ML20212D909

Localized Rod Cluster Control Assembly (Rcca) Wear at PWR Plants, Engineering Evaluation Rept

ML20212D909
Person / Time
Site:	Kewaunee, Point Beach, Haddam Neck, 05000000
Issue date:	12/23/1986
From:	NRC OFFICE FOR ANALYSIS & EVALUATION OF OPERATIONAL DATA (AEOD)
To:
Shared Package
ML20212D903	List: ML20212D900 ML20212D904 ML20212D909
References
TASK-AE, TASK-E613 AEOD-E613, NUDOCS 8701050142
Download: ML20212D909 (12)

Text

AE0D ENGINEERING EVALUATION REPORT

• UNIT: Point Beach Unit 2 EE REPORT NO.: AE0D/E613 Kewaunee, Haddam Neck DATE: December 23, 1986 DOCKET NO.: 50-301; 50-305; 50-213 EVALUATOR / CONTACT: E. J. Brown LICENSEE: Wisconsin Electric Power Company Wisconsin Public Service Corporation Connecticut Yankee Atomic Power Co.,

NSSS/AE: Westinghouse /Bechtel Westinghouse / Fluor Engineers, Inc.

Westinghouse / Stone & Webster

SUBJECT:

LOCALIZED R00 CLUSTER CONTROL ASSEMBLY (RCCA)

WEAR AT PWR PLANTS EVENT DATE: April 24, 1984 (301/84-002); April 1, 1984 (305/84-003);

March 3, 1986 (213/86-015)

SUMMARY

The purpose of this study is to investigate RCCA wear relative to possible generic application to PWR plants. The study was initiated based on the events at Point Beach Unit 2 and Kewaunee. The RCCA degradation at these Westinghouse designed PWRs was found to be related to (1) wear during rod motion associated with startup, shutdown, reactor trip, and load following; (2) flow induced vibration with fretting wear between the rodlet and the guide card when RCCAs are at fixed positions for long periods of time; and (3) rodlet cladding cracking (intergranular stress corrosion cracking) that appears related to absorber / clad interaction. The results suggest that wear should be expected for all RCCAs depending primarily upon the number of fuel cycles of operation and/or the extent of plant load-following service. There was no operational data concerning similar wear problems at Babcock and Wilcox or Combustion Engineering designed PWRs. The safety concerns are that the degradation mechanisms may adversely impact RCCA insertion into the core or lead to f possible loss of absorber material with reduced shutdown margin or reduced e negative reactivity worth.

The extent of degradation can only be determined by inspection of the RCCAs during a refueling outage. Such inspections require relatively long-term planning. Hence, it is believed that the informal notification letters that that have been sent to licensees by Westinghouse may not be adequate to assure timely action by licensees.

• This document supports ongoing AE00 and NRC activities and does not represent the position or requirements of the responsible NRC program office.

8701050142 861229 3 PDR ADOCK 0500 P

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, In the events reviewed, the licensees have indicated that the events were judged as not reportable by the LER requirements (10 CFR 50.73(a)) or to the NPRDS.

Since RCCA degradation has occurred in Westinghouse designed plants, it is recommended that IE issue an Ir. formation Notice to alert licensees about the wear problem, and the need to plan and schedule inspections to determine the condition of all RCCAs. It is suggested that NRR consider whether there is a need to establish periodic inspections (such as a 10-year interval) of RCCAs and assess the feasibility of reactor water chemistry monitoring procedures.

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. INTRODUCTION This study was initiated as a result of two events involving wear of rod cluster control assemblies (RCCAs). The events were reported in LER 301/84-002 for Point Beach Unit 2 and LER 305/84-003-01 for Kewaunee. In each plant, the primary degradation was rodlet cladding wear at the guide card locations. This wear was attributed to flow induced vibration with the RCCAs in fixed positions for long periods of time. The corrective action taken at both plants was to insert the RCCAs a few steps at the fully withdrawn position of the RCCA. Each of these units is a Westinghouse two loop, PWR that began commercial operation in the early 1970s. Although the events were reported by an-LER, the report from Kewaunee indicated the event did not meet the 10 CFR 50.73(a) criteria for an LER. The Point Beach event was only reported because of the manner in which the technical specifications address RCCA control banks for both Units I and 2.

As a result of the corrective action taken to reduce the RCCA wear, Point Beach Units 1 and 2 submitted changes to the plant technical specifications related to the RCCA fully withdrawn position from 228 steps to 225 steps. However, it was determined that Kewaunee did not need a technical specification revision even though the fully withdrawn RCCA position changed from the original 228 steps to 226 steps. Therefore, this study investigated potential RCCA safety concerns and related possible generic implications at PWR plants supplied by Westinghouse. Also, the issue of whether a change in the fully withdrawn position of the RCCAs would result in a technical specification change was reviewed.

DISCUSSION The rod cluster control assemblies provide a rapid means for reactivity control during both normal operating and accident conditions. The RCCAs are divided into control or shutdown categories and are usually operated in banks which may contain several RCCAs. The control banks may be inserted or withdrawn to compensate for various reactivity changes during operation of the reactor and can trip (scram) to provide shutdown capability. The shutdown banks are reserved for shutdown use only and are always fully withdrawn from the core when the reactor is critical. The shutdown banks are inserted only when a reactor trip occurs. The total reactivity worth of the RCCAs must be adequate to meet nuclear requirements of the reactor. The RCCAs also provide shutdown margin so the core would be subcritical at hot shutdown with all RCCAs tripped with the assumption that the highest worth RCCA remains withdrawn.

The issue concerning potential generic concerns and extent of occurrence of RCCA wear was initially explored by a search of the Sequence Coding and Search System (SCSS) and the Nuclear Plant Reliability Data System (NPRDS).

The only reports obtained from the SCSS were the Point Beach, Unit 2 (Ref. 1) report and the kewaunee report (Ref. 2). There were no reports related to RCCAs found in the NPRDS. The only other information available about RCCA wear concerned wear in 900 MWe plants (Ref. 3) in a foreign country, but no details about the nature of the wear were available.

, Thus, the original events that elicited the interest, which were reported in 1984, were the only events identified by normal database searches. An event reported in April 1986 by an LER from Haddam Neck was subsequently discovered.

Since the normal database reviews did not identify any relevant operational experience the investigation process concentrated on discussions with licensee staff for the affected plants (Point Beach, Kewanuee, cnd Haddam Neck), as well as with personnel in the Westinghouse Electric Corporation, the equipment verdor. The primary issues of interest involved the nature of the degradation, the potential extent of the problem and its generic impl: cations, industry awareness of the potential problem, safety implications, reportability require-ments, identification or monitoring guidelines, and the appropriate inspection requirements. The specifics from each inquiry are provided in separate sections to identify and clarify certain issues.

Point Beach, Unit 2 The information on the RCCA wear at Point Beach, Unit 2 was reported in LER 301/84-002 (Ref. 1) on May 24, 1984. It became reportable as an LER because the resident inspector discovered that the fully withdrawn position of the RCCAs at 225 steps was not consistent with the technicci specifications which indicated 228 steps as the fully withdrawn position.

The actual discovery of RCCA wear was determined during Unit 2 refueling (number 9) on May 10, 198? and was reported to NRC by letter on August 18, 1983 (Ref. 4). It appears that a few RCCAs were scheduled for inspection during the refueling as part of a joint effort between the licensee, Westinghouse, and EPRI to assess possible long-term wear of RCCAs. Since degradation was identified in the initial small sample (a crack on one rodlet and more wear than anticipated on certain rodlets in other RCCAs), the inspection was extended to include all 33 RCCAs.

The longitudinal crack on one rodlet was approximately 2 inches long near the tip of the rodlet. It appeared to be a localized tubing defect and was not observed on other RCCAs.

The clad wear was postulated to be of two types. One type was wear resulting from flow induced vibratory interaction between rodlets and guide cards during long periods of steady-state power operation with the RCCAs essentially out of the core (both shutdown banks and control banks). This type of wear is essentially related to RCCAs being kept in fixed positions for long periods of time. The other type of wear appears to be related to sliding of the rodlets over the guide cards during RCCA stepping and trip insertion. The licensee reviewed this wear problem with Westinghouse. It was determined that the fully withdrawn position for RCCAs could be changed to 225 steps from the initially prescribed 228 steps. The primary bases to permit the change were that the three step change had a negligible effect on total rod worth and that sufficient shutdown margin existed at both beginning-of-life and end-of-life to meet safety analysis values. Thus, the August 18, 1983 letter from the licensee (Ref. 4) indicated that Technical Specification 15.3.10.A " Bank Insertion Limits" would be met with the PCCAs at 225 steps for the fully withdrawn position for Unit 2, cycle 10. There did not appear to be a need to change the specifications at that time.

, The discrepancy concerning the fully withdrawn position identified by the resident inspector during cycle 10 operation in April 1984 was related to the manner in which the technical specifications address the RCCA shutdown banks and control banks. The position for RCCA shutdown banks is addressed by Technical Specification 15.3.10.A.1. This item had simply identified that the

" shutdown banks shall be fully withdrawn." However, the control banks were covered by Figure 15.3.10-1 which apparently identified percent withdrawn (up to 100%) as well as the number of steps (228) for 100% withdrawn. This figure was not changed due to an oversight. It was this inconsistency in the number of steps, 225 compared to 228, that was observed by the resident inspector and led to LER 301/84-002. This situation was corrected for Point Beach Units 1 and 2 by a technical specification revision. The revision removed the number of steps for 100% withdrawn from Figure 15.3.10-1 and inserted a note for item 15.3.10. A.1 that stated the fully withdrawn position was " equal to or greater than 225 steps" and that the definition was " applicable to shutdown and control banks". A limited review of the technical specifications for a few older plants seems to indicate that the usual procedure is to provide the position as percentage withdrawn without stating the number of steps at 100%

withdrawn. This may explain why other plants believe a technical specification change is not needed to address this issue. However, plants that have followed the standard technical specifications will have the number of steps rather than the percent withdrawn. Thus, the need for technical specification revisions will be plant specific.

Kewaunee The licensee scheduled a limited inspection of 3 out of 29 RCCAs durino the refueling outage at the end of fuel cycle IX because of wear that was found at another facility (Point Beach). The RCCA assemblies are a spider mounted design which contain 16 rodlets per RCCA to be compatible with the 14 x 14 fuel design used at the Kewaunee plant. The absorber material in the rodlets is silver, indium, and cadmium. Wear was found at the guide card locations that was related to flow induced vibration with the RCCAs at fixed positions for long periods of time (steady-state power operation).

The wear was reviewed by Westinghouse. It was determined that none of the inspected RCCAs had observed wear that exceeded the wear criteria established by Westinghouse for replacement. However, it was sugoested that the RCCA normally parked position be changed by two or three steps. Hence, the licensee revised the normal fully withdrawn position from 228 steps to 226 steps. This would minimize the fretting at the location of wear, provide a clean surface for contact with the guide card, and help to extend the life of the RCCAs.

. The technical specifications for the Kewaunee plant identify the fully with-drawn position in terms of percentage withdrawn (both shutdown and control banks), but do not identify the number of steps at 100% withdrawn. During discussions with the licensee staff, together with the NRR project manager, it was identified that the change in number of steps was handled without a change in the technical specifications. The situation involved a 10 CFR 50.59 review to determine that safety analyses criteria were met when the fully withdrawn position was redefined to 226 steps from 228 steps. The licensee then issued a temporary change request (TCR) to provide administrative control for the 226 step position as the fully withdrawn RCCA position. The licensee would also expect further guidance from Westinohouse if additional wear or concerns were identified.

The LER for this event indicated the event did not meet the reporting criteria of 10 CFR 50.73(a). However, the LER did specify that the event was reportable to NPRDS, but the event was not found in the NPRDS. Subsequent discussions with the licensee staff revealed that the plant procedure did not include review of LERs for possible submittal to NPRDS even though the LER indicated the event was a candidate for such reporting. As a result of this inquiry, the plant procedure for selecting NPRDS reportable events was modified to include a review of LERs that indicate possible NPRDS reportable situations. The event was subsequently reported to NPRDS in late October, 1986.

Westinghouse Since each licensee indicated that RCCA wear criteria and changes in fully withdrawn positions were Westinghouse recommendations, the issue was discussed with a Westinghouse staff member to ascertain the vendor's perspective. It appears that initial vendor interest evolved from the broad issue of establishing expected RCCA lifetime with an initial concern of wear in the tip area. Thus, the longest operating plants, which were the early two loop PWRs, were candidates for review. The wear discovered at Point Beach was determined to be related to flow induced frettino action between the rodlet and guide card surface in the upper core intervals. (The specific wear was previously discussed in the above sections for each plant.)

The suggested action involved censideration of shutdown margin, determination of the extent of wear, and position changes that would modify the wear

, patterns. The proposed two or three step change in RCCA position had negligible effect on rod worth and shutdown margins. Conversely, this small change would reposition the guide card to a rodlet area that had not been worn (especially for shutdown rods that were fixed in the fully withdrawn position). The wear depth criterion was related to rodlet thinning and possible absorber rodlet clad collapse due to reactor pressure.

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. The absorber material for the early plants was silver, indium, and cadmium.

Some newer plants use boron carbide (8 C) as the absorber material. The rodlet cladding for the 8 C is thicker than that used for silver-idium-cadmium.

4 In addition, it is believed that no plant has operated yet for more than three fuel cycles with the B,C absorber. Thus, the overall assessment suggests that all Westinghouse PWR pTants should expect wear in all RCCAs. The specific form of wear appears to be related to both length of time of plant operation (number of refueling cycles) and whether the plant was operated for load following. We understand that information concerning RCCA wear at Point Beach was sent to other licensees with operating Westinghouse plants in a somewhat informal manner. The guidance did suggest visual examination of a sample of RCCAs based on plant operating characteristics related to the number of fuel cycles since initial plant startup. In the event of a breach of the rodlet cladding, the absorber material, such as silver, may leach into the primary coolant and the guidance suggested that monitoring coolant water chemistry for silver may indicate possible degradation of the RCCAs. Additional information provided by Westinghouse appears in reference 5.

Haddam Neck The information concerning RCCA degradation at Haddam Neck was reported in LER 213/86-015-00 (Ref. 6) on April 23, 1986. All 47 (45 in the core and two spares that had been previously used) RCCAs were eddy current and profilometry inspected during refueling at the end of cycle 13. This plant has a four loop cooling system with a 10 foot core (different from most other plants). The RCCAs use stainless steel clad rodlets with silver, indium and cadmium as the absorber material. The reported wear involved sliding, fretting from guide cards, and clad cracking (not previously reported by other licensees).

The sliding wear appeared as long axial grooves in the rodlet cladding that result from RCCA position changes during startup, power reduction, or scrams.

The wear is caused by rodlet clad contact with the guide card plates which maintain the 20 rodlets in each RCCA in a stable array. The eddy current testing showed wide spread wear, but the worst case sliding wear did not exceed established wall thickness criteria for replacement. Guide card fretting wear for fixed position PCCAs was observed as local fretting impact between the rodlet and guide card over a long period of time due to flow induced vibration. The RCCAs kept at a fixed, full out position (320 steps) were most '

susceptible to wear which is a strong function of time (plant operation). Five RCCAs exceeded the minimum wall thickness criterion with four worn into the Silver-Indium-Cadmuim absorber. Guidance from Westinghouse indicated that worst case loading from stepping, scram, and differential pressure could still be accommodated for cycle 14. The other 42 RCCAs had less wear than the worst five.

A significant amount of RCCA rodlet cracking was found just above the end plug of the RCCA rodlet. The cracking is believed to be related to absorber material swelling with subsequent absorber / clad contact that causes high stresses in the stainless steel clad which results in intergranular stress corrosion cracking. The eddy current analysis and profilometry demonstrated that all cracks were axial and that little or no change had occurred in rodlet

A diameter. Thirty two of the 47 RCCAs had signs of rodlet cracking. The worst case RCCA had cracking in 13 of 20 rodlets.

For operations during cycle 14, the normally fully withdrawn RCCA position was changed from 320 steps to 324 steps (further withdrawn rather than inserted as

. at Point Beach and Kewaunee). The two most badly worn RCCAs were replaced with spares. Also, the five high wear RCCAs were restricted from control bank operation to minimize the stresses in badly worn rodlets. As part of the long-term corrective action, the licensee will procure 45 RCCAs for the next fuel cycle.

It was evident from discussions with the licensee staff that inspection and/or monitoring to detect wear or perforation of the rodlet clad requires special effort. The inspection by eddy current and visual confirmation methods appear to be the necessary approach. Although water chemistry monitoring methods to detect severely worn cladding (through-wall) may be desirable, it was not successful at Haddam Neck. Since silver was a constituent of the absorber, the water chemistry of the primary system was monitored for silver. However, silver had been detected in the primary system prior to discovery of the worn RCCAs. The source appears to be the silver coated seal rings of the reactor vessel head. This source of silver masks the effects from RCCA through-wall wear to the absorber material.

The LER for this event indicated that the event did not meet the reporting criteria of 10 CFR 50.73(a). However, the LER did specify the event was reportable to NPRDS, but it was not found in the NPRDS and actually was not reported. Discussions with the licensee staff revealed that the determination for an event to be reported to NPRDS is made by a nuclear operations group at a corporate location (separate from the plant staff). The LER had been reviewed by the corporate staff (unlike the situation at Kcwaunee) and it was determined as not reportable to NPRDS, but there was no feedback communication to the people at the plant who prepared the LER. The licensee now plans to report the event to NPRDS within 60 days. In addition, the procedures pertaining to LER indication about NPRDS reportability will be modified. The procedure will have the LER preparer contact the group that determines NPRDS reportability for a decision so the LER will be marked in a manner consistent with the decision about NPRDS reportability.

Overview The available information suggests that RCCA wear and degradation will occur.

The degradation is related to wear during RCCA motion (at startup, shutdown, reactor trip, and load following), fretting wear due to flow induced vibration with RCCAs at fixed positions for a long period of time, and clad cracking (IGSCC) apparently related to absorber and clad contact forces. The damage mechanisms appear to be long-term degradation and no immediate safety problems are evident.

However, there are potential safety concerns related to RCCA insertion and loss of absorber material. The safety aspects relate to degradation mechanisms that affect rodlet cladding structural integrity due to wall thinning, through wall perforations, and cracking. This may adversely impact RCCA insertion or l

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. 9 lead to possible loss of absorber material with reduced shutdown margin. In addition, it appears that relatively long-term planning is needed to schedule both inspections and replacement of degraded RCCAs. Thus, potential safety concerns due to RCCA degradation are anticipated to vary on a plant specific basis depending upon length of plant operation and whether load following operation was involved.

There were only three RCCA degradation events found in the SCSS database and none in NPRDS. It was stated in the LERs that the RCCA events did not meet the reporting criteria of 10 CFR 50.73(a) and thus these reports were not '

required. It was also discovered that although two of the three LERs indicated the event was reportable to NPRDS the event was not reported. Part of the problem was that different groups within the licensee organization determine whether an event is reportable by LER or to NPRDS or both. Different groups also prepare such reports. In one instance, the procedure to identify NPRDS reportability did not include consideration of LERs. In the other, even though the LER indicated the event was reportable to NPRDS, the decision about reporting to NPRDS was made by a separate group. In this case, those who pre-pared the LER were not aware of the decision about NPRDS reportability. These instances raise several questions and concerns about reportability (where, when or whether) of component related events.

Since the RCCA wear and degradation has been interpreted by licensees to be a non-reportable event, we do not have an effective method to ascertain whether other plants have investigated or found such wear. Further, if the RCCA degradation is not reported, licensees would not have access through the normal programs to effectively utilize operating experience in this area. Moreover, is not clear that informal vendor information is an adequate means to alert licensees of the problem and receive appropriate attention at on organizational level to assure appropriate and timely inspections or monitoring for RCCA degradation.

FINDINGS AND CONCLUSIONS The events reviewed which involved observed RCCA wear and degradation at Westinghouse PWRs raise concerns regarding RCCA insertion, structural integrity, and loss of absorber material. We do not have the information to determine whether similar RCCA wear should be anticipated at PWR plants supplied by other vendors. The following findings are provided:

1. RCCA wear and degradation has been reported by at least three domestic and some forefor. Westinghouse PWR plants. Wear has occurred on both types of RCCAs (i.e., shutdown and control banks).

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. 2. RCCA wear and degradation are related to plant operational time (number of refueling cycles) and whether the plant was operated for load following.

3.

The reportedshutdown, (at startup, RCCA degradation is related reactor trip, to wear and load during)RCCA following , frettingmotion wear due to flow induced vibration with RCCAs at fixed positions for long periods of time, and clad cracking (IGSCC)' apparently caused by stresses resulting from absorber and clad contact forces.

4. RCCA wear and degradation may not be adequately covered by database system reportability requirements because LERs and NPRDS requirements have been interpreted by licensees to exclude the reporting of RCCA degradation.

5. Licensee procedures and methods related to the LER/NPRDS interface may have several variations so that there may not be consistent industry application. This may lead to non-reporting of events that should be reported.

Based on this review and the findings, it is evident that RCCA degradation in Westinghouse PWRs will occur. This leads to the conclusion that there is a need for licensees of Westinghouse PWR units to establish appropriate inspection and/or monitoring programs to determine the status of the RCCAs. In addition, if the replacement RCCAs have an anticipated design life that is similar to the original equipment, it may be desirable to establish a long-term inspection ,

program for RCCAs (perhaps similar to or as part of the ASME ten-year inservice inspection program). Further, RCCA degradation should be reportable tc at leastoneofthedatabasesystems(LERorNPROS).

The issue of whether RCCA wear will result in technical specification changes appears to be plant specific. Those older plants thet identify control rod position as percentage withdrawn, without the number of steps, may not need to change their technical specifications. Conversely, tb plants that follow standard technical specifications, which specify the number of steps, will most likely need to change their technical specifications when the fully withdrawn position is redefined.

SUGGESTED ACTION The informal notification procedure used by Westinghouse to disseminate information about RCCA wear may not be adequate to ensure appropriate action by all licensees for both detection and replacement of degraded RCCAs before potential safety issues develop. Because RCCA degradation will likely occur in at least Westinghouse PWRs, it is recommended that an IE Information Notice be prepared to alert licensees about the need to plan and schedule inspections to determine the condition of RCCAs. The information notice should discuss the wear mechanisms reviewed in this study.

It may also be appropriate for NRR to consider whether there is a need to )

establish a requirement for periodic RCCA inspections (perhaps at 10-year l intervals). This may depend upon understanding the RCCA wear criteria

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currently used relative 4to the impact on structural integrity, assurance of insertion, and possible , joss of absorber material. This effort should also consider whether there _are feasible reactor water chemistry monitoring procedures that would be useful to detect serious degradation (through-wall) of RCCA cladding (the events reviewed in the report. involved silver, indium, and cadmium as the absorber, but some recent plants use 8 C as the absorber 4

material).

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. REFERENCES

1. LER 301/84-002/03L, Submitted for Point Beach Unit 2 on May 24,1984.

2. LER 305/84-003-01, " Rod Cluster Control Assembly Cladding Wear," Submitted for Kewaunee on October 31, 1984.

3. U.S. Nuclear Regulatory Commission, Memorandum from C. J. Heltemes to C. Giroux, " Localized RCCA Wear at PWR Units," dated June 12, 1986.

4. Letter from C. W. Fay to H. R. Denton, " Control Rod Wear, Point Beach Nuclear Plant, Unit 2," dated August 18, 1983.

5. Letter from E. P. Rahe, Jr. to E. J. Brown, " Request for Information on Westinghouse Rod Cluster Control Assembly Performance," dated November 25, 1986.

6. LER 213/86-015-00, " Rod Cluster Control Assembly Wear and Cracking,"

Submitted for Haddam Neck on April 23, 1986.

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