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{{Adams
#REDIRECT [[GNRO-2011/00046, Response to Request for Additional Information Regarding Extended Power Uprate]]
| number = ML111670059
| issue date = 06/15/2011
| title = Response to Request for Additional Information Regarding Extended Power Uprate
| author name = Krupa M A
| author affiliation = Entergy Operations, Inc
| addressee name =
| addressee affiliation = NRC/NRR, NRC/Document Control Desk
| docket = 05000416
| license number = NPF-029
| contact person =
| case reference number = GNRO-2011/00046
| document type = Letter
| page count = 6
}}
 
=Text=
{{#Wiki_filter:GNRO-2011/00046 June 15, 2011
 
U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC  20555
 
==SUBJECT:==
Request for Additional Information Regarding  Extended Power Uprate  Grand Gulf Nuclear Station, Unit 1 Docket No. 50-416 
 
License No.
NPF-29   
 
==REFERENCES:==
: 1. Email from A. Wang to F. Burford dated May 20, 2011, GGNS EPU Request for Additional Information Related to Mechanical and Civil Engineering Branch Review Excluding the Steam Dryer (ME4679) (NRC ADAMS Accession Number ML111400299)  2. License Amendment Request, Extended Power Uprate, dated September 8, 2010 (GNRO-2010/00056, NRC ADAMS Accession Number ML102660403)
 
==Dear Sir or Madam:==
 
The Nuclear Regulatory Commission (NRC) requ ested additional information (Reference 1) regarding certain aspects of the Grand Gulf Nuclear Station, Unit 1 (GGNS) Extended Power Uprate (EPU) License Amendment Request (LAR) (Reference 2). Attachment 1 provides responses to the additional information requested by the Mechanical and Civil Engineering
 
Branch.   
 
No change is needed to the no significant hazards consideration included in the initial LAR (Reference 2) as a result of the additional information provided. There are no new commitments included in this letter.
 
If you have any questions or require additional information, please contact Jerry Burford at
 
601-368-5755. 
 
Entergy Operations, Inc. P. O. Box 756 Port Gibson, MS  39150 Michael A. Krupa Director, Extended Power Uprate Grand Gulf Nuclear Station Tel.  (601) 437-6684 GNRO-2011/00046 Page 2 of 2 I declare under penalty of perjury that the foregoing is true and correct. Executed on June 15, 2011.   
 
Sincerely,
 
MAK/FGB/dm
 
Attachments:
: 1. Response to Request for Additional Information, Mechanical and Civil Engineering Branch -
Non-Steam Dryer
 
cc: Mr. Elmo E. Collins, Jr.
Regional Administrator, Region IV U. S. Nuclear Regulatory Commission 612 East Lamar Blvd., Suite 400 Arlington, TX  76011-4005
 
U. S. Nuclear Regulatory Commission ATTN: Mr. A. B. Wang, NRR/DORL (w/2) ATTN: ADDRESSEE ONLY ATTN: Courier Delivery Only Mail Stop OWFN/8 B1
 
11555 Rockville Pike Rockville, MD  20852-2378 State Health Officer Mississippi Department of Health P. O. Box 1700 Jackson, MS  39215-1700 NRC Senior Resident Inspector Grand Gulf Nuclear Station Port Gibson, MS  39150
 
Attachment 1 GNRO-2011/00046 Grand Gulf Nuclear Station Extended Power Uprate Response to Request for Additional Information Mechanical and Civil Engineering Branch Non-Steam Dryer    to  GNRO-2011/ 00046
 
Page 1 of 3 Response to Request for Additional Information Mechanical and Civil Engineering Branch  Non-Steam Dryer By letter dated September 8, 2010, Entergy Operations, Inc. (Entergy) submitted a license amendment request (LAR) for an Extended Power Uprate (EPU) for Grand Gulf Nuclear Station, Unit 1 (GGNS). By letter dated February 23, 2011, Entergy submitted responses to the initial request for additional information (RAI) from the Mechanical and Civil Engineering Branch (NRC ADAMS Accession No. ML110540545). Subsequently, the U.S. Nuclear Regulatory Commission (NRC) staff has determined that the following additional information requested by the Mechanical and Civil Engineering Branch is needed for the NRC staff to complete their review of the amendment. Entergy's response to each item is also provided below.
RAI # 1 In response to request for additional information (RAI) 1, Reference 1 indicates that the turbine stop valve closure (TSVC) fluid transient loads, used in support of the pipe stress analyses performed for the GGNS EPU, were generated using the STEHAM computer code. Section 3.9.1.2.1.3.3 of the GGNS Updated Final Safety Analysis Report (UFSAR) indicates that the TSVC loads were generated using the TSFOR code, for the current licensed thermal power (CLTP) level. Please provide justification for the use of the STEHAM code in lieu of the TSFOR code, given that the latter is part of the current licensing basis for GGNS. This justification should include results of benchmarking performed to compare the results of the STEHAM code against the TSFOR code, to support the use of the former at GGNS. Additionally, provide any bases for the regulatory acceptance of the STEHAM code, if the STEHAM code has previously been utilized in support of licensing actions which required NRC staff approval.
Response For the CLTP piping evaluations, TSVC forcing functions were generated using the GE - Hitachi Nuclear Energy Americas, LLC (GEH) proprietary TSFOR code for load development in the piping analyses for the Main Steam (MS) system piping inside containment, as indicated in Section 3.9.1.2.1 of the GGNS UFSAR. For the EPU evaluation, the NRC approved ratio method was applied to MS and associated piping system evaluation (inside containment) as stated in GGNS EPU LAR Attachment 5, Section 2.2.2.2.1.2. For the CLTP piping evaluations, TSVC forcing functions were generated using a hand calculation method for load development in the piping analyses for the MS system piping outside containment; this is a separate calculation from the analyses for the MS piping inside containment. For the EPU evaluation of the MS piping outside containment, the NRC approved ratio method was applied to MS and associated piping (outside containment) and the loads resulted in some MS piping and supports exceeding code allowable limits, as described in GGNS EPU LAR Attachment 5, Section 2.2.2.2.2.2. As a result, for EPU, the MS piping was re-evaluated using a more detailed analysis using computer code STEHAM to generate the TSVC    to  GNRO-2011/ 00046
 
Page 2 of 3 forcing function. The computer based analysis provided more realistic results compared to the original hand calculation, which was very conservative. This EPU analysis using the STEHAM derived forcing function resulted in all MS piping and pipe supports outside containment meeting all code criteria. The STEHAM computer program is a SHAW Stone and Webster proprietary code used in all steam hammer computerized analyses for all nuclear plant designs including power uprate projects since 1975, including the Point Beach Units 1 & 2 EPU license amendment application recently approved by the NRC (NRC ADAMS Accession No. ML110880039). The program description is provided in the first round response to RAI #1 (NRC ADAMS Accession No. ML110540545). The software is qualified for use in accordance with Shaw's QA program
 
requirements.
RAI # 2 In response to RAI 3.c, it was stated in Reference 1 that the GGNS feedwater (FW) piping experiences no percent-increase in stresses as a result of the proposed EPU implementation at GGNS. With respect to the response to RAI 3.c, please address the following: a) Please confirm that the piping stress analyses performed in support of the proposed EPU implementation at GGNS were carried out at the limiting conditions for operation of these piping systems. This should include a confirmation that the FW system has been evaluated at the highest temperature at which it is expected to operate under EPU conditions (i.e.,
confirm that the proposed EPU at GGNS will be accomplished with no change in the maximum operating temperature). b) The response to RAI 3.c states that the FW flow at GGNS will increase by approximately 13.1% following EPU implementation. Please provide a quantitative summary of the effects which the higher FW flow has on the occasional loads considered in the FW pipe stress analyses, including water hammer and other transients associated with higher FW flow. As part of this response, please specify which American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code load combinations, for FW pipe stresses and pipe supports, are affected by FW flow transients. c) Please confirm that these increased occasional loads, discussed in RAI 2.b, above, were considered in the pipe stress re-analyses, including affected pipe supports, in accordance with the provisions of the design code of record applicable to the GGNS FW piping and
 
supports. d) Please confirm that the increased loading resulting from FW flow transients at EPU conditions were developed in accordance with the methodologies outlined in Section 3.5 of the constant pressure power uprate licensing topical report (CLTR or Reference 2). If these loads were not developed in accordance with the NRC-approved methodologies discussed in the CLTR, please discuss the methodologies used to evaluate the FW flow transients and    to  GNRO-2011/ 00046
 
Page 3 of 3 provide the regulatory bases for the acceptance of a methodology which is not described in the CLTR.
Response    a) It is confirmed that the piping stress analyses for the feedwater (FW) system performed in support of the proposed EPU implementation at GGNS were carried out at the limiting conditions for operation. The FW system has been evaluated at the highest operating
 
temperature of 420
°F under EPU conditions. b) The EPU flow rate increase of approximately 13.1% does not affect the reactor FW piping system because water hammer loads were not a design load in the original stress analyses.
Thus, the increased EPU flow rate will not cause any new credible design basis transient loading. For EPU, there will not be any new transient loads for the flow increase. c) The GGNS design basis results of the occasional and/or transients loadings for the FW systems are not affected by the EPU flow rate change. As indicated in response 2b) above, the water hammer loads were not a design basis load in the original stress analysis calculation. d) The original design basis FW flow transient loads were not affected by the EPU parameters applicable to the FW system. Therefore, the design basis transient load results remain unchanged for the uprated conditions for the FW system. The FW piping was evaluated in accordance with the methodologies outlined in Section 3.5 of the constant pressure power uprate licensing topical report (CLTR). REFERENCES
: 1) Letter from M. A. Krupa, Entergy Operations, Inc., to NRC Document Control Desk, "Request for Additional Information Regarding Extended Power Uprate - Grand Gulf Nuclear Station, Unit 1 - Docket No. 50-416 - License No. NPF-29," dated February 23, 2011. (ADAMS Accession No.: ML110540545) 2) GE Nuclear Energy, "Constant Pressure Power Uprate," Licensing Topical Report NEDC-33004P-A, Revision 4, Class III (Proprietary), July 2003; and NEDO-33004, Class I (Non-
 
proprietary), July 2003.}}

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