0 0.59 10-20 Pcpulation 8,937 395 747 14,115 327 24,571 Population Dose 0.47 0.013 0.049 0 93 0.018 1.5 20-30 Population 9,651 3,504 1,902 4,623 327 20,007 Population Dose 0.14 0.032 0.038 0.092 0.006 o.31 30 40 Population 22,775 4,081 2,916 4,847 o
34,619 Population Dose 0.14 0.016 0.025 0.043 0.0 0.22 40-50 Population 40,790 8,888 5,873 12,101 o
67,652 Population Dose 0.14 0.018 0.026 0.054 0.0 0.24 0-50 Population 85,866 20,978 11,438 36,447 785 155,409 Population Dose 2.17 2.13 0.14 1.41 0.13 6.o Site Boundary 6.5x10-3 4.2x10-3 7,5xio-3 7,gxio
-3 Dose (Ren)
(
VIII-5
(
r Ci !
(Limiting Dose Rem /Yr)
~
D
=
a l
ii where: D is the individual dose in Rem /yr, a
Ci is the Average concentration in Lake Michigan of the individual nuclides measured, in uCi/m1, MFC is the concentration of each nuclide =easured required to produce the limiting dose at continuous intake in pCi/ml and limiting dose is the dose produced at continuous expcsure to MFC concentrations.
In calculating ingestion dose from the censu=pticn of fish, an equation similar to the one used for drinking water dose is used except that a stanNrd daily diet of 50 grams of fish flesh is used in contrast to the 2,200 ml of fluid consumed daily by the standard man.
' This, in effect, alters the MPCi by 50/2,200 or 0.0227 The calculation of individual doses, both frca drinking water and censuming fish, are per the previcus for. ala while integrated pcpulation doses in man-Rem are calculated utilizing the following parameters:
a.
For drinking water, the individual doses are su:=ed over-the entire population that receives its-drinking water from Lake Michigan with discherEe canal' flow apprcpriately mixed with the lake. This is approxi=ately.10 million pecple of which approximately 7 =$ ilien reside inithe Chicago metropolitan area.
.b.
The population dose'due to drinking water to Charlevoix -
- residents is based en a populaticn.of 3,500 pecple.
c.
- For. fish consu=ption,' the average concentration in Lake.
Michigan water, resulting frem plant releases, is used with a bicaccu:mi-lation factor to determine 'the average concentration-.in fish, d.
Fish do~not' reside continuously'in the discharge canal but migrate. ThisL can be seen in the following table which compares the fish consu=ption dose based on the discharge canal water concentration
- ERG Special, Report No 2,-' " Trace Element Distrib:tiens 'in Lake Michigan
-Fish: A Baseline Study With Calculations of Cc:hentration Factors and I
Equilibrium Radioisotope Distributions," March 3573 VIII-6 a
~. -
i and the appropriate reconcentration factors to the fish consu=ptien dose calculated from actual concentraticns in fish caught in or near the dis-charge canal.
Population doses based upon drinking water from the Charlevoix municipal system were 0.05 persen-Rem and total Lake Michigan drinking water concurptien population dose was 1.0 person-Rems. The censumpticn of all of the Lake Michigan fish harvested resulted in a populaticn dose of 0 37 persen-Bem.
As a measure of total environmental impact, the radioactivt liquid releases frcn the plant are averaged over the entire lake and then used to determine the population dose frc= fish caught through-out the entire lake and total water consumed frem the lake.
Both of the dose calculations are conservative in that:
(1) Equilibrium is not obtained in the human body for most isotcpes released.
(2) No credit is taken for precipitation and dqposit in sediment or uptake by life forms other than fish which are seen to occur by the data shown in Appendix D.
(3) No credit is taken for radioactive decay which for I-131 is significant.
Results are shown in the following tables.
VIII-7
+.
CO NS UM ER S PO WF 4 CowpANY DA TE P1/18/74 R]G ROCn NUCLEAR POWER PLANT CALCUL ATED R ADI ATION 00$ES FROM L IGtJ1D ErrLUFNT 9
-POPUL A TION 3RINM ]NG Wa TE R 009E 1/1/73 TO 12/31/73 AVG CONCF NTR ATION PortiL A T I ON POCF CRITICAL CURIES IN LAKE MICHIGAN
<CI/*PCI)wPD]
00PUL A T ION 00SF CHARLEv0!n.wlCH VECTOR ISOTOPE MPC ORGAN RELE ASE D (UCl/ML1 CI/MPCI
("REw/YR)
(MAN-RFP)
(PAN-RFwt WATER ZN-65 1.00E-04 WHOLE RODY 0.000a 1.57F-16 1.57E-12 7.85E-10
- 0. 0 c r> 01 1.93r-07 WATER
~l-131 3.00E THvp0!D.
0.0504 1.24F-14 4.12F-0R
?.0er-05 0.2r^17 5.n6E "3 i
, ATER CS-134 9.00E-06 WHOLE BODY 0.2493 5.1H -14 5.77E-09 7.89E-06 c.02686 7.r*E-cd W
W A TF R CS-137 2.00E-05 WHOLF RODY 0.9197
- 1. 0 8F -13 5.41E-09 2.71r-no 0.02707 A.edF-ra W A TER ' B AL A-140 2.00E-05 G.I.~ TRACT 0.0004 8.96E-17 4.4PE-12 6.72E-09 0.00007 1.oSF-sa WATER CO-58 1.00E-06 G. I. TRACT 0.02A4 5.92r-15 5.92F-09 8.AAE-06 0.08aA4 7.1 m F-r.4 WA TE R C0-60 3.00E-05 G.1, TRACT 0.2153 4.4 9E -14 1.50E-09 2.24F-06 0.02/43 5.59E-r4 W A TER OTHERS 3.00E-n7-WHOLE BODY 0.9669 2.01F-13 6.71E-07 1.StE-04 1.81787 4.4*F-02 TOTAL WHOLE RODY 1.86880 4.58E-02
<M THYR 0!D 0.20612 5.e6E-03
.. {
G.I. TRACT 0.11134 2.73F-03 RONE 6.n c.c (1) Average concentration in take Michi
= curies released / volume of Iske Michigan DATF 01/18/74 Voluma of Imke Mi-higan 16.8 x 10-liters.
( } Based on a fluid intake of I?00 al/ day.
Population taking its drinking water fr<ns IAke Richigan is approximately 10.000,000 people with T.000,000 in the Chiesgo area.
Using average concentration in discharge canal diluted by 800.
(5)10 Crit 20 Mrc for unknown mixture with certain isotopes not present.
6 This vospares to a. background and sedical radiation done of 0.215 Rem /yr/prson or 2,15 x 10 ann-Rems for the population taking its drinking water from IAke Michip;nn.
CONSUMERS POWEH CowpANY PATF 01/18/74 R!G ROCK NU CL E A R PO WE R PL AN T CALCULATED RADIATION 00SES FROM LIQUID EFFLUENTS - FISH CONSUMPT ION DOSE 1/1/73 TO 12/31/73 AVG CONCENTR AT ION A VG CON CE NT R A TI ON CRITICAL B10 ACCUMULATION IN LAKE MICHIGAN IN rISH (CFf/MPCIIMPOI POPULATION DOSE VECTOR. ISOTOPE MPCI ORGAN FACTOR (UCI/ML)
(UC!/G) twAEw/vp3 (WAN-REw)
FISH ZN-65 4.40E-03 WHCLE RODY 900.
1 57E-16
- 1. 41 E-13 1.61E-00 0.00001 r!SH 3 131 1.32E-05 THYR 0!D 500.
1.24E-14 6.18E-12 2.34E-na 0.33013 FISH CS-134' 3.96E-04 WHOLE BODY 2360.
5.19E-14 1.23E-10 1 55E-04 0.0919M FISH CS-137 8.80E-04 WHOLE RODY 2360.
1 08E-13 2.55E-10 1.45E-04 0.08623
' FISH OALA-140 8.80E-04 G.I. TRACT 365.
R.96E-17 3.27E-14 5.58F.08 0.00003 r!SH CO-58 4.40E-05 G.I.' TRACT 330.
5.9 2E -15 1.95E-12 6.66E-95 0.03959 rtSH Co.60 1.32E-03 G.I. TRACT 330.
4.4 9E -14 1.48F-11 1.6BE n5 0.00990 FISH OTHERS 1.32E-05 WHOLF RODY 80.
2.01E-13 1 61E-11 3.30E-04 0.19570 T OT AL WHOLE RODY 6.30E-04 0.3739P d
d THvR0!D 2.34E-04 n.33913
~'
G.I.
TDACT 8.3 5E -0 5 9.04960 BO NE 0.0 0.0
( }m aim a Permissible Concentration for Fish = MPC *(2200h0).
g ERG Special Report No 2'.
Using 23,873,689 pounds of fish harvested from Lake Michism in 1970. This number includes both connercial and sports catebes as shown in Appendix D minue alevives which are not generally consumed.
5 This compares to en average background and medical radiation dose of 0.215 Rem /yr/ person or 13 x 10 mp,,
for the population necessary to consume the IAke Michigan fish catch at a rate of $0 g/ day / person.
_ _ _ _. _ =. _
? t -~
".(
Occupational Exposure 1
l (7 2.2.1.2.a(1)(h))
I t
l Nu=ber of Persons 'dithin Exposu-e Rance nRe= Dese
-7/ 2/73 - 7/29/73 7/30/73 - 8/26/73 8/27/73 - 9/23/73 0-100
- Maint 3.Oper 8 Maint 5
Oper 8 Maint 3
Oper 7 Supv 15 Tech 7 Supv 16 Tech 6 Supy 16 Tech 5 Others 12 Others 15 Others 2h 101-500
- Maint
~9 Oper 11 Maint 6
Oper 6 Maint 2
Oper k Supy h
Tech 2 Supy 3
Tech 2 Supv 3
Tech 3 Others 5 Others O Cthers 8 501-1250
- Maint 2
Oper o Maint 3
Oper 5 Maint 6
Oper 7 Supy 0
Tech 1 Supv 0
Tech 2 Supy 0
Tech -2 Others O'
Others 0 Others O f
1251-2500 Maint 1
Oper 1 Supy 0
Tech 0 Others 0 Total Nu=ber of People Badged.
79 77 92 l
mrem Dose-.
9/2h/73 - 10/28/73 10/29/73 - 11/25/73 11/26/73 - 12/30/T3 l
O-100
- Maint Oper 8 Maint 2
Oper 10 Maint 2
Oper 10 Supv 16 -Tech 5 Supy 16 Tech 5 Supv 15 Tech 2~
-Others 22 Others 30 Others 18 i
101-500
- Maint--
8' -Oper -T ?
M:. int 8 1Oper 8 Maint 8
Oper 8 4
Supv 3
Tech 3 Supy
'3~
Tech 1
.Supy 1 -Tech 7
'Others 10 Others 11
-Others 17 2
Others include office secretaries, General Office personnel, contract ' personnel, l-vendors, plant guards, information center' personnel,' Region repairmen other than
/'
from Traverse City.and visitors.
- I
'Maint includes ; Region repairmen from Traverse City.
c e4, i
~.
i e
t 4.
4 Nunber of Persons ~ Within Exposure Range (Contd) mRen Dose 9/2h/73 - 10/28/73 10/29/73 - 11/25/73 11/26/73 - 12/30/73 2
501-1250
- Maint 3
Oper 3 Maint 1
oper 10 Paint 5
Oper o Supy 0
Tech 2 Supy 0
Tech 2 Supv 1
Tech 1 Others 5 Others k Others 13 1251-2500 Maint 0
Oper o Maint 0
Oper 0 r
Supv 0
Tech 2 Supv 1
Tech 0 Others 2 Others 10 Total Nunber of People Bad ed 100 105 119 6
The nunber of persons that. received more than.2500 r. Rem during 1973 was 33 and the r.ajor causes were as follows:
1..
Lining of the spent fuel pool with stainless steel.
2.
Refueling shutdown.
4 a.
Head remval and replacement.
b.
Steam drum spool piece flange.
c.
Insulation removal and cleanup.
' d.
Weld inspection.
e.
~ Clean-up demineralizer drain valve.
f.
Changing of rod drives..
g.
Recirculating pu=p overhaul, b.- Primary leak detection system.'
- i. Limitorque motor operator adjustment.
3 Routine maintenance..
l h.
Routine plant surveillance and inspection.
~
Others include office secretaries, General office personnel'. contract personnel,
.' vendors, _ plant' guards, infomation center personnel, Region repairmen other than.
r
' from Traverse City and visitors;
- Ma'nt-includes Region repairmen from Traverse City.
.['
.}*b l
7
n X.
F11DICACTIVF IE*ETE 174 PRI?;0TPI.E F1UID SYSE'S Mini =r.
Averare Mix " -
I A.
Prinary Ccolent Eeactor Water Filtrate (a)
-2
-e
-1 nCi/ 1 1.5 x lo 95x10 2 9 x lo Reacter Water Crud (c)
_1 FCi/el/ Turbidity Unit 2.9 x lo " 1.o x 10 2.5 x 10 ^
Icdine Activity
-h
-2
-1 pCi/ 1 5.0 x lo 3 1 x lo 2 x lo E.
Peacter cooline Water Syste=
Reactor Cooling Water (a) 7
-2
-?
pCi/=1 h.h x lo ' 2 9 x lo h.h x lo -
C.
Stent ?uel Pool Puel Storage Pool (a)
-h
-h
-2 1.2 x 10 c.c x lo 2.9 x 10
-7
-6
-b Fuel Fcol Iodine 3.o x lo 9x10 2 x lo (a)A counter efficiency tasad en a decay scheme consisting cf cne ga--*
photon per disintegration at o.662 MeV used to ccnvert count rate to microctries. All count rates were taken two hcurs after sa=pling.
(k) Eased on efficiency of Iodine 131 two hours after ss=pling.
~
Eased en APHA turbidity units and 500 =1 of filtered sa=ple.
l X-1
.__._______._J
/
~'
ATTACEGTI A Section II D, Page 15, of the Eighteenth Semiannual Report should be corrected to read as follows:
SYSTai CALIBRATION OF CANAL LIQUID PROCESS MONITOR RESULTS The calibration of the canal liquid process monitor is a comparative calibration used to demonstrate operations of the monitor and to-detect gross calibration changes and/or instrument drift. The results of these monthly calibrations ehowed that a monitor drift has occurred since the last calibration which utilized certified standards.
Recalibration of the monitor with liquid standard sources will be completed shortly. Also, an acceptance criteria for all process moni-tor calibrations will be developed.
In Appendix A, Table I, of the Eighteenth Semiannual Report, the figures listed in the percent of Technical Specifications limits for particulates should be corrected to read as indicated in Appendix A, Table I, of this report.
In Appendix B, Table I, of this report, the figures denoted with an asterisk (*) are corrections to Appendix B of the Eighteenth Semiannual Report.
':\\
1 4
AITENDIX A, TABIE I CONMTR3 PMR COMPANY Big Fock Poir.t Flant, tracket !.o 50-155 Atmospheric pleases of Radioactive Material Six-Month Units January February Ma rch April May June Total Total hoble Cases Curies 47,100 45,300 6,280 905 7,530 7,130 11b,000
- Total Halogens 0.15 0.045 4.2 0.0k8 0.0L2 3.0 x 10 4.5 Total Partiestates (8,v) 0.016 2.9 x 10~3 0.098 7.8 x 10 0.23 6.4 x 10' O.34 Total Tritium 7.4 6.9 1.2 39 7.4 7.4 34 Total Particulate - Cross Alpha 3.9 x 10" 1.2 x 10*I 1.7 x 10*I 1.2 x 10' 1.1 x IO~I 1.3 x 10"I 1.0 x 10 3
3 Maximum Noble Gas Release Rate uCI/s 2.1 x 10 2.2 x 10 1.8 x 10 1.4 x 10 1.2 x 10 4.6 x 10 2.2 x 10 Percent of Teah spee Limits for:
Noble Gases 1.8 1.7 0.24 0.03 0.29 0.27 0.72
- Halogens 4.8 1.4 132 1.5 1.3 0.006 24 Particulate +'
O.30 0.004 0.61 0.025 1.3 1.3 0.59 Isotopes Releasemi:
Curies Particulates 3
~3
~g BaIA-lho 7.8 x 10~4 2.9 x 10 1.3 x 10*3 0.2 4.6 x 10 0.24 Cs-134/137 4.1 x 10' 1.1x10g 5.3 x 10 0 Mn-54 0.092
~
1.8 x 10'k*
8.8x10~g 0.0 Net Unidentified Peta 7.8 x 10-3 1.5 x 10*J 7.8 x 10'b Halogens a
k.
- I-131 0.15 0.h5 4.2 O.048 0.0h2 1.9 x 10 1.4 x 10-4 3 9 x 10~7 2.7x10g I-133 9.3 x 10-5 I-135 Gases Xe-19 10,300 9,220 1,450 255 1,820 1,270 24,000 Kr-87 6,2%
6,320 991 94 1,110 1,080 16,000 Kr-88 3,k70 4,920 6h9 41 644 629 10,0m Kr 85m 2,270 2,430 361 18 ya 319 5,800 Xe-135 S,380 9,810 1.h60 68 1,t>80 1,7Ho 23,200 Xe-133 4,350 5,260 813 7.3 1,160 985 12,600 Xe-lh3 0
0 0
0 0
0 0
Kr-94 0
0 0
0 0
0 0
h-93 0
0 0
0 0
0 0
Xe-141 0
0 0
0 0
0 0
Kr-92 0
0 0
0 0
0 0
Kr-91 1.8
<1 0
<1 0
0 4
Xe-140 24 10
<1 3.7
<1 0
ho Kr-90 272 120 2.0 28 7.2 1.1 4 30 Xe-139 404 1R0 3.2 38 11 13 650 Kr $)
1,830 884 32 81 65 20 2,900 Xe-137 4,200 2,070 80 171 156 105 6,800 Xe-135m 3,8A0 2,220 151 85 213 247 6,800 Kr-83n 1,350 1,1 %
139 IL 15?
272 3,100 Xe-133m 109 M5 61 71 56 174 350 Xe-131m h.5 49 11 0
10 10 8'i F1-85 3.6 342 8?
O 74 t'i9 570 N-13 3h7 31?
19 171 h23 L67 1,740
- Reflects Hessured values naltiplied by Three
- Corrected From Eighteenth Semiannual Report
6
- 6. N N 0000VM00000ni0uu0o00)W0 122 l
0 7
7 0
0 M.3 1
713 0000 0t 1
0, 1,
7 o
0 0.100 h 0. O.
5,1,17,k 6, i 94 l, 9'
1 a
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k 1
t, te a,1,o,6,01 t
5, le 0
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0 0 01 oT b
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00 94124$
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)
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APPENDIX B. TABTE I Consumers Power Cmpany Big Rock Point Plant, Docket No 50-155 Radioactive Liquid Ecleases Units January February Msrch April May June
_ Total
' Total Radioactivity Released (Except Tritium, Dissolved Gases and Alpha)
Curies 0.05 0.053 0.93*
0 78*
1.81*
5 5
3 72 x 10 Volume of Waste Discharge Liters 1.9 x 10 1.65 x 10 1.92 x 10 8.97 x 10 Average Concentration of
~g
~3 3 52 x 10-3 h.91 x 10-3 g,9,yo-3, Waste Prior to Discharge uci/ml 2.64 x 10 3.19 x 10 Volume of Circulating 9
9 9
n 9
10 9
Discharge Water I.iters 8.79 x 10 7.9 x 10 8.45 x 10 8.50 x 10 8.79 x IC '
8.79 x 10 5.12 x 10 Average Concentration Released (Except Tritium, Dissolved
-9 Gases and Alpha) pC1/mi 5.7 x 10-9 6.61 x 10 8.01 x 10~
5.18 x 10~
3.5 x 10
- Maximum Concentration (Except Tritium, Dissolved
~7 Gases and Alpha) pC1/mi 1.95 x 10~7 1.11 x 10 9.68 x 10~7 6.8 x 10-7 9.68 x 10-7 Percent of Applicable Limits:
1.26 0.88 17.0 2.0*
3.5+
Tritium Released Curies b.2 x 10' O.215 0 36 3.75 4.37 Average Tritium Concentration a
4
~I
~g Released pCi/ml 4.8 x 10~9 2.7 x 10 h.3 x 10 4.h x 10 8.5 x 10
. Total Gross Alpha Released Curies 1.2 x 10 3.6 x 10 2.58 x 10~
3 9 x 10 Avirage Alpha Concentration pC1/mi 1.5 x 10 4.22 x 10 3 03 x 10~
7.6 x 10~12
~I Isotopes Culte s 3
4.C x 10-2, h.1 t 10~g 3.4 x 10~2 3,g, 1g-2, 9,9,in Cs-134 3 7 x 10-3 1 9 x 10-3.6 x 10 2*
-2 3,g, in-1, I-131 Cs 137 91 x 10~3 1.1 x 10-2 5.8 x 10-2*
2.1 x 10~1 3 0 x 10~1*
2 9 x 10-2, Co 58
-3 1.0 x 10'l 2.7 x 10~p g,9, 19~
Co-60 4.4 x 10-3 2.5 x 10 1 3 x 10,,
Mn-54 1.6xloj*
1.6 x 10 ;*
~
1.6 x If 1.2 x 10"2 2.8 x 10' Ce-14k 2.6 x 10~1 2.6 x 10~1 Cr-51 4.6 x 10~
L.6xloj Sr-89 Sr-90 1.1 x 10 1.1 x 10 Total Identified Released.
~
3 6 x 10 5.5 x 10 3.4 x 10~
9.4 x 10
~1
~1 Radioactivity 1.8 x 10 Percent of Total Identified 36 68 59*
44*
52*
- Correctiens Made to Eighteenth Semiannual Report
n
,. ~
m.
APPENDIX B. TABLE II ColCtNERS POWER COMPANY
-~
Big Rock Point Plant, Docket No 50-155 Radioactive Liquid Releases Six-Month Unite July August Sept aber october Nov mbe n December Total Year Total
%tal Radioactivity Released
~ (Racept Trititsu, Dissolved cases and Alpha)
Curies o.35 0.09 0.033 0.084 0.059 0.22 0.84 2.65 k
5 k
5 Volume of Waste Discharged Liters 2 5 x 10 2.0 x lo" 2.0 x lo 2.0 x 10 3,9,to 8.3 m lo" 4.3 x 10 8.0 x lo)
Average Concentration of waste Prior to Discharge pci/ml 1.4 x 10-2 4.5 x 10-3
-3
,g
-3 2.6 x 10-3 2.o x lo'3 2.6 x lo'3 ~
1,7, to 3,7,to 1,5 x 1o Volme'of Circulating Disenarse Water Liter 8.79 x 10 8.79 x 10 8.5 x 10 8.75 x 10 8.5 x lo 8.6 x 109 1o 11 9
9 9
9 9
5.2 x 10 1.03 x 10 Average Concentration Released (Except Trititas, dissolved cases
,g
- 9 4
. and Alpha)
-pC1/a1 k.o x lo'g 1.0 x 10 39xlo 8.5 x 10-9 6.8 x 10-9
,3 1.6 x 10,g 2.o x 1o 2.5 x 10 Mm=4 - Concentration (Except Tritium, Dissolved cases and Alpha) uct/mi 1.6 x lo 5.7 x lo*7 3.6 x 10-7 6 o x 10~7 3 9 x lo'I 5.2 x lo'7 1.6 x 10-6 1.6 x 10'0 Percent of Applicable Limits 5.o 15 0.3 17 19 5.6 2.7 31 Tritium Released Ctaries 5.5 x lo
4.3 x 10-2 4.4 x lo'#
12 1 1.k 5 14k 15 3 19.7 Average Trititan Concentration
-6 Released wC1/ml 6.3 x 10'9 4.9 x lo'9 5.2 x 10'9 1.4 x 10 1,.g,to-7 1,7,yo-7 g,7,3o-7 1,9, io-T
-5
-5
-5 4
-5
-4
'!btal oroes Alpha Released Curies 1.7 x 10 1.4 x 10-5 6.8 x 10 1.8 x 10 3.8 x 10 2.3 x 10 1,g,to 5 3 x 10' Average Alpha Concentration Released pC1/ml 1.9 x 10-12 1.6 x 10 8.o x 10-12 2.1 x 10-12 4.4 x lo'13 2.7 x 10-12 2.7 x 10-12 5 1 x 10-12
-12 Isotopes Curies I-131 E
2.1 x lo 4 2.k u 10'3 2.3 x 10 6.3 x lo 2
-2
-2 2
-2
-1 C2-134 5.5 x lo'1 1.6 x lo 5.0 x lo 3.o x 10'3 8.2 x 10 1.8 x 10 2 9.8 x 10 2.4-x 10 Ca-137 1.4 x 10' 2.2 x 10 1.8 x 10 3,g,10-g,g,to-3 4.1 x 10-p,3,yo-1
$,1, 1o-1
-2
-2 0
-2 9.hx10'f 9.4 x 10 2 9 x lo Cc-58 3.9 x lo' 1 3 x 10' 8.2 x 10-2 2.1 x 10~2 1
Co-60 2.6 x lo-1 3 x lo" 2.9 x 10 2.3 x lo-Mb-54 9.8xlo'j 9.o x 10' 1.1xloj' 2.7 x lo 2 2.o x 10' 2.0 x lo 2.8 x 10-C5-14%
Cr-53 2.6 x 10'1 k.6 x lo Sr-89 BmLa.lbo k.3 x lo"g 4
1.1 x 10~g Sr-90 4
k.3 x lo 4.3 x lo 7.5 x to 7.5 x 10'g 7.5 x 10'g En #j Total Identified Radioactivity Beleased 0.22 5.1 x 10-2 2.6 x 10 g,1,yo-2 3 2 x 10-2 7,g,1o g,g,yo-1 1.38
-2
-2 Percent of Total Identified 63 57 78 52 53 3h 53 52 Dircolved Noble Cases 1.5 x 10 Ke-133
-3 1,p,to-2 1,3, yo 1,3, to-2
-2
-3
-3
-3 Xs-135 k.3 x 10 g,3, 1o g,3, 10
-/
APPENDIX C Off c'. Le Shipment of Radioactive Material 1
Shipment Transfer No' Date From Transfer To Radioactive Material Disposition 531 7/2/73 DPR-6 GE-Val, CA SNM-960 6 Irradiated Fuel Examination Amend 71-20 Rods, 25,500 Ci 332 7/ 9/73 DPR-6 GE-Val, CA SNM-960 8 Irradiated Fuel Examination Amend 71-20 Rods, 29,500 Ci 333 7/12/73 DPR-6 NFS, CFS-1 NY 8 Irradiated Puel Reprocessing Bundles, 1,528,000 Ci 334 7/16/73 DPR-6 GE-Val, CA SNM-960 7 Irradiated Fuel Examination Amend 71-20 Rods, 30,000 Ci
- 335 7/23/73 DPR-6 GE-Val, CA SNM-960 4 Irradiated Fuel Examination Amend 71-20 Rods, 5,500 Ci 336
.7/30/73' DPR-6 NFS, CFS-1, NY 9 Irradiated Fuel Reprocessing Bundles, 3,882,462 Ci 337:
8/ 3/73 DPR-6 NECO,'16-NSF-1 11 Fuel Support Burial Morehead,=Ky Tube P'. -;ce s, 3 Fuel Ch3nnel Pieces, 166 Ci 338 8/7/73 DPR-6~
'NECO,16-NSF-1 4 Fuel-Support.
Burial-Tube Pieces, 7
~Puel Channel Pieces, 296 Ci
'339 8/9/73 -DPR-6 NECO, 16-NSF-1 7 Fuel Support Burial.
Tube Pieces, 10 Fuel Channel Pieces, 428 Ci
-340 8/13/73 LDPR-6 NECO, 16-NSF-1
' 6 Fuel ' Support
. Burial Tube Pieces,10 Fuel Channel Pieces, 504-Ci 341' 8/15/73_
DPR-6 NFS', CFS-1, NY 6 Irradiated Fuel Reprocessing.
s Bundles, 2,073,413
-(
Ci' C-1 a.t-
I Shipment Transfer No Date From Transfer To Radioactive Material Disposition 3h2 8/15/73 DPR-6 ECO, 16-NSF-1 7 Fuel Support Tube Burial Morehead, Ky Pieces, h Fuel Chan-nel Pieces, 188 Ci 343 8/18/73 DPR-6 NPI, 19-12667-04 1 Irradiated Cobalt Reuse Dickerson, Md Rod, 11,000 Ci 3h4 8/17/73 DPR-6 ECO, 16-NSF-1 3 Fuel Support Tube Burial Morehead, Ky Pieces, 9 Fuel Chan-nel Pieces, 372 Ci 3h5 8/20/73 DPR-6 ECO, 16-NSF-1 12 Fuel Support Tube Burial l
Morehead, Ky Pieces, 12 Fuel Chan-nel Pieces, 528 Ci 3h6 8/22/73 DPR-6 NECO, 16 NSF-1 18 Orifices, 21.
Burial Morehead, Ky Transition Pieces, 780 Ci 3h7 8/2h/73. DB-6 NECO,-16-NSF-1 8 Transition Pieces, Burial
.Morehead, Ky 12 Orifices, Misc,.
j --
.L 7h2 Ci 3h8 8/28/73 DPR-6 NECO, 16 NSF-1 11 Pieces of In-Burial Morehead, Ky Core, 1 Fuel Chan-nel Pieee,-Vacuum 3
Tank, Misc, 555 Ci 349 8/30/73
'DPR-6 NFS, CFS-1, NY 9 Irradiated Fuel Reprocessing Bundles, 3,987,200 j '.
Ci 1
.350
.8/31/73 DFR-6
.NECO, 16 NSF-1 4 Neutron Windows, Burial Morehead, Ky-
- 2 Fuel Rod Storage Cans, 2 Neutron Scurce Sheaths, i
5,000 Ci 351 9/13/73 DPR -NECO, 16 NSF-1 1 Control Rod Blade, Burial.
Morehead, Ky 12 Fuel Pool Filter Socks, 1,766 Ci 352 10/24/73 DPR-6 NECO, 16 NSF-1 8 -~55 Gal Barrels, Burial -
Morehead,-KyJ 48 Ci f.
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-i APPEiDIX D (Contd)
Difference in Average TLD Readings mR/ month Site vs Background Inner Ring vs Background Month Station Station January ND ND February ND ND Phrch ND ND April 2 94 I 2.19 ND
-May 3 43 t 1.85 ND
(
June ND ND (3) July (3) August (3)septenber October 25 1 2.4 ND November ND ND December ND ND Average 0 74 1 0 31 ND (1)ND: No difference at the 95% confidence level
(
Control TLD's were not shipped with others, statistical analysis based on total dose including dose received in shipment (usually 12 to 14 mR)
(3}TLD's removed from field for recalibration for.inberent background,.
environmental film badges show no difference between site, back-ground and inner ring locations.
k m.
.. = = =.
?
t APPENDIX D (Contd)
Big Rock Point Plant Aquatic Biota October 1973 65,
60 kO 34,, 58 137 g
Co K
c c
Co Cs Minnovs Discharge
.16 0.02 1/4 Mile South
<0 1/4 Mile North
.48 56 0.005 Mt McSauba
.Nine Mile Point' Crayfish Discharge
.22
.08
.08
.26 0.22 1/4 Mile South
.28
.06
.05 0.11 1/4 Mile North
.10
.09
.27 0.08 Mt McSauba
.08
.08
.42 0.03 Nine Mile Point
.05
.03 33 0.04 Periphyton
~
Discharge
. 44
.28
.40 2 70 Discharge
.63 38
.43 2.15 1/k Mile South
.48
.13
.09 0.43 1/4 Mile North 56-
.16
.14
.29 0 703 Mt McSauba'
.10
.05 52 0.11 Nine Mile Point
.19
.06
.81 0.167
- Algae-Discharge
.14
.08
.06 0.11 1/4 Mile South
'1/4 Mile North Mt McSauba
.07
.64 0.06
'Nine Mile Point
. 61 0.02 Sm11[ Mouth Bass Discharge
.04
. 07 0.06
}.-
APPENDIX D (Contd)
Sampling and Analysis Sumnry Number of Samples Frequency of Medium Desc ription___
_ Location Collected Type of Analysis Analysis Air Continuous at All 326 Gross Beta, 13 Weekly Approxinately 1 cfm Take Water.
1 Gal Grab ST 24 Gross Beta, Gross Gaw n Monthly 3
37Cs, NMn, Quarterly Sr, Cs, Co, Co, Zn, Fe Well Water 1 Gal Grab St 12 Gross Beta Monthly Gamma-Dose Continuous All 470 Flim Dose Monthly 161 TLD Dose Aquatic Biota. Grab St,IIM, 30 Gross Beta, Gross Gamma Semiannual Mt McSauba Spectrum
3 APPENDIX D-(Contd)
.High, le.i and Average Concentrations For Hig est Average Sampling Location Type Type of Analysis Location High Low Average Air Gross Beta-Gamma TC
.08
.01
.03 I-131 All
< 0.2 40.2 40.2 Lake Water Gross Beta BR ST LWo 64 35 19
. Gross Gan=na BR ST LWO 120 (6
.23
.88
-Cs-137 BR ST LWO 62 13 27 Well Water.
Gross Beta BR ST WW 8.7 0.8 3 05
- TLD Dose' E
11.1 0.4 54 4
4
- In excess of control dosimeter.
AFFENDIX D (Contd)
- Michigan Department of Nstural Resources Fisheries Division Lar. sing, Michigan Sports Catch, Lake Michigan and Anadromous Streams, 1970 Species Number Caught Estimated Total eight (lbs)
Perch 1,700,000 283,333 Walleye 69,000 207,000 Bass 246,000 492,000 Panfish 1,300,000 260,000 Northern Pike 146,000 292,000 Suckers 482,000 1,446,000 Smelt 2,800,000 280,000 Lake Trout 2h5,000 1,715,000 Rainbow Trout 285,000 1,h25,000 Brown Trout 168,000 840,000 Brook Trout 125,000 250',000 Crho Salmon 534,000 5,340,000 Chinook Salmon 180,000 2,700,000 Other Species 368,000 368,000 Total -
8,648,000 15,898,333
- Unpublished 1970 data from postcard census program of the Michigan Department of Natural Resources, Fisheries Division.
l i,,
.,,..,.... i
APPENDIX D (Contd)
- Wichigan Department of Natural Resources Fisheries Division Lansing, Michigan COWERCIAL CATCH, IAKE MICKIGAN - - 1970 Estimated Total Species Weight (lbs)
Alewives 5,981,415 Bullheads 610 Burbet 51,261 Carp 2,394 Chubs 4,028,3h0 Herring 676 Lake Trout 89,939 Menominee 161,987 Perch 22 Pike 65 Rock Bass 35 Sauger 1
Sheepshead 12 Smelt 1,700,365 Suckers 521,807 Walleyes 7
White Bass 1
Whitefish 1,417,834 13,956,771 Taken from: GREAT IAKES FISHERIES l
1970 data taken frca December issue of Michigan, Ohio and Wisconsin Iandings.
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