ML19260B234

Responds to NRC Ltr Re Hot Leg Temp Anomaly.Reactor Vessel Internals Inspected on 791017 Did Not Reveal Cause of Effect of Temp Switching Anomaly.Describes Actions Planned If Test Program Results Call for Action Criteria

ML19260B234
Person / Time
Site:	Arkansas Nuclear
Issue date:	12/03/1979
From:	Trimble D ARKANSAS POWER & LIGHT CO.
To:	Reid R Office of Nuclear Reactor Regulation
Shared Package
ML19260B235	List: ML19260B234 ML19260B236
References
NUDOCS 7912070446
Download: ML19260B234 (13)

Text

, .

u _.

ARKANSAS POWER & LIGHT COMPANY POST OFRCE BOX 551 UTTLE ROCK. ARKANSAS 72203 (501) 371-4000 December 3, 1979 2-129-1 Director of Nuclear Reactor Regulation ATTN: Mr. Robert W. Feid, Chief Operating Reactors Branch #4 U. S. Nuclear Regulatory Commission Washington, D.C. 20555 Sub j e c.t : Arkansas Nuclear One - Unit 2 Docket No. 50-368 License No, h7 F- 6 Hot Leg Temperature Anomaly (File: 2-1520)

Ref: Letter: Reid to Cavanaugh same subject, undated; received November 20, 1979.

Gent 1emen:

The referenced letter requests that we report to you the results of the inspections which were made of the ANO-2 reactor vessel internals to look for causes or effects of our hot leg temperature anomaly. Our report is enclosed as Attachment A. Futher, you have requested that we provide a sumary o f our testing program during startup following the vessel inspections. This summary is requested to include the a:tiot criteria relating to the proper reassembly of the reactor and vessel. This information is included as Attach-ment B. In addition, you havc requested that we identify our proposed criteria for evaluation of the hot leg temperature anomaly as it is related to continued operation, power in-crease and reporting to the NRC. Our action criteria relatad to the hot leg temperature anomaly are given in Attachment C.

In the event that our test program yields results which take us into the action criteria, our actions criteria, our ac-tions will be as follows: The data and results will be Aco/

' 5 1514 233

/

/

MEYEsEA MICOLE SOUTM UTILITIES S EM 7912070 ;f

b .

2-129-1 Mr. R. W. Reid December 3, 1979 reviewed by the Plant Safety Committee and by Combustion Engineering representatives. If this review determines that an unreviewed safety question may exist or that there is a significant deviation f rom ernect ed perf o rmance, the data Tnd results will be presented to .he Safety Review Committee

( S RC) . The NRC resident I&E inspector will be inf ormed of the situation and all pertinent information made available to him. If the SRC determines that an unreviewed safety ques-tion or other significant deviation from expected performance exists, the NRC would be immediately informed. The evalua-tion of test program results might require that further or repeat testing be performed, but escalation to higher power levels would not be made until the questions were satisfactor-ily resolved.

~

Ve ry tru.y yours, M b.

David C. Trimble Manager, Licensing DCT /THC /ew Attachments 1514 234

. s ATTACHMENT A ANO-2 REACTOR VESSEL AND FUEL INSPECTIONS

1. Reactor Vessel and Internals

a. Introduction On October 17, 1979, a visual inspection of the Reactor Vessel Internals and the Reactor Vessel /

Internals Interfaces was completed. The inspec-tion was performed to determine the possible cause and/or effect of the t enpe r a tur e anomaly in the reactor vessel outlet ducts discovered during start-up power testing. The inspection was docu-ment ed wi th video tape and /or pho tographs.

b. _

Summary No indications were found, during this inspection, which would indicate the cause or any effect of the temperature switching anomaly. Internal components show typical light discoloration caused by an oxide film resulting from operating conditions. Inter-face surfaces, such as alignment keys and guide lugs display signs of normal contact. These dis-coloration and contact areas are consistent with data recorded during the post PVMP (Precritical Vibration Moni toring Program) , inspection and can therefore be attributed to the movement of the in-terface surfaces as a result of differential ther-mal expansion between the carbon steel vessel and the stainless steel internals. In addition, be-cause of close clea. ne's between the keys and key-ways, a lateral shift of the internals during assembly could result in contact between one side of the align-ment keys and /or guide lugs and their respective key-slots.

During the inspection, a dimensional survey was made in the flange-holddown ring and outlet nozzle regions. All dimensions are shown to meet the acceptance criteria established. The radial gap be twe en the core support barrel and reactor vessel _

outlet nozzels was shown to be within acceptance limits.

1514 235

I

. 1.

~

c. Description and Results The inspections were conducted in three phases to correspond with the sequerce of reactor disassembly.

The first phase was a general vi sual inspection of the reactor flange areas, including measurements to determine the vertical position of the reactor inter-nals with respect to the reactor vessel-reactor vessel head mating surface.

Hol ddown Ring Contac t wi th Reac t or Vessel Head The holddown ring made continuous contact with the vessel head between the alignment keyslots. The area of the head where ring contact occurred appear-ed as a shiny band approximately 3 /16" wide. Light scratches in the radi al direction are superimposed on this band. These scratches are due to relative motion of the ring and head caused by differential thennal expansion between the carbon steel head and the stainless steel holddown ring and in part to the rolling action of the ring rotation as the head is bolted down. The ring itself showed no unusual conditions.

Dimens ional Survey o f Flange-Hol ddown Region The vertical locations of the Core Support barrel (CSB) and Upper Guide Structure (UGS) flanges and the holddown ring relative to the reactor vessel mating surface were verified by taking the three dimensions shown on Figure A-1.

Dimension A is the distance from the top of the core support barrel upper flange to the reactor vessel mating surface. The average reading of 4.249" found during this inspection matches the 4.248" average found during the post PVNT inspec-tion, well within the acceptance criteria limits of 4.248 .005 inches.

The amount of deflection /preload and holddown ring will experience when the head is bolted to the ,

reactor vessel is given by Dimension C - Dimension A. The average holddown ring de flection calculat ed from Dimension C minus A was .148" which is a slight decrease when compared to compatible data from post PVMP. To some exte.t. this difference can be attri-buted to the flatness effect of the CSB and UGS flanges.

^.u e to fabrication, it is possible to have high and low areas on the interface surfaces of both flanges which are s till wi thin fla tness -al lowab l e s . When the i514 236

measurements are taken in this region with the flanges stacked together, as they were during the PVMP inspection, it is conceivable to have the high area of one on top of the high area of the other causing an increase in dircens ional readings .

During this current inspect!on, measurements were taken with the flanges separated, a result of the fuel spring upli f t on the UGS. This separation eliminated the effect of any flatness effect which might exist during operation, yielding a slightly lower deflection. A calculation based on this measured deflection indicates no significant change in the holddown clamping force. The holddown clamping force provides adequate margin against rocking or sliding of the internals for operating condition.

Dimension A+B-C represents the gap between the UGS and CSB upper flanges. This distance corresponds to the uplift effect of the fuel springs on the UGS. The average gap measured of .864" was well within the acceptance criteria limits of .855

.050".

In an attempt to more accurately assess the con-dition of the fuel springs across the core, an addi tional dimension from the reactor vessel mating surface to the top of the UGS upper flange was taken. Thi s measurenent showed that the UGS was level with respect to the reactor vessel within

.017 inches. Since the UGS rests on top of the fuel springs in their preloaded, undeflected posi-tion, the levelness would indicate the condition of the fuel springs in various areas of the core with respect to one another. The .017" levelness of the UGS is well within drawing tolerances.

Control Element Extension Shafts The control element assembly extension shaits, extension shaft guides, and instrument stalk: were given a general visual inspection. No anomalies were observed, with all shafts and guides appearing .

normal.

The second phase of this inspection was a close visual survey of the upper guide structure and accessible areas of the core support barrel and reactor vessel head. Areas of particular interest and importance are the flange and head keyways, alignment kays, guide lugs, and CEA shroud and instrument 6ube assemblies.

1514 237

~

4_

Core Support Barrel Alignment Keys The present condition of the core support barrel alignment keys shows a close correlation with results recorded during post PVMP inspection. The contact areas are indicated by discoloration, and show no measureable wear. These areas of contact can be attributed to relative thermal expansion and a closing of the gaps between the keys and the head or UGS resulting from internals assembly. .

The 00 alignment key shows no indication of contact with the UGS or vessel head keyways. There is some discoloration of the 270o side of the key, but no indications of wear.

The 90o alignment key has markings on its 180 side in the area of the head keyway. These mar:.ings show no signs of depth and can therefore be attri-buted to relative thermal expansion and not wear.

No indications of contact are evident in the UGS flange region.

The 1800 alignment key exhibits some marks on its 2706 side. In addition, contact was made on the 90o side of the key with the UGS. Since no depth was apparent in these contact regions, they can be attributed to relative t he nna l growth or interfer-ence during assembly.

Light discoloration at the vessle head was found on the 270o alignment key - O side. This contact was not present at the post P NTT inspection. Once again, since no depth was apparent at this surface.

it can be attributed to contact during internals assembly or temperature changes. No indications of contact are evident in the UGS flange keyway.

UGS and Reactor Ve s s e l Al i gnmen t Key Slots The keyslots in the UGS flange and reactor vessel head had marks corresponding to the marks noted on _

the core support carrel alignment keys. The UGS keyslots showed no anomalies or contact areas.

Light wear due to contact between the vessel head and the alignment keys was visible in the 1800 keyway -270o side and 00 keyway - 270o side. These contact areas are consistent wi t h pos t P VMP inspec-tion results and can be attributed to thermal growth. In addition, the 180o ke7way - 270 side 1514 238

contains a 1" long slightly raised dimple, outward from the contact surface. Since this raised metal is away from the slo t-key contact area, it is not attributed to interference during normal assembly or operation. Because of location, the raised dimple will not affect component integrity or c ompo nen t interfaces during normal assembly and operation.

Guide Lug Slots and Guide Lug Inserts All of the guide lug inserts showed contact mark-ings with the fuel alignment plate keyslots.

Slight wear was noted on the Oo, 90o, and 2700 guide lugs. The corresponding 00, 900, and 2700 guide lug slots showed signs of light wear, each estimated to be less than .010 inch in depth. All con'act and wear areas are consistent with post PVM? inspection results. The light wear seen on the guide lugs and guide lug slots are a result of a wearing process. The original assembly clear-ance at the guide lugs were a total of .009 .012 inch, which is slightly less than drawing allow-able. With the smaller clearance and a tighter fit than at the alignment keys by design, a small lateral shift of the internals at assembly will result in contact at the guide lugs. This contact, followed by any movement of the UGS resulting from operating loads will result in a normal wearing-in process.

Control Element Shrouds The peripheral shrouds at the outlet nozzel loca-tions were given a general inspection to verify that no unusual conditions had developed. These shrouds, experiencing the highest cross flow from the outlet nozzles, appeared satisfactory with no evident effects of operation other than the typical discoloration which appears on all internal struc-tures as a result of operation. The instrument -

tubes and ins t:ument tube attachments, adjacent to the outlet nozzles, appeared asmanuiactured with no anomalies observed.

1514 239

t Fuel Alignment Plate The inspection of the underside of the fuel align-ment plate consisted of a general inspection of the outer fuel assembly locations near the outlet nozzle. It was verified that the CEA chroud bolts and lockbars were intact. There was no evidence in the fuel pin holes of the plate of any contact with

, the fuel pins. However, there was a faint light colored outline of the holddown plate visible on t he bo t tom surf ace in some fuel assembly locations.

No indication of any wear was identified in these areas.

Core Support Barrel Outlet Nozzles The core support barrel outlet nozzles were viewed using a right-angle camera lens. No anomalies were observed. The radial nozzle gap between the CSB and RV appeared to be unifonm around the circum-ference of both nozzles.

The gaps at the bottom of the nozzles were measured remotely to compare with the internals field align-ment data. This total measured di ame terical gap of

.130" was smaller than the .170 t .020 inch gap acceptance criteria. This discrepancy can be attributed to the size and shape of the measuring tool. The tool consisted of a stepped feeler gauge in increments of .010 inches, which means that the actual gap could be up to .0095 inches larger radially than measured. In addition, the feeler gauges consisted of 3" wide rectangular cross sections which could not take into account the nozzle curvature. The effect of curvature is up to

.015 inches, radially. Considering the effect of curva ture alone, on top cf the .130 in. measured gap, results in a gap of .160 in, which is within the acceptance criteria. Any additional discre-pancy could be the result of radial thermal growth of the "SG relative to the reactor vessel, closing the outlet nozzle gap.

The third phase of this inspection was a visual review of the four vertical seams in the core shroud. The flow skirt and reac tor vessel bo t tom Faad regions were also inspected,'eing likely

.ocations for any loose or displaced components.

1514 240

Core Shroud The four vertical seams in the core shroud were observed and no anomalies were found. The core shroud face in the 90o-180o quadrant showed signs of contact with the spacer grids from the adjacent fuel assembly; however, no wear was evideat.

Core Platt and Fuel Alignment Pins The core plu te and fuel alignment pins in the regions of the four vertical core shroud seems were viewed.

No anomalies or unusual conditons were observed.

There was a faint light-colored outline of the bottom of the fuel bundle post visible on the core plate around the fuel pins in some fuel assembly locations.

No indications of any wear were evident in any of the locations.

Bottom Head and Flow Skirt The bottom head region of the reactor vessel looked clean with only a f ew small whitish-colored parti-cles floating in the area. These particles h,ve been noted during previcus inspections in the bottom head region of other operating reactors and can be attributed to boron deposits from the pri-marly coolant.

The flow skirt looked satisfactory with no anoma-lies or unusual conditions observed.

2. Fuel Assembly Inspections:

a.

Introduction:

During the period from October 14 through October 17, 1979, a visual examination of pre-; elected fuel element assenbli es and control element assemblies was conducted.

The examination of the fuel assemblies was con-ducted in three phases, all of which were carried -

out ut il zing closed circui t underwater television.

All inspection sequences were recorded on video tape.

b. D?scription:

The first phase of the fuel inspection was con-ducted while all fuel ass emb lies were in place in 1514 241

the core and consisted of a complete core scan noting the condition of the upper end fittings and spacing between assemblies.

The second phase was conducted after certain fuel assemblies were removed from the core to provide visual access to inspect the core shroud. This phase consisted of a vertical scan of the exposed sides of the assemblies adjacent to those removed.

Particular note was taken as to the condition of the upper end fitting springs, the top of fuel rod to flow palte gap, the relative length between fuel rods and poison rods, the conditon of the spacer grids, the uniformity of the flow channels between fuel rods and the spacing between fuel element assemblies.

The third phase of the fuel inspection was con-ducted on six pre-selected assemblies which had been removed from the core. Each assembly, while being suspended from the re-fueling machine grapple was lowered, then raised in front of the underwater television camera. All four sides of each assembly were scanned noting the overall appearance, spacer grid condition, guide post head wear, fuel rod to flow plate gap and end fitting wear.

Two control element ass emblies were vi sually in-spected. With the underwater television camera in a fixed position the CEA was raised from its parent fuel assembly then lowered while all five fingers were viewed.

Filowing complete re-assembly of the reactor core, a final verification of proper f t. e l assembly loca-tions was conducted.

c. Results During the core scan, wi th all assemblies in place in the core, no unusual conditons were found to exist. Fuel assembly spacing was uni form through-out the core. Upper end fitting discoloration was uni f o rm. All upper end fittings appeared in good condition with no observable wear or anomalies.

The vertical scan of the exposed fa.es of the installed fuel at the five locations denoted in Figure A-2 revealed no unusual conditions or anoma-1514 242

_9_

lies. The upper end fittings appeared normal. The fuel rod to flow plate gaps were uniform, the flow channels uniform and no assembly bow was detected.

The detailed, out-of-core inspection was conducted on assemblies AKA 005, AKA 002, AKL 502, AKC 412, AKC 504 and AKA 108. As was the case in the above two pahses of the inspection, no abnormal condi-tions were found to exist. On assemblies AKC 412, AKC 504 and AKC 502 slight grid wear was observed on the faces of the assembly which contact the core shroud wall. This conditon is con idered normal.

During the visual inspection of the core shroud wall, a burnished area was detected at the fuel assembly AKC 503 location at an elevation about mid-core. The north face of this assembly, which contacts the shroud wall, was inspected. The number six spacer grid showed slight wear. This elevation matches the burnished area on the shroud wall.

During the final core scan for verification of fuel assembly locations, a burnished condition was noted on the holddown plates of the upper end fittings.

The assembly identification number is engraved in one corner of the holddown plate. In some cases, it appeared as though part of the number was obs-cured or missing, thus indicating wear. The video tape of this core scan along with the fuel inspec-tion video tapes were taken to CE, Windsor. The tapes were reviewed by experienced Fuel Inspection Personnel. It was detennined that the burnished areas which exist where the holddown plates contacc the core alignment plate were highly reflective and when encountered with the two 500 watt underwater lights had the effect of " washing out" the number in the video picture. No wear other than the burnished surface was noted.

Control Element Ass embly numbers 01 and 3 8 were visually inspected. No wear or abnormal condition -

were found to exist.

1514 243

Figure A-1 CO:G. B111G :s j\

v )

, /\ *

'l UPPER GUIDE STRUCTURE FL1J!GE A

v

%/

\

/

CORE SUPPORT 3ARREL FLA';GE REACTOR VESSEL W -- _ - .

t 4

0 1514 244

Figure A-2 A B C D E F G H J K L M N P R

- m p

/ . . _ _ _ _

N 1 - -- --

7

\

2

~

3 f 5 .

l -_-

6 _ _ ,

7 -

c 270 9

11

>z

\

\ XX. XXX

'N XX' > XIX .

x

/x /

~

TV C a Hole -

0 "X" denotes Fuel Asser;bly to be recoved for internals inspection. 1514 245

. .