ML20125C543
| ML20125C543 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 12/31/1975 |
| From: | NORTHERN STATES POWER CO. |
| To: | |
| Shared Package | |
| ML20125C541 | List: |
| References | |
| NUDOCS 9212110228 | |
| Download: ML20125C543 (78) | |
Text
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NORTilERN STATES POWER COMPANY MONTICELLO NUCLEAR GENERATING PLANT l
DOCKET NO. 50-263 LICENSE NO. DPR-22 i
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i s- REPORT TO UNITED STATES NUCLEAR REGULATORY COMMISSION
, k. SD41-ANNUAL OPERATING REPORT NO. 10 JULY 1, 1975 TO DECDiBER 31, 1975 t-t.
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9212110228 760301 J PDR ADOCK 05000263 t R PDR
t TABLE OF C0f(flNIS I. OPERATIONS SUFfMRY A. Chronological llistory B. Clanges in Plant Design C. Performance Characteristics D. Procedure Changes E. Results of' Surveillance Tests and Inspections F. Containment Leak Rate Tests G. Changes, Tests G Experiments Requiring hHC Authorization
- 11. Changes in Plant Operating Organization I. Occupational Personnel Radiation Exposure J. Relief Valve Operation Sumary
', II. PCWER OPERATION AND SilUTD0h?6 A. Power Generation Statistics B. Slutdowns III. MAINTENANCE SUhf4ARY IV. ClWGES, TESTS AND EXPERIMERTS V. RADI0 ACTIVE EFFLUENT RELEASES A. Gaseous Effluents
- 1. Gross Radioactivity Releases
- 2. Iodine Releases i< 3. Particulate Releases B. Liquid Effluents i U C. Solid Waste r-
- VI. RADIATION EhVIRON41STAL MONITORIl0 l
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I. OPERATION 5 SLM RRY A. Chronological History f
7/1/75 Operated at 83% of rated power. Power was ad-to ministratively limited to minimize background l 7/4/75 radiation and equii nent contamination levels l
within the plant.
On 7/3/75 during a routine surveillance test the A channel reactor high pressure scram switch PS i
2-3-55A, would not trip due to an improperly ad-justed bourdon tube stop plate. The stop plate l wasadjusted.
3 a 7/5/75 Power was reduced to 70% of rated for control rod exercise testing. Upon completion of the
" testing, power was increased to 83% of rated.
a 7/6/75 Operated at 83% of rated power.
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'/7/75 Power was briefly reduced to 82% of rated when a singic control rod inadvertently scramed during i
a surveillancetest due to a faulty scram solenoid l fuse. The fuse was replaced and the rod was with-l drawn.
7/8/75 Operated at 83% of rated power, to 7/10/75 1
7/11/75 Power was decreased to 45% of rated for control to rod sequence interchange. Power was increased 7/17/75 gradually following tie interchange achieving 70%
j of rated by 7/17/75.
On 7/13/75 a small Icak developed on a local pres-
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sure gauge monitoring reactor pressure. he Icak lowered the water level in the reference leg iJr the reactor water IcVel instruments connected to b the same process tap. 'As a consequence, these instmments indicated an abnomally high water r icvel. The root valve for the pressure gauge was
,. closed. The reference leg refilled and water i level indication returned to noma 1.
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7/18/75 to Operated at 70% of rated power.
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v 7/25/75 Power was reduced to 60% of rated for control rod exercise testing. Upon completion of the testing power was increased to 71% of rated, f
7/26/75 to Operated at 71% of rated power.
7/27/75
- On 7/27/75 the "A" FJR torus cooling injection valve (hD-2008) motor failed. The stem clamp I set screws had failed, allowing the stem to rotate. !
This prevented the travel limit switches from deenergizing the motor at the end of the valve i stroke, and caused the motor to overheat. Larger l set screws were installed, and the linit switches were recalibrated. ,
i 7/28/75 Power was decreased to 57% of rated to remove #12
- to recirculation system motor generator (M G) set from ?
7/29/75 service to repair the generator brushes. Power was i increased to 65% of rated while #12 M G set was repaired. Power was decreased to 52% of rated to i
return #12 M G set to operation. Power was then ,
increased to 69% of rated.
7/30/75 Power was reduced to 66% of rated to minimize back- l j ground radiation and equipment contamination IcVels within the plant.
1 7/31/75 Operated at 66% - 63% of rated. power except for j to brief weekly reductions to 55% of rated to perform ,
control rod exercise testing.
8/26/75 i
On 8/18/75 a threaded nipple in a section of 1/2" I seal water line between #12 Rhul pump and seal water I heat-exchanger failed at the thread root. The affected parts were replaced and design. work was i initiated to change from threaded pipe to tubing. <
l On 8/24/75 the "A" RfR torus cooling injection valve i (bO-2008) failed to close completely due to an f impro,erly adjusted torque switch. The torque switchwasreadjusted.
On 8/26/75 control circuit for valve hD-2036, (
steam supply to HPCI Turbine, was found deenergized.
l Investigation revealed that the motor control unit t undervoltage relay coil had failed. The relay {
j coil was replaced.
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8/27/75 A power spike to 68% of rated occurred due to an increase in recirculation flow, caused by a worn bearing on the recirculation FG set scoop tube i positioner. The positioner was locked in position.
Reactor power was returned to 64% of rated.
l 8/28/75
! to Operated at 64% of rated power.
8/30/75 i 8/31/75 Power was raluced to 55% of rated for control rod
! exercise testing. Upon completion of the testing power was increased to 62% of rated.
A reactor scram occurred due to low reactor water level. The "C" condensato demineralizer effluent valve failed to open when placed in service. hhen the effluent valve failed, the condensate dcznineral-izer circuitry drove the other effluent valves closed in an attempt to balance flow. The demineralizer systan bypass valve failed to open in time to prevent
- a low pressure trip of the reactor feodwater pumps.
loss of feedwater resulted in the reactor low water level scram.
9/1/75 Power was returned to 62% of rated. From 9/1/75 to to 9/3/75 control rod adjustments were accomplished to 9/3/75 minimize background radiation and equipment contamina-tion IcVels within the plant.As a result, reactor power was reduced to 57% of rated.
9/4/75 to Operated at 55-57% of rated power.
9/10/75 f
l' 9/11/75 Began scheduled outage to refuel the reactor and
. perfonn miscellaneous plant inspections, modifica-tions and maintenance.
In addition to refueling, major items accomplished t' during the outage (9/11/75 to 11/18/75) included q the following:
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- a. Local Icak rate tests o'f selected primary g containment isolation valves,
- b. In-service inspection activities.
L c. Surveillance tests and inspections.
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9/11/75 d. CRD maintenance including dye penetrant test of i (Cont 'd. ) collet housings.
l e. Shock suppressor maintenance. f j f. Replacement of #12 reactor feedwater pump rotating assembly, i 4
l g. Installation of "G" Torus vacuum breaker. .
. h. Repair and repacking of miscellaneous valves.
i j i. The 4" recirculation bypass lines were changed 4
to 304L material. ,
! j. Installation of a 10" vent line to the con-
- densate demineralizer backwash receiving tank.
- k. Removal of the 2" reactor drain valves.
1.' Replacement of a 6' section of turbine crossaround ,
, piping.
- m. Inspection and maintenance of turbine controls '
and valves.
- n. Replacement of 18 LPRM strings. ;
- o. Safety /ReliefValvemaintenance.
- p. Replacanent of Inboard Steam Line Drain Isola-tion valve M0-2373.
- q. Additional sealing of reactor and cable spreading room electrical penetrations, i
- r. Modifications to provide loss of valve position indication when valve motor overloads trip: ,
- s. Replacement of #12 RhCU pump seal water lines tdth stainless steel tubing.
! t. Completion of modified off-gas system design changes.
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- u. Installation of instrumentation in drywell and i torus for future relief valve actuation tests..
- v. Installation of 16 RTD wells in the torus wall.
- w. Installation of. shock suppressors on the turbine ,
cross around relief valve discharge piping.
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- x. Installation of circulating water pipe break de- i
, tection switches and alarms. s
- y. Installation of restraint on _ condensate water line-to upgrade section of line to seismic class I.. j
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s 9/14/75 A hydraulic slock suppressor located on the B loop LPCI line was found low on oil due to an improperly installed seal. The suppressor was replaced. Sub-sequent tests verified t ut the suppressor was still capable of perfoming its intended function.
9/21/75 During refueling core alterations the count rate of an SD1 decreased below 3 cps. Fuel was reloaded near the detector and the 541 count rate increased.
An administrative limit was established above 3 cps.
, 9/25/75 A crack was observed in the heat affected zone of the junction box to sparger weld on one of the four fecdwater spargers. This probi m had been antici-pated because similar problems had been ebserved at t
other BWR facilities. A contingency plan for replace-ment of all four spargers was activated.
4 10/5/75 Core reloading was completed. Preparations for
. Feedwater sparger replacement began.
10/12/75 Indications of cracking on the inner blend radii of the four reactor vessel feedwater nozzles was re-
- vealed by liquid penetrant examination. The crack-i ing was caused by Icakage of feedwater past the sparger themal sleeve, resulting in a thermal stress at the blend radii. The machined interferrence fit of the replacement themal sleeves is expected to minimize this phenomena.
10/15/75 During an operability test, the #2 starting system for the #11 emergency diesel generator failed due to a loose fitting on the control air line to the air l' start motors. The fitting was tightened and start system operability was demonstrated.
10/21/75 Completed out-of-core fuel sipping of the 268 discharged 7 x 7 fuel assemblics.
l' b i 10/26/75 Grinding to remove all crack indications from the p feedwater nozzles was completed.
U l 11/1/75 Completed replacement of feedwater spargers.
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L A crack was_ditcovered in a 1" inlet nipple to the B 1 l loop LPCI relief valve, RV 2005. The relief valve discharge line and. inlet connection were modified to ,
j ( reduce stress.
l ; 11/12/75 Completed the primary system leakage ter.t.
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During the primary containment integrated leak
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11/13/75
- rate test, No. 3 TIP ball valve was found partially open. 'Ibe valve was tapped and it closed fully.
Investigation revealed that a water-type limit switch stuck preventing the valve from closing.
The valve was replaced.
11/15/75 Completed the primary containment integrated leak rate test.
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11/16/75 Established initial criticality for Cycle 5.
11/16/75 The #12 diesel generator tripped on overspeed during j-a fast start test due to improper adjustment of a governor cam. The governor cam was adjusted and fast start capability was demonstrated. '
During routine surveillance of the APR System, it was found that the Auto Blow Down Time Delay Relay , ,
for the "A" logic Channel failed to energize in the F required time interval. The time delay was reset to allow greater margin for drift.
11/17/75 Conducted operator training, lleatup ccrnmenced and to HPCI and relief valve testing was completed.
11/18/75 On 11/18/75 the RCIC Turbine, failed to start when ,-
given a fast start signal ~due to the steam control valve sticking closed. The valve stem and ?ivots k were polished and lubricated and RCIC operability ,
was demonstrated. i
- l, 11/19/75 Generator on line and reactor power increased to t 20% of rated. [.
11/20/75 Reactor power was reduced to 8% of rated and the gen-erator was taken off the line to perform a turbine f overspeed trip test. The generator was put back on u
, line and power was increased to 50% of rated.
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11/21/75 p to 11/27/75 Power was gradually increased to 93% of rated. [
li 11/28/75 Power was reducal to 74% of rated to allow control h rod repositioning for flux shaping, ,
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!i t I 11/29/75 Power was increased to 91% of rated.
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d 11/30/75 Power was increased to 94% of rated. Power was re-duced to 54% of rated following trip of both off-gas reembiners initiated by a high off-gas flow rate trip. (The flow rate trip was believed to have been initiated by an intennittant control system mal-function, ibwever, on 12/7/75 the true cause was found to be water buildup due to a plugged drain line).
The rec mbiner system was returned to service and power was increaed to 96% of rated.
i 12/1/75 lbwer was increased to 100% of rated.
12/2/75 Power was reduced to 48% of rated following trip of
' both off gas rec mbiners initiated by a high off gas flow rate trip. (The flow rate trip was believed to have been initiated by en intermittant control system malfunction. However, on 12/7/75 the tnie cause was founi to be water buildup due to a plugged line.) The recombiner system was returned to service and power was increased to 96% of rated.
12/3/75 Power was increased to 100% of rated.
12/4/75 to Operated at 100% of rated power.
12/5/75
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, 12/6/75 Power was reduced to 68% of rated for control rod i exercise testing. Upon completion of the testing power was returned to 100% of rated, p.
( 12/7/75 Power was reduced to 50% of rated following trip of both off gas recombiners initiated by a high off-gas
[ flow rate trip. The recombiner system was returned to service and power was increased to 91% of rated.
The trip was caused by pressure oscillations re-sulting from water in the recombiner inlet delay 3 line, i
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The rec mbiner inlet delay line drain line was found
(' to be plugged. The line was opened by pressurizing
- i. the drain tank. No further high off gas flow trips were experienced during the remainder of the report-g- ing period.
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, 12/8/75 Operated at 100% of rated power except for brief weekly I to reductions for control rod exercise testing.
12/26/75 i
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- On12/22/75 the p pe clamp for a Hydraulic Six>ck j Suppressor locat outside primary containment on >
l #12 Core Spray discharge line was found loose '
- and improperly oriented. The pipe clamp was ad-
! justed ani tightened. ,
ri L 12/27/75 Power was reduced to 68% of rated for control rod
l l exercise testing and make Control rod position ad-justments which were expected to increase the margins i to core thennal limits. Upon completion of the test- b l ing power was increased to 96% of rated.
l 12/28/75 :i[
i to Operated at 96-98% of rated power. Control rod i 12/31/75 position adjustments made on 12/27/75 were found y' j to be only partially successful. Power was adminis-
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- tratively limited to assure operation within core-
! thennal limits.
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B.' Changes in Plant Design Changes to the facility as described in the FSAR and other submittals, are included in section IV of this report in accordance with the reouirenents of 10CFR 50:59 (b). These changes include the following:
- 1. A balance damper ard diversion duct was installed in the condenser area discharge duct of V-MZ-5 to facilitate control of airborne activity.
- 2. The reactor vessel bottom head drain line and associated valves were renoved to eliminate a radioactive crud trap.
- 3. The condensate demineralizer backwash receiving tank vent line was modif-
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ied to discharge to the main stack via the 1.75 minute delay line. This modification minimizes the possibility of airborne activity from the vent line leaking into the reactor building vent plenum.
- 4. The Montiello segmented Test Rod Bundle was modified per Addendum I to NEDE - 2019 as subnitted in a letter from Mr. L. O. Mayer to Mr. A.
Giambusso, Director of Reactor Licensing, dated August 20, 1975.
- 5. During the fall 1975 refuelinc outage 268 initial core 7 x 7 fuel assem-blies were replaced with 8 x 8 fuel assemblies enriched to 2.19 wtt U-235.
- 6. A modified torus hatch cover was installed to provide penetrations for instrumentation associated with a special " torus response to relief valve actuation" test.
- 7. As. indicated in section IV.3 of the previous semi-annual report (no. 9),
an air operated isolation valve was installed on the instrument air line to the torus.
Numerous changes were made to the Modified Off Gas System prior to and during the preoperational and startup test program. Due to the number of changes and their sometimes temporary, experimental, overlapping, or superceding nature, it was felt to be practical and prudent to complete the test program prior to reporting any changes. The test program was completed following the fall 1975 refueling outage. Changes to the system are:
- 1. Since adequate redundant charcoal filters were located upstream,the HEPA Stack filter was not replaced with a combined charcoal and HEPA filter, thereby eliminating unnecessary modification, testing and main-
- tenance.
, 2. ASME Section III root valves were added for all instnanent lines associated with the modified off gas system orginating from ASME Section III piping and components. These valves serve as code boundary points.
- 3. The code classification for the recombiner condensate cooling water system was changed from ASME Section III class 3 to ANSI B.31.1. to
'i eliminate construction and procurement problems.
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- 4. Off gas storage tank d,scharge piping specifications were changed from .
SA-106 Grade B to eithar SA-333 Grade 6 or SA-312 type 304 to reduce the ' '
possibility of nil-duc tility failures upstream of the flow restrictors in the event of heate failures. ,
- 5. The material specifir;ation for off gas piping between the recombiner r and after condenser was changed from AS'IM A-312 TyTe 304 to SA-335 Grade P11 to reduce stress levels due to thermal expansion.
- 6. Off Gas storage tank drain traps could not be purchased to ASME Section III Class 2 and 3 requirements. They were therefore encapsulated in vessels which meet the code requirements.
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- 7. A third hydrogen analyzer was added at the outlet of each recombiner train to meet licensing (DRL) requircsents that two analyzers be in operation continubusly. ..
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- 8. The' ventilation exhaust from the off gas storage building was modified to automatically shutoff when the standby gas treatment system is in- '
itiated, thereby assuring adequate SGTS flow.
- 9. The off gas flow control system was modi-fied several times to improve perfomance and reliability. j
- a. The initial modification consisted of:
a.
Adding a pressure control system to maintain a constant pressure ir. ll '
the sixty second N-16 delay line upstream of the recombiner trains.
Adding a second pressure control system to limit inlet off-gas flow I so as to maintain the pressure at the eductor outlet below the design lJ operating limit. ,
Modifying the electronic auctioneer circuitry associated with the re- ,
combiner inlet hydrogen control system to pemit manual section of one hydrogen mass flow rate signal and one N-16 delay line pressure ,
control signal, p
- b. The second modification relocated the off gas inlet flow meter to a
position ahead of the flow control valve to eliminate pnematic feed-back from the dilution eductors. .
- c. The third modification consisted of: n
- 1) Replacing the Eastech inlet off gas flow meters with orifice plates and differential pressure transmitters. p
- 2) Eliminating the inlet hydrogen analyzers.
- 3) Eliminating the hydrogen inventory processors.
This allowed a simplification of the inlet off gas flow control system with reduced, but still acceptable, capacity. :
i 10. Electric heating elements were added to the outside of the recombiner vessels to maintain vessel and catalyst temperature above saturation 1
. during shutdown and to shorten the startup time.
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- 11. Recombiner condenser coniensate cooling water supply and return p ints were changed from the main condenser hotwell to down stream of ti main condensate pumps to provide adequate NPSH for the recombiner corriensate pumps.
- 12. A flow limiting orifice was installed in the off gas storage tank sample purge line to limit tank blowdown rates .into the 30 minute delay pipe, j 13. The off gas compressor cooling water systems were cross-connected to improve reliability.
- 14. Check valves were added in liquid drain lines and gas purge lines re-turning to the 30 minute delay pipe from the off-gas storage building to prevent backflow.
- 15. Pressure relief valves across the recombiner train off-gas outlet iso-lation valves were determined to be unnecessary and were deleted.
- 16. Check valves were added downstream of drain traps on the off-gas storage tanks to preclude inter-tank flow.
- 17. Discharges from the compressed off-gas relief and compressor unload-ing valves were rerouted to a point upstream of the off-gas filters to
. preclude backflow through the filters.
- 18. Isolation valves were installed downstream of the compressor relief and
, unloading valves to pennit isolation with the other compressor operating.
- 19. Off-gas compressor sealing air low pressure alarm switches were relo-cated to monitor other components in addition to the sealing air compressor, V
- 20. 'fbe 9ff= gas stack drain collec+1ng tank piping was modified to eliminate a direct path from the off-gas compressors to the stack.
- 21. Dual (parallel) steam pressure control valves were installed because of difficulty in procurement of a single suitable valve.
- 22. Redundant hydrogen inventory processors were added to the recombiner inlet flow control system to assure a conservative concentration of hydrogen is
! used at all times for inlet flow control. The processors were subsequently i- eliminated from the system (see item 9 above).
23 Electric heaters were added to bypass lines returning off gas to the main u, condenser to prevent freezing. They were later determined to be unnecessary and were deleted, r.
. 24. A number of alarms were added to the off gas control panel in the control room to alert operators to abnormal conditions.
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- i. 25 The recombiner condenser outlet flow meters were determined to be unnecessary and were deleted.
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- 26. A reembiner inlet line shutdown pressurization system was installed to , .
prevent hydrogen entering the rec mbiner when the system is shutdown.
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- 27. A program was conducted to reconstitute the 'A' recmbiner vessel pressure ,
ooundary after a temperature excursion in excess of the ASME code design temperature.
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- 28. Demisters were installed in the recombiner outlet hydrogen analyzers sample system to improve moisture rmoval.
f' 29 The valve disc, pins aM all other loose internal parts were removed from check valves located in the off gas recirculation lines to the main con-denser to eliminate potential friction or sparking. It was subsequently t detemined that this action was unnecessary and the parts were reinstalled. l
- 30. . Internals of several control and check valves were changed to non-sparking '
materials to eliminate potential ignition sources.
- 31. A short portion of the drain line from the 6" off gas pipe in the Storage .
Building to drain collecting tank V-811 was enlarged and sloped to im- I prove drainage and minimize line plugging, f
l 32. A code deviation was obtained from the Minnesota Chief Boiler Inspector i to allow liquid penetrant examination and radiograph in lieu of hydro-I static test on new welds in the off gas inlet line. '
l 33 The reembiner outlet hydrogen analyzer trap drains were routed to the flash tanks to improve moisture removal. '
o l 34. Thermocouples were installed downstream of the off gas flow control valves l to provide indication of pipe wall temperatures in the control room, t 4
35 To improve reliability, the differential pressure indicating switches which provided level indication and control for drain collecting tanks V-810 !
and V-813 vere replaced with capacitance probes in external standpipes and t l
gage glasses.
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- 36. The sample lines for the modified off gas system inlet hydrogen analyzers i connected to the air ejector discharge piping were capped to prevent off gas frm leaking into the turbine building. 1 37 The off gas empressor coolant flowpath was modified to provide higher I
temperature at the first and second stages, thereby minimizing moisture. [l'
- 38. The modified off gas system was reclassified in accordance with Regulatory ,
Guide 1.26.
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! 39 The low off gas inlet pressure trip of the recombiner system was determined to be unnecessary and was eliminated to improve reliability, j I-12 ;
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- 40. The off gas storage tank subsystm discharge flow meter was detemined to be unnecessary and was eliminated.
41 The off gas storage tank radiation imnitors were detemined to be unnecess-ary and were eliminated.
- 42. The discharge of Drain Pumps P-900A and P-900B was routed to the turbine building floor drain sump to reduce the volume of high conductivity water that must be processed.
43 A low point drain line, with isolation valve and strainer, was installed on the suction piping for each of the off gas eductors to faciliate mois-ture removal, i
C. Performance Characteristics Most plant performance characteristics during this reporting period continued to be as predicted.
Mministrative limits on reactor power to minimize in-plant background and contamination levels were continued from the previous reporting period until September 13th, at which time the plant was shutdown for refueling. Prior to shutting down, power was limited to approximately 56% of rated.
Fuel sip?ing operations identified 77 leaking fuel assemblies. All leaks fourrl to date mye been in initial core 7 x 7 fuel. All of the remaining initial core 7.x 7 fuel was removed from the reactor during this outage. Off-gas activity following the outage was at least a factor of 12 less than pre-outage t
levels and remained low throughout the rest of the reporting period.
During a visual inspection using underwater television, a crack was found in the heat affected zone of a sparger to junction box weld on one of the four feedwater spargers. This prob 1cm had been anticipated because of similar problems at other BWR facilities and a contingency plan to replace all four feedwater spargers had been prepared. Recent experience at other facilities also included discovery of cracking at the inner blend radii of the reactor vessel feedwater nozzles. Dye penetrant examination of our nozzles showed many linear indications on each of the four nozzles. All of the indicatons were completely removed by grinding. The amount of metal remaining after grinding was found to be in excess of the requirements of ASME Section III. me re-placement spargers are of an improved design utilizing heavier material,
, a forged junction box, and machined themal sleeves. The interference fit of the themal sleeves in the nozzle bores should minimize leakage during power operation, which is expected to minimize thermal fatigue of the nozzle i
blend radii.
Other significant equipnent malfunctions or performance deviations are briefly summarized in Section I.A.
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4 D. Procedure Changes The following changes in procedures were necessitated by itens B and C e above, or were required to improve the safety of plant operation:
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- 1. Surveillance and o? crating procedures were issued or revised as necessary
! to satisfy new Tec1nical Specifications requirenents and to acccanodate I:
equipnent modifications, t
- 2. A surveillance test procedure was issued to assure that the RCIC pump bearings are properly lubricated prior to testing. l-
- 3. Functional test procedures were issued for various Radiation Protection Instrunents, p
- 4. Several procedures were issued in response to actions requested by NRC-IE Bulletins 75-04 and 75-04A. They include: 1 3
j a. Several surveillance test procedures which provide for more thorough inspections of various portions of the fire protection system. +
- b. A surveillance test procedure for the smoke detector systen located in l the cable spreading roan, ,
, c. A surveillance test procedure for determining perfonnance of pumps I-associated with the Fire Protection System.
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- d. A detailed procedure for sealing and testing of electrical penetrations.
- 5. The procedure for monitoring the tenperature of the head and vessel during 3 tensioning and detensioning of bolting studs was revised to include specific locations to be monitored. '
- 6. The preventive maintenance procedure for the HPCI Turbine was revised to 5 include verification of proper adjustment of the overspeed trip reset spring. I.
! 7. A Rodworth Minimizer Bypass Checklist w.'s issued outlining steps to be taken f if the RhM becones inoperable. i,
- 8. A preventive maintenance procedure was issued for the Main Steam Isolation <-
Valves. ,
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- 9. A special procedure for Snubber Replacanent was issued for use during the Fall 1975 refueling outage. I, i
- 10. The procedure for internal inspection of the Pressure Suppression Chamber was revised to assure all pertinent areas are inspected.
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- 11. U A procedure for periodically testing the Off Gas Systen Heat Tracing and High Tenperature Trips was issued. ,
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. 12. A procedure for the rmoval, relocation and reinstallation of the mechanical stop on the fuel preparation machine was issued.
- 13. A procedure for core reload verification was issued.
- 14. An operating procedure was issued to provide guidance on the preferred sequence of water addition to and heat rmoval fra the vessel.
, 15. The procedure for plant shutdown outside the control rom was revised.
- 16. Emergency fill and discharge procedures for the off gas storage tanks were revised to assure that design flow rates could be met.
E. Results of Surveillance Tests and Inspections The results of all surviellance tests ana inspections required by the Technical Specifications were satisfactory except as noted below:
- 1. During routine surveillance, one of the reactor high pressure scram switches failed to trip due to an improperly adjusted bourdon tube stop plate, as reported in Mr. L. O. bhyer's letter to A. Giambusso, dated July 11, 1975. After adjustment, the switch operated properly.
- 2. During routine monthly surveillance the motor for "A" RER torus cooling injection valve . overheated and failed, as reported in Mr. L. O. Mayer's letter to Mr. A. Giambusso, dated August 6, 1975. The stm clamp set-screws had sheared, allowing the stm to rotate. This prevented the 4
limit switches fra demergizing the motor at the end of the valve stroke and the motor overheated. The motor was replaced, and larger set screws were installed.
- 3. During routine quarterly surveillance, the "A" RFR torus cooling injection valve failed to close completely due to an incorrect torque switch setting 4
as reported in Mr. L. O. Mayer's letter to Mr. R. S. Boyd, dated Sept aber 3, 1975. The torque switch was adjusted.
- 4. hhile performing a routine surveillance inspection, it was discovered that a hydraulic shock suppressor on "B" Loop LPCI Line was low on oil as reported i-in Mr. L. O. Mayer's letter to Mr. R. S Boyd, dated Septmber 24, 1975.
The shock suppressor was replaced.
- 5. hhile perfonning a monthly Standby Diesel Generator Test, #2 starting i
. . system for #11 diesel generator failed to operate due to a loose connection on a contr61 air line to the air start motors, as reported in Mr. L. O. Mayer's letter to Mr. R. S. Boyd, dated October 24, 1975. i The loose connector was tightened. j
- 6. hhile preparing for the performance of a Primary Containment Integrated I I,
Leak Rate Test, #3 TIP ball valve was found to be partially open as reported in Mr. L. O. Mayer's letter to Mr. R. S. Boyd, dated Novmber 26, 1975. Investigation revealed a stuck wiper on a wafer-type limit switch.
The valve was replaced, i4 I-15
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, e i 7. During routine surveillance of the Auto pressure Relief Systs, it
/ was found that the Auto Blowdown Time Delay Relay for the "A" logic d
channel failed to energize in the required time interval. This was.
j reported in Mr. L. O. Mayer's letter to Mr. R. S. Boyd, dated Novmber 26, U 1975. The time delay relay was reset with a greater margin allowed for i j drift.
(
i 8. During routine surveillance, the RCIC turbine failed to start when given [
a fast start signal as reported in Mr. L. O. Mayer's letter to Mr. R. S.
- Boyd, dated Novmber 26, 1975. The failure was due to the steam control e i- valve sticking closed. The stem and pivots were polished and-lubricated.
- f j 9. During routine monthly surveillance, the #12 Diesel Generator tripped on -
! overspeed during a fast start as reported in Mr. L. O. Mayer's letter to 't l Mr. R. S. Boyd, dated Novmber 26, 1975. The failure was due to an t-j improperly adjusted cam operated switch in the fast start control circuit.
i The cam was adjusted, r i
l
! F. Containment Leak Rate Tests Test Date Reason for Test Results of Test j 1. Septmber 12, 1975 Secondary Containment Capability Demonstrated capability Test perfonned in accordance to maintain at Icast with Section 4'.7.C.1.a of the 1/4" water vacuum under i :
! Technical Specifications to calm wind conditions, 1 ensure that the secondary with less than 4000 cfm j containment system meets leak- filter train flow.
t age acceptance criteria prior j to refueling, a 2. Sept aber - Type B local leak rate tests Sixty-one components- L Nov mber 1975 perfonned in accordance with were tested. All met Technical Specification the acceptance criteria, 4.7.A.2.e to ensure that however, three (3) were o j the primary containment- repaired to rcxiuce their syst m meets leakage accept- leakage.
- ance criteria.
1-
[ 3. Septanber - Type C local leak rate tests. Sixty-eight components ,
! November 1975 performed in accordance with were tested of which Technical Specification twelve (12) required b 4.7.A.2.e to ensure that the repairs to meet the primary contaiment system _ criteria. M
, meets leakage acceptance a 1
criteria.
- 4. Septonber - Local leak rate tests of the Four of the eight (8)
October 1975 MSIV's performed:in accord- valves required repair ance with Technical Spec- to meet the criteria.
ification 4.7.A.2.c to ensure that the primary containment l system meets leakage accept-ance criteria.
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[ Test Date_ Reason for Test Results of Test i 5. Novm ber 12, 1975 Air lock leak rate test I The measured-leakage prior to integrated leak was 4.94-SCFH.
rate test.
e
- 6. Novmber 14 515, Type "A" primary contain- Measured leakage was j 1975 ment integrated leak rate 0.2456 weight percent of '
j test perfomed in accord-
- - ' contained air per 24 ance with Technical hour period. (The
{
i Specification 4.7.A.2.a to Technical. Specifications
- ensure that the primary - require that the_oper-contaiment systm meets ational_ leakage not operaticnal leakage accept- exceed 0.90 weight
- ance criteria, percent of_ the contained -
l air per 24-hour period.)
! 7. Novmber _19,1975 Air: lock leak rate test during l - The measured leakage-
- startup, was 1,61 SCFH.
i G. Chang'es, Tests and Experiments Requiring NRC Authorization p
There were no changes, tests or experiments during this report period which l required prior NRC Authorization;
}
H. Changes in Plant Operating Organization l .
There were no changes in the plant operating staff for those positions which 1
are designated as key supervisory personnel on Figure 6.1.2 of the Technical Specifications.
i 1
!_ I. Occupaticnal Personnel Radiation Exposure i
{ 1. Radiation Exposure by Incr ments in Reporting Period:
! Exposure Number of Personnel Pemanent T mporary Incr eents (Site NSP Employees)
(hRBO (Off-Site NSP-Employees) 1 4100
- 11 65 t
+
100 -'240 4 13
!- -250'- 500 -3 i- 500 - 750 14" 8 17:
!t 750 - 1000 16 13 1000 --2000 ..43 41 F 2000 - 3000 '15 3000 - 4000 12 10 7 4000.- 5000- -
4 3
> 5000 0 0 d ,
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O w e--,- - - ~=p- r ,,e,,a .wr- ,newc-,.--.x -. ww w er-- % ,,w r .<- + o ~,.sww-...-m.. comw..-rw.. <+s*w%v-r--+-e-
/
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- 2. Exposure by Duty Punction in Excess of 500 mrem in Reporting Period: ,
Duty Permanent Number of Exposures >_500 mrm
- T mporary j
Function (Site NSP Ihployees) (Off-Site NSP Ihployees) -
Routine Plant '
Sutveillance and Inspection 61 Routine Refueling i' Operations 33 Routine Plant hhintenance 26 b Special Plant ** ,
hhintenance 44 93 i Radwaste System ,,
hhintenance 23 '
5
-l.
Radwaste Preparation and '
Shipnent 19 l Off Gas Systs hhintenance 10
- 3. Exposures in Excess of 3000 mrm for the Year 1975: I V
Number of Job Major ***
Personnel Category Causes [
l 4 Radiation Protection Routine Plant Surveillance Radwaste Shipments j 16 Site hhintenance Special Plant hhintenance Routine Plant thintenance ,,
Radwaste hbintenance Radwaste Shipnents [
13 Operations Radwaste Preparation 1 Routine Plant Inspections d Refueling Operations 1 Engineering Off Gas Maintenance ,"
Routine Plant Inspection t
U I-18 ll L
4 Number of Job Major m Persmnel Category Causes 1 Instrument 6 Special Plant Maintenance Controls Off Gas Maintenance Routine Surveillance 16 Off-Site Special Plant Maintenance Maintenance Radwaste hhintenance
- Some individuals appear in more than one duty function.
- Special plant maintenance is maintenance performed in primary containment during plant shutdown.
- Causes are listed in order of greatest percentage of exposure received.
J. Relief Valve Operation Sumary Following is a sumary of relief valve operation during this report period:
Relief Valve Number of Number Date How Initiated Initiations RV 2-71A 11/18/75 Manually during 1 Operability Testing RV 2-71B 11/18/75 bhnually during Operability Testing 1 L
RV 2-71C 11/18/75 bbnually during l Operability Testing 1 RV 2-71D 11/18/75 Manually during Operability Testing 1 RV 2-71E 11/18/75 Manually during Operability Testing 1 RV 2-71F 11/18/75 Manually during i u Operability Testing 1 RV 2-71G 11/18/75 bhnually during l
, Operability Testing -1 RV 2-71H 11/18/75 Manually during-j ;*,
Operability Testing 1 I-19 f
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YEAR: 1975 JULY AUGUST SEPTBBER OCTOBER NOVBBER DffBBER TOTALS
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G R p pg 911,424 796,126 230,074 0 325,042 1,214,714 3,477,380 hg m sa
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303,600 263,340 73,800 109,970 GM @N 0 417,280 1,167,990 0
g 286,368 246,766 67,448 -1944 103,828 401,288 1,103,75A BER OF M REACTOR CRITICAL 744 742 254 0 321 744 2,805 OF H M 744 742 247 0 266 744 2,743 GENERATOR ON LINE t
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B. Shutdowns A " Routine Operating Report Shutdown Record" form, containing all of the information required by the Technical Specifications, has been established to facilitate reporting of outages. This section contains copies of the record forms for this report period. 't I l 4 a t 5 Id
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_ _ _. _ _ _ ~ __ . _ . .. __ _ __ _.-_ --____ _ __ P4 ? 6002 )
- FOUTINF OPERATIfB RFPORT SHUT [DWN RECORJ , r
[ ( i 1. - Cause of _ the outage: The "C" condensate deineralizer effluent valve failed e to open when placed in service. This' caused the conden- sate d eineralizer controls to close the other effluent valves, resulting'in a low pressure trip of the reactor F, 4 feedwater pumps. Ioss of feedwater caused a low reactor ., j water level scram. *
- 2. Me thod o f shu t t i ng down _ t he reacto r; e. g. , T r i p au toma t i c rundown, or l' manually controlled deliberate shutdown:
i T Reactor scram due to low reactor water level. l [ 4 i [ 3. - Duration of outage: Date Time v ! 1 [ Generator off 'line: 8/31/75 2216 Generator on line: 9/1/75 0530 'i j l- !- 4. Plant Status during the ' outage (e.g., cold shutdown or hot standby): Hot' shutdown. j 't'
- i i
i 1
- 5. Corrective Action taken to prevent repetition, i f appropriate:
r ! Effluent valve was disassembled, inspected and new teflon seats- [ j installed.
- . .- r The condensate demineralizer controls were modified'so that a f
i high system differential pressure will override.the nonnal flow control signal and prevent valve closure.' 4 J b - l l t
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_. . . . - . . .-. . - - . . - _ .- _ ,,._% . . . , . . , . ~ . . . . . . . - - _ - _ . _ , . . . - . . . - . . - , . . . . . . . . . . . . . , , . . - , - _ . - . . -
i 6002 . j i 1 FOUTI ff OPERAT I f6 RFPORT SHliT[O AN RFCORD 4 l
- 1. Cause of the outaBe Scheduled refueling outage I
i
- 2. Method of shutting down the reactor; e.g., T rip automat ic rundown, or
- manually controlled deliberate shutdown
^
- Manually controlled deliberate shutdown.
- 3. Duration of outage: Date Tim Generator off 1ine: 9/11/75 1716 1
- Generator on line
- 11/19/75 1435
- 4. Plant Status during the outage (e'.g., cold shutdown or hot standby):
Cold Shutdown . t
- 5. Corrective Action taken to prevent repeti t an, i f appropriate:
f L, 4 f' k r .
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i 11-5
y ) ) 4001 Rev. 2 { , MAINfENANCE SLM4ARY Summary of Work Performed, Including Corrective Action Yaken to Prevent Repetition (if applicable) and Pre-System or Cause of the Results and Effect cautions Taken to Provide for Reactor Component Malfunction on Safe Operation Safety During Repair (if applicable). Neutron Monitoring Randan Component Failure Failed Downscale - Replaced Failed Ion Chamber Power Supply System-IPRM Channel Downscale Alarm 28-05B Received Neutron Monitoring Unknown Excessive Noise-SRM Replaced both detectors. Rod Block System-SRM Channels Counts would.not go to initiated while repairs were in progress. 21 and 22 zero with nonrial discri-minator setting and
- detectors withdrawn.
O a Neutron Monitoring Random Component Failum Channel Output Bypassed channel and replaced relays K1 Systen-IRM Channel Read High when placed and K4 in the Amplifier Attenuator Circuit. 15 in '40' Test position. Test circuits calibrated and IRM returned to nomal operation. Neutron bbnitoring Under Vessel Connector No Channel response Bypassed channel, repaired under vessel Systen-IRM Channel Damaged during Plant to Control Rod movanent. connector, then returned channel to 11 Outage normal operation. Rod Worth Random Component Failure RWM Initiated Rod Replaced Failed. Relay in Steam Flow Alam Minimizer Blocks at 86% Power. Unit which automatically bypasses RIN when steam flow greater than 20%. 4 Reactor Manual Normal Component Wear Timer Malfunction Alann Replaced wom clutch parts in Rod Drive - Control Syston- and Rod Block received. Timer Rod Drive Timer
-- -- - . . . . - - -, - -. .. ... -- - - - . . . . .. . - - . . -- .. - . - . ~ . . - ~.. . --- - . - . .. . .. .-
t 1 !' 4001 Rev. 2 i MAIhTENANCE S(f f L\RY Sumnary of Work Performed, Including t Corrective Action Taken_to Prevent Repetition (if applicable) anni Pre- , System or Cause of the Results and Effect cautions Taken to Provide for Reactor Component' Malfunction on Safe Operation Safety During Repair (if applicable). Traveling In-Core ' Sticking Limit Switch Ball Valve Failed to Replaced Ball Valve with new spare unit. Probe - #3 TIP close completely. Reactor was shut down and primary contain-
. Ball Valve ment integrity was not required at the time.
Emergency Diesel: Nonnal- _ Excessive Setpoint Replaced Diaphragm in pressure switch. Generator - #12 Drift - Compressor Air tanks pumped to required air pressure
. Component Weak Diesel, #11 Air - Small liole in pressure loading point drifted prior to initiating repairs.
.[ Compressor switch diaphragm. high.
RIR Service' Water ' Vibration caused wear on Small air. leak developed Replaced copper air line with stainless System Control copper' air line. in air supplyg steel. Reactor shutdown at time, RHR SW Valve, CV 1728. system not required to be operable. Modified Off-Gas Moisture in Sample systen- 11 and 2% Alanas . Components repaired or replaced as required. 11oldup System- and Randan Component received. Outlet flydrogen - Failums Analyzers
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4001 ~ Rev. 2
%\IhTFlm'CE SU'It\RY Sumary of Work Performed, Including Corrective Action Taken to Prevent Repetition (if applicabic) and Pre-System or Cause of the Results and Effect cautions Taken to Provide for Reactor Component Malfunction on Safe Operation Safety During Repair (if applicable).
HPCI Minimum Flow The valve seat was slightly Water leakage through Lapped valve disc. to seat and adjusted Valve CV 2065 cut. the valve when closed. the valve stroke. Reactor was in Cold Shutdown condition. i$ #11 SBIC Pump Monnal use. Distortion of the nylon Installed new packing followers.
, packing followers.
The redundant pump was demonstrated to be operable before each pomp was repaired.
"B" Core Spray Snub- [ncorrect Installation. The snubber was inoper- loosened the snubber clamp around the core her SS-41 able.
spray line. Rotated the clamp for proper location and tightened the bolt'.s SBLC check valve Crud on the disc and seat Did not meet leakage Cleaned and adjusted the disc. XP .;urfaces. test requirements. Reactor was in Cold Shutdown condition. Drywell equipment Foreign Material under the Did not meet leakage test Cleaned and lapped seating surfaces. drain sump isola- seat, requirements. Reactor was in cold shutdown condition. tion valve A0-756,1A.
=
4001 Rev. 2 MAIKfE NANCE SLDtuRY _ . . . _ _ . - _ - _ _ _ . ._g_.. Sumnary of h'ork Performed, Including Corrective Action Taken to Prevent Repetition (if applicabic) and Pre-Sy t.s er Ortsu oJ tir P,esults a d Effect cautiens Taken to Provide for Reactor Co.penent :".W cietion on Safe Operation Safety During Rcpair. (if applicable). HPCI - Steam Supply The undervoltage relay coil The valve was electri- Replace the undervoltage relay coil Valve FD- 2036 opened. cally inoperable. The IIPCI system had been made inoperable foi testing and redundant equipment operability had been verified. A. RHR Valve 50-2008 Stan guide set screws shear- 'Ihe valve was electri- ' Drilled and tapped the stem guide for largen ed allowing stem to turn. cally inoperable. set screws and set than in loctite. Replace i burned out motor and installed a new limit switch assenbly. I F W vbnt Valves on the Systen were closed j during repair. t - 1 Off Gas recombiner Scratches on the disc's' Valveswould fail to open. Lapped the disc of each valve, system valvesTCV- cause the valves to stick. 7516 A6B, and TCV i Reactor was in cold shutdown condition. 7517 AGB. Core Spray check Crud on the seating sur- Did not meet leakage Cleaned seating surfaces ani reassembled . valves 14-13 A6B. faces. test requirements. with new gaskets. Reactor was in cold shutdown condition. . 4 6 m eg_ e . ws.*w- em mus-es e e - enum.==iumum, =
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4001 Rev. 2 , MAIhTENANCE SGT4ARY t
- t Stur.ary of Work Performed, Including Corrective Action Taken to Prevent -i Repetition (if applicabic) and Pre- i System or Cause.of the Results and Effect cautions Taken to Provide for Reactor ,
Safety During Repair (if applicable). Component Malfunction on Safe Operation. , CRD check valve Crud on.the seating sur- Did not meet leakage tes1. Cleaned,' lapped and reassembled with new i CRD-31. faces, requirements. bonnet seal ring. . Reactor was in Cold Shutdown condition. i MSIV 86A Scratches on the' pilot valve Did not meet leakage tes1. Iapped the main valve seat and cleaned the .j g requirements. , main poppet and seat. l 7
" g Reactor was in cold shutdown condition. !
i Scratches on the pilot seat Did not meet leakage test lapped main valve seat and cleaned the main MSIV 80A requirements. poppet. Iapped the pilot poppet and seat. + Reactor was'in cold shutdown condition. 1 bSIV 86C Scratches on the pilot valve,Did i not meet leakage Iapped the main valve seat and cleaned the
- test requirements. main poppet. Iapped the pilot poppet and
- seat. t 1
- Reactor was in Cold shutdown condition.
NSIV 80C Scratches in the. pilot seat. Did not meet leakage Iapped the main seat and cleaned the main
- test requirements. poppet. Impped the pilot poppet and seat.-
l 4 l Reactor was in Cold sluitdown condition. i. 3 -
1 b
.4001 '
t Rev. 2 MAINTENANCE SQf4ARY ; i Sumary of Work Performed, Including Corrective Action Taken to Prevent i Repetitien (if applicable) and Pre- ' System er Cause of the Results and.Effect cautions Tahen to Provide for Reactor Cenponent 'b1 function on Safe Operation Safety During Repair (if applicable). i e j .Off Gas Flow con- Crud on the plug and cage. Valve striking Cleaned th'e plug and cage. ; trol Valve 7489B Reactor was in cold slutdown condition. Diesel Generator Unknown Speed drifted ' Replaced the governor on #12 diesel.
#12 Reactor was in cold shutdown condition.
, :0 Ye Off Gas Cmbiner Foreign material in the Valve would not close. Replaced air solenoid valve. Valve E V solenoid. Reactor was in cold slutdown condition. i:. I #12 Stack Gas Defective starting switch.. Motor would not start. Replaced the motor. '!
- SamPl e P NP' Reactor was in cold shutdown condition. ,
#12 Diesel Generator Still under investigation Heat exchangers tubes Installed two new cooling systen heat ex-1eaking. changers. ,
Reactor was in cold shutdown condition. t i LPCI relief valve Vibration caused a crack at small leak Modified the connection'to the LPCI line. 4 RV 2004 the thread connection Modified the drain line from the relief valves. , ! Reactor was in cold . shutdown condition. - l r-- . f. - q . 7. s , ,m* 1.-<- y- -4 , .. - ~7 7- - - - - - , y - - .q - ~~3 - 3__- ;f
.tC ; J. ;J ; .; . .. . , , - 4001 - Rev. 2
~.
MAINTENANCE SU.itt\RY Stenary of Work Performed, Including Corrective Action Taken to Prevent. Repetition (if applicable) and Pre-System er . Cause of the Results and Effect cautions Taken to Provide for Reactor Component. Malfunction on Safe Operation Safety During Repair (if applicable). Scram inlet valves Normal deterioration of Slight leakage through Replaced the seats. N 's'30-47 and seats. valve. Reactor was in cold shutdown condition. Auto blowlown timingThe relay was set to high Did not time out within Reset the relays to time out at 106 seconds. g- 2E-K5A and~2E-K5B for it's repetitive accur- specified limit. Reactor was in cold sMtdown condition. acy of + 10%. w7 RHR Check Valve Hinge pin plug was not Leakage of hinge pin Replaced bonnet gasket and welded a cap RHR-21, properly torqued when plug. over the hinge pin plug. installed. Reactor was in cold'slutdown condition. CRD Module 46-35. . The chrome plating on accum- Accianulator would not Replaced the water accumulator. ulator cylinder wall flaked remain charged. off and pitted. Reactor was in cold slutdown condition.
.= .- - _ - - _ .-.
IV. CH00ES, TESTS AND EXPERIMENTS The following sections include a brief description and summary of the safety evaluation for those changes, tests and experiments which were carried out without prior NRC approval, pursuant to the requirements of 10CFR 50:59 (b):
- 1. DAMPER AND DIVERSION DUCT FOR CONDENSER AREA AIR SUPPL 3 DUCT (M75-42)
Description of Change A balance damper and diversion duct was installed in the condenser area , discharge duct of V-hE-5 to facilitate the regulation of air flow to the condenser area when desired for control of airborne activity. The excess air supply not required for the condenser area is discharged to a cable tray area of the turbine building. Sumary of Safety Evaluation 4 Flow frm the pot:ntially contaminated condenser area to the clean cable
, tray area is not possible, even without V-FE-5 nmning, since the conden-ser area is maintained at a negative pressure with respect to the cable tray area.
- 2. RB10 VAL OF THE REACTOR VESSEL B0lT04 HEAD DRAIN LINE (M75-49)
Description of Change
, The reactor vessel bottom head drain line and associated valves were re-moved to eliminate a cmd trap.
Summary of Safety Evaluation The modification was perfomed in accordance with AEI B31.1 - 1975, with quality assurance requirements in accordance with Article NA 4000, ASME BSPVC Section III-1974 edition, including winter 1974 Addenda. A waiver from the State of Minnesota was obtained to permit a 1000 psi hydrosatic leak test arri liquid penetrant examination of the new socket welds in lieu
. of a hydrostatic leak test at 1.1 times the operating pressure.
- 3. BACKWASH RECEIVING TANK VENT LINE TO 1.75 MINJFE DELAY LINE (M75-64)
Description of Change The condensate demineralizer backwash receiving tank vent line which
,, discharged into the Steam Jet Air Ejector Room Ventilation Systm duct work was replaced with piping and an isolation valve to allow discharge
.to the main stack via the 1.75 minute delay line. This minimizes the possi-bility of airborne activity fra the vent line leaking into the reactor
~ '
building vent plenum and initiating meaningless alarms or isolation of
.. reactor building ventilation.
Summary of Safety Evaluation The final point of discharge, the plant stack, is not changed. The modifi-1 7 cation meets the requirements of ANSI B.31.1 Power Piping Code. The opera-i tion of the vent line and associated systems is limited by interlocks and
' procedures to prevent overpressurizing the backwash receiving tank. In addition, overpressure relief capability is provided by the tank overflow
! line.
IV-1
1 Il b i
- 4. h0NTICELLO SEGIENTED TEST ROD BUNDLE h0DIFICATION (M75-64) U The description and safety evaluation of the modification to the hbnticello p
! segmented Test Rod Bundle was submitted as Addendum I to NEDE-2019 in a Ictter from Mr. L. O. Mayer to Mr. A. Giambusso, Director of Reactor Li-j censing, dated August 20, 1975. The modification was perfomed as de- ,
- scribed during the fall, 1975 refueling outage.
1 e j 5. CYCLE 5, REACTOR CORE REIDAD, (M75-66, M75-87) t i During the refueling outage 268 initial core 7 x 7 fuel assemblies were replaced with 268 8 x 8 fuel assmblies enriched to 2.19 wtt, U-235. One hundred and four assemblies with finger springs were preferentially - < loaded symetrically 'in the central orifice zone of the ' core, while the -g i remaining 164 assemblies were loaded to fill the remaining core locations j designated for reload - 4 fuel. . s l One fuel assembly loaded into the reactor core during the fall refuel-
- ing outage utilizes a 14 tab water-capture-spacer rod. The remaining l
267 reload assemblies utilize a 7 tab water-capture-spacer rod. i Sumary of Safety Evaluation l The safety analysis for 8 x 8 fuel is provided in Licensing Topical
- R6 port NEDO-20360 and NEDE-20360-1P " General Electric Boiling Water I.
Reactor Generic Reload Application for 8 x 8 Fuel."
- 6. INSTALLATION OF TOMJS INSTRUMENT AIR, ISOLATION VALVE (M74-45)
Description of Change ! As noted in Semi-Annual Report No. 9, a 3/4 inch air operated isolation i valve was installed on the instrument air line to the torus. < f Summary of Safety Evaluation 1 L The valve fails closed on. loss of air or power. Origirsi design codes, ! specifications and seismic requirments are satisfied. l-l 7. bODIFIED TORUS HATCH COVER (M75-76) t Description of Change
- A modified torus hatch cover containing penetrations'to accommodate 7 multipin bulkhead connectors, 24 Conax connectors and 6 sample lines *
(5 of 3/8" diameter and 3 of 1/2" diameter) was installed to provide the penetrations necessary for instrumentation associated with a -special : j Torus Response to Relief Valve Actuation" test. 1 1 Summary of ~ Safety Evaluation f ! 1 The suppression chamber is. designed for 56 psig internal pressure, 2' ; psig external pressure and a maximum tmperature of 281 F. The design t i code is' ASME Code Section III Subsection B. i t l IV p i
,n,n.,,m n , , , . . ,-- .-
j The modified hatch cover was analyzed per the design codes and shown to be adequate for the above design conditions. In addition all connect-
- ors, fittings and valves used on the hatch have design specifications i which exceed the design limits for the torus. The hatch was also placed
- on a test stand and pressure tested at 70 psig and leak tested at 43 l psig to verify a leakage rate less than or equal to 17.2 SCFH.
1 , All changes following are associated with the Modified Off Gas System. Re-vised P4ID's for the Modified Off Gas System are provided at the end of the
- section.
- 1. DELETION OF THE CHARCOAL FILTER IN 'IHE OFF GAS STACK (SRI 107)
- The HEPA stack filter was not replaced with a charcoal and HEPA filter as orginally proposed, thereby eliminating unnecessary modification, testing and maintenance.
Summary of Safety Evaluation i Two charcoal filter beds are installed at the suction _of the off gas j ' compressou. These filter beds provide adequate iodine removal capability, j consistent with the AEC-DL safety evaluation for Technical Specification j Change No. 2.
- 2. CHANGES DUE 'ID CODE AND FRTERIAL RBQUIREMENTS (SRI 130A) l Description of Change i
- a. ASME Section III root valves were added for all instrument lines q originating frca ASME Section III piping and components. These
- valves serve as the code break point and were required due to i
general unavailability of Section III instruments.
- b. The code classification for the condensate cooling water system was changed from ASME Section III Class 3 to ANSI B31.1 because of the difficulty in procuring Section III pumps and the need to cast the cooling water piping into the shield walls as part of the struct-ural construction contract.
- c. Off-Gas storage tank discharge piping specifications were changed
' from SA-106 Grade B to either SA-333 Grade 6 or SA-312 Type 304 to reduce the possibility of nil-ductility failures upstream of the
- flow restrictors in the event of heater failures. To protect down-l stream piping, low pipewall temperature-sensors were added which re-
{' sult in alarms and in automatic termination of discharge flow.
- d. The material specification for off-gas piping between the recomb-iner and the after condenser was. changed from ASIM A-312 Type 304 to SA-335 Grade P11 to reduce stress levels'due to thermal expansion.
- e. Off-gas storage tank drain traps and recombiner steam supply traps could not be purchased to ASME Section III Class 2 and-3 requirements respectively. They were therefore encapsulated in vessels meeting =
the code requirements. The discharge sides of the traps are exposed j to the encapsulating volumes to preclude over-pressurization potential due to confined liquids. IV ._ -
- -. _ - - - _ _ . ~ ~ . _ . _ _ _ . ~ . __ __ __
Sumary of Safety Evaluation Each of these changes have been reviewed and found acceptabic with re-spect to code and material requirments.
- 3. ADDITION OF A 1111RD IffDROGEN A)MLYZER TO Tile OLITLLT OF FACH OFF CAS RIE04BINER TRAIN SRI 130B)
Description of Change A third hydrogen analyzer was added at the outlet of each recombener train to meet licensing (DRL) requirements that two analyzers be in operation continuously. 8;umary of Safety Evaluation Addition of a third analyzer increases the reliability of the hydrogen analysis.
- 4. ALTTQ4ATION OF VFNTIIATION IXHAUST FR04 IIIE GAS S7DRAGE BUILDING (SRI 130C)
Description of Change The ventilation exhaust from the off gas storage building was modified to automatically shutoff whenever the standby gas treatment system (SGTS) is initiated. This minimizes the back pressure on the SGTS and essures that adequate SGTS flow can be maintained. Provision is made to restart the off-gas storage building ventilation in a full recirculation mode if full SGTS flow is essential, or with partial exhaust (0-500 cfm) if full SGTS flow is not required, such as during routine containment purging. Sumary of Safety Evaluation This change improves the flow capability of the SGTS.
- 5. bDDIFICATIONS TO CONTROL SYSTEM OF OFF GAS INLIT Flui CONIROL VALVE (SRI 130D)
Description of Change A pressure control system was added to maintain a constant pressure in the sixty second N-16 delay line upstream of the recombiner trains. This control prevents reduced delay times due to evacuation of the delay pipe and improves system response to air ejector flow transients. A second pressure control systan was added to limit inlet off-gas flow so as to maintain the pressure at the eductor outlet below the design oper-ating limit of 20 psia. This control serves to prevent recombiner train shutdown if off-gas flow exceeds availabic compressor capacity during , transient conditions such as startup without compressor bypass, l IV-4
The ciectronic auctioneer circuitry associated with the recombiner in-Ict hydrogen control s/ stem was changed to pemit manual selection of one hydrogen mass flow rate signal and one N 16 delay pipe pressure control signal. These selected signals and the eductor outlet pressure control signal are input to the auctioneer, which selects the signal requiring the greatest amount of inlet flow restriction arxl directs it to the valve controller. Sumnary of Safety Evaluation If owrating limits are exceeded either the signal selected for control or tae redundant signal for each parameter will result in automatic train shutdown.
- 6. RE1LCATION OF OFF GAS IN1ET F1MIETERS (SRI 130P) j Description of Change
- During the early stages of recombiner system testing, the off gas inlet I flowmeters were found to operate erratically at dilution steam flows above 3000 lbs/ hour. This erratic behavior was detemined to be caused by a pneumatic feedback from the dilution eductors to the inlet flow-meters.
To eliminate the feedback the inlet flowmeter and inlet flow control valve were relocated to place the valve between the flowneter and the eductor. Summary of Safety Evaluation , This change eliminates the feedback problem and does not in any way affect the procc s. The modification was made in accordance with applicable codes.
- 7. REPLACDIENT OF OFF GAS SYSTD1 FIIM METERS AND ELIMINATION OF INLET IffDROGEN ANALYZERS AND INVENf0RY PROCESSORS (SRI 130DDD)
Description of Change The Eastech inlet off gas flow meters were replaced with orifice plates and differential pressure transmitters. The inlet hydrogen analyzers and associated inventory processors were eliminated. This allowed a simplification of the inlet off gas flow control system with reduced, but still acceptable, capacity. Sumary of Safety Evaluation The modified control system will continue to limit the off gas flow rate such that an explosive off gas mixture does not reach the recombiner. The modification was performed in accordance with applicable codes. Testing was conducted to verify proper operation of the modified flow control systm. t 4 IV-5
.a - -
r - , . -_ ,
i i i .
- 8. ADDITION OF E1ILTRIC llEATI?E ElB(FRIS 1U 11E GITSIDE OF 11IE RImBIhTR I VESSELS (SRI 130E) l Description of Change ,
Electric heating elements were added to the outside of the recombiner l vessels to maintain the vessel and catalyst temperatures above saturation during slutdown and to shorten the startup time. Sumary of Safety Evaluation The heater elements are external to the recombiner vessel and have themo-static controls and safety switches. The heater elements have a maximum element temperature well below the hydrogen-oxygen autoignition temperature. 1 9. MODIFICATION OF OFF GAS ODNDliNSATE COOLI?C KATER SUPPIX AND RE'IURN PolhTS l(SRI 130F) I Description of Change Recombiner Condenser Condensate cooling water supply and return points were changed fr a the main condenser hotwell to downstream of the main condensate pumps. This provides an adequate NPS!! for the recombiner con-densate pumps. Strainers were deleted from the recombiner condensato suction lines since they are now downstream of the main condensate pump strainers. Summary of Safety Evaluation The small additional condensate temperature increase as a result of leat
, removed from the recombiner condenser (x 21/2*F) does not significantly affect the air ejector or condensate dominerali:cr performance.
- 10. MISCELLANEOUS PIPI?E GIA?CES (SRI 130G)
Description of Changes
- a. An orifice was installed in the off gas storage tank sample purge line
, to limit tank blowdown rates into the 30 minute delay pipe. This avoids exposing the piping to temperatures below NUIT.
- b. The off gas compressor cooling water systems were cross-connected for improved system reliability.
- c. Check valves were added in liquid drain lines and gas purge lines re-turning to the 30 minute delay pipe from the off-gas storage building to prevent backflow,
- d. Pressure relief valves across the recombiner train off-gas outlet isolation valves were deleted. They were unnecessary since the system is designed to withstand all conceivabic pressure excursions in the isolated condition,
- c. Check valves were added downstream of drain traps on the off-gas storage tanks to preclude inter-tank flow.
IV-6
_ _ - - _ _ - _ . _ - - . _ . _ . _ . ~ . . _ - - _ _ - - _ _ .__ ._ _ _ __ _ _ _ _ _. T. Discharges frm the cunpressed off gas relief and cmpressor unload-ing valves were rerouted to a point upstream of the off-gas filters to preclude backflow thropgh the filters. , j g. Isolation valves wre installed downstream of the relief and unloading i valves to pemit isolation with the other campressor operating. l h. Off-gas compressor scaling air low pressure alarm switches were re-
- located to downstream of the pressure relief valves in addition to the sealing air compressor system.
- , Sumary of Safety Evaluation These changes meet applicable code and process requirements.
- 11. BLOhlXXN lODIFICATION OF lilE OFF GAS STACK DRAIN COLLECTING TANK (SRI 13011)
Description of Change _ d The off gas stack drain collecting tank (V-811) piping was modified to i eliminate a direct path frm the off gas compressors to the stack. Sealing j air is used to blowdown the tank rather than compressed off gas. Sumary of Safety Evaluation Although the scaling air pressure is normally much higher than the drain tank pressure, backflow is prevented by a check valve and a pressure re-ducing valve.
- 12. FDDIFICATION OF PRESSURE CONTROL FOR OFF GAS TRAIN STF#! SUPPLY (SRI 1301)
Description of Change i Because of procurement problems, single pressure control valve for the off gas steam supply was changed to two parallel pressure control valves. Summary of Safety Evaluation The dual pressure control valves fully meet the process requirements and satisfy the design and ASME Code requirements.
- 13. ADDITION OF REIUNDtWT INDROGEN IhVEh%RY PROCESSORS TO THE REC 04BIh'ER INLET CCNTROL SYSTaT-(SRI 130J)
Description of Clunge Redundant hydrogen inventory processors were added to the recombiner inlet control system to account for the varying delay times in the 60 second
, N 16 decay pipe during startup, shutdown, and operational transients.
LUTE: The processors were subsequently climinated (see item 7 above). i IV-7 1.
J
~
I j . . ! Stmary of Safety Evaluation . These processors assure conservatively safe hydrogen concentration signals l at all times, including power interruptions. il
- 14. ADDITION OF MTRIC HEATING ELD 4WIS 'ID ' IRE OFF GAS BYFASS LINES (SRI 130K p i and 130ZZ) j Description of Change b 0
Electric heaters were added to the bypass lines which return off gas to the l main condenser during flow transients. The_ heaters were installed to pre-j f! ] vent line or valve freezing due to tic refrigeration effects of expanding d i
- the off gas to the high vacuums in the condenser. ,
NOTE: Subsequent testing disclosed the heaters were unnecessary and they l f were rmnoved. ) 4 i Summary of Safety Evaluation b 0 i The heaters have m:ximum temperatures well below tie auto-ignition temper-i ature of hydrogen-oxygen mixtures. {
.1l
, 15. ADDITION OF ComROL ROG1 ALARMS FOR OFF GAS SYSTfM (SRI 130L) .j 1 . f: Description of Change [ f ! A number of alarms were added to 'tle off gas control panel in tle control i room. \ i Summary of Safety Evaluation The additional alanns enhance process operability and safety. j ! 16. ELIMINATION OF RECOBINER CONDENSER OlJTLLT FIN METERS (SRI 1300) i Description of Change The recombiner condenser outlet flow meters were elianinated. i Summary of Safety Evaluation I Recombiner outlet flow indication is not required, however, a method of . { matching the storage facility supply and discharge rates is desirabic for 4 inventory control. This is accomplished by flow adjusting the discharge-i rate such that the sum of the pressures of the tank being filled plus the tank being' emptied remains less than 285 psig.. -f -
=
- 17. INSTALLATION OF RECOBINER TRAIN SIUTDOWN PRESSURIZATION SYSTIN (SRI 130Q) -Q
" Description of Change - j f To avoid the possibility of hydrogen Icakage past the inlet off gas-valves { and into the recombiner when a train is in the shutdown condition, equip-L ment was installed to automatically pressurize the portion of the system , between the train inlet valve and recanbiner inlet valve during shutdowni c L Pressurization to about 20 psia is' accomplished with instrument air. ,
, L i-l IV-8 f t
,,.._ _ ,, _ _ _ . . , _ , , , _ _ _ _ _ , _ _ .,_,.,a. ;.__________._.
4 The air is introduced through the existing supplementary air line. Hard and soft seated check valves in the air line prevent backflow in the event that theprocess line is pressurized with steam. The recombiner bypass valve KV-7516A (B) control was modified to vent the process line to the main
. condenser if the pressure exceeds 25 psig.
1 Summary of Safety Evaluation These changes are beneficial to safe and efficient operation of the system and are adequate to preclude steam from entering the instrument air system.
- 18. OFF GAS A TRAIN RECOGINER VESSEL REWORK (SRI 13(TT)
- Description of Change A failure of the heater control system for the "A" recombiner vessel strip heaters occurred prior to final connection of the off gas system to the plant. A temperature excursion in excess of the AShE Code design temperature was experienced by the recombiner vessel shell material.
A program was conducted to reconstitute the vessel pressure boundary to meet the requirements of the original design. Summary of Safety Evaluation The modification of this pressure vessel does not affect its basic design pressure or temperature and will not require any change to operating parameters. The vessel no longer complies with the requirements of ASME Section III Class 3 in the following areas:
- 1. The vessel shell material does not comply precisely with the re-4 quirements of SA 387 grade C specification. However, mechanical properties of the shell material are well defined and have been used in the stress calculations.
- 2. ASbE Section III requires that radiographs be provided after the final vessel heat treatment. The vessel has received a field stress relief heat treatment in addition to the temperature trans-ient experienced during the heater incident after the shop radiographs were taken.
- , The rewrked vessel meets all system design parameters and should be considered operationally equivalent to the original vessel. The AShE Jurisdiction Authority (Minnesota Chief Boiler Inspector) has approved the reworked vessel for service in Minnesota, s
- 19. INSTALLATION OF DEMISTERS IN SAMPLE SUCTION lEATER TO CUTLET HYDROGEN ANALYZERS (SRI 130Y) 4 Description of Change During operational testing of the recombiner trains the outlet hydrogen anlayzers on the B train tripped due to water collection in the sample suction header following shutdown of the train. To preclude future re-IV-9 4 -
--*--v- y y- m . . ,.y-_ - , - --.wg- a
-n - >--w- w
. r currences of such water collection, demisters were installed in the comon i sample suction headers for both the A and B train analyzers.
Sunmary of Safety Evaluation I The demisters were fabricated using code material and qualified welders. ,.
- 20. RI}DVAL OF IhTERNAL PARTS FR04 REC 04BINER SYSTD1 CHECK VALVES (SRI 130BB) .
Description of Change r The valve discs, pins, and all other loose internal parts were removed from check valves OG-51-1, and 2, located in the off gas recirculation lines. The purpose of removing the valve internals was to eliminate any source of flow restriction, sparking or mechanical friction. , NOTE: Subsequent testing proved that these valves were not the cause of hydrogen detonations and they were returned to nomal status. Summary of Safety Evaluation Since the recirculation valves are rarely open, the steam relief valves are rarely lifted, and the warmup bypass line flowrates are not sufficient to create significant backpressure, the probability of backflow is remote. If backflow did occur, it would be adding diluent steam into the off gas which would not constitute an unsafe condition. The off gas piping that would be exposed to such backflow of steam is designed for pressures arxl temperatures in excess of those which could be developed. 1-
- 21. CHANGE TO ION-SPARKING VALVE INTERNALS (SRI 1300)
Description of Change The plug material of recombiner train "A" and "B" flow control valves FCV ' 7489A and B, was changed from stellited stainless steel to aluminum bronze. i-The plug cages and seats of the off-gas pressure control valves, PCV 7496A and B were changed from stellited stainless steel to bronze. Check valves (OG 51-1 and 2) downstream of the pressure control valves were modified by , welding a silicon bronze alloy material on the strike point of the disc hanger. . These changes were perfomed to eliminate a possible source of off-gas ' ignitions by making the valves non-sparking under operating conditions. IUTE: A stainless steel plug was subsequently reinstalled in one of the i, flow control valves following investigations that determined the cause of previous off gas ignitions was catalyst in the inlet piping. { Summary of Safety Evaluation These changes enhanced the safe operation of the recombiner system. , ( IV-10 4 a
l
,' 1
- 22. FDDIFICATION OF DRAIN COLIJCT!?C TANK V-811 (SRI 130GG)
Description of Change A short ,ortion of the drain line from the 6" off gas pipe in the l Storage 3uilding to drain collecting tank V 811 was modified to increase the pipe size to 2" and to provide a continuously downward sloped line to the drain tank. 4 ~ Summary of Safety Evaluation The modified pipe meets applicabic code requirements.
- 23. CODE DEVIATIONS ON OFF GAS SYSTEM INLLT PIPING AS A RESLJLT OF REC 04BINER CXTALYST RIMNAL OPERATIONS (SRI 13010I) 4 Description of Change During recombiner catalyst removal operations on the portion of the modified off gas syst m between the air ejector discharge tie-in and the recom-biner inlet, several code pressure boundries were opened and subsequently welded closed. A code deviation was obtained to perform liquid penetrenent examination and radiograph of the new welds in lieu of a hydrostatic test.
Sumary of Safety Evaluation This deviation was accepted by the hiinnesota Chief Boiler Inspector.
- 24. RECOMBINER ClTTLLT HYDROGEN ANALYZER SAMPLE SYSTIM FDDIFICATION (SRI 130bt4)
Description of Change ! The recombiner outlet hydrogen analyzer sample systm moisture trap drains were routed to the flash tanks to improve moisture removal. Summary of Safety Evaluation The modification improves the perfomance of the hydrogen analyzers by , providing better moisture removal from the sample systm. The isolation f valve for the drain line meets the applicable code requirements,
- 25. INSTALLATION OF PERMANENT 'IIIERM0 COUPLES ON OFF GAS INLET PIPE (SR1 130NN)
Dcscription of Change Thennocouples were installed domstream of the off gas Flow Control Valves (FCV 7489 A and B) to provide indication of pipe wall temperatures in the control room. l Summary of Safety Evaluation l The thermocouples are mounted external to the piping and do not provide ( i any control functions. l l l IV-11 l . l
g,o
- 26. FDDIFICATION OF LEVF1 INSTIUMENTATION FOR DRAIN COLLirTING TANKS V 810 * "
AND V 813 (SRI 13000) Description of Change r The differential pressure indicating switches which provide level indica-tion and control for drain collecting tanks V-810 and V-813 were replaced with capacitance probes in external stand pipes and gage glasses. 7 i Ste.ary of Safety Evaluation
'Ihe modification does not alter existing alarm and control circuitry. The C i capacitance probe instruments utilize solid state relays with zener diodes ! , to limit the probe voltage to a maxinen of 10 vdc.
4
- 27. CAPPING TIIE INLET I!YDROGEN ANAIXZERS SAMPLE LINES (SRI 130QQ) I 4
Description of Change l-The sample lines for the modified off gas system inlet hydrogen analyzers connected to the air ejector discharge piping were capped to prevent off , gas leaking' into the turbine building. , Summary of Safety Evaluation The inlet hydrogen analyzers were not required' to be operable at this 1: tine since a " Worst Case" hydrogen concentrention signal was inserted into i the hydrogen inventory processors in accordance with Technical Specifica- r-l tion 3.8.E.3. l: NOTE: The inlet ifydrogen analyzers were subsequently deleted (sce item 7 , above). i=
- 28. FDDIFICATION TO OFF GAS C04 PRESSORS COOLING h'ATER PIPING (SRI 130SS)
I Description of Change i. The flow path of coolant through the off gas compressors was changed from ,: a series arrangement to two parallel paths to allow the first and second g. Stages of the compressors to operate at higher temperature, thereby minimiz-
- ing moisture problems. ,..
Summary of Safety Evaluation I There is no safety significance associated with this change. The coolant 0 system for the off gas compressors is not a safety system. b
- 29. hDDIFIED OFF GAS SYSflN CODE RECIASSIFICATION (SRI 130W) r.
I Description of Change '- The Modified Off Gas System was reclassified in accordance with NRC Regula- j tory Guide 1.26. The major changes were: l; f G 4 IV . i:
. , _ _ . _ _ . . ~ _ . ._ - _
r ! l
- 1. All of the recombiner system orginally classified to ASMB Section j
! III Class 3 was changed to AEC Quality Group D (Augmented). ! 1 i j 2. The Category I Seismic Classification of tic recmbiner systs te- ' j; tween valves OG 50-1/0G 50 2 and OG 551/0G-55-2 was eliminated. ; j 3. The storage syst s , with the exception of the tanks through their 1 i ' boundary valves was changed to AEC Quality Group D (Augmented) as l 1 defined by Regulatory Guide 1.26 ani ESTB Position I, dated October ;
- 9, 1974.
i l 4 The compressed gas storage tanks, through their boundary valves, were ! changed from ASME Section III, Class 2 to ASME Section III, Class . I 3. I 1 ! l Sumary of Safety Evaluation i i { The code reclassification is in accordance with Regulatory Guide 1.26. j ! 30. " ELIMINATION OF RECOUlINER SUBSYSTIN TRIP ON IIM OFF GAS IN1Er PRESSURE ! , (SRI 130XX) Description of Change .4 4
-i
! The low off gas inlet pressure trip of the recombiner subsystm was clinin- ) j ated. 1 i Summary of Safety Evaluation ' l The actual operating ranges of off gas flows and pressures are such that i the N 16 dose level in the reembiner building is not excessive at low off , gas pressures. The high off gas flow rate trip of the recombiner subsystm 4 provides protection in the event of failure of the off gas flow control valve to the full open position. 1 j
- 31. ELD 11 NATION OF THE OFF GAS SYSTD1 DISCHARGE FLOWMETER (SRI 130YY) l
. Description of Change
- The dff gas storage tank subsystem discharge flowmeter was eliminated, I
j Safety Evaluation { , The discharge flowmeter is not necessary since tank inventory control is
- achieved by maintaining the sum of the pressures for filling and dis-charging tanks below 285 psig ' Also, discharge flow rates can be easily calculated fra the tank depressurization rate.
4
'32. ELIMINATION OF OFF GAS STORAGE TANK RADIATION MONITORS (SRI 130AAA) l' Description of Change n
The Off Gas Storage tanks radiation monitors were eliminated. The'radia- ' tion monitors did not provide a satisfactory indication of. total activity of.the off gas, and thus provided no useful information. IV-13'
Summary of Safety Pvaluation - Tank samples are analyzed to assure that tle storage tank activity limit is , satisfied. 1
- 33. RE RCUTE OF DRAIN IUMP P 900A AND P 900B DISQlARGE PIPItC (SRI 130EEE)
Description of Change {. The discharge of Drain Pumps P-900A and P-900B was re-routed frm the turbine building equipment drain sump to tie turbine building floor drain sump to ' reduce the volume of water that is mixed with tie hig1 conductivity delay line drain water which must be processed by the radwaste system. Smmary of Safety Evaluation ( The new piping was designed and installed to the same requirments as the ' j orginal piping.
- 34. ADDITION OF DRAIN LINES IN OFF GAS EIUCTOR SlCTION PIPING (SRI 130FFF)
Description of Change f A low point drain line, with isolation valve and strainer, was installed on , the off gas eductor suction piping to facilitate moisture removal. Summary of Safety Evaluation The drain line was installed and tested in accordance with applicabic codes. I
- 35. IMERG12CY FILL AND DIS 01ARGE OF THE OFF GAS STORAGE TANKS USING S1DRAGE TANK DRAIN TRAP BYPASS LINLS (SRI 149)_
Description of Change , ~ The original concept for emergency fill and discharge of the gas storage i tanks was to use a path which included the gas sampling headers. Test , results indicated, however, that the flow restrictulon through the smaller sample lines was so excessive that the compressors would cycle repeatedly . and the design fill and discharge rates of 28 scfm could not be met. Thus the procedure was revised to use the drain trap bypass lines as r follows: , (a) The outlet isolation valve of the common drain trap discharge line p is closed, t (b) The disabled tank and an operable tank are connected in parallel via the trap drain bypass valves. (c) The fill and discharge valves of the operable tank are used to fill and discharge the two tanks in parallel, e k IV-14
. . . , _ , _ . - - _ ,_ . . ~ . _ . . , .
1
, Stanary of Safety LYaluation,,
i The procedure was demonstrated to be satisfactory even at flow rates above the design flow of 28 scfm. Existing piping and valves are used, therefore changing the procedure does not introduce any safety considerations different than the orginal concept. )
- 36. htNDtEt(f 0F IRRADIATED FUEL ASSIMBLIES INDIVIIUAL FUEL PINS AND UDIER IRRA-IHATEri cct4PONINrs WImIN 'IIE SPENT FUEL POOL (SRI 168)
Description of Change 4 The FSAR in Vol. II. Sec. X, page 10 2.3 states that ". . . and fuel being moved is at all times covered by a minimum depth of eight feet of water". Contrary to this statement, there are several movements of irradiated j components which may allow these components to come closer than 8 feet to the surface. These movements include, but are not limited to: a) placement of irradiated fuel on a fuel preparation machine and tio subsequent raising of this fuel to the full up position. J Irradiated channel inspection in a special channel inspection rack t b) mounted on the north side of the fuel pool, c) Reconstitution of leaking fuel assemblics (During this operation individual fuel pins may come to within 6 feet of the water surface.) d) Special fuel and component inspections. Suninary of Safety Evaluation An adequate amount of water to shield all radioactive ccrnponents is assured by the application of a) appropriate mechanical stops on selected eculp-ment, b) proper use of tools and c) carefully formulated and reviewed , procedures. Radiation Protection Specialist coverage is provided on the refueling floor whenever activities are b61ng conducted which involve or have tie potential to involve, personnel exposure. Radiation levels on the refulcing floor are constantly monitored by Four Area Radiation Monitors, (equipped with audible alanns) and two airborne activity monitors (also equipped with audible alanns). IV-15 J
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m - I B. LI,QUIC, EFFLUENTS y -- - - _ - - I NO LIQUID EFFLUENT RELEASES bERE PA0E CURING THIS REPORTING PERIOD. I I I I I I I I I I 3 I J F L ; o ,
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Co SOL 10 h ASTE SHIPPED , . A
- 1. TOT AL AMOLNT OF SOL IC k A STE SHIPPEC (CUBIC FEETI '
I 5 l JUL AUG SEP OCT NOV Dhc i TOTAL E ___ __ _ ___ _ _ _ _ _ _ ________ __ -- _ _ _ _ _ _ _ ___ g 0.0 1413.0 570.0 2043.C 3113.2 1500.0 8639.2 2 TOTAL ESTIMATED ACTIVITY (CURIESI JUL AUG SEF OCT NOV CEC TOTAL 0.0 1 E 7. 60 795.60 2.30 265.66 1863.00 3114 15
- 3. DATES AND DISPOSITICN 8 /12/75 NUCLEAR EhGihEERIAG CCNPANY, SHEFFIELD, ILLINOIS g 8/13/75 AUCLEAR E AG IhEER i hG CCPP ANY, SHE FF I ELD , ILLINOIS 8/14 /75 5 NUC L E A R ENG ihEER !hG CCP P ANY, MOORHEAD, K EN TUCK Y 8/26/75 NUCLEAR ENGlhEERihG CCPPANY, SHE FF I ELD , ILLINDIS 9/ 2/75 CHEM NLCLEAR INC., BARhhELL, SOUTH CARCLINA 9/ 4/75 AUC L E A R ENGIhEER ihG CC FP ANY, SHEFFIELD, ILLIh0!S
- 9/ 9/75 NUC L E AR EhG IhEER ihG CCFP ANY, SHEFFIELD, ILLINDIS 9/11/75 CHEP NLCLEAR INC., BARhWELL, SOUTH CAROLINA 10 / 7/75 NUCLEAR EhG lhEER !hG COPP ANY, SHEFFIELO, ILLINDIS 10/21/75 NUCLEAR ENGINEERihG CCPPANY. SHEFFIELD, ILLINDIS 11/ 3/75 CHEM NLCLEAR INC., BARhhELL, SOUTH C ARCLINA 11/ 5/75 g huCLEAR ENGINEERIhG CCNPANY, SHEFFIELO, ILLINOIS 5 l 11/12/75 NUCLEAR ENGlhEERIhG CCNPANY, SHEFFIELO, ILLINOIS 11/14/75 NUC L E A R ENGlhEERIhG CCPP/NY, SHEFFIELD, ILLINDIS i 11/19/75 hDCLEAR ENG!hEER!hG CCFPANY, SHEFFIELO, ILLINDIS -
t 11/25/75 CHEM NLCLEAR INC . , B ARhk ELL , SOUTH C ARCLIN A i 12/ 4/75 CHEM hlCLEAR INC., B ARhWELL, SOUTH CARCLINA ! 12/ 5/75 CHEM NLCLEAR INC., BARhhELL, SOUTH C AR OLIN A 12 /10 /75 NUCLEAR ENG lhEER i hG CCMP ANY, SHEFFIELD, ILLINDIS 12/11/75 NUCLEAR ENGINEER!bG CCFPANY, SHE FF I ELD , ILLINOIS 12/12/75 CHEM NLCLEAR INC., BARhWELL, SOUTH C AR CLIN A i 12/13/75 NUCLEAR ENGlhEERlhG COMPANY, SHE FF I ELD, ILLINDIS 12/15/75 CHEP hlCLEAR INC. , B ARhh ELL, SOUTH C AROLIN A 12/16/75 CHE M ALC LE A R I NC. , B ARhh ELL , SOUTH C AROLINA ! 12/19/75 CHEM NLCLEAR INC., BARhWELL,. SOUTH C AROLIN A 12/22/75 CHEM hlCLEAR INC., BARhWELL, SOUTH CAROLINA r i I I ! V-4
NOTE $s O CUANT IT IES REPORTED AS PERCENT CF TECHNICAL SPECIFICATION LlilT FOR GASEOLS FOLLOWING MANAERI EFFLUENT S HAVE BEEN CETERPihEC IN THE JULY 1 - SEPT 30 PERCENT CF AANUAL AVERAGE RELE ASE I LiklT ESTABLISHEC th THE ORIGINAL TE CFhlC AL SPECIF IC AT IONS ASSUMihG RELEASE RATE CONT!hlES AT THE i SAFE RATE FOR THE Eh TIRE YE AR OCT 1 - CEC 31 PERCENT CF MA*!MLN RELE ASE R AT ES 110 CFR 20 LIMITS) EST ABLISHED BY hRC TECFhlC AL SPECIF ICATICN CFANCES 12 Aht 19 THIS O!FFERENCE IN REPORTING PERCEhTAGE CF TECFAICAL SPECIFICATICh L IMITS h AS NECESSITA1ED BY CH ANGES TO THE TECHNICAL OCT0 DER L. SPECIFICATIONS 1 S T S. kHICH elC APE EFFECTIVE ON Bi l i t V- 5
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Table 1 provides a summary of the Radiation Environmental Monitoring Program for the Monticello Nuclear Generating Plant as required by Technical Specification Table k.8.1. The hi8hest activity monitoring location for each sample medium is included in Table 2. The analytical , results for each medium are summarized in the following sections with the six-month average activity levels for each location and medium presented in Table 3. Table k provides an explanation for each required sample that was not collected or analyzed. With the possible exception of milk, none of the sample media indicated any elevated radioactivity attributable to plant operation i during the period of this report. l Milk Milk, collected during the first month of each quarter f rom eight farms five or more miles from the plant sit e, and monthly from four I fams near the plant, was analyzed for iodine-131 and other gamma emitters. During the grazing season (May 30 to October 15), veekly an=ples from the four nearby farms were analyzed for iodine-131. All - iodine-131 deteminations were made utilizing a highly sensitive ion-exchange resin technique and gamma spectrometry (NaI). Quarterly, a I strontium-90 determination was performed on a sample from each farm. I I I VI-2
E . I VI. RADIATION ENVID0NMENTAL MONITORING Data presented in this report is primarily in summary and average form to allow more effective evaluation. However all of the individual sample analysis data (including garna scans) has been organized and tabulated in a reference document. This data document is available for review at the Northern States Power Company General Office. L E VI-1
I With the possible exception of the iodine-131 levels observed in milk from nine of the twelve farms sampled, no contribution from the plant operations could be detected. The iodine-131 levels detected are equal or greater than those observed at other locations during this six-month period but are consistent with levels observed during 197h. Aquatic Radioactivity I Gross beta and gamma emitter concentrations in the aquatic environment were determined by analysis of water, bottom sediments, algae, aquatic insects and aquatic vegetation from the Mississippi River and two lakes. Analyses were also performed on fish and clams taken from the Mississippi River, and topsoil from fields irrigated with water taken from the Mississippi River downstream of the plant. I Tritium concentrations were determined on _ monthly composites of the river water and on the lake water samples. Strontium-90 concentrations vere determined on the insect and algae samples. No increased concentrations of radioactivity attributable to plant operation vere detected in any of the samples. Ground Water Ground water samples, collected from two wells on the plant site and five other vells, did not show any measureable increase in radio-activity due to plant operation. VI-3
Terrestrial Deposition Surface deposition of radioactivity was measured by analysis of i precipitation, field vegetation, topsoil, and edible cultivated crop samples. All levels found were consistent with pre-operational data. I Airborne Radioiodine and particulate Radioactivity Air samples were collected weekly at eight locations surrounding , the plant site. Iodine-131 was collected using carbon impregnated air filters exposed in tandem with the particulate filter. The eight carbon filters were analyzed simultaneously for iodine-131 using gamma spectrometry (Hal). If iodine-131 van detected in the composite, the filters vere analyzed separately. The individual particulate filters were analyzed for gross beta activity and a monthly composite of all filters was analyzed for gamma emitters using gamma spectrometry (GeLi). All gamma and beta emitter concentrations were consistemt with data of previous years. Gamma Radiation The external radiation dose was measured at the fourteen locations required by the Technical Specifications as well as fifteen other sites using calcium fluoride thermoluminescent dosimeters (TLD's) that were housed in an aluminum-lead-tin shield which compensates for the detection i I VI L I I
f g . I over-response occurring from low-energy gam::.a radiation. To provide information on radiation levels that may be attributable to plant operation, the locations were grouped as " reference" or " control" (over 2.0 miles away from the Monticello stack)," indicator" (less than 2.0 miles from the Monticello stack), and "Monticello"(within the city of Fbnticello, Minnesota). The six-month accumulated gamma dose is 30.3 mr for all reference locations and 31.1 mr for all indicator locations yielding a net six-month dose of 0.8 mr. The highest dose was measured at location #1. The six-month measured net dose for this location, was h.8 mr above the average for all reference locations. This level is not considered to be attributable to plant Operations because station 1 is located 11.1 miles from the plant. These dose values show that the TLD network did not measure
! radiation levels attributable to plant operation during the period of this report. See Figure 1 for a graph of net gamma exposure vs. off-gas release rates from the plant.
Film Badge Results During the report period, none of the film badges shoved any
, ,eaeings above M- (1ess than 1e m1111,em).
I VI-5
, I Small Game Animals No concentrations of radioactivity attributable to plant operation were detected in the indicator or reference small game animal samples collected.
I f P
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VI-6 l
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_ _ _ _ _ - _ _ . _ _ _ l l
ll Table 1 SUhNARY OF RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Monticello July through December, 1975 Number of Radiation Number of Number of Locations Above Attributable tc Medium Locations Samples Backg.ound Plant Operatior River Water 3 81 0 0 I Lake Water Well Water 2 7 1h h 0 0 0 0 Precipitation 2 12 0 0 (Terrestrial Deposition) Lake & River Bottom h h 0 0 Sediment I ilankton, Algae, or Insects k 10 0 0 Aquatic Vegetation 4 8 0 0 Cla=s 2 2* 0 0 { Fish 2 7* O O Milk (Regions 1-h) 8 16 ** ** l Milk (Nearby dairy farms) h 72 ** ** Topsoil 6 6 0 0 Vegetation 3 3 0 0 1 Edible Cultivated Crops b k 0 0 Air Samples (Particulate) 8 202" 0 0 Air Samples ( Airborne Iodine-131) 8 202* O O TLD's 29 69h* O O Film Badges 38 229 0 0 Small Oame Animal 2 2 0 0 1
- Cee Table #h I
**See Narrative Section i
VI-7 I
_ _ _ _ _ _ __ _ _ _ _ - _ . - - - - - - -- -" - - ~ ' Table 2
SUMMARY
REPORT OF RADIOACTIVITY IN THE ENVIRONIEh"I HIGHEST ACTIVITY IOCATION FOR EACH MEDIUM Monticello (July 1, 1975 - December 31, 1975) Distance Direction AcWhy Media Miles Degrees Units High Low Average Analysis River Water (St. Paul Water Intake) pCi/1 16 h T.T Gross Beta Lake Water (Locke Lake) 5.6 292 " 10 " " 13 12 Well Water (Plant Well #1) " h " " 5 h.5 Precipitation (Terrestrial 2 LO 130 pCi/m h000 h20 1800 " " Deposition) (MDH) Bottom Sediment (Big Lake) 4.6 90 pCi/g 56 - 56 " " Aquatic Insects (Mississippi River- " 29 23 26 " " Downstream) Alga (Mississippi River-Upstream) " 77 _ 77 a a Aquatic Vegetation (Mississ 1ppi ' " 76 y River-Upstream) 60 68 " " Fish (Mississippi River-Downstream) - " 8.5 6.6 T.T " " Milk (Nelson) 2.2 2Th pCi/1 h.70 <0.10 0.T8 131 Milk (Dvinger) 13.0 335 " 7 19 3 11 13T Cs Milk (Peterson) 2. 3 111 " 31 10 20 90 Sr Topsoil (Vegetation Field #1) 0.8 100 pCi/g 57 - 57 Gross Beta Field Vegetation.(Field #3) 1.1 280 " h9 - h9 " " Air Pa--ticulates (Station #2) 3.8 33h pC1/m3 0.086 0.011 " " 0.oh3 TLD's (Station #1) 11.1 306 mr/h weeks 6.0 h.8 5.h camma Small Game Animal (Flesh) (On - - pCi/g Plant site) 0.22 - 0.22 137Cs Edible Cultivated Crops 0.7 160 22 - 22 Gross Beta ,
7 7--~g--- g- g g~g g g~g g 3 gn g j n Table 3
SUMMARY
REPORT OF RADIOACTIVITY IN THE ENVIRONMENT Monticello 6 MONTH AVERAGE ACTIVITY LEVELS Sample / Location Gross Beta Cesium-137
- Tritium River Water pCi/1 pCi/1 pCi/1 Upstream of Plant (1000 ft. ) 7.6 1 0.7 (1.8 470 ! 100 Downstream of Plant (1000 ft. ) 7.2 0.7 1.1 0.2 360 ! 80
- St. Paul Water Intake 7 7 1 0.7 0 9 1 0.2 360 1 90 Lake Water pC1/1 pCi/1 pC1/1
< Big Lake (Mitchell) (h.6 mi @ 90 ) 10 2 < 1. 8 h50 140
[ Locke Lake (5.6 mi 6 292 ) 12 !3 < 1. 6 a 560 130 Well Water pCi/1 pCi/1 pCi/1 Plax Well #1 (On site) h.5 ! 2 5 < 1. 8 320 180 Plant Well #2 (On site) 2.5 1.8 < 1. 5 270 ! 160 City of Monticello (3.2 mi @ 128 ) 3.2 2.1 ( 1. 4 270 160 Gauthier (1 3 mi 0 130 ) 43.0 < 1. 7 460 ! 180 Schultz (1.2 mi @ 249 ) 2 5 ! 1.8 < 2.2 < 2ho Swanson (1 5 mi @ 277 ) < 3.0 < 2.1 h20 1ho Trunnel (0.3 mi @ 21h ) 3.0 2.0 42.0 220 ! 150
*All other isotopes were below the Ic.rer limit of detection unless noted otherwise.
i 1
- - - . . - -~ --- - - -
tble 3 (continued)
SUMMARY
REPORT OF RADIOACTIVITY IN THE ENVIRONMENT Monticello 6 MONTH AVERAGE ACTIVITY LEVELS Sample / Location Gross Beta Cesium-137*' Bottom Sediment pCi/gm pCi/gm Big Lake (Mitchell) (h.6 mi e 90 ) 56
- 3 27 2 0.1 Locke Lake (5.6 mi 8 292 ) 22 2 1.1 1 0.1 Mississippi River, Upstream (1000 ft.) 23 1 2 0.085
- 0.013 Mississippi River, Downstream (1000 ft.) hh 1 2 0.069 1 0.010 Aquatic Vegetation pCi/gm' pCi/gm Big Lake (Mitchell) (h.6 mi 8 90 ) h7 2 1.04 0.06 h Locke Lake (5.6 mi 8 292 ) 58 2 0.12 0.03 5 Mississippi Rivar, Upstream (1000 ft.) 68 2 0.15 0.02 Mississippi River, Downstream (1000 ft.) 60 1 2 0.11 0.02 Clams pCi/gm pCi/gm Mississippi River, Upstream (1000 ft.) 2.31 0.3- < 0.022 Mississippi River, Downstream (1000 ft.) 2.11 0.2 4 0.024 Fish (Flesh) pCi/gm pCi/gm Mississippi River, Upstream (1000 ft.) 7.6 0.1 0.027 i 0.009 Mississippi River, Downstream (1000 ft.) 7.7i 0.1 0.033 1 0.00h Topsoil pCi/gm pCi/gm Vegetation Field #1 (0.8 mi 8 100*) 57 1'3 0.hD i 0.02 Vegetation Field #2 (0.7 mi 8 198 ) 53 3- 4 0.01 ,
Vegetation Field #3 (1.1 mi 8 280 ) 49 33 0.h0
- 0.02 Dechene Potato Co. (4 7.mi @ 118 ) h9 12 0.13 1 0.02 .
Dechene Potato Co. (h.9 mi 8 1150) 49 13' O.2h
- 0.02 #
Ewing Potato Co. (5.7 mi 8 111 ) 51 13 0.16 1 0.02 - eAll other isotopes were below the lower limit of detection unless noted otherwise.
-M' W W W W M M M M M M M M M M M M W W
M M M M M M M M M M M M M M M M M W- gr
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Table 3 (continued) ' SUM 4ARY REPORT OF RADIOACTIVITY IN THE ENVIRON!ENT Monticello 6 MONTH AVERAGE ACTIVITY LEVELS Sample / Location Gross Beta Cesium-137* Iodine-131 Edible Cultivated Crops pCi/gm pCi/gm Dechene Potato Co.-Potatoes (4.7 mi @ 118 ) 7 1 40.005 0.031 i 0.018 i Dechene Potato Co.-Corn (4.9 mi @ 115 ) 11
- 2 Ewing Potato Co.-Potatoes (5.7 mi @ 111 ) 1111 40.004 i Brooks Farm-Corn (0.7 mi @ 160 ) 22 2 0.027 ! 0.004 Vegetation pCi/gm pCi/gm pCi/gm Vegetation Field #1 (0.8 mi @ 100 ) 31 1 3 0.13 0.02 0.51 1 0.04 i Vegetation Field #2 (0.7 mi 8 198 ) 31 3 0.08 0.01 0.91 0.03 Vegetation Field #3 (1.1 mi @ 280 ) 49 i 3 0 30 0.02 0.40 t 0.0h Air Particulates & Airborne Iodine-131 pCi/m3 pCi/m3 Station #1 (11.1 mi @ 306 ) 0.032 1 0.001 0.016 0.003 Station #2 (3.8 mi 8 334 ) 0.043 0.001 0.010 0.005 Station #3 (6.h mi @ 290 ) 0.022 0.001 0.008 1 0.002 Station Wh (0.6 mi @ 100 ) 0.038 0.001 0.013 1 0.002 Station #5 (10.h mi @ 106 ) 0.021 0.001 < 0.013 Station #6 (12.4 mi 8 133 ) 0.032 i 0.001 ( 0.013 Station #7 (8.8 mi @ 39 ) 0.031 i 0.001 < 0.013 Station #8 (10.3 mi 8 227 ) 0.017
- 0.001 0.008 2 0.002 Call other isotopes were below the lower limit of detection unless noted otherwise.
t' . --
Table.3 (continued)
SUMMARY
REPORT OF RADIOACTIVITY IN THE ENVIRONFENT Monticello 6 MONTH AVERAGE ACTIVITY LEVELS Sample / Location Gross Beta Cesium-137* Iodine-131 Otherd's Strontium-90 Tritium i Milk pCi/1 pCi/1 pC1/1 i i Dwinger (13.0 mi @ 335 ) 11.0 i 1.0 0.3h 0.09 12.0 1 1.0 Kirchenbauer (11.5 mi 8 323 ) 5.8 1 1.4 0.31 0.10 16.0 1 1.0 Kotilinek (5.6 mi @ 230 ) 5.1
- 1.h 0.17 0.07 5.6 1 0.6 Vandergon (8.3 mi @ 247 ) 53115 40.13 3.2 ! 0.5 Holland (8.1 mi 6 199 ) 3.2 1 1.7 0.17 i 0.10 3.0 i 0.5 Hopkins (7.6 mi 8 193 ) 4.8
- 1.6 4.0.19 2.8 i 0.5 Becker (10 mi @ 130 ) h.h 1. 7 - 0.20 1 0.10 3.2 1 0 5 f Vetsch (9.4 mi @ 128 ) 3.4 i 1.6 40.19 h.6 05 Y Nelson (2.4 mi @ 269 ) 5.1 1 0.9 0.78 0.08 7.6 i 0.6 01 eon (2.5 mi @ 2h ) 6.0 09 0.72 1 0.0h 11.0 2 2.0 Peterson (2.3 mi 8 111 ) 8.6 09 0 53 0.0h 20.0 1.0 Shovelain (3.0 mi 8 250 ) 5.6 ! 0.8 0.35 0.03 14.0 1 1.0 Precipitation (Terrestrial deposition) pCi/m2 pCi/m2 pCi/m2 pCi/m2 pCi/m2 pCi/m2 On-site Station (0.6 mi @ 100 ) 1300 1 100 43 ! 4 14 5 150 30 13 1 250 110 Minn. Dept. of Health (40 mi 6130 ) 1800 100 43 h 419 120 10 .12 1 1 220 ! 120
~
Call other isotopes were below the lower limit of detection unless noted otherwise.
?
ggg g g g M M--M M M M M M M M M M E E[ ' E
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~i Table 3 (continned)
SUMMARY
REPORT OF RADIOACTIVITY IN THE ENVIRONMLiT Monticello 6 MONTH AVERAGE ACTIVITY LEVELS Sample / Location Gross Beta Cesium-137
- Other[' s, Strontium-90 Algae pCi/gm pCi/gm pCi/gm pCi/gm Big Lake (Mitchell) (4.6 mi 8 90 ) 49 3 29 *09 0.6 0.7 40.2T Locke Lake (5.6 mi 8 292 ) 73!3 1.9 0.7 0.8 ! 0.5 0.1h 0.10 Mississippi River, Upstream (1000 ft) 77 3 1.3 0.8 13 1.2 0.h8 0.14 Mississippi River, Downstream,1000 ft) 59 ! 3 1.2 1 0.9 2.2 2.0 0.74 0.16 Aquatic Insects pCi/gm pCi/gm pCi/gm pCi/gm Big Lake (Mitchell) (h.6 mi @ 90 ) 812 2.S 0.3 0.7 0.6 2.80 2 0 50 Locke Lake (5.6 mi 8 292 ) 13 2 1.8 2 0.3 1.1 ! 0.5 0 92 ! 0.23 Mississippi River, Upstream (1000 ft) 2612 3.6 0.3 5.3 ! 1.8 0.19 0.10 Mississippi River, Downstream (1000 ft) 28 ! 2 1.2 0.3 3.6 2.k 0.36 0.11 Air Particulate (Composite) pCi/m3 pC1/m3 Eight Station Analysic (8.8 i O.7)XE-h (3.8 0.2)XE-3 Small Game Animals pCi/gm pCi/gm Indicator (On site) Flesh 0.22 1 0.05 0.0h 0.03 Liver 0.31 ! 0.05 0.23 1 0.0L Reference (12 mi @ 258 ) Flesh 0.14 1 0.03 0.22 1 0.10 Liver 0.33 1 0.0T 0.11 0.11
#All other isotopes were below the lover limit of detection unless noted otherwi se.
M. - . qeg
Table 3 (continued) ' iSUERRY REPORT OF RADI0 ACTIVITY IN THE ENVIRONMENT Monticello 6 MONTH AVERAGE ACTIVITY LEVELS (6-11-75 to 12-23-75) Sample / Location Gamma Dose
- I Thermoluminescent Dosimeters mr/6 months REFERENCE I Station #1 (11.1 mi. @ 306 ) 35.1 Station #2 (3.8 mi. @ 334 ) 29.9 2.0 Station #3 (6.h mi. @ 2900) 29.9 g
Station #5 (10.4 mi. @ 106 ) 31.8 Station #6 (12.h mi. @ 133 ) 32.5 Station #7 (8.8 mi. @ 39 ) E 31.2 3 Station #8 (10 3 mi. @ 227 ) 29.2 Station #h2 (2 7 mi. @ 345 ) 29.2 t 1.3 Station #h3 (2.6 mi. @ 10 ) Station #44 (2 9 mi. @ 27 ) 31.8 f 0.6 28.6 1 0.6 l Station #h5 (3 5 mi. G 34h ) 28.6 Station #h7 (8.1 mi. @ 357 ) 29.2 t 0.6 Station #48 (8 7 mi @ 10 ) 30.6 Station #55 (b.7 mi. 6126 ) 31.8 INDICATOR I Station #h (0.6 mi, @ 100 ) 29.9 0.6 Station #13 (0.6 mi. @ 242 ) 29.2 Station #1h (0.4 mi. @ 1550) 31.2 0.6 Station #15 (0.8 mi. @ 100 ) 28.6 Station #16 (0 9 mi. @ 321 ) 30.6 Station #17 (0.7 mi. @ 250 ) 31.2 1 2.0 Station #18 (1.7 mi. @ 110 ) 29.2 Statiori #2h (0.6 mi. @ 208 ) 32.5 0.6 Station #25 (0.8 mi. @ 157 ) 3 31.2 0.6 3 Station #26 (0.8 mi. @ 1h0 ) 32.5 1.3 ' Station #27 (0.9 mi. @ 104 ) 31.8 0.6 Station #50 (1.8 mi. @ 132 ) MONTICELLO 28.6 i 1.3 l Station #51 (2.7 mi. @ 132 ) 30.6 0.6 1 Station #53 (3.h mi. @ 130 ) 29 9 0.o Station #54 (3.0 mi. @ 122 ) 30.6 ! 0.6
- Includes background I
I l li VI-lh g t n l 1 k% . ,
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] Table 4
SUMMARY
REPORT OF RADI0 ACTIVITY IN THE ENVIRONMENT Monticello Samples Not Collected Medium Location Period Reason
- Clams Mississippi River attempted sampling with Upstream & Downstream 4th quarter no success
. Fish, game Mississippi River attempted sampling with Upstream 4th quarter no success Air Filters Station #3 7/1-7/15/75 Equipment malfunction Air Filters Station #4 7/22-7/29/75 Personnel error
- Air Filters Station #5 7/1-7/8/75 Equipment malfunction Air Filters Station #6 7/29-8/5/75 Personnel Error 4
Air Filters Station #8 7/1-7/8-75 Equipment malfunction $ TLD's Station #1 6/11-7/8/75 lost in field TLD's Station #3 9/3-9/30/75 Lost in field TLD's Station #8 7/9-8/5/75 Lost in field TLD's Station #15 i 11/26-12/11/75 Weather / road conditions precluded collection. TLD's Station #16 10/29-12/11/75 Weather / road conditions precluded collection. TLD's Station #17 10/29-12/11/75 Weather / road conditions precluded collection. TLD's Station #18 8/6-9/2/75 Lost in field I i VI-15
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R A010 ACT IVE EFFLUE AT RELEASES FCR Iv3 .p. ..-- - - - . _ - - - - - - - - 6/30/15 THROUGH 12/31/75
--------=_ _
I 'A . GASEOL S EFFL UENT S
- 1. NOBLE GAS RELEASES A. TOTAL 09055 R ADIC ACTI%ITY (CURIES)
JLL AUG SEP OCT NOV OEC TOTAL 19669. 21151. 10908. 290. 1072. 1364. 54453. B. MAXIMUM G9 CSS RADIDACTIVITY RELE ASE IN ANY ONE HCUR PER IOD (UCI/ JUL AUG SEP OCT NOV CEC TOTAL 15090. 4(520. 69160. 240. 1930. 1345. 69160. C. TOTAL GROSS RADICAC11%IT). BY SUCLICE (CURIES) JUL AUG SEF 1 OCT NOV CEC TOTAL 14900. 15455. 8689. 290. 166. 59500. IXE133 XE135 KR85M 38. 7.
- 45. 18. O. 8.
1.
- 12. 120.
- 8. 3. O. 1. 2.
i KR88 15. 19. E. O. 22. KR 87 29. 5. 8. 55. XE138
- 29. 13. O. 8. 16.
406. 573. 170. 94. KR90 O. 209.
- 16. 19. 7 O.
366. 1724. XE 139 47. 47. 31. 120.
- 56. 20. O. 52.
KR89 461. 555. 202. O.
- 73. 248.
XE13 7 602. 245 366. 1830. 1085. 264. O. 310. XE135M 36. 43. 16. 462. 2724. O. 17. 135. KR 83 M 5. 6. 2. O. 24. XE133M 64. 2. 3. 19.
- 72. 36. O. O. O.
XE131M 65. 103. 172.
- 50. O. O.
KR 85 2979. 3083. O. 217. 1410. 0-. O. O. 7472. U. PERCENT OF T ECHNIC AL SPEC IF IC ATICb LIFIT JUL AUG SEF OCT NOV DEC TOTAL 5.47 6.94 2.95 0.07 0.66 1.34 2.92 L V-1 l L ___ . _ _ _ _ . _ - _ _
t IOOINE RELEASES 2. h, 4
, A.
s0TAL IODINE RADIDACTIVY EY NLCLICE (CURIES) JUL AUG SEP OCT NOV CEC
- TOTAL I-131 1-133 0.738 1.765 0.579 0.652 0.033 0.003 0.009 Ii '
1.534 0.359 0.000 2.013 1- 135 1.048 0.017 0.055 3.731 0.898 0.341 0.0 0.012 0.022 2.321 1 1 B. PERCENT OF T ECHNIC AL SPECIFIC ATIOh LIFIT
; JUL AUG SEP OCT NOV CEC TOTAL 244.57 192.01 132.60 0.42 0 04 0 10 95.27
- 3. PARTICUL A TE RELE A SE S A. CRCSS RADIO ACTIVI1Y I EET A Ah0 GANP A) (CURIES)
JLL AUG SEP OCT NOV CEC ! TOTAL I.72E-01 9. C 8 E- 0 2 4.56E-02 1.28E-02 4. 86 E-0 3 9.64E-03 3.36E-01 B. GROSS ALPHA R ADIO AC11 %IT) (CURIESI i JUL AUG SEF _ OCT NOV OEC _ _ TOTAL l ___
- 2. SiE-04 1.25E-04 1. 80 E- 05 1.90E-04 1 27E-05 8.02E-06 6.10E-04 i
C. TOT AL GROSS R A0!O ACTIVITY RELE ASED CF NUCL10ES hl f b HALF L IFE GRE ATER TF AN 8 DAYS (CLRIES) __=- l JUL AUG SEF OCT NOV OEC TOTAL 3.80E-02 2.42E-02 3. 51 E-02 1.76E-02 3.10E-03 3.21E-03 1.21E-01 D. PERCENT OF TECHNIC AL SPE CIFIC ATION L 's T i JUL AUG SEP _ _ - - - - OCT NOV DEC i TOTAL 13.11 7.45 5.25 t.72 1.10 0.49 5.72 V-2
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