05000285/LER-2003-001, Re Lack of Guidance Results in Noncompliance with Technical Specification Surveillance Requirement

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Re Lack of Guidance Results in Noncompliance with Technical Specification Surveillance Requirement
ML031340479
Person / Time
Site: Fort Calhoun Omaha Public Power District icon.png
Issue date: 05/05/2003
From: Ridenoure R
Omaha Public Power District
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
LIC-03-0059 LER 03-001-00
Download: ML031340479 (6)


LER-2003-001, Re Lack of Guidance Results in Noncompliance with Technical Specification Surveillance Requirement
Event date:
Report date:
Reporting criterion: 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications
2852003001R00 - NRC Website

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wrsH Omaha Public Power Distnct 444 Southt 16thi Street Mall Omaha NE 68102-2247 May 5, 2003 LIC-03-0059 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001

Reference:

Subject:

Docket No. 50-285 Licensee Event Report 2003-001 Revision 0 for the Fort Calhoun Station Please find attached Licensee Event Report 2003-001, Revision 0, dated May 5, 2003. This report is'being submitted pursuant to 10 CFR 50.73(a)(2)(i)(B).

Attachment c:

E. W. Merschoff, NRC Regional Administrator, Region IV A. B. Wang, NRC Project Manager J. G. Kramer, NRC Senior Resident Inspector INPO Records Center Winston and Strawn Employment wvithl Equal Opportunity I

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Abstract

During the week of March 3, 2003 an evaluation to determine the adequacy of the Fort Calhoun Station (FCS) boric acid program was conducted. As part of the evaluation, one of the evaluation team members requested information on the results of the VT-2 ispections on the lower portion of the reactor vessel. A review of the inspection results indicated that the VT-2 examination had not been accomplished which is not in compliance with Section Xl of the ASME Boiler and Pressure Vessel Code as required by Technical Specification section 3.3.1.a.

The most probable cause of this event is lack of procedural guidance, caused by poor human factors in the FCS procedure that is used to inspect the rest of the reactor coolant system. There appears to have been a mind set among individuals that the room housing the reactor vessel was then, and had always been, inaccessible because of radiological dose considerations. Further, there appears to be a mind set that "inaccessible" because of radiological dose considerations is equivalent to "inaccessible" as defined in the code.

FCS will perform the examination as required by the code if the relief request submitted for processing prior to the 2003 refueling outage is delayed or denied.

NRC FORN 366 (7-2001)

(If more space is required, use additional copies of (If more space is required, use additional copies of NRC Forn 366A)

Reactor o

Head to Vessel Flange o

ICI Grayloc Flanges o

CEDM Tool Access Flanges o

CEDM Seal Housing Flanges" OP-ST-RC-3007 step 7.7.3 states, "Perform VT-2 visual inspection on all Class I piping associated with the items listed in Attachment 3". Attachment 3 lists numerous specific.valves associated with the RCS or associated systems. Nowhere in OP-ST-RC-3007 is there a specific requirement to perform a VT-2 visual inspection on the reactor vessel after a four hour "soaking" period. There are specific instructions to perform VT-2 inspections of unique locations of the RCS.

Earlier versions of the OP-ST-RC-3007 were reviewed. Revision 0 has an equivalent step to step 7.7.2 of revision 12 that states, "Inspect all RCS Class I components and piping including, but not limited to, those associated with the following:

Reactor Coolant Pumps (all)

Steam Generators (both)

Pressurizer (including heaters)

Reactor" In May, 1990, there was an effort to ensure that Inconel 600 penetrations were emphasized during the VT-2 inspections. Therefore, under each piece of major equipment listed in the step reflecting the four hour hold, subsections were added indicating specific portions of the major equipment in an attempt to improve OP-ST-RC-3007. Revision 2 was issued in an attempt to further clarify and standardize OP-ST-RC-3007. The step was changed, and the word "component" was apparently inadvertently left out of the step. The step only required the inspection of piping.

On March 7, 2003, a review of the reportability of this event was completed. This is being reported pursuant to 10 CFR 50.73(a)(2)(i)(B).

SAFETY SIGNIFICANCE

This issue has a negligible impact on the health and safety of the public as discussed below:

There is little possibility of an unacceptable flaw or flaw propagation in the region of the reactor vessel not examined by the surveillance test. A visual examination would only have detected a through-wall flaw, and would have provided no additional assurance for less than through-wall flaws of any size. This conclusion is supported by the following:

Possibility of Undiscovered Flaw The presence of an undiscovered flaw resulting in leakage from the reactor vessel is not probable. The portion of the reactor shell in question does not contain any materials other than the reactor shell material (i.e., there are no mechanical penetrations in the bottom of the FCS reactor pressure vessel). That is, there are no materials present that are susceptible to rapid crack initiation and propagation mechanisms under the reactor operating conditions, such as the Ni-Cr-Fe alloy 600 family of materials. Consequently, there is considered a low likelihood of an active cracking mechanism that would produce flaw initiation and growth.

(If more space is required, use additional copies of (If more space is required, use additional copies of NRC Form 366A) 1992 Reactor Vessel Inservice Inspection Per ASME section Xl requirements, the reactor vessel weld inspection was performed during the 1992 refueling outage. This inspection consisted of 176 automated ultrasonic exams and an interior visual exam of the reactor vessel welds and heat affected zones. The inspection found no rejectable indications in any of the reactor pressure vessel welds or heat affected zones. Twenty minor indications were found that were initially categorized as code recordable.

twelve (12) of these indications-were small-laminar-indications.that did not obscure the-backwall UT signal and could be immediately dispositioned as code acceptable.

two (2) were determined to have code acceptable signal level when re-exam was performed.

three (3) were found to have been caused by transducer lift-off due to a local surface irregularity.

three (3) were sized and found to be well within the code allowable acceptance criteria for weld indications.

Based on these results the vessel was found to be sound and satisfactory to be returned to service.

RCS Leak Rate Trending The FCS staff has assembled 9 cycles of unknown leakage data that demonstrate RCS operating threshold for boundary integrity. This trending of unknown leakage based on a threshold definition and category zones has been verified by previous cycles and validated by known leakage issues. These two categories are grouped into either connection (range from 0.075 to 0.2 gpm) or boundary integrity leakage (greater than 0.2 gpm) in facilitating an assessment of system performance. In monitoring of cycles 19, 20 and 21 (current) it is evident the boundary leakage threshold has not been challenged in the previous and/or current cycle. During the current cycle, the RCS leak rate has been below the threshold baseline.

Therefore, this issue has a negligible impact on the health and safety of the public.

CONCLUSION The most probable cause of this event is lack of procedural guidance, caused by poor human factors in procedure OP-ST-RC-3007. There appears to have been a mind set among many individuals that the room housing the reactor vessel was inaccessible. When questioned why they thought the room was inaccessible, most individuals stated that it-was because of-radiological dose reasons.-Further,-there appears.to be a-mind -set that--

"inaccessible" because of radiological dose considerations is, in some way, equivalent to "inaccessible" as defined in the code. Some individuals assumed there was an engineering argument that documented the inaccessibility of the room as related to the code. There was an additional argument/mind set from some individuals that, since the word "component" was not included in the test, the reactor vessel inspection was not required. Therefore, it appears that a contributing cause for this event was a number of inappropriate mind sets.

CORRECTIVE ACTIONS

FCS will perform the examination as required by the code if the relief request submitted for processing prior to the 2003 refueling outage is delayed or denied. Other corrective actions related to this issue are being completed in accordance with the FCS corrective action program.

SAFETY SYSTEM FUNCTIONAL FAILURE This event did not result in a safety system functional failure in accordance with NEI 99-02.

PREVIOUS SIMILAR EVENTS

FCS has not had any previous similar events.

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