GNRO-2011/00011, Request for Additional Information Regarding Extended Power Uprate
| ML110540545 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  |
| Issue date: | 02/23/2011 |
| From: | Krupa M Entergy Operations |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML110540544 | List: |
| References | |
| GNRO-2011/00011, TAC ME4679 | |
| Download: ML110540545 (21) | |
Text
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 Attachment 1 contains proprietary information.
GNRO-2011/00011 February 23, 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 January 27, 2011, GGNS EPU Request for Additional Information Related to Mechanical and Civil Engineering Branch Review Excluding the Steam Dryer (ME4679)
(Accession Number ML110270187)
- 2. License Amendment Request, Extended Power Uprate, dated September 8, 2010 (GNRO-2010/00056, Accession Number ML102660403)
Dear Sir or Madam:
The Nuclear Regulatory Commission (NRC) requested 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.
GE-Hitachi Nuclear Energy Americas, LLC (GEH) consider portions of the information provided in support of the responses to the request for additional information (RAI) in Attachment 1 to be proprietary and therefore exempt from public disclosure pursuant to 10 CFR 2.390. An affidavit for withholding information, executed by GEH, is provided in Attachment 3. The proprietary information was provided to Entergy in a GEH transmittal that is referenced in the affidavit.
Therefore, on behalf of GEH, Entergy requests to withhold Attachment 1 from public disclosure in accordance with 10 CFR 2.390(b)(1). A non-proprietary version of the RAI responses is provided in Attachment 2.
When Attachment 1 is removed, the entire letter is non-proprietary.
GNRO-2011/00011 Page 2 of 2 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.
I declare under penalty of perjury that the foregoing is true and correct. Executed on February 23, 2011.
Sincerely, MAK/FGB/dm Attachments:
- 1. Response to Request for Additional Information, Mechanical and Civil Engineering Branch - Proprietary
- 2. Response to Request for Additional Information, Mechanical and Civil Engineering Branch - Non-Proprietary
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 2 GNRO-2011/00011 Grand Gulf Nuclear Station Extended Power Uprate Response to Request for Additional Information Mechanical and Civil Engineering Branch Non-Proprietary This is a non-proprietary version of Attachment 1 from which the proprietary information has been removed. The proprietary portions that have been removed are indicated by double square brackets as shown here: (( )).
to GNRO-2011/ 00011 Page 1 of 14 Non-Proprietary Response to Request for Additional Information Mechanical and Civil Engineering Branch 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). The U.S. Nuclear Regulatory Commission (NRC) staff has determined that the following additional information requested by the Mechanical and Civil Engineering Branch by correspondence dated January 27, 2011 (Accession Number ML110270187) is needed for the NRC staff to complete their review of the amendment. Entergys response to each item is also provided below.
GE-Hitachi Nuclear Energy Americas, LLC (GEH) consider portions of the information provided in support of the responses to the request for additional information (RAI) to be proprietary and therefore exempt from public disclosure pursuant to 10 CFR 2.390. An affidavit for withholding information, executed by GEH, is provided in Attachment 3. The proprietary information was provided to Entergy in a GEH transmittal that is referenced in the affidavit. Therefore, on behalf of GEH, Entergy requests to withhold this Attachment from public disclosure in accordance with 10 CFR 2.390(b)(1). A non-proprietary version of the RAI responses is provided in .
RAI # 1 Section 2.2.2.2.1.2 of Attachment 5A to Reference 1, Structural Evaluation for Affected Safety-Related Piping, indicates that for the stress analyses of the main steam (MS) piping outside containment, the turbine stop valve closure (TSVC) loads used in these analyses were developed using a methodology which contrasts with the methodology found in the NRC-approved Constant Pressure Power Uprate Licensing Topical Report (CLTR or Reference 2),
the latter of which utilizes scaling factors to develop these loads under EPU conditions. Please describe the development of the forcing functions applied in the stress analyses to model the TSVC loads and provide the bases for the regulatory acceptance of this methodology in place of the NRC-approved methodology. Additionally, please indicate whether this forcing function was utilized in lieu of the NRC-approved methodology to evaluate the acceptability of other piping systems, including supports, for the proposed EPU implementation at GGNS.
Response
Forcing functions for TSVC fluid transient loads are generated using the STEHAM computer code. The STEHAM computer program is a general fluid transient analysis code that is used to perform steady state and transient analyses of a steam filled flow network. The program has the capability to model any compressible fluid flow network containing valves, safety / relief Non-Proprietary to GNRO-2011/ 00011 Page 2 of 14 Non-Proprietary valves, reservoirs, branch piping, and steam chests. The steam is modeled as an ideal gas with homogenous and adiabatic fluid properties.
The program uses the method of characteristics numerical algorithm with a finite difference approximation in both space and time in order to solve the governing partial differential equations. The spatial characteristics of the system are defined by the flow parameters that are assigned to each node. One node is placed at each end of a pipe and the remaining nodes are equally distributed along the pipe. The nodal span is governed by the integration time step, together with the criteria:
t = x / (a + v) where
t = time step (sec),
x = nodal span (ft),
v = velocity (fps), and a = speed of sound (fps).
The fluid flow piping model is developed based on the system flow diagram, corresponding piping / isometric drawings, valve and equipment functions and heat balance drawings.
The reactor pressure vessel is modeled as a constant pressure reservoir. The inlet nozzle to the MSRs, high pressure turbine, condenser, and reactor feed-pump turbine are modeled as choked or non-choked valves with constant back pressure. Pipes are modeled as pipe components with interior nodes set at specific spans. The friction factor for each pipe is determined. Tee / Branch connections and pipe area changes are modeled as steam chests with the volume calculated as the sum of the attached nodal volumes. A nodal volume is equal to the product of the attached pipe nodal span and flow area. The Turbine Stop valves (TSVs) and Turbine Bypass valves (TBVs) are modeled as choked valves. The flow element is simulated as the passive isolation valve located outside containment.
Forcing functions are determined for each pipe segment in the computer model.
The forcing function developed using this methodology was used in the piping analyses of the main steam (MS) piping outside containment only. The supports associated with the affected main steam piping were evaluated to the resulting TSVC loads.
This methodology was not applied to any other BOP piping system.
Non-Proprietary to GNRO-2011/ 00011 Page 3 of 14 Non-Proprietary RAI # 2 Section 2.2.2.2.1.2 of Attachment 5A to Reference 1 also indicates that the MS piping outside containment was evaluated for acceptability at EPU conditions by demonstrating compliance with the allowable limits of the code of record. The code of record, as indicated in this section, is provided as the:
[American Society of Mechanical Engineers] ASME [Boiler & Pressure Vessel]
B&PV Code,Section III, Subsection NB, 1974 Edition with Addenda through Summer 1975, with exceptions and use of some sub-sections from the 1977 Edition with Addenda through Summer 1979 and the 1980 Edition with Addenda through Summer 1981.
Please confirm that the exceptions noted above are consistent with the current licensing basis (CLB) requirements of the MS piping outside containment. If these exceptions are not part of the GGNS CLB for this system, please provide justification for evaluating these components against code requirements which are not part of the CLB for this system.
Response
The exceptions noted above are part of the current design basis.
RAI # 3 Section 2.2.2.2.1.2 of Attachment 5A to Reference 1 also describes the structural evaluation of the feedwater (FW) piping system, which is affected by the proposed EPU implementation at GGNS. Please address the following items related to the evaluations performed for the FW piping system:
a) Please provide the Edition(s) and Addenda of the ASME B&PV Code utilized to demonstrate the acceptability of the FW piping inside containment. Additionally, please confirm that the Edition(s) and Addenda utilized to demonstrate the compliance of the FW system inside containment at EPU conditions is/are consistent with the GGNS CLB.
b) The description of the structural evaluation of the FW piping outside containment indicates that this portion of the FW piping system, including branch piping, was evaluated for acceptability at EPU conditions by demonstrating compliance with the allowable limits of the code of record. The code of record, as indicated in this section, is provided as the:
ASME B&PV Code,Section III, Subsection NB, 1974 Edition with Addenda through Summer 1975, with exceptions and use of some sub-sections from the Non-Proprietary to GNRO-2011/ 00011 Page 4 of 14 Non-Proprietary 1977 Edition with Addenda through Summer 1979 and the 1980 Edition with Addenda through Summer 1981.
Please confirm that the exceptions noted above are consistent with the CLB requirements of the FW piping outside containment. If these exceptions are not part of the GGNS CLB for this system, please provide justification for evaluating these components against code requirements which are not part of the CLB for this system.
c) Table 2.2-3a of Attachment 5A to Reference 1 indicates that the FW piping experiences no percent-increase in stresses due to the implementation of an EPU at GGNS. Please provide additional details regarding these results, including a discussion of the impact that the EPU has on the individual loads which make up the ASME Code equations for which compliance must be demonstrated. Additionally, please discuss the assumptions which were made regarding the operating conditions of the FW system when the stress analyses for the system were performed for EPU conditions. The discussion of any assumptions should include, but not be limited to, the FW heaters out of service (FWHOOS) assumption included on page 1-13 and Note 9 to Table 2.2-7, which indicates that normal operations do not include the FWHOOS assumption. If the FWHOOS assumption has been included as part of the FW piping structural analyses, please provide justification that this assumption provides for a bounding FW piping structural analysis at EPU conditions.
Response
a) The GGNS FW piping was analyzed in accordance with the rules of NB-3600 of ASME Section III Boiler and Pressure Vessel Code 1974 Edition and Addenda through Summer 1975 to demonstrate the acceptability of the FW piping inside containment. This is the current licensing basis (CLB) and the Edition and Addenda utilized to demonstrate the compliance of the FW system inside containment at EPU conditions.
b) The exceptions noted above are part of the current design basis.
c) Table 2.2-3a of Attachment 5A to Reference 1 indicates that the FW piping experiences no percent-increase in stresses due to the implementation of an EPU at GGNS because there is no change in temperature due to EPU from the current licensing thermal power (CLTP) temperature and also pressure change for EPU for the reactor feedwater (FW) system is less than 1% compared to CLTP pressure. The flow change of approximately 13.1% does not affect the FW piping system. FWHOOS at EPU conditions is acceptable because the current licensing basis is not affected by the EPU conditions as discussed above.
The following table provides the quantitative CLTP and EPU temperature and pressure values as discussed above.
Non-Proprietary to GNRO-2011/ 00011 Page 5 of 14 Non-Proprietary FW Piping Analysis Input EPU - CLTP Parameter CLTP EPU Remarks
(%)
Temperature (qF) 420 420 0.0% No Change Pressure (psia) 1065.0 1072.0 0.7% < 1% change With respect to the Feedwater Nozzle Analysis, calculations were performed according to EPU methodology with additional exceptions, considerations, and calculations noted in RAI 8.
Table 2.2-7 of Attachment 5A of the EPU LAR contains stress and fatigue values for the Feedwater Nozzle analysis, not the Feedwater Piping system which was analyzed separately. Note 9 of Table 2.2-7 indicates that the reported cumulative usage factor (CUF) value is only for normal operation of the Feedwater Nozzle. The Feedwater Heater Out-of-Service flexibility option was considered in a separate fatigue calculation and found acceptable (CUF < 1.0) based on a limited operational basis per year.
RAI # 4 Section 2.2.2.2.2.2 of Attachment 5A to Reference 1 discusses the structural evaluations performed for balance-of-plant (BOP) piping affected by the proposed EPU implementation at GGNS. The affected BOP piping systems were evaluated for acceptability at EPU conditions by demonstrating compliance with the allowable limits of the applicable codes of record. The code of record for the Class 2 and Class 3 piping, as indicated in this section, is provided as the, ASME B&PV Code,Section III, Division I, 1974 Edition with Addenda through Summer 1975, with some exceptions and use of the 1977 Edition with Addenda through Winter 1979 and the 1980 Edition with Addenda through Winter 1981. Please confirm that the exceptions noted above are consistent with the CLB requirements of the affected Class 2 and 3 BOP piping systems. If these exceptions are not part of the GGNS CLB for this system, please provide justification for evaluating these components against code requirements which are not part of the CLB for this system.
Response
The exceptions noted above are part of the current design basis.
Non-Proprietary to GNRO-2011/ 00011 Page 6 of 14 Non-Proprietary RAI # 5 The conclusions regarding the structural evaluations for all piping systems affected (safety-related and BOP) by the proposed EPU implementation at GGNS indicate that all pipe supports on the affected piping systems will maintain adequate design margin to accommodate the additional loads imposed by the EPU implementation. However, no information was provided to support this conclusion. Please provide the code(s) of record for the aforementioned piping supports and confirm that these supports were evaluated against the applicable provisions of their respective code(s) of record as part of the structural evaluation of the piping systems affected by the proposed EPU implementation at GGNS.
Response
The affected safety related piping inside containment was analyzed for impacts for the extended power uprate (EPU) parameters. The pipe support load increases due to the EPU parameters were found to be bounded within the maximum design load and structural capacity for each of the affected supports. Therefore, the pipe support structures for the inside containment piping did not require any design modifications.
Based on existing margins available for the BOP (safety-related and non-safety related) piping supports outside containment, it was concluded that EPU does not result in reactions on existing structures in excess of the current design capacity.
The Codes of record for the safety-related pipe supports are listed in Chapter 3, Sub-Section 3.2, Table 3.2-4 of the GGNS UFSAR. The Codes of record delineated for pipe supports components in the table are as follows:
- a. Hangers and support (Excluding Snubbers): ASME Section III, 1974 Edition, no addenda
- b. Snubbers: ASME Section III, 1974 Edition, through Summer 1976
- c. Fabrication of steels for pipe supports: ASME Section III, 1974 Edition, with Summer 1974 addenda The code of record for the BOP non-safety related piping supports is ANSI B31.1 Power Piping, 1973 Edition and all addenda up to and including Winter 1973.
Non-Proprietary
Attachment 2 to GNRO-2011/ 00011 Page 7 of 14 Non-Proprietary RAI # 6 As part of the piping structural evaluations performed in support of the proposed EPU implementation at GGNS, Tables 2.2-4a and 2.2-4c of Attachment 5A to Reference 1 appear to provide a quantitative summary of the most limiting nodes resulting from the stress analyses of the ASME Class 1 MS piping. In order to demonstrate compliance of these portions of the MS piping with the applicable code of record, the stresses and CUF values coupled with these limiting nodes were compared to the corresponding ASME Section III Subsection NB allowable values. Furthermore, Tables 2.2-4b and 2.2-4d evaluated each limiting node against the code allowable values from Equations 10, 12, 13, and 14 in an effort to demonstrate that the pipe stresses satisfied the criteria used at GGNS to postulate an intermediate pipe break. However, Table 2.2-4d (for MS lines B & C) does not include information demonstrating that nodes 128 and 020 satisfy the GGNS pipe stress criteria for pipe break postulation though these nodes were denoted on Table 2.2-4c as limiting nodes with respect to Equations 10 and 14, respectively. Please provide the quantitative results for nodes 128 and 020 which supports the conclusion found in Section 2.2.1 of Attachment 5A to Reference 1, which states that no new break or crack locations are required to be postulated as a result of EPU implementation.
Response
Table 2.2-4d (for MS lines B & C) has been copied below and supplemented with information demonstrating that nodes 128 and 020 satisfy the GGNS pipe stress criteria for pipe break postulation. The quantitative results shown below for nodes 128 and 020 support the conclusion found in Section 2.2.1 of Attachment 5A to the EPU LAR that no new break or crack locations are required to be postulated as a result of EPU implementation.
Table 2.2-4d Checks for Pipe Breaks Against All Nodes Above Maximum Code Stresses as Shown In Table 2.2-4c MS Lines B & C Node EQ. 10 EQ. 12 EQ.13 EQ. 14 Remarks 31,821 x 1.023 =
022 5,550 < 45,960 15,931 < 45,960 0.0414 x 1.0115 = 0.0419 < 0.1 Ok 32,552 < 45,960 54,539 x 1.023 =
002N 37,486 < 45,960 18,699 < 45,960 0.0110 x 1.0115 = 0.0111 < 0.1 Eq 12/13 satisfy.
55,793 > 45,960 56,675 x 1.023 125 13,704 < 43,680 27,101 < 43,680 0.0697 x 1.0115 = 0.0705 < 0.1 Eq 12/13 satisfy.
57,978 > 43,680 34,030 x 1.023
030 416 < 42,480 29,706 < 42,480 0.0061 x 1.0115 =0.0062 < 0.1 Ok 34,812 < 42,480 56,937 x 1.023 =
128 13,411 < 54,600 27,152 < 54,600 0.0685 x 1.0115 = 0.0693 < 0.1 Eq 12/13 satisfy.
58,247 > 54,600 28,772 x 1.023 =
020 4,733 < 57,450 25,770 < 54,600 0.0732 x 1.0115 =0.0740 < 0.1 Ok 29,434 < 54,600 Non-Proprietary to GNRO-2011/ 00011 Page 8 of 14 Non-Proprietary RAI # 7 In Section 2.2.2.3 of Attachment 5A to Reference 1, a number of components are determined to be consistent with the dispositions provided in a number of topical reports, including Reference
- 2. As such, it is concluded that these components are acceptable for operation at EPU conditions. While the NRC has approved the use of the methodologies within the three topical reports cited in Section 2.2.2.3 for dispositioning a number of reactor pressure vessel components with respect to EPU implementation evaluation, it is not clear which of these methodologies have been applied in dispositioning the components listed on pages 2-63 and 2-64 of Attachment 5A to Reference 1. For each component, please provide the specific provision(s) from the applicable topical report(s) cited which provides the basis for a components acceptability at EPU conditions.
Response
NEDC-33004P-A, Licensing Topical Report Constant Pressure Power Uprate, Section 3.2, Reactor Pressure Vessel (RPV) and Internals states in part, ((
)) the plant specific evaluation will be performed consistent with the methods documented in Appendix I of ELTR1.
NEDC-32424P-A, Licensing Topical Report Generic Guidelines for General Electric Boiling Water Reactor Extended Power Uprate, (i.e., ELTR1)Section I.2 Reactor Vessel Analysis states in part ((
))
NEDC-32424P-A,Section I.2 Reactor Vessel Analysis, Subsection I.2.2 Normal and Upset Conditions states in part ((
))
PUSAR Section 2.2.2.3 components dispositioned for EPU ((
))
Non-Proprietary to GNRO-2011/ 00011 Page 9 of 14 Non-Proprietary PUSAR Section 2.2.2.3 components with ((
))
For other pressure boundary components not covered by LTR generic dispositions; ((
)) Plant specific analysis is performed to qualify these components for EPU. The results of plant specific analysis can be seen in PUSAR Table 2.2-7.
RAI # 8 Table 2.2-7 of Attachment 5A to Reference 1 provides the stresses and CUFs for limiting components of the reactor pressure vessel. Specifically, this table illustrates the stresses and CUFs of these limiting components at the current power level, the EPU power level, and compares these values against the allowable stresses and CUF limit. Regarding Table 2.2-7, please address the following:
a) For the first component, the FW Nozzle - Carbon Steel Replacement Safe End, the peak stress of this component decreases at EPU conditions when compared to the stress realized at current operating conditions. This appears discrepant based on the fact that the loading conditions at EPU conditions are more limiting than those at the current power level. Please provide a technical justification, including a description of any revised analyses performed for this component, which rectify this apparent discrepancy and justify its acceptability for operation at the proposed EPU power level.
b) For the second component, the FW Nozzle - Stainless Steel Clad Replacement Safe End, the stress at EPU conditions rises while the overall CUF at EPU decreases. However, Note 5 to Table 2.2-7 indicates that a reduced number of cycles and a finite element analysis (FEA) calculation of critical transients were considered in order to arrive at the EPU CUF for this component. Please provide justification, including the incorporation of operating Non-Proprietary to GNRO-2011/ 00011 Page 10 of 14 Non-Proprietary experience, for reducing the number of cycles considered in evaluating the fatigue of this component at EPU conditions. Additionally, please provide additional information regarding the FEA (summary of analysis) performed on this component and how this was factored into the fatigue evaluation.
Response
a) Using EPU scaling methodology the maximum stress for the FW Nozzle - Carbon Steel Replacement Safe End exceeds ASME Code allowable stress limits. Therefore, a finite element analysis (FEA) was performed to demonstrate acceptability of the Feedwater Nozzle at this location. See the response for 8 b) for a general description of the FEA methods. The question of why the maximum stress decreased despite increases in operating conditions requires consideration of: (1) the particular transient in question and (2) comparison of CLTP and EPU FEA methods.
(1) The reason the stress did not increase with increased operating conditions is due to the transient under consideration. ((
)).
(2) The FEA reduced stresses from the CLTP evaluation mainly due to more modern FEA tools and techniques. Examples of CLTP FEA issues that are no longer in the EPU FEA evaluation include the following:
((
Non-Proprietary to GNRO-2011/ 00011 Page 11 of 14 Non-Proprietary
))
b) Grand Gulf has instituted a fatigue monitoring program, documentation of which shows projected 40-yr life thermal cycle counts for some transients. During the course of the EPU evaluation, it was found that the basis for the CLTP evaluation used thermal cycle counts, in some cases several times greater than the actual Grand Gulf thermal cycle documents.
Therefore, in the EPU calculation, for some critical transients, ((
)), the design cycles in the cumulative usage factor (CUF) calculation were reduced to the cycle counts in the Grand Gulf thermal cycle documents.
Reductions in considered design cycles were only implemented after reviewing fatigue monitoring documentation for thermal cycle projections exceeding Grand Gulf design cycles.
The FEA analysis was performed on those transients that most critically effected the CUF calculation. Unlike the Feedwater Nozzle - Carbon Steel Replacement Safe End calculation, EPU scaling methodology was used to qualify the primary plus secondary stress intensity of the Feedwater Nozzle - Stainless Steel Clad Replacement Safe End as compared to the ASME code allowable values.
The FEA analysis was performed using an axisymmetric model, created and run in ANSYS 11.0 SP1 with geometry and material properties nearly identical to the CLTP stress report.
Loadings were also applied in a similar manner to the CLTP stress report, though they were modified for EPU conditions. Stress values were extracted from the calculation for use in the Feedwater Nozzle - Stainless Steel Clad Replacement Safe End CUF calculation at the critical location from the CLTP analysis at the designated times in those critical transients.
Through the use of modern FEA tools and an enhanced mesh, certain FEA issues of the CLTP analysis were inherently removed from the calculation, such as those outlined in RAI Response 8 a). After combining stresses according to CLTP transient stress combinations and ASME methods, the primary plus secondary plus peak (P+Q+F) stresses could be used Non-Proprietary to GNRO-2011/ 00011 Page 12 of 14 Non-Proprietary in the Salt calculation and primary plus secondary (P+Q) stresses were ready for use in the Ke factor calculation. The FEA calculated P+Q+F and P+Q values supplanted EPU scaled P+Q+F and P+Q values which would have been otherwise used in Salt and Ke calculations.
RAI # 9 0 to Reference 1, Vibration Analysis and Testing Program, describes the procedures being implemented at GGNS to evaluate the effects of flow-induced vibration (FIV) on those piping systems which will be affected by flow increases as a result of the proposed EPU implementation at GGNS. With respect to the proposed EPU vibration monitoring program described in this attachment, please address the following:
a) Section 5.2 indicates that numerous modal analyses were performed on piping system models in an effort to determine additional locations susceptible to FIV following EPU implementation. Please discuss how the inertial and elastic properties of the piping systems were captured given that it is stated, Static loads, such as weight, were neglected.
b) Please indicate what computer program(s) and/or code(s) were utilized to perform the aforementioned modal analyses.
c) Reference 7.3 of Attachment 10 to Reference 1 indicates that the provisions of the ASME OM-S/G, Standards and Guides for Operation and Maintenance of Nuclear Power Plants, Part 3, 1987 Edition, were used in order to demonstrate the acceptability of the affected piping systems susceptible to FIV. Please confirm that the use of this version of the OM Code is in accordance with the CLB requirements at GGNS. If this is not the OM code of record for GGNS, please provide a technical justification for the use of a version which is not part of the GGNS CLB.
Response
a) The flow induced vibration loading is a long term high cycle fatigue alternating stress , which is evaluated based on the guidance of the ASME OM S/G Part 3, which states that for steady state vibration, the maximum calculated alternating stress shall not exceed Sel/,
where the loading considered is due to vibration only, where static and other inertia loading from seismic etc., are not included and is considered separately in the ASME code primary and secondary stress evaluation against different allowable values.
The governing equation from OM S/G Part 3 for the alternating stress criteria is given below:
Salt = C2 x K2 x M / Z Sel /
Non-Proprietary to GNRO-2011/ 00011 Page 13 of 14 Non-Proprietary Where:
Salt = Alternating stress intensity C2 = Secondary stress index as defined in ASME III Code K2 = Local stress intensification factor as defined in ASME III Code M = Maximum zero to peak dynamic moment loading due to vibration only Z = Section Modulus of the pipe Sel = 0.8 SA, where SA is the alternating stress at 106 cycle from Figure I-9.1, or SA at 1011 cycles from Figure I-9.2.2 of the ASME Code,Section III
= Allowable stress reduction factor, 1.3 for material covered by Figure I-9.1 or 1.0 for material covered by Figure I-9.2.1 or 9.2.2 of ASME Code,Section III For ASME III Class 2 and 3 Piping, or ANSI B31.1, the C2 x K2 = 2*i and i is the stress intensification factor, as defined in Sub-Section NC and ND of ASME Section III or ANSI B31.1. The maximum allowed alternating stress intensity is:
- Stainless steel material, SA = 13,600 psi, is unity, then, Salt = 0.8*13,600 = 10,880 psi.
b) The NUPIPE SWPC program is used to perform analyses in accordance with the ASME Boiler and Pressure Vessel Code,Section III Nuclear Power Plant Components. The program performs a linear elastic analysis of three dimensional piping systems subjected to various thermal, static and dynamic loading conditions including uniform spectra analyses.
The NUPIPE SWPC analysis was performed on an IBM compatible desktop computer system.
c) ASME OM-S/G, Standards and Guides for Operation and Maintenance of Nuclear Power Plants, Part 3, Requirements for Preoperational and Initial Start-up Vibration Testing of Nuclear Power Plant Piping Systems provides guidance for evaluating the alternating stress due to flow induced vibration and was generally adopted by the nuclear industry. The 1987 Edition was the first version in which a guide for flow induced vibration evaluation was available. The acceptance criteria for vibration testing in the 1987 version of ASME OM-Non-Proprietary to GNRO-2011/ 00011 Page 14 of 14 Non-Proprietary S/G, Part 3 remains unchanged and is the same as the current guidance provided in ASME OM-2009 which incorporated and superseded ASME OM-S/G-2007.
RAI # 10 From Section 2.3.4, Table 2.2-10 of Attachment 5A to Reference 1 presents the reactor internal pressure differences (RIPDs) for a number of reactor vessel internal components under faulted conditions at the current power level and under the proposed EPU power level. Note 1 to Table 2.2-10 provides additional context on the results presented in this table, including information concerning the FWHOOS assumption which results in a reduced FW temperature (RFWT).
Please indicate whether normal operating conditions or RFWT conditions result in bounding RIPD values and confirm that the bounding RIPDs were incorporated into the structural analyses of the reactor vessel internals in support of the proposed EPU implementation at GGNS.
Response
As specified in Note 1 to Table 2.2-10 of Attachment 5A to Reference 1, the RIPD values reported in this table at both CLTP and EPU conditions are the maximum values from the conditions covered, including both normal feedwater temperature (NFWT) and reduced feedwater temperature (RFWT) conditions. The maximum RIPD values for the Shroud Support Ring and Lower Shroud, Core Plate and Guide Tube, Upper Shroud, Top Guide, and Fuel Channel Wall (Maximum Power Bundle) are associated with NFWT. The Shroud Head and Shroud Head to Water Level (Irreversible) maximum RIPDs are associated with RFWT. The Shroud Head to Water Level dP at RFWT is the same as that at NFWT.
The bounding RIPDs were incorporated into the structural analyses of the reactor vessel internals in support of the proposed EPU implementation at GGNS.
REFERENCES
- 1) Letter from M. A. Krupa, Entergy Operations, Inc., to NRC Document Control Desk, License Amendment Request - Extended Power Uprate - Grand Gulf Nuclear Station, Unit 1 - Docket No. 50-416 - License No. NPF-29, dated September 8, 2010. (ADAMS Accession No.: ML102660403)
- 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.
Non-Proprietary
Attachment 3 GNRO-2011/00011 Grand Gulf Nuclear Station Extended Power Uprate Response to Request for Additional Information Mechanical and Civil Engineering Branch GEH Affidavit for Withholding Information from Public Disclosure
GE-Hitachi Nuclear Energy Americas LLC AFFIDAVIT I, Edward D. Schrull, PE, state as follows:
(1) I am the Vice President, Regulatory Affairs, Services Licensing, GE-Hitachi Nuclear Energy Americas LLC (GEH). I have been delegated the function of reviewing the information described in paragraph (2) which is sought to be withheld, and have been authorized to apply for its withholding.
(2) The information sought to be withheld is contained in GEH letter, GEH-GGNS-AEP-427, Larry King (GEH) to Brian Newell (Entergy), NRC Mechanical and Civil Engineering Branch RAIs, dated February 20, 2011. The proprietary information in Enclosure 1 entitled, Responses to GGNS NRC EMCB RAIs (Proprietary), is identified by a dotted underline inside double square brackets. ((This sentence is an example.{3})). In each case, the superscript notation {3} refers to Paragraph (3) of this affidavit that provides the basis for the proprietary determination (3) In making this application for withholding of proprietary information of which it is the owner or licensee, GEH relies upon the exemption from disclosure set forth in the Freedom of Information Act (FOIA), 5 USC Sec. 552(b)(4), and the Trade Secrets Act, 18 USC Sec. 1905, and NRC regulations 10 CFR 9.17(a)(4), and 2.390(a)(4) for trade secrets (Exemption 4). The material for which exemption from disclosure is here sought also qualifies under the narrower definition of trade secret, within the meanings assigned to those terms for purposes of FOIA Exemption 4 in, respectively, Critical Mass Energy Project v. Nuclear Regulatory Commission, 975 F2d 871 (DC Cir. 1992), and Public Citizen Health Research Group v. FDA, 704 F2d 1280 (DC Cir. 1983).
(4) The information sought to be withheld is considered to be proprietary for the reasons set forth in paragraphs (4)a. and (4)b. Some examples of categories of information that fit into the definition of proprietary information are:
- a. Information that discloses a process, method, or apparatus, including supporting data and analyses, where prevention of its use by GEH's competitors without license from GEH constitutes a competitive economic advantage over GEH and/or other companies.
- b. Information that, if used by a competitor, would reduce their expenditure of resources or improve their competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing of a similar product.
- c. Information that reveals aspects of past, present, or future GEH customer-funded development plans and programs, that may include potential products of GEH.
- d. Information that discloses trade secret and/or potentially patentable subject matter for which it may be desirable to obtain patent protection.
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(5) To address 10 CFR 2.390(b)(4), the information sought to be withheld is being submitted to the NRC in confidence. The information is of a sort customarily held in confidence by GEH, and is in fact so held. The information sought to be withheld has, to the best of my knowledge and belief, consistently been held in confidence by GEH, not been disclosed publicly, and not been made available in public sources. All disclosures to third parties, including any required transmittals to the NRC, have been made, or must be made, pursuant to regulatory provisions or proprietary and/or confidentiality agreements that provide for maintaining the information in confidence. The initial designation of this information as proprietary information, and the subsequent steps taken to prevent its unauthorized disclosure are as set forth in the following paragraphs (6) and (7).
(6) Initial approval of proprietary treatment of a document is made by the manager of the originating component, who is the person most likely to be acquainted with the value and sensitivity of the information in relation to industry knowledge, or who is the person most likely to be subject to the terms under which it was licensed to GEH. Access to such documents within GEH is limited to a need to know basis.
(7) The procedure for approval of external release of such a document typically requires review by the staff manager, project manager, principal scientist, or other equivalent authority for technical content, competitive effect, and determination of the accuracy of the proprietary designation. Disclosures outside GEH are limited to regulatory bodies, customers, and potential customers, and their agents, suppliers, and licensees, and others with a legitimate need for the information, and then only in accordance with appropriate regulatory provisions or proprietary and/or confidentiality agreements.
(8) The information identified in paragraph (2) above is classified as proprietary because it contains results of an analysis performed by GEH to support the Grand Gulf Nuclear Station Extended Power Uprate (EPU) license application. This analysis is part of the GEH EPU methodology. Development of the EPU methodology and the supporting analysis techniques and information, and their application to the design, modification, and processes were achieved at a significant cost to GEH.
The development of the evaluation methodology along with the interpretation and application of the analytical results is derived from the extensive experience database that constitutes a major GEH asset.
(9) Public disclosure of the information sought to be withheld is likely to cause substantial harm to GEH's competitive position and foreclose or reduce the availability of profit-making opportunities. The information is part of GEH's comprehensive BWR safety and technology base, and its commercial value extends beyond the original development cost.
The value of the technology base goes beyond the extensive physical database and analytical methodology and includes development of the expertise to determine and apply the appropriate evaluation process. In addition, the technology base includes the value derived from providing analyses done with NRC-approved methods.
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The research, development, engineering, analytical and NRC review costs comprise a substantial investment of time and money by GEH. The precise value of the expertise to devise an evaluation process and apply the correct analytical methodology is difficult to quantify, but it clearly is substantial. GEH's competitive advantage will be lost if its competitors are able to use the results of the GEH experience to normalize or verify their own process or if they are able to claim an equivalent understanding by demonstrating that they can arrive at the same or similar conclusions.
The value of this information to GEH would be lost if the information were disclosed to the public. Making such information available to competitors without their having been required to undertake a similar expenditure of resources would unfairly provide competitors with a windfall, and deprive GEH of the opportunity to exercise its competitive advantage to seek an adequate return on its large investment in developing and obtaining these very valuable analytical tools.
I declare under penalty of perjury that the foregoing affidavit and the matters stated therein are true and correct to the best of my knowledge, information, and belief.
Executed on this 20th day of February 2011.
Edward D. Schrull, PE Vice President, Regulatory Affairs Services Licensing GE-Hitachi Nuclear Energy Americas LLC 3901 Castle Hayne Rd.
Wilmington, NC 28401 edward.schrull@ge.com Affidavit for GEH-GGNS-AEP-427 Affidavit Page 3 of 3