ML071360081
| ML071360081 | |
| Person / Time | |
|---|---|
| Site: | Palo Verde  |
| Issue date: | 05/03/2007 |
| From: | Mims D Arizona Public Service Co |
| To: | Document Control Desk, NRC/NRR/ADRO |
| References | |
| 102-05696-DCM/RKR | |
| Download: ML071360081 (6) | |
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subsidiary of Pinnacle West Capital Corporation Dwight C. Mims Mail Station 7605 Palo Verde Nuclear Vice President Tel. 623-393-5403 P.O. Box 52034 Generating Station Regulatory Affairs and Plant Improvement Fax 623-393-6077 Phoenix, Arizona 85072-2034 102-05696-DCM/RKR May 03, 2007 Attn: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001
Dear Sirs:
Subject:
Palo Verde Nuclear Generating Station (PVNGS)
Units 1, 2 and 3 Docket Nos. STN 50-528, 50-529, and 50-530 Response to NRC Request for Additional Information Regarding Request for Amendment to Technical Specification 3.1.6, "Shutdown Control Element Assembly (CEA) Insertion Limits" By letter no. 102-05533, dated July 20, 2006, Arizona Public Service Company (APS) submitted a request to change PVNGS Technical (TS) Section 3.1.6, "Shutdown Control Element Assembly (CEA) Insertion Limits." The proposed change would require Shutdown CEAs to be withdrawn to _>147.75 inches, instead of the current limit of
_>144.75 inches.
By letter dated March 6, 2007, the NRC requested additional information (RAI) related to APS's July 20, 2006, amendment request, due within 60 days of the date of the RAI letter. The Enclosure to this letter contains APS's response to the NRC RAI.
There are no commitments made to the NRC by this letter. If you have any questions, please contact Thomas N. Weber at (623) 393-5764.
I declare under penalty of perjury that the foregoing is true and correct.
Executed on
/o /7 Sincerely, A member of the STARS (Strategic Teaming and Resource Sharing) Alliance Callaway
- Comanche Peak
- Diablo Canyon 4 Palo Verde
- South Texas Project 0 Wolf Creek
ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Response to NRC Request for Additional Information Regarding Request for Amendment to Technical Specification 3.1.6, Shutdown Control Element Assembly (CEA) Insertion Limits Page 2 DCM/TNW/RKR/gt
Enclosure:
As stated cc:
B. S. Mallett M.T. Markley G. G. Warnick A. V. Godwin T. Morales NRC Region IV Regional Administrator NRC NRR Project Manager NRC Senior Resident Inspector for PVNGS Arizona Radiation Regulatory Agency (ARRA)
Arizona Radiation Regulatory Agency (ARRA)
Enclosure Response to NRC Request for Additional Information Regarding Request for Amendment to Technical Specification 3.1.6, "Shutdown Control Element Assembly (CEA) Insertion Limits" NRC Question I Section 4.0, TECHNICAL ANALYSIS: Safety Analysis, (Paragraph 1)
[a] Please identify the "Specific parameters" that were analyzed to determine if, with the rods positioned at the new limit, the power distribution in the core was still within the assumptions made in the safety analyses. [b] Paragraph 1 states
"[t]o assess the effect control rod position would have on parameters with the rods positioned at the new limit, several events and specific parameters were analyzed. The events were chosen because of their sensitivity to rod position."
What transients/accidents, if any, other than steam line break and control rod ejection were analyzed? [c] Identify the specific parameters, and the assumptions made in the safety analyses.
APS Response to Question 1.a The "Specific parameters" that were analyzed were the axial peak and saddle index.
These parameters are also discussed in Section 4.0, Paragraph 4 of the original submittal.
APS Response to Question 1.b The Physics Assessment Checklist (PAC) methodology dictates the assessment of key physics parameters. Verification of these parameters assures that the physics assumptions in the safety analysis are valid. Therefore, when the key physics parameters of a new core design meet the criteria set in the PAC methodology the existing limiting physics input to safety analysis remain valid.
To examine the effect the new Technical Specification (TS) Shutdown Control Rod Insertion Limit would have on the safety analysis, the PAC parameters were surveyed to find those that had the most potential to be significantly affected. These were identified to be the SCRAM worths, axial peak, saddle index and ejected rod worth.
To assess the validity of SCRAM worth and SCRAM worth versus position, the methodology requires the assessment of four SCRAM worths, axial peak, and saddle index. The SCRAM worths relate to hot full power (HFP) at beginning-of-cycle (BOC) and end-of-cycle (EOC) and hot zero power (HZP) at the same times in cycle life. Axial peak is evaluated at BOC and saddle index is evaluated at EOC. To assess the effect the new TS limit on shutdown control rods would have on SCRAM worth the HFP 1
SCRAM worth was calculated at EOC conditions with the rods initially at the new TS limit. Normally, at this time in life rods would be fully withdrawn in accordance with the Guide Tube Wear Program. This condition was chosen because according to the checklist HFP EOC Steam Line Break conditions require the largest value of SCRAM worth. Very conservative assumptions were used in the Steam Line Break analysis and the result was only a 1% change in SCRAM worth. Since the change was within the margin between the applicable historical cycle specific calculations and the limiting value used in safety analysis the effect was considered negligible. The change in the other SCRAM worths would be of the same magnitude and thus also be negligible. Due to the potential of future fuel design changes which could raise the active fuel height, hence decreasing existing SCRAM worth margin, it was decided to modify the automated physics assessment process to explicitly account for such changes on a cycle specific basis in the future.
The SCRAM worth versus position curve used in various safety analyses is verified by examining two parameters, axial peak and saddle index, and the EOC Steam Line Break SCRAM worth. If these calculated values are within the limiting values, the SCRAM worth versus position curve is valid.
The ejected rod worth was also analyzed at the new limit. Rods positioned at the new limit would affect the flux of the ejected rod. The result found a negligible increase in ejected rod worth as described in the response to question 3.
In summary, the Steam Line Break SCRAM worth, BOC axial peak, and EOC saddle index were analyzed. They were chosen because they confirm the SCRAM worth and SCRAM worth versus rod position used in the safety analysis remain valid. The ejected rod worth was also analyzed since the initial power distribution would be slightly affected resulting in increased ejected rod worth. Since the small change in these parameters most sensitive to initial CEA position verify that the values of SCRAM worth, SCRAM worth versus rod position, and ejected rod worth used in the safety analysis do not significantly change, the effect of the TS change on the less sensitive physics input to safety analysis would also be insignificant.
APS Response to Question 1.c As noted above, the specific parameters are the axial peak and saddle index. The assumptions made in the safety analyses being referred to are the limiting axial power distributions used by safety analysis. These are verified on a cycle by cycle basis.
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NRC Question 2 Section 4.0, TECHNICAL ANALYSIS, Safety Analysis:
(Paragraph 2)
Paragraph 2 states that "[t]he checklist has three limiting parameters for rod worth. Each of the three parameters applies to a different power level and cycle length." Please explain these points and the subsequent statements that are made in Paragraph 2.
APS Response to Question 2 These limiting parameters are the SCRAM worths used in the safety analysis. These parameters are evaluated for Hot Full Power and Hot Zero Power at Beginning of Cycle and End of Cycle (there are actually four SCRAM parameters checked). Verifying that the calculated values are larger than the limiting values used in safety analysis provides assurance that the assumptions made in the safety analysis remain valid. The largest required limiting value is the EOC HFP SCRAM worth associated with the Steam Line Break event that was evaluated as described in the response to question 1.
NRC Question 3 Section
4.0 TECHNICAL ANALYSIS
- (Paragraph 3)
The second sentence states that, "[t]he case having the smallest difference between the calculated value and the limiting value was selected." Also, the fourth sentence states that, "[t]he results of the simulation found a reduction in the margin to be less than one half of 1%." Please identify the parameters that are being referred to in each of these statements.
APS Response to Question 3 The parameter referred to in both sentences is the ejected rod worth. The calculated value is determined by the control rod ejection analysis. The limiting value is used in the safety analysis. The calculated ejected rod worth data from a recent operating cycle was surveyed and the case having the smallest difference between the calculated value and the limiting value was evaluated with the rods initially positioned at the new Technical Specification Shutdown Control Rod Insertion Limit. The analysis found a reduction in margin of less than one half of 1%. The corresponding change in ejected rod worth was 0.0003 %Ap. This is insignificant.
3
NRC Question 4 Tech Spec Bases, Page B 3.1.6-4 Please provide clarifying information (e.g., title, etc.) for Reference Number 4, Calculation 13-JC-SF-0202.
APS Response to Question 4 The title of calculation 13-JC-SF-0202 (revisions 0 - 4) is "Control Element Assembly (CEA) Position Uncertainty Calculation." This calculation covers the control element assembly (CEA) reed switch position transmitters (RSPT) instrument loops in the reactor control system and their inputs to the control element assembly calculator (CEAC) and core protection calculator (CPC) in the reactor protection system.
4