ML20024B202
| ML20024B202 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 02/24/1975 |
| From: | J. J. Barton, Toole R GENERAL PUBLIC UTILITIES CORP. |
| To: | GENERAL PUBLIC UTILITIES CORP. |
| References | |
| TASK-*, TASK-GB GPU-2480, NUDOCS 8307070314 | |
| Download: ML20024B202 (12) | |
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AEC CCCU.E*li REVIEli Plant / Unit The attached AEC document has been reviewed for test program and design modification requirements for the above Plant / Unit.
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DOCUMENT:
Operating Experience, dated:
1 Current Events - Power Reactors, dated: /2 7'/
e Other
, dated:
Rem' w of t e attached document has concluded.that no action is required.
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r 73 ar up & Test ips(ager
, Date/
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2 - 2'I - 7C-Test S0jerintencent.
Date Review of the attached document has concluded that action is re-red by:
Problem Report (s) has/have been issued.
Startup 4 rest Manager Date
/
Test Superintencent Date W
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DISTRIBUTION:
R.W. Howard, Jr.
W.T. Gunn E.D. Mc" vitt J.E. Kunkel M.A. Nelson
..o.g J.T. Faulkner File 449 8307070314 750224 d
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w POTE'iTIALLY SIGNIFICANT EVENTS SELEt.a.u FROM REFORTS SUEMITTED TO THE UNITED STATES NUCLEAR RECULA! CRY CCMMISSION AS OF:
DECEMBER 1974
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res:d 1-J UNCCUPLED CONTROL RODS
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During startup of Unit No.1 of the Dresden Nuclear Power Station after 9lll5' a 1973 fall refueling outage, instru=ents did not verify that four control-rods were properly coupled to their control rod drives (CRD's), so each af fected control red was fully inserted and the control red drive electrically
- j disarmed and re=oved from service. Af ter reactor shutdown on August 31, 1974, it was deter =ined that each of the suspected control rods had not been feoperly coupled in the 1973 refueling outage; they were found lodged between their associated fuel asse=blies. No damage was noted.
I-It was concluded the control rods beca=e uncoupled because procedures followed sequence.during the refueling outage were not perfor ed in the proper Although a satisfactory pull test had been completed for each of the eighty control rods in the core, the test was completed prior to loading the four associated fuel aesemblies. As a result, it was possible for a control rod to rotate 90* and become unlatched from the control map rod drive coupling spud.
The minimum friction pressure required to begin movement of a CRD with a
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new seals is approx 1=stely 48 psig with control rod loading. Without control rod loading, approximtely 30 psig of pressure is required.. None k
of the CRD's soved at friction pressures belce 48 psig. However, because m
1 friction testing was completed prior co' scram testing, the uncoupled a
h control rods had not been given the opportunity to dimensionally separate
}P from their respective CRD's.
The control rods were s1= ply pushed by the CRD to provide the same total frie:1onsi resistance =easure=ent as a properly coupled CRD.
Following completion of the friction tests, scram testing was initiated.
With this test, the uncoupled control rods were thrust into their associated fuel assemblies at approxi=ately 4.5 ft/sec. The control rods displaced approx 1=ately 1.5 to 2.0 inches af ter leaving the control rod drive spud. Each control rod was held in the fully inserted position at four poincs because of the geometric configuration of the assembly base i'
between the orifice and the control rod hub.
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.7-e 2-A The cold critical worths of the four control rods did not reach the peak fuel enthalpy limit during reactor operation.
When it was suspected that the control rocs were not properly latched, and the four CRD's were fully inserted, electrically disar=ed, and re=oved from service, a daily control rod verification surveillance test was initiated. The re=aining seventy-six CRD's were limited to =ove=ents of three notches or less without verification of response of nuclear instru=entation during =ove=ent.
With these precautions, there was no danger to the health and safety of the public. No other CRD abnor=nlities were noted during subsequent reactor operation.
After reactor shutdown, to assure that the control rods were properly coupled to the CRD, each of the eighty control rods was withdrawn one notch on an individual basis and =achanically pulled back into position.
Following the pull test, the CRD's were friction tested. All CRD's exhibited pressures in excess of 60 psig. The reactor was brought critical on October 16, 1974 af ter a 46 day =aintenance outage.J-j OFF GAS EXPLOSIONS 5
Quad-Cities 1 & 2 a
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Unic No. 1 of the Quad-Cities Nuclear Power Station was operating at 520
.We and Unit No. 2 was at 620 We; both units were increasing power at the race of 3 We/hr. Several people in the vicinity of the =ain chl=ney in the off gas filter building heard an explosion, and five alar =s were received in the control room to indicate a problem in the off gas Shortly thereafter, the =easurec:ent of =ain stack radioactivity syste=.
increased from 30,000 to 100,000 uCi/sec, indicating an off gas detonation had occurred; station procedures for an off gas detonation were followed.
Approxi=ately 30 =inutes later, the =ain stack radiation =onitors decreased to pre-incident levels and the area radiation =onitor alar =s cleared.
The off gas explosion was caused by construction personnel grinding through an unpurged and i= properly isolated section of the off gas piping.
The personnel working on the job were under the i=pression that
- the piping had never been in service, and thus, would not cont 2in off gas, Specifically, the contributing factors leading to the detonation were i
failure of construction personnel to notify Operations of the work being
- done, failure to isolate the line properly before beginning work, and failure to purge the line before beginning work.
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The detonation resulted in ruptures of the steas jet air ejector rupture disc and a 2-inch rupture disc on the off gas discharge pipe upstrea= of the holdup line. One off gas filter was demolished, and fire ds= age occurred to another.
I==ediately following the incident, it was made clear to personnel that standard station procedures could have prevented this incident.
There were no personnel injuries, and environmental i=plications on the health and safety of the public were mini =al.2 Soil, vegetation and air particulate samples collected on the day of the detonation indicated plant operations were a possible but not probable source of the measured radioactivity. However, gross beta radioactivity concentrations in the air particulate filters did not indicate the
=easured radioactivity was attributable to station operations.3 Dresden 3 Unit No. 3 of the Dresden Nuclear Power Station was ' operating at an electrical load of 190 W when one of the off gas recombiners was valved into service. A noise resembling a water ha==er or a detonation was heard; stack gas activity increased and vacuum began to drop. An alternate steam jet air ejector was valved into service, but vacuus continued to decrease, and it was necessary to trip the turbine =anually.
Following the turbine trip, a reactor scram occurred.
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revealed the noise was a detonation in the off gas system. The off gas l
filter was da= aged and the air ejector rupture disc had blevn.
Apparently, the off gas detonation occurred when the recombiner outlet valve was opened; when the valve plug came off the closed seat, a spark apparently caused the detonation.
u Pl During and after the detonation, the safety of the plant and public was not in jeopardy. At no time was the technical specification release i
Ll limit of 700,000 uCi/sec exceeded. It was calculated that, for a two hour period, the release reached approx 1=ately 640,000 uC1/sec. Isotopic analyses performed on upvind and downwind air sa=ples, and soil and grass samples did not indicate any radioactivity attributable to this occurrence.4 TIP BALI. VALVE FAILURE I
l While Unit No.1 of the Quad-Cities Nuclear Power Station was operating at a power level of 735 We, the reactor operator discovered one of the I
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y Ag ball valves for the traversing incore probe (TIP) =achine in the open g
position. The manual control switch failed to close the valve and, 4
after an unsuccessful atte=pt to repair it in place, the ball valve was b
declared inoperable and an orderly shutdown of the reactor was initiated.
il lk The ball valve was re=oved, partially dismantled, and inspected for an j?
accumulation of the dry type lubricant used in the TIP tubes; a 4:ause of previous ball valve failures. No accu =ulation was found. When the j :
f valve was actuated on the bench; it was discovered that the limit switch M
for the closed valve position was not actuating. All =oveable parts of
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the valve were cleaned and lubricated, the spring tension on the valve g
was re-adjusted, and the li=it switch was adjusted. The ball valve was a
re-installed and functionally tested satisfactorily.
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It was not known how long the ball valve had been in the op'en position; k_
the latest recorded operation of the TIP =achine was three days earlier.
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Under accident conditions, with the drywell pressurited to 62 psig, the 6
flow through the open penetration would have been 72.1 SCFM.
Since the 5
standby gas treatment system (SBGT) would start automatically in the h
event of a loss of coolant accident, this leakage would have 'seen y
processed through it.
A 72.1 SCni represents only a s=all percentage of h
the rated flow of the SBGT system. Therefore, there would have been no 7;
significant amount of radioactive materials released, and the public j t health and safety would not have been endangered.5 i
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[ r DiPROPER REASSDiBLY OF CONTARDiENT ISOLATION VAL 7E i
l At Unit 2 of the Three Mile Island Nuclear Station, the split wedge gate
! I assembly of the prdg sampling contain=ent isolation valve was discovered 3:
missing. The valve had been disassembled about five =enths earlier to 2
replace a leaking bonnet gasket, and the work authori:ation detailed only j L valve body to bonnet gasket replacement. It did not specifically h
require personnel to check the valve internals prior to reassembly. The LT valve gate assembly was not replaced on the valve stem when the valve 3
was reassembled, and without it, the isolation valve was inoperable.
5 The valve was reassembled with a new gate assembly and leak rate tested.
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7 Station personnel now check valve internals prior to reassembly to verify a
che presence and correct orientation of all valve internals prior to Q
valve reassembly.
_7 During the period the pri=ary sampling containment isolation valve was i
inoperable, other valves would have provided the necessary isolation had there been an actuation of the Engineered Safeguards System. Therefore, I
the inability of the primary sampling contain=ent isolation valve to T
function did not constitute a threat to the health or safety of the public.6 7
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- MANUAL OVER-TORQUDIG OF VALVE During a maintenance test on a contain=ent cooling valve at Unit No. 1 of the Millstone Nuclear Power Station, the valve was operable for only two-thirds of its fun stroke travel because of a corsional bow in the valve stem.
The valve is contro n ed by a Limitorque =otor operator with a 40 to 1 driving gear ratio. Overtorquing during =otor operation was suspected, but the set points of the torque switches operated properly. This led to the conclusion that the over-torque condition resulted from excessive force when the valve was closed manually.
All operations and maintenance personnel were cautioned about using hand wheel wrenches on motor operated valves because of the danger of a damaging over-torque condition through the gear ratios on Limitorque operators.
The plant was in the cold shutdown condition, so there was no ha:ard to the health or saf ety of the public.7 UNPLANNED RELEASES OF PREL\\RY *JATER At Zica Generating Station, both Units 1 and 2 pri=ary water storage tanks were discovered to be overflowing. Water from the boric acid monitor tank was being pumped into the storage tanks; the level of the storage tanks had not been checked before starting the operation.
Approximately 500 ganons of primary water from the Unit 1 tank and 100 ganons of water from Unit 2 tank were spilled to the ground.
The initial action was to stop pumping the water to storage and to lower the level of the storage tanks by pumping the contents to the lake discharge tank. A gravel dike was built around the storage area to limit the spread of water, and versiculite was used to absorb as =uch spinage as possible. The vermiculite and top one-inch of ground were then conected and sealed in barrels. A survey of the area revealed no detectable activity above a background measurement.
It was estimated that approximately 250 ganons from Unit 1 and a small amount of liquid from Unit 2 seeped into the sand and entered the water table approximately 150 feet from the shoreline of Lake Michigan. The 3 uC1/cc of tritium (2.87 calculated release concentration was 8.6 x 10 3 uC1/cc times the 10 CFR 20 unrestricted release limit) and 6.9 x 10 7 uC1/cc limit if Ra 226 and 228 are of gross beta gamma (below the 10 not present; these nuclides were not detected). Samples of Lake Michigan water were taken five hours after the release and, later, an additional sample was taken at the County water supply point at the lake the following 9
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Wahia :ees E3G me sE day. The release had no detectable effect on the lalde water, hence 8
%j public health and safety was not affected.
F CRACK IN TORUS /DRWELL PURCE LINE I.
p=~ ode and the depell was open for repairs
% nile Unit No. 3 of the Dresden Nuclear Power Station was in the shutdown
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conducted on the boundary for=ed by air-operated valves on the corus/drpell urge piping. The test involved pressuri:ing the piping to 50 psig and
=enitoring pressure decay. A rapid decay was noted; investigation revealed a crack in the 13-inch diameter steel torus /drywell purge line.
9 The crack extended through the pipe wall, for approxi=ately 170 degrees M
around the lower half of the pipe, running through the weld joint for a ERT 3/4-inch bushing that served as a flow test line connection.* A si=ilar bushing, located on the opposite tide, was plugged and unused. The crack followed the are joining these two bushings.
The cause of the pipe crack is unknown..A fifteen-inch long seg=ent containing the crack was cut frc= the line and a new pipe section was welded in place; the welds were radiographed.
Thirty other welds on the sa=e piping were magnetic particle tested and found sound. The corresponding pipe on Unit No. 2 was visually checked and pressuri:ed with air; no visible or audible leaks were detected. A subsequent =agnetic particle check of 34 welds did not disclose any 3
indications of cracking. (Three surface indications, which disappeared i
with filing, were noted).
Wile the reactor was shut down and with the dryvell open, there was no danger to plant personnel or to the public. Operation with this pipe 3
crack would have posed no i==ediate danger. However, in the event of a W
serious accident involving drywell and corus pressurization, the statistical a
probability of a release into the secondary contain=ent would have been increased because of lack of redundancy in the isolation valve lineup.9 FAILURES OF 30RIC ACID TRANSFER PUMP SHAFT l
9 l
On August 15, 1973 at Unit No. 2 of the H. 3. Robinson Plant, the "3"
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boric acid transfer pu=p shaft, Crane che=pu=p Model GE-20K, broke at i
che i=peller while recirculating the Boron Injection Tank. After replacing N
the rotor, bearings, front bearing housing and impeller, the pu=p was returned to service. During operation on Dece=ber 4, 1973, the sa=e j
shaft broke at the impeller in the same =anner as August 15 incident.
9 "a The break was attributed to fatigue as a result of cavitation. Cavitation
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?n would cause the i=peller-to-shaft key to loosen and result in torquing a
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.evvay apparently continued until fatigue failure of the shaft occurrad.
i, Sere was no evidence of corrosion of the shaf t or the i=peller.
i
.f a:taren 20, 1974, The "3" boric acid transfer pu=p shaft broke again. It
..is rebuilt with certified parts.
he "A" boric acid transfer pu=p had a modified pu=p shaf t with a rounded
.wyvay, also as a result of previous pu=p failures. The "B" boric acid transfer pu=p was scheduled to be replaced with a modified shaf t.
'n April 6,1974, the "A" boric acid transfer pu=p shaf t broke in the etcinity of the keyway. The pu=p was replaced with a spare one from tock. The redesigned shaft keyway, frc= a square to a round configura-
- ion to reduce fatigue in the area of the keyway, did not appear to be
- he ulti= ate solution.
'n August 8 and 15, 1974, the "B" boric acid transfer pu=p shaf t broke again. It was replaced each time, but failed again on Dece=ber 4,1974.
)n this latner occasion, the pu=p bearing showed excessive wear and the
- art of the shaft housed by the bearing was galled and scored.
fhe "B" boric acid transfer pu=p was replaced with a spare and returned
- o service. A new pu=p, nodel C7H-10K, was reco ended by Che=pu=p and
.as investigated as a replace =ent.
- n each case, upon determination that a pu=p would not meet its design function, the alternate boric acid transfer pu=p was tested and operability verified. Systa= capability was not i= paired and the plant continued to operate at the same power level. Technien1 Specifications require the failed boric acid transfer pu=p to be returned to service within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to permit continued full power operation.
In each instance, the pu=p was returned to service satisfactorily within this time period. These occurrences did not result in a release of radioactive =aterial, nor did it endanger public health and safety.10,11 FOREIGN OBJECTS Di REACTOR SYSTEiS Nine Mile Point 1 At Unit No.1 of the Nine Mile Point Nuclear Station, during routine l
testing following a calibration of flow transmitters, one containment spray pu=p developed the required head, but the flow indication was approxi=ately 2600 gp= rather than design flow of 3000 gp=.
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<&. E was taken out of service and disassembled; a piece of wood (1-3/4" x 5-1/4" x 3-1/4") was found wedged in the eye of the pu=p.
The piece of wood restricted suction flow and li=ited pu=p performance. After re= oval of the object and reasse=bly of the pu=p, the perfor=ance was satisfactory.
E satisfactorily met the performance require =ent.
Operation with one conta1==ent spray pu=p is sufficient to re=ove post
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accident core energy releases and will it:ric pressure and te=perature rises in the pressure suppression system to below design values. The contain=ent spray system provides 400% redundancy. Three other pu=ps Therefore, there would have been no hatard to the public or station had the contain=ent spray system been required.12 B_
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Peach Bottom 3
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During surveillance testing at 1% power at the Peach Bottom Accsic Power Station Unit 3, the reactor core inj ection cooling (RCIC) steam line i
drain valve failed to close. A piece of weld slag had entered the valve and beca=e wedged below the seat stem guide.
ij The valve was disassembled and cleaned. The disc and seat were =achined 3
to remove indentations caused by the foreign object and then lapped.
i N The valve was reassembled, leak tested and returned to service.
The RCIC sain steam line drain system has two air operated isolation g
valves in series. Both of these isolation valves close automatically on RCIC initiation. Since only one valve was aff ected by this occurrence, 3':
system operability was not affected.13 3l 91ll al Theodore C. Cintula John J. Ria=o Office of Operations Evaluation U.S. Nuclear Regulatory Cc= mission 1
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h-REFERENCES l
1.
DRESDEN UNIT 1 CONTROL ROD UNCOUPLING, NOVEMBER 7,1974, J. W. Wujciga.
U' Docket No. 50-10.
2.
Letter, N. J. Kalivianakis (Coc=enwealth Edison) to J. F. O' Leary, USNRC, Directorate of Licensing, October 18, 1974. AOR No. 74-29, Docket No. 50-254.
3.
Letter, M. Traur-an (Eberline Instru:: ent Corporation) to N. J.
Kalivianakis, (Co=onwealth Edison), October 15, 1974.
4.
Letter, B. B. Stephenson (Co==onwealth Edison) to J. G. Keppler, USNRC, Directorate of Regulatory Operations - Region III, Nove=ber 19, 1974. AOR No. 74-34, Docket No. 50-249.
5.
Letter, N. J. Kalivianakis (Cor. onwealth Edison) to J. F. O' Leary, USNRC, Directorate of Licensing, November 18, 1974. AOR No. 74-37, Docket No. 50-254.
6.
Letter, R. C. Arnold (Metropolitan Edison Cc=pany) to USNRC, 15, 1974, A'R No. 74-22, Directorate of Licensing, November O
Docket No. 50-289.
7.
Letter, W. G. Council (Northeast Nuclear Energy Cc=pany) to A. Gia=busso, USNRC, Deputy Director for Reactor Projects, Nove=ber 6,1974. AOR No. 74-8, Docket No. 50-245.
8.
Letter, J.S. Bitel (Commonwealth Edison) to J. F. O' Leary, USNRC, Directorate of Licensing, July 31, 1974. AOR No. 74-30, Docket No. 50-295.
9.
B. B. Stephenson (Comonwealth Edison) to J. G. Keppler, USNRC, Directorate of Regulatory Operations - Region III, October 2, 1974.
AOR No. 74-29, Docket No. 50-249.
10.
Letters, E. E. Utley (Carolina Public & Light Company) to J. F.
O' Leary, USNRC, Directorate of Licensing, December 31, 1973, March 25, and April 12, 1974. AOR Nos. 74-7 and 74-8, Docket No. 50-261.
- 11. Letters, E. E. Utley (Carolina Public & Light Ccmpany) to N. C.
Moseley, USNRC, Directorate of Regulatory Operations - Region II, August 14, August 23, 1974 and December 16, 1974. AOR Nos. 74-17 and 74-18, Docket No. 50-261.
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- 12. Letter, R. R. Schneider (Niagara Mohawk Power Corporation) to D. J.
Skovholt, USNRC, Assistant Director of Reactor Operations, Septe=ber 20, 1974. AOR No. 74-13, Docket No. 50-220.
13.
Letter, M. J. Cooney (Philadelphia Electric Cc=pany) to A. Gia=busso, USNRC, Deputy Director of Reactor Projects. October 14, 1974.
AOR No. 74-11, Docket No. 50-278.
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