L-12-071, 10 CFR 50.46 Report of Significant Changes or Errors in the Loss of Coolant Accident Evaluation Model
| ML12076A237 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 03/16/2012 |
| From: | Allen B FirstEnergy Nuclear Operating Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| L-12-071 | |
| Download: ML12076A237 (3) | |
Text
FENOC FirstEnergy Nuclear Operating Company 5501 North State Route 2 Oak Harbor, Ohio 43449 Barry S. Allen Vice President - Nuclear 419-321-7676 Fax: 419-321-7582 March 16,2012 L-12-071 10 CFR 50.46 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001
SUBJECT:
Davis-Besse Nuclear Power Station Docket No. 50-346, License No. NPF-3 10 CFR 50.46 Report of Significant Changes or Errors in the Loss of Coolant Accident Evaluation Model By correspondence dated February 20, 2012, FirstEnergy Nuclear Operating Company (FENOC) was notified of two issues in the loss of coolant accident evaluation model used for Davis-Besse Nuclear Power Station (DBNPS). The first issue is characterized as an application error; the second issue is characterized as a modeling change. Both issues satisfy the criteria of a significant change or error as defined by 10 CFR 50.46(a)(3)(i).
Details of the two issues and their consequences to DBNPS are provided in the attachment. In accordance with 10 CFR 50.46(a)(3)(ii), FENOC is required to report significant changes or errors to the Nuclear Regulatory Commission (NRC) within 30 days.
There are no regulatory commitments contained in this submittal. If there are any questions or additional information is required, please contact Mr. Phil H. Lashley, Supervisor - Fleet Licensing, at (330) 315-6808.
Sincerely,
~~
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rg,/hL..,J Barry S. Allen
Attachment:
10 CFR 50.46 Report of Significant Changes or Errors to the Loss of Coolant Accident Evaluation Model cc:
NRC Region III Administrator NRC Resident Inspector NRC Project Manager Utility Radiological Safety Board
Attachment L-12-071 10 CFR 50.46 Report of Significant Changes or Errors to the Loss of Coolant Accident Evaluation Model Page 1 of 2
Background:
By correspondence dated February 20,2012, AREVA NP INC. (AREVA) notified FirstEnergy Nuclear Operating Company (FENOC) of two issues in the emergency core cooling system (ECCS) large break loss-of-coolant-accident (LBLOCA) evaluation model (EM) used for Davis-Besse Nuclear Power Station (DBNPS). This model is described in BAW-10192P-A, "BWNT Loss-of-Coolant Accident Evaluation Model for Once-Through Steam Generator Plants," Revision 0, June 1998. As reported by AREVA, recent work related to restarting a 205 fuel assembly (FA) plant resulted in the identification of two issues in the LBLOCA EM. Emergency core cooling system analyses for DBNPS are performed by this same nuclear fuel vendor, AREVA. In evaluating the extent of condition for the issues, AREVA identified similar issues for the 177 FA plants, including DBNPS.
The first issue involves the use of incorrect control variables; the second issue involves a weldment modeling change. Both issues satisfy the criteria of a significant change or error as defined by 10 CFR 50.46(a)(3)(i).
In accordance with 10 CFR 50.46(a)(3)(ii), FENOC is required to report significant changes or errors to the Nuclear Regulatory Commission (NRC) within 30 days.
Error Involving Incorrect Control Variables:
While performing a LBLOCA sensitivity study for the 205 FA plant, an application error was identified in the RELAP5/MOD2-B&W blowdown model control variables that calculate the time for total end of bypass. In the model, the end of bypass calculations determine when an 80 percent condensation efficiency for the core flooding tank (CFT) injected liquid could condense all of the steam reaching the upper downcomer region in the reactor pressure vessel (RPV). The control variables used to calculate this steam energy flowing into the upper downcomer region were incorrect. When the control variables were corrected, the predicted end of bypass time was approXimately two seconds earlier, resulting in a shorter lower plenum refill period and a more rapid onset of lower core quench. This also resulted in a lower peak cladding temperature (PCT) for the 205 FA plant's limiting LBLOCA analysis. In evaluating the extent-of-condition for this error, a similar error was identified for the 177 FA plants, including DBNPS.
AREVA corrected the control variable error and performed a new limiting LBLOCA analysis for a 177 FA lowered-loop plant (DBNPS is a raised-loop plant). The correction shortened the lower plenum refill period by approximately two seconds. This change decreased the ruptured segment fuel cladding temperature by approximately 80 degrees Fahrenheit compared to the previously calculated value that included the control variable ECCS bypass error. The limiting unruptured segment cladding temperature also decreased by approximately 40 degrees Fahrenheit. Since the error is common to all 177 FA plants (both lowered-loop and raised-loop) and their models, and the CFT flows and plant geometry are similar for all 177 FA plants and their models, the refill period will shorten by approximately the same interval. Consequently, the cladding temperatures are expected to decrease similarly for all 177 FA plants. Therefore, a generic estimated LBLOCA PCT reduction of
Attachment L-12-071 Page 2 of 2 80 degrees Fahrenheit has been assigned to the ruptured cladding segments and a generic estimated 40 degrees Fahrenheit reduction has been assigned to the limiting unruptured cladding segments to account for changes associated with the error.
Emergency core cooling system bypass is not modeled by small-break LOCA (SBLOCA) analyses. As a result, the SBLOCA analyses are not affected by the control variable error.
Change in Weldment Modeling:
This issue was identified during the evaluation of the control variable error. As reported, an additional LBLOCA sensitivity study was performed for the 205 FA plant with revised upper RPV plenum and upper head modeling that considered the changes in core cooling when the upper plenum column weldments are explicitly modeled. This revised modeling provides a more detailed noding arrangement in the RPV upper plenum than was used and approved for application by the LBLOCA EM in BAW-10192P-A.
AREVA developed a simplified column weldment model for the 177 FA plants based on the 205 FA model. With the simplified model, the scoping case with the column weldment modeled over the top of the hot channel resulted in reduced cooling during portions of the blowdown phase and an increase in the fuel temperatures at the end of the blowdown.
These changes result in an estimated increase of 40 degrees Fahrenheit in the limiting unruptured segment PCT, while the ruptured segment was increased by an estimated 80 degrees Fahrenheit. The column weldment modeling change was then considered for SBLOCAs. It will not affect the limiting results because the SBLOCA is a slower evolving transient with up-flows in the core hot bundles. As a result, there is no change to the SBLOCA PCT due to the addition of a column weldment in the upper plenum model.
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Conclusion:==
By correspondence dated May 27,2011, FENOC reported to the NRC a LBLOCA EM estimated PCT of 2,119 degrees Fahrenheit. The change in PCT due to the control variable error [-80 degrees Fahrenheit] is approximately equivalent in magnitude to the PCT change due to the change in weldment modeling [+80 degrees Fahrenheit]. As a result, they offset each other with no net change to the current estimated PCT of 2,119 degrees Fahrenheit.
Since the PCT is unchanged, maximum cladding oxidation and whole-core maximum hydrogen generation remain unchanged. Also, the ability of the ECCS to maintain a coolable core geometry and provide long-term cooling is not impacted. Therefore, for DBNPS, compliance with 10 CFR 50.46 acceptance criteria is maintained.
As stated within 10 CFR 50.46(a)(3)(ii), a schedule for reanalysis or taking other actions needed to show compliance with 10 CFR 50.46 requirements may be required. Since compliance with 10 CFR 50.46 criteria have been satisfied, a schedule is not required.