ML20141F271

Forwards Copy of Summary of Meeting on Control & Anomalies at South Texas Plant Per Request

ML20141F271
Person / Time
Site:	South Texas
Issue date:	05/19/1997
From:	Charemagne Grimes NRC (Affiliation Not Assigned)
To:	Hoehn J INTERNATIONAL ATOMIC ENERGY AGENCY
References
IEB-96-001, IEB-96-1, NUDOCS 9705210259
Download: ML20141F271 (2)

Text

{{#Wiki_filter:. Q l-(9( Y May 19,1997. Joachim Hoehn Division of Nuclear Installation Safety international Atomic Energy Agency P. O. Box 100 A-1400, Vienna, Austria ?

Dear Joachim:

I As requested, I have enclosed a copy of the summary of the meeting on the control rod t ' anomalies at the South Texas plant. In addition, I understand that the NRC is preparing a supplement to the related NRC Bulletin 96-01 which will summarize the results of the findings thus far; as soon as that supplement is available, I will forward that to you as well. If you have any questions regarding this material, please'do not hesitate to contact me, Sincerely.,inal Signed by: Ong C. l. Grimes Christopher 1. Grimes, Chief Technical Specifications Branch Office of Nuclear. Reactor Regulation

Enclosure:

as stated DISTRIBUTION -\\ FILE CENTER l PUBLIC TSB R/F MCullingford SSchuyler-Hayes i DOCUMENT NAME: HOEHN2.LTR l l OFFICE . C:TSB:NRR NAME. CIGrime DATE. 5/19/97 OFFICIAL RECORD COPY 9 7-M'/ DR D K 05 90 k P PDR

y - ~~ug .4 UNITED STATES .p s g NUCLEAR REGULATORY COMMISSION 't WASHINGTON. D.C. 20555M1 - g... - May 19,1997 Joachim Hoehn -[ ~ Division of Nuclear. installation Safety International Atomic Energy Agency P. O. Box 100

A 1400, Vienna, Austria

Dear Joachim:

. As requested, I have enclosed a copy of the summary of the meeting on the control rod anomalies at t'. e South Texas plant. In addition, I understand that the NRC is preparing a supplement to the related NRC Bulletin 96-01 which will summarize the results of the findings thus far; as soon as that supplement is available, I will forward that to you as well. If you l have any questions regarding this material, please do not hesitate to contact me. Sincerely, /L e \\ L N .b LLL Christopher 1. Grimes, Chief Technical Specifications Branch Office of Nuclear Reactor Regulation

Enclosure:

as stated t ) i q-y-

d \\. UNITED STATES j g (' - NUCLEAR REGULATORY COMMISSION '.**g ~ WASNINGTON, D.C. 2000HOO1 4... j January 29, 1996 I LICENSEE: Houston Lighting and Power Company (HL&P), et al. FACILITY:- South Texas Project,' Units 1 and 2 (STP)

SUMMARY

OF JANUARY ~18, 1996< MEETING ON THE UNIT 1 CONTROL R00 L INSERTION ANOMALY AND THE RESULTING ACTION PLANS I On Janu'ary 18, 1996, representatives of HL&P and NRC staff met to. discuss the above subject. Meeting attendees are listed in Attachment 1. Handouts provided by the licensee are in Attachment 2. -The 1icensee's presentation consisted of an. overview of the anomaly, s i background.Information, trip recovery actions, investigation status, and action. plans. The anomaly occurred during a December 18, 1995, reactor trip 3 i i when three control rods stopped at six steps from bottom, and during subsequent testing when the same three rods plus one additional rod also stopped at six steps. The affected control rods were over high burnup } assemblies. The licensee performed a safety evaluation which bounded the condition and developed plans to determine the cause and corrective action. l o l By way of background, the licensee informed the staff that the control rods i are of silver-indium-cadmium composition with a stainless steel cladding. i They have a 15-year calendar life and have been in service since October 1989, The lower guide tube geometry incorporates a double dashpot feature (Indian ~, Point 2 and Ginna also have this feature) and the four fuel assemblies that i experienced the anomaly are of the Westinghouse (W) XLR fuel design. In their investigation of the STP design versus the standard W design, the licensee i determined that the STP fuel assemblies are about 30 inches longer and their j hold down forces in the vessel are about 200 pounds greater, that the lower i guide tube dashpot region length is about 10 inches longer and most plants i have only a single dashpot, that the control rod radial clearances are about j the same, that the STP fuel design has 2 addition grids, and that the fuel discharge burnup at STP is in the top half of W plants. .The licensee's investigation o.f industry experien,e of control rods failing to i fully insert found-the causes to include foreign material, control rod / drive I line degradation,. corrosion products, and guide tube bow. The licensee does not know the cause'at STP, but has developed action plans to determine the cause and apply corrective action. These plans include inspecting the 4 high ..burnup Unit 2 assemblies in the spent fuel pool, performing hot full flow rod i drop testing during the next. :,hutdown in either unit (but no later than by 60 to 75 effective full power days (EFPDs)' after the December 18, 1995, reactor i trip.on Unit 1;.this was recently done on' Unit 2 since Unit 2 was down for other reasons, and all Unit 2 control rods tested satisfactorily), performing Unit -1 hot, full flow rod drop testing at end-of-life during plant shutdown at .the start of Refueling Outage 6 (IRE 06), revising the Unit 1 Cycle 7 loading . pattern to minimize or eliminate high burnup fuel assemblies, and inspecting .and testing discharged Unit I fuel assemblies during and after 1RE06. i b2 bh[{ e

o The staff thanked the licensee for the meeting and indicated that it was useful in keeping the staff up-to-date with the anomaly. The staff intends to closely evaluate the licensee progress on the anomaly, including the planned testing and inspection results. / @JW WIN ~ Thomas W. Alexion, Project Manager Project Directorate IV-1 Division of Reactor Projects III/IV Office of Nuclear Reactor Regulation Docket Nos. 50-498 and 50-499 Attachments:

1. List of Meeting Attendees

2. HL&P Meeting Handouts cc w/ attachments:

See next page 1 a

Houston Lighting & Power Company South Texas, Units 1 & 2 cc: Mr. David P. Loveless Jack R. Newman, Esq. Senior Resident Inspector Morgan, Lewis & Bockius U.S. Nuclear Regulatory Commission 1800 M Street, N.W. P. O. Box 910 Washington, DC 20036-5869 Bay City, TX 77414 Mr. J. C. Lanier/M. B. Lee Licensing Representative City of Austin Houston Lighting and Power Company Electric Utility Department Suite 610 721 Barton Springs Road Three Metro Center Austin, TX 78704 Bethesda, MD 20814 Mr. K. J. Fiedler Rufus S. Scott Mr. M. T. Hardt Associate General Counsel Central Public Service Board Houston Lighting and Power Company P. O. Box 1771 P. O. Box 61867 San Antonio, TX 78296 Houston, TX 77208 M' G. E. Vaughn/C. A. Johnson Joseph R. Egan, Esq. Central Power and Light Company Egan & Associates, P.C. P. O. Box 289 2300 N Street, N.W. Mail Code: N5012 Washington, DC 20037 Wadsworth, TX 74483 Office of the Governor INPO ATTN: Andy Barrett, Director Records Center Environmental Policy 700 Galleria Parkway P. O. Box 12428 Atlanta, GA 30339-3064 Austin, TX 78711 Regional Administrator, Region IV A-thur C. Tate, Director U.S. Nuclear Regulatory Commission Division of Compliance & Inspection 611 Ryan Plaza Drive, Suite 1000 Bureau of Radiation Control Arlington, TX 76011 Texas Department of Health ) 1100 West 49th Street Mr. Joseph M. Hendrie Austin, TX 78756 50 Bellport Lane Bellport, NY 11713 Mr. William T. Cottle Group Vice President Nuclear Judge, Matagorda County Houston Lighting & Power Company Matagorda County Courthouse South Texas Project Electric 1700 Seventh Street Generating Station Bay City, TX 77414 P, O. Box 289 Edsworth, TX 77483 Mr. Lawrence E. Martin General Manager, Nuclear Assurance Lic. J. W. Beck Houston Lighting and Power Company Little Harbor Consultants, Inc. P. O. Box 289 44 Nichols Road i Wadsworth, TX 77483 Cohasset, MA 02025-1166 i

l i MEETING BETWEEN HL&P AND NRC ON CONTROL ROD ACTION PLANS January 18, 1996 R&mg Oraanization T. Cloninger HL&P D. Leazar HL&P D. Hoppes HL&P R. Dunn HL&P S. Head HL&P J. Sparrow Westinghouse H. Menke Westinghouse S. Ray Westinghouse W. Beckner NRC T. Alexion NRC L. Kopp NRC D. Powers NRC S. Matthews NRC E. Kendrick NRC H. Richings NRC H. Ornstein NRC S. Koenick NRC K. Naidu NRC B. McCabe NRC E. Goodwin NRC ATTACHMENT 1

[ l SOUTH TEXAS PROJECT l l l l i PRESENTATION: l L UNIT 1 CONTROL ROD INSERTION ANOMALY l TO NRC January 18,1996 t l I %., A T. a i 2 i 5- ~ VISION: STP -- A WORLD-CLASS POWER PRODUCER i l..,.....,.. _..

STP TEAM STP Team Ted H. Cloninger - Vice President, Nuclear Engineering David A. Leazar-Manager, Nuclear Fuel & Analysis David F. Hoppes - Supervisor, Nuclear Fuels Roland F. Dunn - Supervisor, Reactor Engineering Scott M. Head - Supervisor, Nuclear Licensing Howard Menke - Manager, Product Performance Engineering, Westinghouse Sumit Ray - Manager, Fuel Licensing Integration,. Westinghouse Jim Sparrow - Principal Engineer, Product Design, Westinghouse SCOD.O Technical issues relating to control rod insertion anomaly 4 4 ,th,. m, -

MEETING AGENDA ~ Overview Background information STP fuel and control rod designs Rod position indication Rod testing Safety limits Event Description Trip Recovery Actions Rod testing i i Safety evaluation i l investigation Status Facts Fuel design comparisons i industry experience Potential causes Action Plans 1 18 pedi-1$ 2'1

OVERVIEW December 18,1995, Reactor Trip, three control rods stopped at six steps from the bottom Testing performed one additional control rod stopped at six steps Affected control rods are over high burnup assemblies Performed safety evaluation which bounded current condition Monitoring plan to ensure conditions remain bounded by safeiy ^ evaluation Plan established to determine cause and corrective action i 1 I 1 teswett.tSekt

BACKGROUND Plant 4-Loop Westinghouse NSSS, PWR 3800 Mwt licensed reactor thermal power 193 fuel assemblies - Westinghouse designed and fabricated 14 foot active fuel length 57 silver-indium-cadmium control rods (RCCAs) Rapid refueling package - unrodded refuelings i i l' %,.i. mima ~

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j i L BACKGROUND L u Fuel l l 3 typical design variations: XL, XLR, V5H I XL -Inconel anti-snag grids i .XLR - Removable top nozzle, debris filter bottom j nozzle, extended burnup capability i i V5H - Zirconium mid grids with XLR features, Integrated Fuel Burnable Absorbers i Common features 14 foot active fuel Stainless steel top / bottom nozzles 10 grids Zirconium guide tubes Double Dashpot for RCCA deceleration Typical initial enrichments: 3.8 to 4.2 w/o U-235 4 W,..

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=- l j BACKGROUND i RCCAs - Silver-indium-Cadmium (80%Ag,15%ln, 5%Cd) Stainless steel cladding Spring hub to absorb trip energy 15 calendar year design life - in-service since 10/89 Fully inserted is O steps or rod bottom Fully withdrawn is 259 steps Rod steps are 5/8 inch . m.,, -, _,

1 ROD CLUSTER CONTROL ASSEMBLY OUTLINE e 4 4 F k N - /* i Q + \\ +

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l I LOWER GUIDE TUBE GEOMETRY i e.s..s. *'"' 180 in Top Nozzle 3 / h 33.5 in O.450ID Start top dashpot 38 steps / 0.397 ID f,! / 1 I 22in 21.2 in l End top dashpot = HH!HlHlHH H.H;H HH H[ 20 steps orid #2 N 18in Start lower dashpot 8 14 steps 'N O.450 ID Fuel pin O.374OD*n'b 15.0 in +9 steps 13 in (~4 in above l Rod Bottom) +6 steps 11.2in +3 steps 9.36 in Rod Bottom - O steps 3.71 in H!HH,H H f lHlHHlHH,'HlH i.5 in d Start of active fuel 0in waxxwasxwsx Top of Bottom Nozzle

BACKGROUND Rod position indication L ' Digital Rod Position Indication (DRPI) at STP Drive rod connected to RCCA penetrates energized DRPI coil stack Impedance changes Provides true indication of actual rod position l l Indication in six (6) step increments with 3 step accuracy i DRPI is used to measure rod drop times l i 2:.7,,.~,

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BACKGROUND Rod Drop Time Testing F Typically at BOL, hot standby, full flow conditions One at a time, bank at a time, or all rods simultaneously 4 Dropped from 259 step position Trip breaker opened, rod drops by gravity Record electrical signal from de-energized DRPI coils Rod drop time - delta time between gripper power loss and dashpot entry No adverse trends noted Typical rod drop times approximately 1.55 to 1.65 seconds

TYPICAL ROD DROP TRACE Turnaround time Technical Specification Rod Drop Time <- Dashpot-* ~ Excess ' Start timing when <- energy h< Stationary Gripper removal . voltage drops g-f3) ~ End timing 7k p when dashpot E is entered E / \\ / \\ n / 9 (NsJ Stationary Gripper . Voltage (normalized) u 4~_ J Voltage induced \\ . in DRF'l coils C Spider hub spring +,, compression starts ,/ L-. Turnaround point t (O steps,9.36 in) Time (sec) l3."..-,. s.

l 4 UNIT 1 ROD DROP TIME HISTORY t i l Shutdown Banks 1.75 i 1.70 ^ t 1.65 2-i \\ J p -{l I I 1.60 i 9 l 1 l l I 1.45 i l D2 812 M14 P4 B4 014 P12 M2 G3 C9 J13 N7 C7 G13 N9 J3 E3 C11 L13 N5 C5 E13 N11 L3 H4 D8 H12 M8 i Cycle 1 Cycle 2 Cycle 3 - Cycle 4 l I Cycle 5 Cycle 5 (SG) Cycle 6 Control Banks 1.75 e. 1.65 j

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L BACKGROUND l l Safety limits Tech Spec rod drop time: < 2.8 seconds Shutdown Margin: > 1.3% A k/k or 1300 pcm l l i l l ~ f f l l 1 l t A 1 l euew[t tepet tseN1

~ EVENT DESCRIPTION Unit 1 Cycle 6 reactor trip on 12/18/95 3 of 57 control rods not at rod bottom - DRPI indicated 6 steps fcom rod bottom Core locations F-10, C-9, and N-7 N-7 DRPI indication changed to rod bottom within about 1 hour No indications from Loose Parts Monitoring System DRPI indication was correct ? I l 1",.. esw

.. T .L -. :.: ~. J J --. T ^ TRIP RECOVERY ACTIONS u l l l Rod Testing l Hot, full flow rod drop timing Tested Control Bank C and Shutdown Bank B (16 RCCAs) l Control rod drive mechanism currents monitored on 3 affected rods Same 3 rods indicated 6 steps from rod bottom after drops New rod, N-9, also stopped at 6 steps Both N-7 and N-9 went to rod bottom within about 1 hour t No significant change in dashpot entry time Repeatable rod sticking at 6 steps F-10 and C-9 were manually inserted to rod bottom No unusual indications during subsequent reactor startup eUue, enSome

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m. I ROD DROP TRACES GraPhtcal File Ove rl aw InFornation Sun Jan 14 16:07:18 1996 2.99 l f,. , sv%,,,,m,4 t s -j [ ] 1 l i '/ i f l l 4 i '- 1 l ? l l 1 i 4 2.24 - 1l 1 s j I l i! I I h I f ai 1 i i'} i l f j / .( .( l.49 I l i \\\\ 2 \\\\ i i l 1 l 1 4 l ~ l i k M. 'ni -. l \\ i H e j ~ l t l l e.cn I, l e See Jeep I !.H e9 2000 2 bee AddH ? Une t : 1 Tim (ns> Rod : rae Pte F6 N6 Eror: 1 1 1 1 rate: F1912951.RDA F1H12951.RDO F612951.RDA N612951.RDA DathPet: 1591 1561 159e 1613 T Av een nd : 2730 2571 2705 2234 TenP: 568.8 Dev r 568.H Dev F 568.8 Des T $68.9 Des F Press: 2235.8 PSIC 2235.8 PSIC 2235.6 PSIG 2235.8 PSIC l VI mate: SPD.O ine.O 'A ann.n

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d CONTROL ROD DROP CHARACTERISTICS SOUTH TEIAS UNIT 1 vs RINGHALS UNIT 4 I F10 SOUTH TEXAS UNIT 1

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u. . _ _ u_. 1... T_. ~- _T_ ~- ~ 1 TRIP RECOVERY ACTIONS Safety Evaluation Affected rod's safety function maintained - shutdown margin (SDM) met through EOL Accident analyses not adversely impacted Conservative assumptions: -All rodded twice-burned core locations affected (32 of 57) -32 rods stop at 12 steps from bottom after presumed trip -Rod drop times < 2.8 seconds (actuals are about 1.6 seconds) i l Recent shutdown margin calculations, SDM met through EOL for: 57 control rods at 12 steps 28 control rods at 20 steps (in 28 highest rodded burned fuel) 20 control rods at 38 steps (in 20 highest rodded burned fuel) i sapens esest

INVESTIGATION STP Facts 4 affected core locations - twice burned (43000 mwd /mtu) XLR Fuel assemblies F26 (F-10), F25 (N-9), F64 (N-7), F59 (C-9) Highest U1 burnups to date U2 C4 comparable at EOL - 4 rodded assemblies at 43000 mwd /mtu 1st fuel region in Unit 1 with XLR features U1 C6 BOL rod drops sat U1 trip on 8/29/95 - all rods on bottom - 3900 mwd /mtu delta between 8/29 and 12/18 1",. ~, i

i l l INVESTIGATION l STP Facts U1 current and previous core - defect free Manufacturing reviews - no indications -No indications from site receipt drag testing of affected rods t No correlation from previous U1 C1 cladding wear measurements in affected locations Rods in-service since C2 (10/89) Pic.fi is-i ista.i

1 INVESTIGATION STP Facts Satisfactory monthly rod exercise just before trip Rods not significantly slowing down until dashpot Recoil trends may be an indicator Cycle-to-cycle rod drop times 1.55 to 1.65 seconds No adverse flux map trends No chemistry excursions in U1 C6 M,. -,,, =

l U1 RECOIL TRENDS: L (16 TESTED RCCAs) ~ RCCA BOL C5 BOL C6 12/19/95 LEGEND c10 0 x x X-No Recoil O-Recoil 0 N7 0 X X N9 0 X X K10 0 X X K6 0 0 0 F6 0 0 X J3 0 0 0 C7 0 0 X G3 0 0 0 J13 0 0 0 G13 0 0 0 H2 0 0 0 B8 0 0 0 H14 0 0 0 P8 0 0 0 .~...-v

INVESTIGATION t l STP Design vs. Standard Fuel assembly length l Hold down forces in vessel l -Dash pot region length Double dashpot Radial clearances Grids ~ i Discharge burnup 4 m C r - - - - - - = - - - - - - - - - - - - - - ^ ^ ^ - " ^ - - - - ~^ ~ ~ ~ ~

CURRENTWESTINGHOUSE 17x17 FUEL BURNUP EXPERIENCE 55000 50000 g y/y-45000 Unit 2 Cycle S Unit 1 Cycle 6 3 40000 _____g Unit 2 Cycle 4 E l 35000 - Unit 2 Cycle f - '~ Unit 1 Cycle 5 3 r Und 1 Cycle 4 gE ,/ t 25000 / B' 20000 i E Other WRecent Discharge Batch Burnups E STP Recent Cycle Discharge Batch Burnup 4 Wi 18 poshi 16 1 -~ =

o GENERAL FUEL FEATURES COMPARISON [ Feature XL (5 ea) XLR (136 ea) V5H (52 ea) Standard 12'(N/A) i Grid material v Pottom, Top Inconel inconel Inconel inconel Mid inconel inconel Zirc inconel Guide tube Sleeve SS SS Zire SS Mid Grid Pitch (in) 19.81 19.81 19.81 20.55 y Grid height Bottom, top (in) 1.322 1.522 1.322 1.522 Mid (in) 1.322 1.322 1.500 1.322 Guide tube sleeve length (in) 2.590 2.590 0.750 2.590 Control rod travel (in) 162.2 162.2 162.2 143.125 (From 259 steps) Reactor Vessel engagement: Top: Pins on fuel Pins on fuel Pins on fuel Pins on Upper Core Plate' Bottom: Pins on Pins on Pins on Pins on Core Plate Core Plate Core Plate Core Plate M,. ~.,,.

GENERAL FUEL FEATURES COMPARISON (CONT.) Feature XL (5 ea) XLR (136 ea) V5H (52 ea) Standard 12'(N/A) Guide tube overalllength (in) 178.695 179.055 179.055 154.117 Guide tube 0.017 0.017 0.017 0.017 wall thickness (in) Guide Tube OD (in) 0.484 0.484 0.476 0.484 Control rod radial clearance Dashpot (in) 0.006 0.006 0.006 0.006 Above Dashpot 0.035 0.035 0.031 0.035 (in) Dashpot length 33.845 33.845 33.845 23.995 (total) (in) t l Dashpot 0.017 0.017 0.017 0.017 l wall thickness (in) Guide tube material Zirc Zirc Zirc Zirc 1",. -,,s

GENERAL FUEL FEATURES COMPARISON (CONT.) Feature XLR STD 12' Fragema (12') Average Spring force (ibf) 1372 1072 1372 Hydraulic lift force (Ibf) 1652 1387 611 Weight (Ibf) 1692 1445 1445 Net holdown force (buoyancy 1242 1000 2076 included) (Ibf) Average guide tube normal 3490 2732 5023 Stress (psi) % Difference (Fragema = 1.0) -30.5 -45.6 0.0 $) l . T J,.,,,,

xy 4 - ~ ~ ,pa GENERAL FUEL FEATURES COMPARISON (CONT.). 4 l Feature XL , Sta idard g_ Spider assembly l l Material 304 SS 304 SS Spnng inconel inconel Travel 1.05 0.88 1 Poison Material Ag-In-Cd Ag-!n-Cd i Length 158 87 142.0 Diarneter 0.341 0.341 SprF, inconel Inconel Clad ID 0.344 0.344 OD 0.381 0.381 Inserted length 170.370 150.585 J P Number of rodlets 24 24 Shuming Rapid Non-Rapid refueling refueling Axial repositioning Normally Various quarterly .s 2,. ~,,s --J

f _ ~ Z. ~ T._ "_ __ R INDUSTRY EXPERIENCE Events of Control Rod Failure to Fully insert Foreign Material -Surry 1 (6/84)- spring clamp from Fuel Assembly Top Nczzle lodged in rodlets -Palo Verde 1 (1/88)-ball bearing lodged in rod guide tube STP 1 (9/87)- ball bearing lodged in rod guide card ANO2 (3/90)- unidentified material in rod guide card Braidwood 2 (4/94)- dowel pin lodged in rod guide card ASCO (1995) - foreign material in guide tube French 900 Mwe plants (1/90,1/92) - foreign material in rod guide tube, rod guide card - French 1300 Mwe Plant (5/94) - foreign material in rod guide card

- ~. 3 ~ INDUSTRY EXPERIENCE DEGRADED CONTROL ROD INSERTION Control Rod - Drive Line Degradation Maanshan 1 (9/88) - hafnium rod swelling SONGS 1 (12/81 - rod assembly weld failure Foreign plant (11/84) - loose CRDM breach guide screw lodged and prevented motion (NRC IE Information Notice 85-14) French 900 Mwe plants (6/83,12/88, 7/89, 9/89) - guide card deformation o French 900 Mwe plant (1989) - broken rodlet Westinghouse plants (1986-1981) - RCCA spider vane separation (3 events) Corrosion Products ANO2 (2/91) - corrosion products in rod guide tube BW plants in 1993 (Oconee 2, Oconee 1, TMI 1) - crud deposits in upper drive line. This resulted in slower rod drop times, but no failure to fully insert. ' s ",,. -, m.

. L. l ~' ~ ~~ INDUSTRY EXPERIENCE DEGRADED CONTROL ROD INSERTION Guide Tube Bow Ringhals 4 (8/94)- 12' 17x17 Fragema fuel with burnup of 37,000 to 41,000 mwd /mtu -Rod stuck at 18 steps out; 4 rods stuck during subsequent testing -Rod drop testing showed rod speed affected well before dashpot -S-shape;l bow observed in fuel under problem control rods Doel 3 (Fall 1994)-112' 17x17 Siemens fuel with burnup of 37,000 mwd /mtu -Rod could not be fully inserted; moved freely in top half of travel -Fuel assembly was discharged; no exam of fuel or control rod French 1300 Mwe Plant (10/92) - 14' fuel, rod stuck at 7 steps; fuel assembly was bowed (25mm) Cause Not Determined WWE-1000 (Russian plants, after 1992) - rod stuck near the bottom of the core Koziodny NPP Unit 6 (Oct.1994) - rod stuck near the bottom of the core zA,.-,.,,.

POTENTIAL CAUSES AND CHARACTERISTICS Debris - Foreign Material Sudden decrease in speed Rods hard to move Reproducible Burnup/ time independent

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I i POTENTIAL CAUSES AND CHARACTERISTICS l l Control Rod or Drive Line Degradation Core location dependent - no radial shuffling at STP Likely to be repeatable Ag-110m trends - show no changes t s ucwwet.tS pot.15aari

l t POTENTIAL CAUSES AND l CHARACTERISTICS l l Corrosion Products Guide tubes, rod cladding, guide cards-Gradual effect i Worsens with time 1 Not core location dependent Tight clearances required t t [ l M,...,... i

POTENTIAL CAUSES AND CHARACTERISTICS Guide Tube Bow Banana bow -Prevelant -Not known to degrade RCCA insertion -Not burnup dependent S-shape bow 9 -Rare t -Can cause RCCA interference. -May be burnup dependent 1 IMi-i$ % bihN$ ^- - - --

ACTION PLANS Approve safety evaluation for Unit 2 Cycle 5 by 2/8/96 inspect (drag testing, visuals) the 4 high bumup Unit 2 fuel assemblies discharged to SFP during 2RE04 - proposed start date is 1/22/96 Complete review of fuel and control rod manufacturing records Perform hot, full flow rod drop testing during next shutdown in either unit Perform Unit 1 hot, full flow rod drop testing approximately 60 to 75 EFPD after the 12/18/95 reactor trip (subject to change if forced outage occurs) Perform Unit 1 hot, full flow rod drop testing at EOL during plant shutdown at the start of 1RE06 Evaluate performing cold, full flow rod drop testing during 1RE06 plant shutdown on affected rod banks, and perform if necessary. Decide by 3/18/96. t 2 1. ,,s

.. _ _ _ ~ _ _ ____'i_^~ ACTION PLANS, CONTINUED l Revise Unit 1 Cycle 7 loading pattern to minimize or eliminate high burnup fuel assemblies (> 40000 MWD /MTU at EOL) in rodded locations Evaluate performing RCCA testing 1 ~ Visually inspect affected Unit 1 fuel assemblies (scheduled for discharge) during l 1RE06 Complete further testing of affected Unit 1 fuel assemblies as soon as practical following 1RE06. Testing under consideration includes: drag testing of guide tubes, boroscopic examination of guide tubes, feeler gauge measurements in guide tubes, and assembly axial growth measurements. .}}