ML19319D382
| ML19319D382 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 05/01/1978 |
| From: | FLORIDA POWER CORP. |
| To: | |
| Shared Package | |
| ML19319D380 | List: |
| References | |
| FOIA-79-149 NUDOCS 8003160215 | |
| Download: ML19319D382 (15) | |
Text
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9 CRISTAL RIVER UNIT THREE STATUS REPORT MAI 1, 1978
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Introduction Crystcl Riv2r Unit 3 is praccntly in an extsndsd eteg$ as th2 racult of a Burnable Poison Rod Assembly (BPRA) separating from its Fuel Assembly and subsequently damaging the Once Through Steam Generator (OTSG).
Below is a brief description of events that lead to.the April 6 meeting to' discuss this topic v the NRC.
O:. 2farch 3,1978, at 2337 hours0.027 days <br />0.649 hours <br />0.00386 weeks <br />8.892285e-4 months <br />, Crystal River Unit 3 was shut down to investigate an apparent loose part(s) detected by the Loose Parts Monitoring System (LPMS) in the upper tube sheet region of the "B" OTSG.
Inspection confirmed that several pieces of a Burnable Poison Rod Ass t y (BPRA) were i
in ths "B" OTSG and subsequent video inspection revealed damage in the tube and 'ube sheet area. Ongoing inspections of the Reactor Vessel Internals t
have shown an additional adjacent BPRA had become separated from its fuel assembly and was in the, plenum. To date, a variety of BPRA parts have been identified in "B" OTSG, Plenum, Core Support Assembly, various fuel assemblies and at the bottom of the Reactor Vessel. ' The Reactor Vessel has been completely defueled and video inspection of the fuel and reactor internals has taken place.
Over the past few months, there have been several events which it is felt were signs of the BPRA separation.
On December 12th, an unexpected power _ tilt _ developed _following a reactor trip from 70% power.
As early as this, serious concern was given to the possibility of a BPRA having cladding failure. Subsequ~nt Florida Power Corporation (TPC) analysis of available physics data and Reactor Coolant activities, however, did not support this theory.
In separate incidents on January 1st and 3rd, alarms on the LPMS sounded,.
-indicating a potential problem in'"B" GTSG. Although data was accumulated by the CR-3 staff and forwarded to FPC Plant Engineering, a positive deter-mination of a loose part could not be substantiated.
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During the period from January 19th through February 15th, rapid changing of makeup prefilters because of high AP occurred. This was indicative of
~ aither a filter er reactar coolant pntciculats prsblan.
Chemical staplso cf the RC syatem did not reveal any evidence that the RC system was the cause; however, the filter system itself also seemed in order.
The investigation into this perplexha. Problem was still going on when on February 17th the I.PMS again went into alarm, indicating a definite problem with "B" OTSG.
Unlike the earlier event, though, there was no longer any doubt that a loose part existed and action was necessary.
The purpose of this report is to give the present status, developments since April 6, future plans and schedules.
II.
Present Status and Plans A.
Fuel 1.
Fuel Assemblies 3C35 and 3C37 s
Fuel assemblies 3C35 and 3C37 from which BPRAs B-47 and B-52 separated, have undergone a series of tests and inspections to verify their structural integrity and capabilih for successful operation during subsequent fuel cycles.
Both were visually inspected in the reactor.and spent fuel pool buildings and all observed debris has been removed.
a.
Visual Examination The fuel assemblies were====hed in datail from the top, bottom and the sides to assure that no mechanical damage had occurred, the flow channels were free of debris and no anomaly exists.
Fuel assembly 3C35 was clean while fuel assembly 3C37 had a piece of burnable poison rod cladding, approximately 1/2 inch in length, lodged in the lower end fitting grillage.
The piece of clad b
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.wes su sequent y removed.
b.
Probe Test Each guide tube of these fuel assemblies was probed full length through the top of the guide tube to the lower end plug to determine if any debris was present.
The results of this examination revealed that 14 of the 16 guide tubes in 3C35 and 9 of the 16 guide tubes in 3C37 were free of any debris Through the use of specially designed tooling, all debris from the guide tubes l
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- was removed.
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Bubbis' % t
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Each guide tube was then subjected to a bubble test to determine
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if cracks or holes were present. The bubble test involves sealing one end of the guide tube underwater and graducily injecting high pressure air at that end. Pressure level at which air bubbles are observed in the pool determines the condition of the guide tube.
Twenty-five (25) of the thirty-two (32) guide tubes tested had bubbles appear at a pressure at which it should have for a normal guide tube with the flow holes at the lower end. Four (4) guide tubes in fuel assembly 3C35 and three (3) guide tubes in fuel assembly.3C37 had air bubbles appear at pressura lower than that for.a normal guide tube. The result is inconclusive since there may have been leakage around the seal at the guide cube nut.
Further inspections will be perforined to determine the structural integrity of the fuel assemblies.
All the debris located to date in fuel assemblies 3C35 and 3C37 has been removed. A detailed visual enmination.of the upper end fittings will be performed again to insure no debris is present.
All visible debris will be removed. Pending the s.uccessful completion of all tests, inspections and confirmation of their structural integrity, these two fuel assemblies will be returned to the core to be used as unrodded assemblies during subsequent cycles.
2.
BPRAs and Their Fuel Assembly Coupling (Holddown Iatch Assembiv)
The inspections and tests in this area are summarized below.
a) ANO-1:
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Thirty-six holddown latches on the fuel assembly were inspected at Arkansas Nuclear One, Unit 1.
Preliminary results indicate that clie wear areas at ANO-1 are less than that reported on Oconee 2 in the
- April 6,1978 meeting between Florida Power Corporation and the NRC.
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b) CR-3 Innocctionst Fonowing defueling at Crystal River 3 (CR-3), all 66 remaining BPRAs were subjected to a lock test, and a n were found to be locked in their respective fuel assemblies. During removal of the 66 BPRAs, an ball-lock couplings were visually evamined; nothing unusual was seen. Nine (9) of the BPRAs were visually ernmined full-length and 360* around. Nothing unusual was seen. All fuel assembly holddown latch assemblies (68) containing BPRAs were' visually examined 360* around on the inside. Two wear areas were seen on each latch assembly, oriented at 180* to each other.
Three fuel assemblies had wear in the holddown latches which
~ approximated that observed in the holddown latches of fuel assemblies 3C35 and 3C37. While the results of the holddown latch inspectio'as are still being evaluated, preliminary results indicate the wear in the latch assemblies at CR-3 is much higher than the wear observed at Oconee or ANO.
3.
ORAs and Their Fuel Assembly Counling (Holddown I. acch Assemblies).
Orifice Rod Assemblies (ORA) at Crystal River 3 were avamined as well as the corresponding holddown latches in each fuel assembly. No evidence of wear or any abnormal condition was seen.
Each of the forty ORAs were identified and checked for orientation with respect to the fuel assemblies. The ball latching mechanisms were examined for ball orientation and condition.
The ho1Tdown latches were examined for
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evidence of wear, and general visual appearance. Each ORA was reinserted into its corresponding fuel assembly and was verified to be locked in place.
None of the holddown latch assemblies had wear marks, or any features except for two tiny rpherical dimples corresponding to the location of the latching balls.
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Inspection of ORAs at Oconee, and holddown latches at Oconee and ANO-1 provide additional verification of the observations at CR-3.
No evidence of wear or abno'rmal operat. ion has been seen for any of the 0 ras and ORA holddown latches.
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These results show that ORAs have been ussd with no failures and no degradation of any kind.
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l The results of the recent ORA latch mechanism evamination firmly supports the current plans of reusing present ORAs.
This same ORA design will also be used to replace, as required BPRAs which are removed.
Adeinistrative steps will be taken at Crystal River to assure that in inserting the ORAs the locking balls are oriented in a different direction than that in which the
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BPEA locking balls were oriented.
4.
Remaining Fuel Assemblies The lover end fittings of the r===4n4ng fuel assemblies were inspected during defueling. Eighty-five (85) fuel assemblies had debris lodged into the lower end fitting. All observed ' debris was removed and no damage detrimental to the structural integrity of the fuel assembly lower end fitting was observed. The majority of t$e debris was from the broken pieces of the separated burnable poison rods assemblies (BPRAs).
In addition to the side view inspection of the fuel assemblies 3C35 and 3C37, side view inspection of eight.(8) other selected fuel assemblies for possible debris or other anomalies was performed.
The fuel assemblies were selected from the different core locations and all had debris collected from their lower end fittings.
Inspection of all four sides of the assemblies showed no debris or other anomalies.
As discussed in Section II-2 and 3 the holddown latches of all the fuel assemblies containing BPRA or ORA were examined. A total of nine (9)
BPRAs were visually examined in detail to look for any unusual wear mark or crud pattern. The nine BPRAs were chosen from the fuel assemblies which cover the range of wear damage in the holddown latch assemblies. No wear marks on l
the burnable poison. rods were observed. Hence, no significant wear is expected in the guide tubes of the remaining fuel assemblies which contained BPRAs.
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Visual insi tien of tha top of the cor..
parformed bsfera the
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defueling and observed debris was removed. A visual inspection of the top of all fuel assemblies will be made again to assure no debris is present. Debris will be removed if found.
All fuel assemblies will be reloa.ed in their' original cycle 1 locations.
5.
Hydraulic BPRA Lift Analysis
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An extensive analysis has been perf'ormed to determine the lift force on the BPRA.
In addition, a lift test has been completed at ARC, using full size prototyfe components to determine the flow at which the BPRA lifts off the fuel assambly. This information was used to benchmark the hydraulic models and form loss coefficients used to calculate. the hydraulic loads on the BPRA. The core flow distribution was calculated using the crossflow code LTNIl.
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This code was used with the most probable core conditions to obtain the most accurate calculation possible.
The calculation shows that. a BPRA in a peripheral core location has a lift force approximately two pounds higher than its nom 4= 1 weight in water at a system flow rate of l12% of design flow (the best estimate flow rate for CR-3).
The corresponding lift force for an interior core location would be one pound higher.
Using the same analysis techniques [the lift force for an orifice rod assembly was calculated to be 40 pounds greater than its weight in water.
This net upward lif t does not allow the ORAs to become free floating and there-fore they are not susceptible to vibratory wear.
. B.
Reactor Vessel and Reactor Vessel Internals The ini*.ial' inspection of the Internals was primarily to look for any obstructions that would hang up or possibly damage the internals of fuel during
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the plenum removal.
It was during this inspection that the second BPRA (B-52) was discovered to be out of its fuel assembly.
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The inspectio. ias cecomplished using an unda ator TV camers. Tin camera was lowered both inside at.1 outside the planum cylinder, 360 around.
The loose BPRA (B-52) was discovered in the Y-Z quadrant resting on one of the large plenum cylinder flow holes.
Broken BPRA pins were also sighted, estending from the fuel assemblies.
Debris was only sighted in the T-Z quadrant.
Prior to the plenum removal, the B-52 ZPRA was extracted through the plenum cover.
The planum was lifted with no irregularities and placed oc its storage stand in the deep end of the refueling canal.
Detail inspections are plannad for the plenum assembly, core support
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and reactor vessel. The purpose is to look for debris and structural damage.
Any debris found will be removed. The inspections will be performed using an underwater TV camera.
1.
Planum:
During the initial inspection described above, the only noticeable damage to the plenum was ding marks in the lowest third of a flow hole in the plenum cylinder. No damage was noted on any of the control rod guide tubes.
Additional inspections are planned for the interior and exterior of the plenum with special consideration given to the following areas:
a) Outsiae Diameter of control rod guide tubes (5 total) in the
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' Y-Z quadrant. The brazement mounting screws and lockwelds will be inspected for any evidence of damage.
I b)
Plow holes in the Y-Z quadrant, concentrating mainly on the large hole where some damage was previously noted.
c) The " porcupine" structure at the outlet regions.
d) Plenum cylinder to plenum cover bolts and lockings clips.
3 e) Top surface of upper grid in the T-Z quadrant.
f) A free path check will be performed for each of the 69 control l
rod guide tubes.
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2.
Corm Supptre Asstably Initial intpectiens of th2 CSA chow:d no damag2,'n11. Gight of th2 vent valves were inspected. After the planum was removed, the fuel assembly from position K-9 was removed and a camera lowered into the lower part of the internals and vessel head.
Several pieces of BPRA pins were spotted between the lower grid support forging and flow distribution plate. No damage to the internals was noted in this inspection.
!Adicional inspections of the core suppo'rt will look for debris and structural damage, plus the operability of each vent valve will be verified. The following areas will be inspected:
a)
Outlet nozzles
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b)
Inside surface of the core suppert shield - Y-Z quadrant
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l c) Porcupine structure around the outlet nozzles d)
Core support shield lower flant,e, especially the bolt heads and locking clips e) Top of the core barrel formers f)
Thermal shis~~
- raint blacks, bolts and locIking clips g)
Each SSHT w cotal) h) Each guide, block (12 locs.tions - 24 blocks) i)
The area between the lower grid rib section and the flow dis-tributor plate j)
The area between bottom of flow distributor plate and top of
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support forging.
k) Area between incora guide support plate and flow distributor head 1)
Each of the eight (8) vent valves will be exercised using standard procedures 3.
Reactor Vessel During the video inspection through position K-9 de, scribed above, j
several pieces of BPRA pins were spotted in the lower head region; the largest piece was spproximately one foot long. No damage to the reactor vessel was_noted on this inspection.
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i Fallowing 2 sval of th3 cera suppsrt, ths
.ceter vescol will be
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visuelly inspected for debris and damage. Areas of special consideration are:
a)
Inside of inlet and outlet nozzles b)
Guide lugs (12 total) c) Lower head d)
Incore nozzles (exterior)
C.
Once Through Steam Generator 1.
OTSG A The upper manway of the "A" generator was opened and the generator examined by looking into the upper head. No loose parts on the upper tubesheet, nor any tube end damage, nor any tube-to-tubasheet veld damage was " observed. -
A free path check of all tubes is planned.
This check will consist of running an addy current probe or equivalent through the tubes.
Debris that blocks free p"assage will be removed or the tube removed from service by standard tube plugging procedures. The lower head wiki be opened and inspected.
2.
OTSG B - Tube Leaks i
Prior to shutdown of the Crystal River 't plant, primary tc secondary system leakage of less than 1 gallon per day was detected. At presen't the leakage 2.s assumed to be a result of damage sun.41ned by the upper tv se ends and their welds in the "B" generator. As of 4/30/78, the inspection is complete and no leaks have been detected.
The secondary side was pressurized
- with Helium gas within the limitations of this unit's Technical Spec'afications.
A search for gas leaks on the primary side was conducted via mass spoetrometry.
The inspection probe was remotely passed from one tube weld to the eaxt using a modified addy current manipulator.
Additional gas bubble testing may be undertaken in the "B" generator. This testing differs from mass spectrometry in that the bubbles of leaking gas are trapped beneath Acrylic she'eting.
Small'itivisible bubbles then are allow 2d to accumulate into a large visible bubble.
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The decicion to pr::cacs.ith cdditionc1 leak tecting du, ends on tha cits capabilities to raise the primary water level to cover the upper tubesheet.
l If leaks are detected, leakage through the weld will be confirmed by the equivalent of a soap bubble test of thei eld. Welds will be manually repaired.
Leaks not confirmed to be weld leaks will require those tubes to s
be removed from service with the standard tube plugging techniques.
3.
OTSG B - Tube End and Tube Sheet Darare A preliminary video inspec. tion of the upper head was made immediately on opening the manway. A summary of this inspection is:
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1.
Seventeen (17) pieces were observed on the tubesheet surface.
a.
One large control component, MK-B47 from coret location N13.
Identified as " coupling spider assy".
NOTE: The sixteen (16) burnable poisoa rods were not on the OTSG tubesheet.
b.
Sixteen (16) smaller pieces ranging in size from 1/8" thk x 1 1/4" lg. x 3/4" wd. to 1/8" thk x 1 1/2" ig. x 2 1/2" wd.
Thes,e pieces were later identified as parts ofthespiderandburnah5epoisonrodcladding.
2.
Radiation Levels:
Tubesheet readings were 7-8 R/HR., except for large piece which was 10 R/HR at six (6) feet.
3.
Observed tube ends were damaged.
Demaged tube to tubesheet welds were in evidence.
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A detailnd. d20 intpcetitn of tha tuba eew has bacn p rformed.
This inspection consisted of "stop action" television recordings from two T.V.
cameras remotely positioned over the tube ends by a modified addy current manipulator,. Each camera simultaneously viewed a typical seven tube pattern (one camera's line of view normal to the tubesheet and the other 30' with the tubasheet face). Viewing these recordings on a 21 inch television provides a magnification factor of 5.
All tubes have been inspected and each tube categorized into the following:
Class I (55% of Tubes)
Impact or roll over of tube ends may exist on 0.D. or I.D. defcuned material does not include veld metal.
Class II (7% of Tubes)
Partially separated chip. May exist vitis Class III or IV damage.
Class III (26% of Tubes)
Minor veld damage including dents anywhere od weld and damage extending into upper 1/3 of weld metal (may include Class II type dam. age).
Class IV (17% Iubes)
Damage to tube end weld in excess of Class III.
(May include Class II type damage).
In addition to manual weld repairs and tube plugging of leaking tubes, all Class II category damage, partially separated chips on about 1000 tubes, will be removed.
Shielding will be installed after remotely marking tubes'with Class II damage.
Small areas of the tubesheet will be uncovered and chip removal completed with a variety of manual operated tools. No other
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.c repair is plann d et c' s cutaga. The following is a maary of the plannsd OTSG repair of the "B" generator.
Video Inspection & Categorize Tubes Locate Leaks & Identify Install Shielding & J Leg Screens Repair Isaks, NDE Repairs, & Dress Tubes Banove Shielding 100% Free Path Check Clean Obstructed Tubes & Eddy Current Test Explosive Plug & Leak Test 100% Free Path Check OTSG-A Banove Screens & Close Up OTSGs A Contingency Repair basically differs from the planned repair
- tn that all Class IV tube-to-tubesheet welds will be rewelded.
If the contingency repair is not undertaken at'this outage, a similar repair may be performed at the next refueling outage. The following is the tabulation of the contingency rapsir.
Video Inspection / Cat. Tubes Setup Machine Carriage Machine / Spot Face Tubas (57%/43%)
Deburr Tube Ends & Vacuum Cisan Spot Face as.Necessary Roll Expand as Necessary Bemove Mach. Carriage Setup & Preheat Tubesheet Install / Setup Weld Carriage e
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Weld Sp;t Fcco Tuba- (2800) i Vides Inspect Welds Tean for Pt Inspection Famove Wald Carriage IT/ Rework /RE-PT Clean / Flush /Go-No Go STSG Q1canup/ Remove Bladders hak Test 3/C Tubes as Required Explosive Flug & Laak/ Closeup OTSG-B Tree Fath OTSG-A h Debris Receval As discussed above inspections have been performed or are planned gf (uel assemblies SC35 and 3C37 (BPRA guide tubes), the top and bottom and fittings of all. (uel assemblies, the reactor vessel and internals and the OTSG heads. All debris found in these areas was or will be removed. Addi-tional plans to insure all debris has been removed calls for b A free path check of all Control Rod Guide Tubes b A.(tee path. check of all OTSG tubes' R is expected that. only a very few small pieces of debris could remain in the system Block. age. studies have indicated that the increased turbulence in the. area. of a.Alockage produces sufficiently better heat
.stansfer conditions that. effset the. loss of flow due to the blockage.
Arkansas Nuclear Unit One and. Oconee Two operated for several months with similar. size. debri.s in. the system with no adverse effect.
The coolant activi_ty showed. no increase. as would have occurred. if there were DNB induced failures.
.e III.
Schedulo of Further.:tivitim rnd Repnrts Commence OTSG tube repairs 15 May Commence Refueling (all Rasctor Vessel work is complete)
Submit Reload Report to the NRC 2 Jun Complete 075G repairs 7 Jun Submit final report to NRC 1 Jul Complete all OTSG vork
.7 Jul Commence Unit startup 22 Jul e
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