RBG-46182, Supplement to Amendment Request, Changes to the Analytical Methods Referenced in Technical Specification 5.6.5, Core Operating Limits Report (COLR)
| ML072190208 | |
| Person / Time | |
|---|---|
| Site: | River Bend  |
| Issue date: | 07/30/2007 |
| From: | Roberts J Entergy Operations |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| RBF1-07-0138, RBG-46182, TAC MD3293 | |
| Download: ML072190208 (5) | |
Text
'Entergy Entergy Operations, Inc.
River Bend Station 5485 U. S. Highway 61 N St. Francisville, LA 70775 Tel 225 381 4149 Fax 225 635 5068 jrober3@entergy.com Jerry C. Roberts Director, Nuclear Safety Assurance July 30, 2007 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555
SUBJECT:
REFERENCE:
Supplement to Amendment Request Changes to the Analytical Methods Referenced in Technical Specification 5.6.5, "Core Operating Limits Report (COLR)," TAC No. MD3293 River Bend Station, Unit 1 Docket No. 50-458 License No. NPF-47 Letter RBG-46583, "License Amendment Request: Changes to the Analytical Methods Referenced in Technical Specification 5.6.5, "Core Operating Limits Report (COLR)" dated October 16, 2006 (ADAMS Accession No. ML062960299)
RBG-46182 RBF1-07-0138
Dear Sir or Madam:
By the above referenced letter, Entergy Operations, Inc. (Entergy) proposed a change to the River Bend Station, Unit 1 (RBS) Technical Specifications (TS) to add a NRC previously approved topical report to the analytical methods referenced in Technical Specification (TS) section 5.6.5, "Core Operating Limits Report (COLR)."
On May 4, 2007, Entergy and members of your staff held a call to discuss questions concerning the proposed change. As a result of the call, the NRC staff requested Entergy to provide additional information on five items in a formal response. An additional call was held on June 14, 2007 in which the NRC provided the following clarifications regarding Entergy's planned response:
- 1. Your response to item (3) mentions that the precise PCT impact for each of the changes has not been quantified because a through sensitivity analysis is not practical. The licensee should know all the impact to this higher PCT results due to each influential parameter with some other's consulting services.
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- 2. Your response to item (4) mentions that since the initial MCPR value of 1.16 remains non-limiting and is conservatively bounding for the current and expected future cycles, the same initial value is used in the EXEM BWR-2000 LOCA analysis. How can a non-limiting MCPR value become conservatively bounding for the current and expected future cycles?
Additional calls were held on July 2 and July 18, 2007 to clarify the above requests. Entergy's response to the original five items is contained in Attachment 1. The response to items 3 and 4 also address the above request for clarification.
There are no technical changes proposed. The original no significant hazards consideration included in Reference 1 is not affected by any information contained in the supplemental letter.
The proposed change does not include any new commitments. If you have any questions or require additional information, please contact Ron Byrd at 601-368-5792.
I declare under penalty of perjury that the foregoing is true and correct. Executed on July 30, 2007.
Sincerely, JryC. Roberts Director - Nuclear Safety Assurance JCR/DHW
Attachment:
Analysis of Proposed Technical Specification Change cc:
U. S. Nuclear Regulatory Commission Region IV 611 Ryan Plaza Drive, Suite 400 Arlington, TX 76011 NRC Senior Resident Inspector P. 0. Box 1050 St. Francisville, LA 70775 U.S. Nuclear Regulatory Commission Attn: Mr. Bhalchandra K. Vaidya MS 0-7D1 Washington, DC 20555-0001 Mr. Jeff Meyers Louisiana Dept. of Environmental Quality Office of Environmental Compliance P. O. Box 4312 Baton Rouge, LA. 70821-4312 to RBG-46182 Page 1 of 3 Response to Request for Additional Information Related to Proposed Changes to the Analytical Methods Referenced in TS 5.6.5 Question:
With respect to the use of the proposed EXEM BWR-2000 model, please provide:
(1) the detailed explanation why the increase in PCT is beyond 50°F using the proposed EXEM BWR-2000 model. Since the 940F PCT increase is a significant increase above the 500F, a more detailed explanation is required;
Response
The change in Peak Clad Temperature (PCT) is the result of the change in methodology as well as more conservative assumptions for two input parameters as stated in Section 4.0 of the RBS License Amendment Request (Reference 1). The input parameters used for both analyses are provided in Tables 1 through 5 of to Reference 1. Processes in place at RBS ensure that the input parameters bound actual plant performance.
The input parameters changed are the number of Automatic Depressurization System (ADS) valves available and the reactor vessel low pressure Emergency Core Cooling System (ECCS) permissive value. Specifically, the new analyses assume operation of only 4 ADS valves; whereas, the current analyses assume operation of 5 ADS valves.
The reactor vessel low pressure ECCS permissive was changed from 450 to 350 psia.
The ADS consists of 7 of the 16 Main Steam Safety Relief Valves. The ADS is designed to provide depressurization of the reactor pressure vessel (RPV) during a small break Loss of Coolant Accident (LOCA) if the High Pressure Core Spray (HPCS) fails or is unable to maintain required water level in the RPV. ADS operation reduces the RPV pressure to within the operating pressure range of the low pressure ECCS systems (Low Pressure Core Spray (LPCS) and Low Pressure Coolant Injection (LPCI) systems), so that these systems can provide core cooling. The Technical Specifications require 7 ADS valves to be operable. Assuming that only 4 ADS valves operate is conservative but does not have a significant impact on PCT because the RBS limiting analysis for PCT is a large break LOCA. The RPV depressurization rate during a large break is not significantly affected by whether 4 or 5 ADS valves are assumed to function.
The reactor vessel low pressure permissive signals ensure that, prior to opening the LPCI and LPCS injection valves, the reactor pressure has fallen to a value below these systems' maximum design pressure. The Technical Specifications require the permissive setpoint to be at or above 472 psig. Using a pressure permissive value of less than 472 psig for the analysis is conservative because the ECCS injection is delayed until the reactor vessel pressure falls to the low pressure value.
to RBG-46182 Page 2 of 3 Both of these changes represent more conservative conditions. As such, the input parameters for the new analyses continue to bound actual plant performance.
With respect to the change in PCT, the impact associated with the operation of one less ADS valve is not significant since the limiting break for RBS is the large break.
Thus, the observed increase in PCT is due primarily to the conservative lower reactor pressure ECCS permissive assumptions.
(2) which parameter such as reactor pressure permissive for operating valves or number of valves available contributes the most increase in PCT of 94 OF due to changing the old EXEM BWR model to EXEM BWR-2000 model in the River Bend LOCA analysis;
Response
As discussed above, the lower reactor pressure permissive signal assumed for the LPCI and LPCS injection valves is the most significant contributor to the increase in PCT.
(3) the results of the calculation based on the same assumptions used for EXEM BWR model
Response
As explained in responses (1) and (2) above, the primary contributor to the increased PCT is the conservative assumption of a lower reactor pressure permissive parameter (from 450 psia to 350 psia). It is expected that without any changes to the analysis inputs, the EXEM BWR-2000 methodology would have produced a slightly lower PCT than the current EXEM BWR methodology used (i.e., the resulting PCT would be less than 18750 F). The precise PCT impact for each of the changes has not been quantified by formal calculation due to the extent of the permutations involved.
However, through discussions with the fuel vendor, we have concluded that the PCT impact for each of the changes can be reasonably estimated. It is estimated that the increase in PCT due to the conservative assumption of a lower reactor pressure permissive parameter is more than 1000 F. It should be noted that since the TS limit on the reactor pressure permissive is unchanged, actual low pressure ECCS performance during a LOCA would continue to be as currently designed resulting in no increase in PCT.
(4) the rational for not using the lower initial MCPR input value in the EXEM BWR-2000 model
Response
The initial hot assembly MCPR input selected for the new analysis is consistent with the current analysis. For AREVA LOCA analyses, the initial MCPR chosen for the analyses is a value that is above the MCPR safety limit (SLMCPR), but is below the expected MCPR Operating Limit (OLMCPR) as determined from the limiting analyzed Anticipated Operational Occurrences (AOO's). OLMCPRs may vary slightly from cycle to cycle depending on factors such as core designs, operations, etc. For RBS, to RBG-46182 Page 3 of 3 the OLMCPR as determined by AOO analysis is not expected to be less than 1.16.
For example, the current cycle lowest OLMCPR determined by AOO analysis is 1.19.
An initial MCPR of 1.16 provides sufficient margin to the OLMCPR and thus was not changed for the new LOCA analysis. If the assumed MCPR was above the AOO OLMCPR, then the actual OLMCPR would be set by the initial MCPR assumed in the LOCA analysis.
The assumed MCPR of the RBS LOCA analyses of 1.16 does not result in reduced operating margins, since the actual OLMCPR is set by AQOs which are higher. If the AREVA LOCA analyses were re-analyzed at the AOO OLMCPR, the PCT would decrease. Therefore, it is concluded that the LOCA initial MCPR of 1.16 is reasonable for evaluating 1 OCFR50.46 criteria, and does not impact OLMCPR margins.
(5) consistent common assumptions used for the analysis to support APLHGR.
Response
The LOCA analysis to support APLHGR uses the NRC approved EXEM BWR-2000 LOCA methodology and the plant -specific input parameters such as reactor power, core flow, and ECCS parameters. The plant -specific input parameters used in the LOCA analysis are selected to be bounding for actual plant performance within the RBS licensing basis.
References :
- 1.
Letter RBG-46583, "License Amendment Request: Changes to the Analytical Methods Referenced in Technical Specification 5.6.5, "Core Operating Limits Report (COLR)"
dated October 16, 2006 (ADAMS Accession No. ML062960299)