NRC-86-3181, Forwards Listed Nonproprietary Documents Re Rod Cluster Control Assembly Performance,Per Request.Topical Rept WCAP-8183, Operational Experience W/Westinghouse Cores Issued for Input to NRC Annual Fuel Performance Rept

Forwards Listed Nonproprietary Documents Re Rod Cluster Control Assembly Performance,Per Request.Topical Rept WCAP-8183, Operational Experience W/Westinghouse Cores Issued for Input to NRC Annual Fuel Performance Rept

ML20214X125
Person / Time
Site:	Point Beach
Issue date:	11/25/1986
From:	Rahe E WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:	Ellen Brown NRC OFFICE FOR ANALYSIS & EVALUATION OF OPERATIONAL DATA (AEOD)
References
NS-NRC-86-3181, NUDOCS 8612100556
Download: ML20214X125 (14)

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Westinghouse PowerSystems pisdnpennsyson,a 15230 03s5 Electric Corporation November 25, 1986 NS-NRC-86-3181 Mr. Earl J. Brown Office for AEOD U. S. Nuclear Regulatory Commission N~ W Washington D.C. 20555

Subject:

Request for Information on Westinghouse Rod Cluster Control Assembly Performance

Dear Mr. Brown:

p p In response to your request to Wiscons-In Electric Power Company for information related to the Point Beach LER on RCCA cladding wear, enclosed is documentation summarizing Westinghouse RCCA performance.

The following non-proprietary documents attached for your information are representative of technical information on RCCA wear submirted to utilities:

1) Summary of visual inspections of the RCCAs at the Point Beach Nuclear Plant (Unit 1, Cycle 11/12, Unit 2, Cycle 9/10).

2) Westinghouse customer information letter on RCCA wear.

3) Extracts from a brochure containing a summary of Westinghouse current design experience regarding RCCA wear.

In the past, the vehicle used to transmit information on RCCA performance has been the topical WCAP-8183 " Operational Experience with Westinghouse Cores". This is updated annually and issued for input to the NRC Annual Fuel Performance Report.

Very truly yours, ww.,

. Rahe , Jr. , Manager Nuclear Safety Department

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Sum dry of Visual Inspections of ECCAs at the Point Beach Nuclear Plant During the Cycle 9-10 refueline shutdown at Point Beach Unit 2 in June, 1983, site p.Iw 1 observed what appeared to be t lam 4 tug wear in excess of the current Westi, pause critaria on sane Rod Cluster Centrol Assablies Doms). 'Jhe K%2s in cuestion were of the spider urx:nted design with 16 redlets per IC2 and are ocupatible with the 14x14 fuel design used at Point Beach. '!he abscIber material is Silver - Indlun -

Cadmium. All of the Unit 2 KIms were examined with a pe.%, and ho w on the observed wear, Westinghouse reo:mnended rep 1=m of one KIA and axial repositioning of the,norml parked position of the run61ni:vy XI2s by 2-3 steps in order to m4M=49e additional wear. A further rer===rx5ation was to inspect the Unit 1 KI2s for wear indi-caticms during the Cycle 11/12 r=*=14=' shutdown (Unit I had appmci-nately 2 more cycles of % i.icn than Unit 2). 'Jhis inspecticri was recently ccupleted and resulted in a Westir-?+1- reconnendaticrt to

( replace 21 of the 33 KI2s with an inferred e =1m1=ted wear depth in excess of the current wear depth criteria.

WEAR CESERVATICES

'Ihe redlets were pv4 4-:4 using a 35m camera coupled to a peri-scope. An analytical nodel was developed which was used to infer the depth of wear indications fmn the observed wear scar widths as measured fmn the p /4 Mis. Quantitative measurunents of wear depth have not been pere ctned to date.

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'No types of war patterns were c$setvie and are characterized as fol-laws: (1) Axially ocntinuous wear scar on the redlet of varvbg depth l t

probably caused by RIA axial stesping notien and scrams, (2) Approxi-nately ene-inch long smre marks at elevaticms coitwiding to the guide plate surfaces.

The wear indicaticms are r--- mmtly en the portion of the cladding surface facing the center or hub of the ICCA. 'Ihe attached sketch shows ,

l wear patterns which are typical of the worst d=2ve wear. Note that, I by design, scue wear is expex+ad due to contact between the RIA redlets and the guide surfaces.

l MIX 2T WEAR CRITERIA

'Ihe um.a.A limiting wear depth crittricn is related to =>- . : -r redlet Clad collapse. 'Ble allmbl* Wear depth was determined based Cn auto-clave tests (at taiperature and pressure) wherein maald, hollow tubi.w samples with similated wear scars were r . sized until collapse hw. .d. '!his test of a hollow tube did not take into account the sup-port provided by the =hr material, and therefore is conservative. -

Westinghouse roommenr%d that the Point Beach RIAs be replaced based on this critericn and on plant specific operating history.

A preliminary evaluation by Westinohouse indicated that an unreviewed safety questicm does not exist.

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, . NFD CUSTOMER INFORMATION LETTER ON RCCA WEAR

Subject:

NFD Customer Information Letter RCCA. Wear

Reference:

March, 1984 letter TO: Customer -

Dear Sir:

The referenced letter issted to you earlier this year provided a summary of visual inspections of the Rod Cluster Control Assemblies (RCCAs) at the Point Beach Nuclear Plant and recommendations for future actions. Those inspections identified the existence of wear marks on the cladding of the RCCA absorber rodlets. Some of the wear scars were estimated to exceed the existing Westinghouse maximum wear depth limit.

As a result the affected RCCAs were replaced based on our recommendation.

The Point Beach RCCAs had operated for 13 years from Cycle 1 initiation with approximately 80% availability.

New developments related to RCCA rodlet cladding wear have been completIed.

These include a revision to the wear depth limit and a new method of obtain-

ing wear depths from photographs. I have attached an update to inform you of the status of on-going efforts by Westinghouse in the area of RCCA wear. I would be happy to answer any questions you may have on this subject.

Signed Fuel Projects Engineer o

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RCCA WEAR UPDATE Introduction ,,

In our last information letter on Silver-Indium-Cadmium Rod C1'uster Control Assembly (RCCA) wear, we brought to your attention in-reactor RCCA wear experience. The following sumarizes that letter.

The RCCAs at Point Beach Unit 2 (End of Cycle 9-June, 1983) and Unit 1 (end of Cycle 11-December,1983) showed significant levels of wear. Photographs of the wear scars were taken and a method was developed to estimate wear depth. Some of the RCCAs had estimated wear depths exceeding the then applicable wear depth limit and were therefore replaced. Two types of wear scars were seen : (1) Stepping Wear - an axially continuous scar on the rodlet caused by RCCA

- axial stepping motion and scrams, and (2) Fretting Wear - an approximately one-inch long score mark at elevations corresponding to guide cards (see Figure 1). Our letter on this subject also described the basis of the Westinghouse wear depth limit and presented preliminary recommendations for future plant specific action. ,

Note that it was concluded that no unreviewed safety question existed at the Point Beach units.

During the inspections it was observed that both types of wear occurred at the interface between the RCCA and the upper internals guide tube guide plates. Based on this study,as well as' previous studies, no unusual or significant wear has. occurred at the interface between the RCCA rodlet tip and the fuel assembly guide thimble tube. Therefore, the observed RCCA wear has no adverse impact on the operation or safe performance of Westinghouse fuel assemblies. .

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The following sections provide updated information on.further inspections, t /

new developments, and plans for hot cell work on the Point Beach RCCAs. -

l RCCA INSPECTIONS PERFORMED -

In addition to the two Point Beach (2 loop, .14x14 fuel) inspections r

described in our previous letter, Westinghouse ha> visually . inspected, , /

l RCCAs at three other reactors, two 2 loop'14x14 and one 4-loop 15x15 plant.I. ~ '

l As a result of these inspections we havehoncluded that the types of r I wear observed at other reactors are similar to those observed at Point '

i Beach. The wear did not appear as severe as that at Poirt Beach;- ' '

l however, the RCCAs were several years you'nger.

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NEW DEVELOPMENTS -

. Since the last information letter Westinghouse has completed a more exhaustive analysis in three areas described below: (1) the wear depth

_ limit; (2) the interpretation of photographs to estimate wear depths; and (3) axial repositioning of RCCAs. The following briefly describes these new approaches.

First, a new approach has been taken to establish a Westinghouse RCCA cladding wear depth limit which is used to meet the existing FSAR critiria. )

The criteria related to control rod clad wear are that the absorber material shall be isolated from the coolant and that the RCCA must be capable of moving freely into the fuel assembly. The impact of absorber clad wear on these criteria for the Ag-In-Cd RCCA was previously considered by Westinghouse by basing the allowable wear on prevention of clad collapse.

For conservatism no credit was taken for the support provided by the absorber material. Two dimensional elastic-plastic finite element analyses subsequent to the original Point Beach evaluation have demonstrated that the absorber material can prevent the clad from collapsing in the observed wear geometry.

Where the absorber material backs the cladding, the allowable wear depth has been increas'ed by several mils (4-6 mils depending on the RCCA design) while still meeting the existing FSAR criteria.

The second area of improvement is the wear scar photograph interpreta-tion method. Based on test data and a thorough review of the wear phe-nomenon a new method of estimating wear depth from wear scar width has been developed. The data are derived from out-of-pile tests performed in reactor grade water at reactor operating temperature and pressure and include measurements of wear scar width and maximum wear depth. Results of the application of the new method suggest that our original method of interpreting photographs may yield over estimates of wear depths by

. several mils at most elevations above the continuous guide area.

• Because the geometry of the aforementioned tests is not identical to the

. reactor geometry, an effort is also being made under EPRI sponsorship to obtain hot cell measurements of worn RCCAs to improve our understanding of the RCCA wear phenomenon. ,

The third area which has been addressed is axial repositioning of RCCAs.

If a significant level of wear is observed at your reactor, a redefi-l nition of the axial parking elevation of RCCAs by 2-3 steps can extend the service life of RCCAs for at least one additional cycle. The bene-fit is derived from spreading the fretting weer over more clad surface, thus reducing the maximum wear depth achieved. An evaluation shows that the RCCAs can be inserted several steps deeper than the normal parking position without adversely impacting operation or safety of the plant.

It should be noted that a Technical Specification change may be i required. If a change to parking elevations becomes necessary, Techni-cal and licensing support can be obtained through your NFD Fuel Projects Engineer.

With the increase in the wear depth limit and the application of the new i photograph interpretation method, the rejected RCCAs at Point Beach Unit I appear to be acceptable for continued use. Tnose RCCAs had oper-ated for 13 years with about 80% availability before the examination.

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ATTACHMENT 3 Westinghouse Rod .

R uluster Control .

Assembly Performance Review May 1986 S

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Introduction

. . As nuclear reactor systems log service rodlet cladding has occurred. In order beyond a decade, components not sub- to obtain a more comprehensive lect to periodic maintenance or evaluation of these observations, replacemen; may be expected to show detailed examinations of RCCA rodlets indications of aging. An ever-increasing having seen extended service were number of Westinghouse-supp!!sd PWRs performed by Westinghouse at its have now been in operation for such Research Laboratory in Churchill, Pa.,

extended periods, and some plant under a contract with the Electric Power operators have observed indications of Research Institute. This report provides reduced component integrity. Recently, a summary of the laboratory Rod Cluster Control Assemblies (RCCAs) examinations and an evaluation of the have shown signs that some wear of information obtained.

Cumulative Westinghouse RCCAs in Domestic Service 2800 2600 2400 2200 2000 -

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Background -

As active components in reactors, con- RCCA/ Upper Internals Layout trol rods are expected to have a lifetime limit in all light water reactors. It is -

Important to monitor and evaluate the operational experience of control rods in order to assess component performance and provide an updated basis for predic- Spider assembly-ting service lifetime. Wisconsin Electric l}

Power Company's Point Beach Nuclear i i hg Units 1 and 2 are among the leaders in RCCA rodlet j l accumulated service life for Westinghouse- \ ,

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supplied PWRs. After years of successful operation, Wisconsin Electric Card 8 Power replaced the RCCAs in Point Card 7 Beach and examinations were performed on the RCCAs taken out of service. Card 6 Westinghouse, along with Wisconsin Electric Power and EPRI, evaluated Card 5 O Guide tube selected RCCAs from Point Beach Unit 1 Card 4 I- l )] surfaces in order to determine a basis for Card 3 predicting operationallifetime reflecting experience in representative plant 2R operation. A campaign initiated in Card 1 October 1983 to investigate wear pheno- -

mena provided visual and photographic Continuous guide region ' ' "

information on 32 RCCAs from Point Beach Unit 1. - -

p T of p op One RCCA from Point Beach Unit I was fuel assembly selected for detailed examination and testing. Three rodlets were selected from this RCCA for sectioning and detailed testing in the Westinghouse R&D Hot Cell Facility in Churchill, Pa.

Ten segments of rodlets were taken for testing. Selection of the specific RCCA and rodiets was based on predeter-mined criteria including RCCA service history, visual observations of wear, and cracks in rodlet cladding.

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Description of Indications Examined ,

RCCA Performance are the result of control rod withdrawal i

Three potentist mechanical performance and insertion movements.

effects on RCCA cladding were studied. 3. Cracking l 1. Fretting Wear Short hairline cracks at the lower extrem-

Wear spots at specific locations along Ity of the cladding were observed on the length of RCCA rodlets were noted. some rodlets. The cracks extended These locations corresponded to loca- axially for approximately four inches, tions of the control rod guide cards in and penetrated through the cladding.

i the upper reactor internals when the RCCA would be fully withdrawn from  :

the core (l.a. the parked position). 'T There are eight guide cards in each '

control rod guide tube, which function to provide lateral support for individual RCCA rodlets when the RCCA is .:

withdrawn. There is also a longer ' "#

continuous guide directly above the upper core plate.

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i Service History The indications were observed after I approximately eleven years of "in plant .

• operation". Measurable wear is consis-tent with expected effects following 1

, service of such extended duration.

Wear was originally expected to be life limiting for Westinghouse RCCAs, and a

2. Sliding Wear life of approximately fifteen years had Wear indications appearing as scratches been predicted based on sliding wear in oriented axially on the surface of the conjunction with anticipated plant load rodlets were noted. These scratches change operations. [

correspond to interaction between RCCA rodlets and the continuous guide, and j

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Laboratory and Engineering Analysis A. Fretting Weer stepping of the RCCA during load Fretting wear is a result of vibrational c ange maneuvers.

contact between RCCA rodlets and the

. guide cards which provide lateral sup- Wear due to RCCA tripping occurs along port for the rodlets when RCCAs are the length of the cladding and, for the rodlets examined, averaged less than withdrawn from the core. Vibration is hydraulically induced by flow of the ten percent of the cladding thickness.

reactor coolant, and is therefore a Wear due to stepping is most pronounced continuous process when the reactor at the lowest portion of the control coolant pumps are in operation. In rodlet from sliding in the continuous general, wear scars resulting from guide region. The combined tripping and stepping wear at the worst location fretting were found to be approximately one-half Inch in length and included measured less than 20% of the cladding approximately one-third of the thickness. RCCA contact with the circumference of the rodlet. Depths of continuous guide causes most of the clad penetration observed varied in sliding wear, and sliding wear between approximate correspondence with the the RCCA and guide cards is minimal, amount of time the RCCAs were in the parked position and location. The C. Cracking degree of wear varied at different guide card elevations. The worst measured Several RCCA rodlets experienced hair-i depth of clad penetration from fretting line cracks in the cladding at the lower I

wear was approximately 65% of nominal extremity of the rodlet. The cracks were cladding thickness. Since there is a typically about four inches long and ,

large margin in load-bearing capability axially oriented. No circumferential for RCCA cladding, the immediate component of cracking was found. The concern if breach were to occur would lower extremity of the rodlet typicaliy l be deposition of radioactive isotopes experiences the highest fluence.

Into the reactor coolant.

Destructive examination of cracked Based on the history of this particular rodlets was undertaken to identify the RCCA, a wear rate of approximately 10% probable cause. Both metallurgical of cladding thickness per 10,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> analysis and dimensional examinations reactor operation was calculated. of the absorber and cladding were performed.

e exa Ina ns s we ence of B Slidin9 Wear Irradiation assisted intergranular The longitudinal scratches observed cracking and stresses due to interaction I

were determined to be the result of between absorber and cladding. The interaction between rodlets and the cladding material was type 304 stainless continuous guide located near the upper steel and the absorber was composed core plate. The wear is caused by slid. of silver (Ag), indium (in), and cadmium l ing movement and has two compon- (Cd) In proportions of 80,15, and 5 e'its. One component is attributed to percent, respectively. Negligible i RCCA trips and the other is due to deterioration of the absorber occurred due to the presence of the crack.

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a Scanning Electron Microscope (SEM) Corrosion tests were also performed to l results confirmed the Intergranular determine the extent of Intergranular nature of the cracks. Cladding tensile attack. The results provided evidence of i tests were also performed to correlate irradiation-enhanced segregation of mechanical properties with the SEM impurities at grain bounderles of the data. These tests provided evidence of stainless steel cladding.

a reduction in cladding ductility i consistent with high neutron fluence Dimensional and density measurements l near the tips. confirmed that irradiation-induced swelling of the absorber was the principal cause of tensile stress in the

The tensile tests were performed in an cladding, which resulted in cracking inert environment so as to avoid possi- after substantial irradiation.

ble effects due to corrosive agents.

Fracture surfaces of both the tensile test samples and the cracked cladding were intergranular and were consistent with other irradiation-induced embrittlement and stress corrosion cracking experience. Based on a

, number of EPRI programs, such fracture t

characteristics are considered to be associated with migration of sliicon and phosphorus impurities.

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