1CAN031202, Unit 1, 10 CFR 50.46 Report – Significant Change in Peak Cladding Temperature
| ML12080A120 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 03/20/2012 |
| From: | Pyle S Entergy Operations |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| 1CAN031202, BAW-10192P-A, Rev 0 | |
| Download: ML12080A120 (5) | |
Text
1CAN031202 March 20, 2012 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555
SUBJECT:
10 CFR 50.46 Report - Significant Change in Peak Cladding Temperature Arkansas Nuclear One - Unit 1 Docket No. 50-313 License No. DPR-51
REFERENCE:
AREVA NP Topical Report BAW-10192P-A, Revision 0, BWNT LOCA - BWNT Loss-of-Coolant Accident Evaluation Model for Once-Through Steam Generator Plants, June 1998
Dear Sir or Madam:
On February 24, 2012, AREVA NP Inc. (AREVA) notified Entergy Operations, Inc. (Entergy) of two errors in the Arkansas Nuclear One, Unit 1 (ANO-1) Emergency Core Cooling System (ECCS) evaluation model (EM). The two errors offset each other with no net change to the current Peak Clad Temperature (PCT) calculated for the Large-Break Loss-of-Coolant Accident (LBLOCA). However, each individual error as well as the sum of the absolute values of the two errors is greater than the requirement of 10 CFR 50.46(a)(3)(ii) for a 30-day report. The errors are discussed below.
The errors were discovered as a result of work being performed for the 205 fuel assembly (FA) Bellefonte plant. In evaluating the extent of condition for the two errors, similar errors were found to exist in the 177 FA plant (e.g., ANO-1) LBLOCA analyses as well. One error occurred in the ECCS bypass calculational model that, when corrected, decreased the PCT because the lower plenum refill period was shortened. The other error found that for the 205 FA Bellefonte plant, modeling a column weldment over the hot channel increased the PCT. This EM deficiency was assessed with scoping analyses to determine how much cladding temperatures could increase. The scoping analyses used the corrected ECCS bypass model to assess the resultant PCT. The scoping analyses indicated little change from the original PCT, that is, the net decrease in PCT from the ECCS bypass error correction is roughly equivalent to the PCT increase when the column weldments are modeled in the 177 FA plants.
While performing a LBLOCA sensitivity study for the Bellefonte plant, a mathematical error was discovered in the RELAP5/MOD2-B&W blowdown model control variables that calculate Entergy Operations, Inc.
1448 S.R. 333 Russellville, AR 72802 Tel 479-858-4704 Stephenie L. Pyle Manager, Licensing Arkansas Nuclear One
1CAN031202 Page 2 of 3 the time for total end of bypass. AREVA identified in the reference report that the end of bypass calculations determine when an 80% condensation efficiency on the Core Flood Tank (CFT) injected liquid could condense all the steam reaching the upper downcomer region.
The control variables incorrectly calculated the steam energy flowing into the upper downcomer region. When the control variables were corrected, the end of bypass time was predicted to occur approximately two seconds earlier, resulting in a shorter lower plenum refill period with a quicker onset of lower core quench and lower PCT for the Bellefonte plant limiting LBLOCA analyses. In evaluating the extent of condition for this error, a similar error was found to exist in all 177 FA plant LBLOCA analyses as well.
AREVA corrected the control variable error and performed a new limiting LBLOCA analysis for the Oconee 177 FA plant. The correction also shortened the lower plenum refill period by roughly two seconds for the 177 FA plant and decreased the ruptured segment cladding temperature by approximately 80F from the previously calculated value with the ECCS bypass error. The limiting unruptured segment cladding decreased by approximately 40F.
Since the same error was in all 177 FA models and the CFT flows and plant geometry are similar in all models, the refill period will shorten by approximately the same interval with the expectation that cladding temperatures decrease similarly for all 177 FA plants. Therefore, a generic LBLOCA PCT change of -80F (reduction) is assigned to the ruptured cladding segments and a -40F is assigned to the limiting unruptured cladding segments to account for changes associated with the ECCS bypass error. ECCS bypass is not used for Small-Break LOCA (SBLOCA) so these analyses are not affected by this error.
During the assessment of the ECCS bypass error, another LBLOCA sensitivity study was being performed for the 205 FA Bellefonte plant with a revised upper plenum and upper head modeling that considered the changes in core cooling when upper plenum column weldments are explicitly modeled. This revised modeling reflects a more detailed noding arrangement in the reactor vessel upper plenum than was used and approved for application in the reference report. A simplified column weldment model was developed for the 177 FA plants based on approximations from the 205 FA model. When this simplified model was used, 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. As a result, the end of blowdown fuel temperatures increased, translating into an approximate 40F increase in the unruptured segment PCTs and 80F in the ruptured segment PCTs. This modeling change was also considered for SBLOCA and it was concluded that it does not affect the limiting results because the SBLOCA is a slower evolving transient with up-flows in the core hot bundles such that there is no net change from the presence of a column weldment in the upper plenum.
The attachment tabulates the effect of the ECCS bypass error correction and column weldment modeling on the limiting LOCA PCTs for the ANO-1 LOCA licensing basis. The attached table with the estimated PCT changes and the final PCT is based on a full core of Mark-B-HTP design fuel which ANO-1 currently uses. AREVA has also reported that correction of the two errors results in similar estimated PCT impacts for other fuel designs.
Thus, for all fuel designs used by ANO-1 currently or in the past, there is no net LBLOCA estimated changes to previously reported PCTs. Since the PCTs are unchanged, the oxidation and whole core hydrogen generation are similar as well and the previously reported values remain applicable.
1CAN031202 Page 3 of 3 As stated above, there is no net LBLOCA estimated change to the previously reported PCT based on several preliminary Mark-B-HTP scoping analyses by AREVA. An evaluation was also performed for the SBLOCA cases and it was concluded that there is no change on the limiting SBLOCA analysis results. Since there is no net change in the LOCA PCTs for these two errors, there are no plans for future reanalysis. AREVA does intend to correct the ECCS bypass model and develop the upper plenum modeling changes which will then be included in any future LOCA analyses.
This submittal fulfills the 30-day reporting requirements of 10 CFR 50.46(a)(3)(ii).
This letter contains no new regulatory commitments.
Should you have any questions, please contact me.
Sincerely, Original Signed by Stephenie L. Pyle SLP/rwc
Attachment:
Loss-Of-Coolant-Accident (LOCA) Licensing Activity for ANO-1 cc:
Mr. Elmo E. Collins Regional Administrator U. S. Nuclear Regulatory Commission, Region IV 1600 East Lamar Boulevard Arlington, TX 76011-4511 NRC Senior Resident Inspector Arkansas Nuclear One P. O. Box 310 London, AR 72847 U. S. Nuclear Regulatory Commission Attn: Mr. Kaly Kalyanam MS O-8 B1 One White Flint North 11555 Rockville Pike Rockville, MD 20852
ATTACHMENT TO 1CAN031202 Loss-Of-Coolant-Accident (LOCA) Licensing Activity for ANO-1
Attachment to 1CAN031202 Page 1 of 1 Loss-Of-Coolant-Accident (LOCA) Licensing Activity for ANO-1 Plant Name:
Arkansas Nuclear One - Unit 1 Mark-B-HTP LOCA Spectrum Utility Name:
Entergy Large-Break LOCA Full Core Small-Break LOCA Reporting Category Description PCT or (PCT Change)
Previous Licensing Basis 2008.1F Estimated 1459F Analyzed Application Error Error in Emergency Core Cooling System Bypass Calculation Estimated (-80F) for Ruptured Segment Estimated (-40F) for Peak Unruptured Segments N/A Evaluation Model Modeling Change Upper Plenum Column Weldment Modeling Estimated (+80F) for Ruptured Segment Estimated (+40F) for Peak Unruptured Segments Estimated (0F)
Current Licensing Basis 2008.1F Estimated 1459F Estimated