ML20248G856
| ML20248G856 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 06/02/1998 |
| From: | Cruse C BALTIMORE GAS & ELECTRIC CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| 50-317-98-01, 50-317-98-1, NUDOCS 9806050366 | |
| Download: ML20248G856 (9) | |
Text
l CHARLES II. CRUSE Baltimore Gas and Electric Company l
Vice President Calven Cliffs Nuclear Power Plant Nuclear Energy 1650 Calven Cliffs Parkway Lusby. Maryland 20657 410 495-4455 June 2,1998 U. S. Nuclear Regulatory Commission Washington, DC 20555 ATTENTION:
Document Control Desk
SUBJECT:
Calvert Cliffs Nuclear Power Plant; Unit Nos.1 & 2; Docket Nos. 50-317 & 50-318 independent Spent Fuel Storage Installation Docket No. 72-8 Reply to Notice of Violation -- NRC Inspection Report Nos. 50-317/98-01 and 50-318/98-01
REFERENCES:
(a) Letter from Mr. L. T. Doerflein (NRC) to Mr. C. H. Cruse (BGE), dated April 15,1998, NRC Region I Integrated Inspection Report Nos. 50-317/98-01 and 50-318/98-01 and Notice of Violation 0 Letter from Mr. C. H. Cruse (BGE) to NRC Document Control Desk, dated May 1,1998, Clarification of Notice of Violation and Request for Extension of Response Time Regarding Inspection Report Nos. 50-317/98-01 and 50-318/98-01 (c) April 27,1998 Teleconference Between BGE and NRC Staff This letter provides Baltimore Ga.:.md Electric Company's (BGE's) responses to the Notice of Violation issued by Reference (a) containing two violations of NRC requirements. The first cited violation (Violation A) concerns NRC regulation 10 CFR 72.11 due to failure to provide the NRC complete and accurate information in the 1989 site-specific license application for an independent spent fuel storage installation. The second cited violation (Violation B) concerns NRC regulation 10 CFR 72.48 for making a change to the evaluation for the dry shielded canister top end drop accident, which the NRC deems involves an unreviewed safety question, without prior NRC approval.
The details of Violations A and B, followed by BGE's responses, are provided in Attachments (1) and (2), respectively. Please note that, by Reference (b), BGE had requested an extension et the response time to June 3,1998. This extension request was discussed with the NRC Resident Inspectors and in Reference (c).
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9806050366 980602 PDR ADOCK 05000317 o
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Document Control Desk Jun'e 2,1998 Page 2 j.
Should you have questions regarding this matter, we will be pleased to discuss them with you.
I Very truly yours, for C. H. Cruse Vice President - Nuclear Energy
- CHC/GT/ dim Attachments (1)
NRC Inspection Report Nos. 50-317/98-01 & 50-318/98-01, Violation A l
(2)
NRC Inspection Report Nos. 50-317/98-01 & 50-318/98-01, Violation B cc:
R. S. Fleishman, Esquire H. J. Miller, NRC J. E. Silberg, Esquire Resident Inspector, NRC S. S. Bajwa, NRC R. I. McLean, DNR A. W. Dromerick, NRC J. H. Walter, PSC l
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ATTACHMENT (1) l I
NRC INSPECTION REPORT NOS. 50-317/98-01 & 50-318/98-01, VIOLATION A I
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l Baltimore Gas and Electric Company l
Calvert Cliffs Nuclear Power Plant June 2,1998 a
ATTACHMENT (1)
NRC INSPECTION REPORT NOS. 50-317/98-01 & 318/98-01 VIOLATION A ALLEGED VIOLATION 10 CFR 72.11 requires that information provided to the Commission by an applicantfor a license shall be complete andaccurate in allmaterialrespects.
Contrary to the above, on December 21,1989, Baltimore Gas and Electric Company (BGE) failed to provide complete and accurate information regarding the behavior ofthe dry shielded canister (DSC) in the site specific license applicationfor use of the NUHOMS system at the Calvert Chfs independent Spent Fuel Storage Installation (ISFSI). Specifically, the original calculations performed to support the Updated Safety Analysis Report (USAR) noted that during a vertical top end drop accident the clip angles, which attach the guide sleeves to the bottom supportplate, wouldfailin shear at an acceleration of 35 G. The USAR indicates that the maximum spacer disk deflection was computed as 0.077 inches and that the guide sleeve to fuel assembly gap was computed to exceed 0.5 inches. However, BGE determined through analysis and testing that the clips actuallyfailed by bending (not shearing) at approximately 43 G and that the guide sleeve wouldpinch thefuel assemblies.
1.
ADMISSION OR DENIAL OF THE AII.FGED VIOLATION Baltimore Gas and Electric Company accepts the violation as stated.
II.
REASON FOR THE VIOLATION in 1989, BGE submitted numerous documents to the NRC for a site-specific ISFSI license for the Calvert Cliffs Nuclear Power Plant (CCNPP). Part of the submittal contained an analysis for a vertical Transfer Cask (TC) drop. The design assumed the guide sleeve clips inside a DSC would fait during a vertical TC drop. These clips hold the guide sleeves in place during normal operation. The drop analysis assumed the clips would fail by shearing at a force of 35 Gs or less to prevent the canister from being damaged.
Also, the drop analysis assumed a gap would exist between the fuel assembly and the guide sleeve after the accident. These design requirements were turned into a mechanical design and fabrication specification by the ISFSI vendor without any physical testing to ensure that the clips would perform as stated in the drop analysis.
Late last year, BGE was informed by our ISFSI vendor, Transnuclear West (TNW), that the guide sleeve clips did not fail as described in the BGE Safety Analysis Report. Baltimore Gas and Electric Company performed a 10 CFR 72.48 safety evaluation that allowed the current gu*de sleeve clip design to meet the TC drop analysis for the nine empty DSCs. The 72.48 safety evaluatie i showed that the clips would actually fail at 43 Gs or less by bsMing causing the gap between the fuel assembly and guide sleeve to be eliminated. The supporting analysis and physical testing by BGE proved that the fuel assemblies could be retrieved from the DSC after a vertical drop using the normal fuel handling equipment and met our ISFSI licensing requirements. The 72.48 safety evaluation was approved by the CCNPP's Plant General Manager cn February 17,1998.
In late February and early March 1998, the NRC reviewed the 72 48 report. The NRC inspectors determined that BGE's original license submittal did not contain complete and accurate information i
I regarding the behavior of the DSC under all accident scenarios. Specifically, the elimination of the fuel assembly to guide sleeve gap and the higher G force to bend, not shear, the clips was not described in BGE's original ISFSI license submittal.
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ATTACIIMENT (1)
NRC INSPECTION REPORT NOS. 50-317/98-01 & 318/98-01 VIOLATION A As discussed in detail in Attachment (2), BGE believes that the difference in NRC and BGE's perspectives of whether the revised behavior of the components constituted a new malfunction is a major contributed to this violation, also. If the elimination of the gap had not been declared a new malfunction requiring NRC review, it would not have been considered consequential to the accuracy and the completeness of the original license application.
III. CORRECTIVE STEPS THAT HAVE BEEN TAKEN AND RESULTS ACHIEVED As a result of BGE's experience with the DSC guide sleeve clips issue, TNW has completed a review of its design process. As part of the new process, procedures have been revised to ensure all components that are required to fail will have a mockup test performed. All components that are required to fail will be tested with their supporting components installed before full scale manufacturing begins.
The implementation of the revised design process was witnessed by BGE when TNW recently tested a new DSC guide sleeve clip design.
IV. CORRECTIVE STEPS THAT WHL BE TAKEN TO AVOID FURTHER VIOLATIONS Baltimore Gas and Electric Company has completed the owner acceptance review of all design calculational / technical ISFSI documents. A third-party independent review is currently underway.
Issues identif'ied by these reviews will be dispositioned appropriately.
V.
DAIE WHEN FULL COMPLIANCE WILL BE ACHIEVED Full compliance will be achieved by December 31,1998.
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ATTACHMENT (2)
NRC INSPECTION REPORT NOS. 50-317/98-01 & 50-318/98-01, VIOLATION B I
Baltimore Gas and Electric Company Calvert Cliffs Nuclear Power Plant June 2,1998
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ATTACHMENT (2)
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' NRC INSPECTION REPORT NOS. 50-317/98-d1 & 318/98-01 I
VIOLATION B ALLEGED VIOLATION I
10 CFR 72.48(a)(1) allows, in part, license holders to make changes to the Independent Spent Fuel Storage Installation (ISFSI) orprocedures as described in the Updated Safety Analysis Report (USAR) 1 unless the proposed change involves a change in the license conditions or an unreviewed safety question (USQ). 10 CFR 72.48(a)(2) requires, in part, that a proposed change be deemed to involve a USQ if a possibilityfor a malfunction of a diferent type than any evaluatedpreviously in the USAR may be
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. created.
10 CFR 72.48(c) requires submittal of an applicationfor a license amendmentfor changes to the ISFSI orprocedures described in the USAR which involve a USQ.
Contrary to the above, as of February 17,1998, BGE made a change to the evaluation for the dry shielded canister (DSC) vertical top end drop accident, without prior NRC approval, which involved a USQ. Specylcally, BGE safety evaluation, ES199601368, Supplement 2, Revision 0, did not identify that a new malfunction was created when the gap between the guide sleeve an the fuel assembly was d
eliminated by the change to the evaluationfor the DSC vertical top end drop accident such that afailure of the guide sleeve clip angles by bending vice shearing would cause the canister guides sleeves to deform and impinge on the installed spentfuel assemblies. In discussing drop accidents, Section 8.2.5.2
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' of the USAR states that the maximum spacer disk deflection was computed to be 0.077 inches and that l
. the gap between the guide sleeve and thefuel assemb!y exceeds [0.5] inches I.
ADMISSION OR DENIAL OF THE ALLEGED VIOLATION Baltimore Gas and Electric Company accepts the violation as stated.
IL REASON FOR THE VIOLATION This violation resulted from a difference in NRC and BGE's perspectives of whether the revised i
I behavior of the components within the DSC after a vertical drop event constituted a new malfunction.
While some calculational reservations are identified in the Inspection Report, discussion with NRC staff 1
has indicated that these differences are not the basis for the violation (Reference 1). There is substantial l
agreement regarding the physical phenomena that are expated after an event.
The Inspection Report accurately notes that the USAR description of the DSC guide sleeve makes it clear that no contact between the guide sleeve and the fuel assembly (specifically, the grid spacers) was expected. Additionally, detailed analysis of DSC spacer disk deflection was not provided based on the existence of a.0.5" gap between the guide sleeve and the fuel grid spacer. Nuclear Regulatory l
Commission's position is that the possibility of impingement (or pinching)'of the fuel assembly grid t
spacer because of the dimpling of the guide sleeve should be viewed as a new malfunction.
Baltimore Gas and Electric Company's perspective, reflected in its 10 CLR 72.48 safety evaluation, was l
that designing the DSC in a way that maintained a gap between the guide sleeve and the fuel assembly was a simple means of assuring that deformation created by a cask drop event would not require excessive withdrawal forces or jeopardize keeping the structure of the fuel assemblies, and particularly the fuel rods, intact. Discovery that the gap would be eliminated under certain scenarios required additional evaluation to determine whether the revised behavior created a safety concern. When this analysis showed that no adverse outcomes resulted from the revised behavior, BGE recorded its 1
ATTACHMENT (2)
NRC INSPECTION REPORT NOS. 50-317/98-01 & 318/98-01 VIOLATION B 1
conclusion that the interaction between the guide sleeve and grid spacer, although different, did not constitute a new malfunction.
We have evaluated the difference in threshold between that stated in the Inspection Report and BGE's 72.48 evaluation. By way of clarification, NRC appears to indicate that after appropriate analysis, BGE could have reduced the gap to a value less than the 0.5" described in the USAR without creating a USQ, but that the unintended contact between the components constituted the new malfunction. Baltimore Gas and Electric Company's threshold for declaring the change a USQ would have been reached if we had I
concluded that the change had substantially increased the extraction force required for post-event recovery of the fuel to the point that different equipment would have been required, or if the impingement had resulted in contact with the fuel rods instead of the fuel assembly grid spacer. The difference, then, reverted to, "What level of physical interaction or phenomenon described in the SAR can be changed by the licensee without prior NRC review." As we understand the NRC's position in this l
violation, that answer appears to be, "None. A change in expected performance-unless clearly j
beneficial--it a new malfunction. Subsequent analysis may determine whether it is acceptable, but it l
will not substitute for NRC review." Baltimore Gas and Electric Company has hesitated to accept this interpretation because it seems likely to capture many beneficial changes as USQs and thereby invoke delays and expense for NRC review where the change in behavior is inconsequential.
Baltimore Gas and Electric Company has been a close observer during the ongoing discussions between NRC and the industry regarding 10 CFR 50.59 (and by implication,10 CFR 72.48) policy issues and potential rulemaking. This instance appears to fall directly into the area of issues where the current threshold for declaring a USQ is under review. Our engineering processes and training are structured to minimize disputes arising out of the current interpretation discussions by directing our engineers to treat potential increases in consequences or probability of accidents or malfunctions conservatively. This policy has been manifested in submittal of USQs on technical concerns, such as our resolution of Generic Letter 96-06 Service Water System modifications and in the revised design of our Saltwater System.
These have served as conspicuous examples to reiterate our intention that engineers should escalate borderline issues for supervisory attention and that we will exercise the appropriate regulatory processes to attain the right safety results.
SAFETY SIGNIFICANCE We do not consider the technical or programmatic aspects of this occurrence to have significantly affected safety.
i While the 72.48 evaluation resulting in this violation was reviewed by supervision without its being declared a USQ, the technical aspects of the issue were closely monitored and extensively reviewed. As noted in the Inspection Report, BGE had a comprehensive program to review ISFSI issues, and we had employed outside technical organizations, both for analysis and for quality reviews, to ensure that the safety concerns were appropriately dispositioned. Likewise, although the 72.48 conclusions reached were different than those reached by NRC staff, the underlying technical issues were self-identified and pursued to resolution under BGE's corrective action program. Conservative decisions were made with l
l respect to conducting spent fuel movements, and we initiated frequent communications with NRC staff to assure a clear understanding of our status.
Subsequent treatment of this interaction between the DSC and fuel assembly as a degraded condition under the guidance of Generic Letter 91-18, Revision 1, was a beneficial instance in demonstrating to our 2
f ATTACIIMENT (2)
NRC INSPECTION REPORT NOS. 50-317/98-01 & 318/98-01 VIOLATION B staff how the regulatory process could be followed while still allowing the appropriate end result of enhancing safety through spent fuel offload.
IIL CORRECTIVE STEPS THAT HAVE BEEN TAKEN AND RESULTS ACHIEVED We have published training guidance using this instance as an example of a new malfunction to our 50.59/72.48 screeners and evaluators and to our Plant Operating Safety Review Committee.
IV.
CORRECTIVE STEPS THAT WILL BE TAKEN TO AVOID FURTHER VIOLATIONS We will include training guidance incorporating the lessons of this instance as an example of a new malfunction in our lesson plan for initial and periodic requalification for personnel preparing 50.59/72.48 safety evaluations and screens.
We will review all 72.48 safety evaluations and a sampling of 50.59 safety evaluations conducted this year to identify any other similar cases that should be reconsidered. Upon completion, we will submit this and any other related USQs, which will not be returned to full compliance, as license amendment requests.
We will review the impact of this interpretation of the regulation after approximately one year. If there are meaningful lessons learned or implications warranting NRC attention, we will communicate them to your staff.
V.
DATE WHEN FULL COMPLIANCE WII L BE ACHIEVED Full compliance will be achieved by December 31,1998.
REFERENCE (1) Letter from Mr. C. H. Cruse (BGE) to NRC Document Control Desk, dated May 1,1998, Clarification of Notice of Violation and Request for Extension of Response Time Regarding Inspection Report Nos. 50-317/98-01 and 50-318/98-01 f
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