View source for RS-11-055, Additional Information Related to License Amendment Request to Revise Technical Specifications Limiting Condition for Operation 3.7.6, Main Turbine Bypass System.

{{Adams
| number = ML111150091
| issue date = 04/22/2011
| title = Additional Information Related to License Amendment Request to Revise Technical Specifications Limiting Condition for Operation 3.7.6, Main Turbine Bypass System.
| author name = Hansen J
| author affiliation = Exelon Nuclear, Exelon Generation Co, LLC
| addressee name = 
| addressee affiliation = NRC/NRR, NRC/Document Control Desk
| docket = 05000461
| license number = NPF-062
| contact person = 
| case reference number = RS-11-055
| package number = ML111150088
| document type = Letter type:RS
| page count = 27
| project = 
| stage = Request
}}

=Text=
{{#Wiki_filter:555 Attachment 2, Enclosure 1, Contains Proprietary Information.
Withhold from Public Disclosure Under 10 CFR 2.390(a)(4).
When separated from Attachment 2, Enclosure 1, this cover letter is decontrolled.
RS-1 1-055 10 CFR 50.90 April 22, 2011 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Clinton Power Station, Unit 1 Facility Operating License No. NPF-62 NRC Docket No. 50-461

==Subject:==
Additional Information Related to License Amendment Request to Revise Technical Specifications Limiting Condition for Operation 3.7.6, "Main Turbine Bypass System"

==References:==
: 1. Letter from J. L. Hansen (Exelon Generation Company, LLC (EGC)) to U. S. NRC, "License Amendment Request to Revise Technical Specifications Limiting Condition for Operation 3.7.6, 'Main Turbine Bypass System,"' dated September 23, 2010
: 2. Letter from N. J. Di Francesco (U. S. NRC) to Mr. M. J. Pacilio (EGC),
                       "Clinton Power Station, Unit No.1 -Request for Additional Information Related to License Amendment Request to Revise Technical Specifications Limiting Condition for Operation 3.7.6, 'Main Turbine Bypass System' (TAC NO. ME4771)," dated March 24, 2011 In Reference 1, Exelon Generation Company, LLC (EGC) requested an amendment to Appendix A, Technical Specifications (TS), of Facility Operating License No. NPF-62 for Clinton Power Station, Unit 1 (CPS). The proposed amendment would modify the CPS TS Limiting Condition for Operation (LCO) 3.7.6, "Main Turbine Bypass System," by allowing revision of the reactor operational limits, as specified in the CPS Core Operating Limits Report (COLR), to compensate for the inoperability of the Main Turbine Bypass System (MTBS). In Reference 2, the NRC requested that EGC provide additional information in support of their review of Reference 1. The information requested in Reference 2 is provided in Attachment 1 to this letter.
Attachment 2, Enclosure 1, Contains Proprietary Information.
Withhold from Public Disclosure Under 10 CFR 2.390(a)(4).
When separated from Attachment 2, Enclosure 1, this cover letter is decontrolled.

April 22, 2011 U. S. Nuclear Regulatory Commission Page 2  of Attachment 2 contains proprietary information as defined by 10 CFR 2.390.
Global Nuclear Fuel (GNF), as the owner of the proprietary information, has executed an affidavit (i.e., Enclosure 3 of Attachment 2) which identifies that the enclosed proprietary information has been handled and classified as proprietary, is customarily held in confidence, and has been withheld from public disclosure. The proprietary information was provided to Exelon Nuclear in a GNF transmittal that is referenced by the affidavit. The proprietary information has been faithfully reproduced in the enclosed RAI responses such that the affidavit remains applicable. EGC hereby requests that the enclosed proprietary information be withheld from public disclosure in accordance with the provisions of 10 CFR 2.390 and 9.17. A non-proprietary version of the RAI responses also is provided as Enclosure 2 of Attachment 2. This submittal is subdivided as follows:
* Attachment 1 provides a response to the NRC request for additional information
* Attachment 2 provides the Global Nuclear Fuel response to NRC Request Nos. 1.a through 1.c and contains three enclosures:
: 1. Response to NRC RAI 1 for Clinton Power Station TBVOOS Submittal - GNF Proprietary Information - Class III (Confidential)
: 2. Response to NRC RAI 1 for Clinton Power Station TBVOOS Submittal -
Non-Proprietary Information - Class I (Public)
: 3. Affidavit for Enclosure 1 The information provided in this letter does not affect the No Significant Hazards Consideration, or the Environmental Consideration provided in Attachment 1 of the original license amendment request as described in the Reference 1 submittal.
In accordance with 10 CFR 50.91(b), "State consultation," EGC is providing the State of Illinois with a copy of this letter and its attachment to the designated State Official.
This letter contains no new regulatory commitments. If you have any questions concerning this letter, please contact Mr. Mitchel A. Mathews at (630) 657-2819.
I declare under penalty of perjury that the foregoing is true and correct. Executed on the 22nd day of April, 2011.
Jeffrey LC.Var en Manager Licensing Exelon Generation Company, LLC Attachments:
: 1. Additional Information Related to License Amendment Request to Revise Technical Specifications Limiting Condition for Operation 3.7.6, "Main Turbine Bypass System"
: 2. Global Nuclear Fuel (GNF) Response to NRC RAI No. 1 for Clinton Power Station (CPS)
Turbine Bypass Valve Out-of-Service (TBVOOS)

Attachment I Additional Information Related to License Amendment Request to Revise Technical Specifications Limiting Condition for Operation 3.7.6, "Main Turbine Bypass System" In reviewing the Exelon Generation Company's (Exelon's) submittal dated September 23, 2010, related to a request to revise Technical Specifications [TS] Limiting Condition for Operation 3.7.6, "Main Turbine Bypass System [MTBS]". for the Clinton Power Station (Clinton), Unit No. 1, the Nuclear Regulatory Commission (NRC) staff has determined that the following information is needed in order to complete its review:
NRC Request No. 1. Section 5 of Attachment 5 to the licensee's letter of September 23, 2010, discusses the results of the analysis of the following events:
(1) the feedwater controller failure, (2) rod withdrawal error, (3) loss of feedwater heating, and (4) slow recirculation flow increase.
NRC Request No. 1.a.      Discuss the methods and computer codes used in the analysis for each of the above referred events, verify that the methods and codes used were previously approved by the NRC, and address compliance with restrictions and conditions specified in the NRC safety evaluation report approving the methods and codes.
Exelon Generation Company, LLC (EGC) Response to NRC Request No. 1.a Feedwater Controller Failure The Feedwater Controller Failure (FWCF) is performed with what is commonly called the ODYN methodology. ODYN is the key transient analysis code for dynamic fast transient events. The set of codes and their approval basis is described in Table 1 below. The set of codes (PANACEA, ISCOR, ODYN, and TASC) applied for this methodology is also known as GEMINI methods. Refer to the response for RAI 1.b for a description of the interfaces among these codes.
Page 1 of 6

Attachment 1 Additional Information Related to License Amendment Request to Revise Technical Specifications Limiting Condition for Operation 3.7.6, "Main Turbine Bypass System" Table 1: NRC Approved Codes and References for Feedwater Controller Failure NRC- Approved Codes           References Letter from S. Richards (NRC) to G. Watford (GE), "Amendment 26 to GE Licensing Topical Report NEDE-2401 1-P-A,'GESTAR II' -
PANACEA Implementing Improved GE Steady-State Methods (TAC No.
MA6481)," MFN 99-035, November 10, 1999 The NRC Safety Evaluation supporting approval of NEDE-2401 1 P Revision 0 by the May 12, 1978 letter from D. G. Eisenhut (NRC) to R. Gridley (GE) finds the models and methods acceptable, and ISCOR                         mentions the use of a digital computer code. The referenced digital computer code is ISCOR. The use of ISCOR to provide core thermal-hydraulic information in transient applications is consistent with the approved models and methods.
Licensing Topical Report, "Qualification of the One-Dimensional Core Transient Model (ODYN) for Boiling Water Reactors," NEDO-ODYN 24154-A, Volumes 1 - 3, August 1986, NEDC-24154P-A Supplement 1, Volume 4, February 2000 "TASC-03A: A Computer Program for Transient Analysis of a Single TASC Channel," NEDC-32084P-A, Revision 2, July 2002 The following provides the limitations from the NRC Safety Evaluation for the latest approval of ODYN (NEDC-24154P-A Supplement 1, Volume 4, February 2000):
: 1. The downcomer level must remain above the jet pump suction, and no prolonged level in the active channel is allowed.
: 2. The duration of the simulation after the upper plenum subcools should be limited.
: 3. The mass in the separators should not remain zero and, therefore, the code is restricted to applications where the water level remains at or above the top of active fuel plus 5 feet.
: 4. The code is not presently qualified to perform stability calculations.
: 5. No lower plenum voiding is allowed.
All of these code limitations are met for the FWCF transient. Although there were no specific limitations identified for the other codes for Anticipated Operational Occurrence (AOO) application, the use of the codes is within the capabilities of PANAC, ODYN, ISCOR and TASC.
Page 2 of 6

Attachment 1 Additional Information Related to License Amendment Request to Revise Technical Specifications Limiting Condition for Operation 3.7.6, "Main Turbine Bypass System" Rod Withdrawal Error and Loss of Feedwater Heating The Rod Withdrawal Error and Loss of Feedwater Heating events are performed with the PANACEA three dimensional core simulator code. The PANACEA code and its approval basis is described in the Table 2 below.
Table 2: NRC Approved Codes and References for Rod Withdrawal Error and Loss of Feedwater Heatin NRC- Approved Reference Codes Letter from S. Richards (NRC) to G. Watford (GE), "Amendment 26 to GE Licensing Topical Report NEDE-2401 1 -P-A, 'GESTAR II'-
PANACEA Implementing Improved GE Steady State Methods (TAC No.
MA6481)", MFN-99-035, November 10, 1999 There are no specific limitations applied to PANACEA as a result of the NRC Safety Evaluation.
These events are performed within the capabilities of the PANACEA code.
Slow Flow Runout (i.e., Slow Recirculation Flow Increase)
The slow flow runout event is analyzed with the PANACEA three dimensional core simulator code and the ISCOR code. The PANACEA and ISCOR codes and their approval basis are described in the Table 3 below.
Table 3: NRC Approved Codes and References for Slow Recirculation Flow Increase NRC- Approved References Codes Letter from S. Richards (NRC) to G. Watford (GE), "Amendment 26 to GE Licensing Topical Report NEDE-2401 1 -P-A, 'GESTAR II' -
PANACEA Implementing Improved GE Steady State Methods (TAC No.
MA6481)", MFN-99-035, November 10, 1999 The NRC Safety Evaluation supporting approval of NEDE-2401 1 P Revision 0 by the May 12, 1978 letter from D. G. Eisenhut (NRC) to R. Gridley (GE) finds the models and methods acceptable, and ISCOR                    mentions the use of a digital computer code. The referenced digital computer code is ISCOR. The use of ISCOR to provide core thermal-hydraulic information in transient applications is consistent with the approved models and methods.
There are no specific limitations applied to PANACEA or ISCOR as a result of the NRC Safety Evaluation. These events are performed within the capabilities of the PANACEA and ISCOR codes.
Page 3 of 6

Attachment 1 Additional Information Related to License Amendment Request to Revise Technical Specifications Limiting Condition for Operation 3.7.6, "Main Turbine Bypass System" NRC Request No. 1.b.       Provide information to show the interfaces (inputs and outputs) of PANAC, ISCOR, ODYN, and TASC used in the calculation of the MCPR and LHGR for the analysis of the feedwater controllers failure events.
Exelon Generation Company, LLC (EGC) Response to NRC Request No. 1.b The PANACEA, ISCOR, ODYN, and TASC transient analysis methodology as it is used in the boiling water reactor (BWR) safety analysis process is shown in Figure 1-1 of "TASC-03A: A Computer Program for Transient Analysis of a Single Channel," NEDC-32084P-A, Revision 2, July 2002. The transient analysis methodology is used to perform the required transient analyses which result in establishing the operating limit minimum critical power ratio (OLMCPR),
demonstrating conformance to the reactor dome pressure vessel safety limit, and satisfying the fuel rod plastic strain and centerline melt limits. The primary parts of the transient analysis process include: (1) the lattice nuclear design methodology (TGBLA); (2) the three-dimensional BWR simulator (PANACEA); (3) the one-dimensional transient analysis model (ODYN); (4) the steady-state hydraulics and hot channel analysis methodology (ISCOR); (5) the transient critical power calculational methodology (TASC); (6) the fuel rod thermal-mechanical design methodology (GSTRM); and (7) the critical quality - boiling length correlation (GEXL).
The transient analysis process begins with the use of the lattice physics methods to develop the two-dimensional nuclear libraries which are required as input to the three-dimensional BWR simulator. To perform the required analyses, the lattice physics methods require a fuel assembly description and data from the cross section library. The lattice physics methods also provide the local peaking patterns used to determine the R-factors for use in the GEXL correlation.
The three-dimensional simulator (PANACEA) is used to define the core state and three-dimensional nuclear parameters used as input to the one-dimensional transient analysis model.
In addition to the inputs from the lattice physics methods, the three-dimensional simulator requires the reference core loading pattern, core operating state as inputs, and the steady-state thermal-hydraulic loss coefficients. These loss coefficients are developed using the steady-state thermal-hydraulics methodology (ISCOR) and are derived from fuel assembly specific pressure drop data as a function of power and flow. With the GEXL correlation as input, the three-dimensional simulator is used to predict the anticipated MCPR throughout the operating cycle and can also be used in the analysis of slow transients to determine the change in critical power ratio (OCPR) for these events.
ODYN is used to determine the peak transient pressure, the transient change in power, and the transient heat flux and thermal-hydraulic parameter changes required as input to both the hot channel analysis and to the transient critical power methodology (TASC). The transient peak pressure is used to demonstrate conformance to the reactor pressure vessel safety limit, which is based on the reactor pressure vessel design pressure. The transient change in power is used to demonstrate conformance to the fuel rod plastic strain and centerline melt limits.
Compliance with these limits confirms the acceptability of the LHGR limits. The fuel rod plastic strain and centerline melt limits are developed from the fuel rod design analyses using the fuel rod thermal - mechanical methodology and are based on the fuel physical parameters inputs.
Page 4 of 6

Attachment I Additional Information Related to License Amendment Request to Revise Technical Specifications Limiting Condition for Operation 3.7.6, "Main Turbine Bypass System" The one-dimensional transient analysis methodology collapses the three-dimensional nuclear parameters from 3D PANACEA state point to one dimension for use in the ODYN 1 D kinetics model and requires the plant configuration and performance parameters as inputs through the transient base deck.
The hot channel analysis (TASC) is performed by using the predicted flow distribution in the core during the transient to establish the flow to the limiting channels of each type in the core to be analyzed using the transient critical power methodology. The hot channel analysis is based on the inputs for the core-wide transient parameter changes provided by the one-dimensional transient analysis model (ODYN). The hot channel analysis requires the GEXL correlation, axial surface heat flux shape, local peaking factors, R-factor, initial critical power ratio (ICPR), and number of fuel assemblies of each type in the core and assembly description for each fuel type as input. The transient critical power calculational methodology is used to calculate the BCPR from the ICPR assumed as an initial condition for the transient being evaluated. This defines the minimum CPR during the transient. The transient critical power calculational methodology requires the GEXL correlation, hot channel hydraulic description, and the ICPR as input.
NRC Request No. 1.c.       Discuss the plant parameters considered in the analysis, identify the major input initial conditions and the worst single failure used in the analysis, discuss the bases used to select the numerical parameters and demonstrate that the numerical values with consideration of the uncertainties and fluctuations around the nominal values are conservative, resulting in minimum margins to the Minimum Critical Power Ratio (MCPR) and Linear Heat Generation Rate (LHGR) safety limits.
Exelon Generation Company, LLC (EGC) Response to NRC Request No. 1.c The response to NRC Request No. 1.c includes Global Nuclear Fuel Proprietary Information.
Therefore, the response to NRC Request No. 1.c is included in Enclosure 1 of Attachment 2, beginning on Page 7 of 11. A non-proprietary version of the response is also included in  of Attachment 2.
NRC Request No. 2. NRC Generic Letter (GL) 88- 16, "Removal of Cycle-Specific Parameter Limits From Technical Specifications," dated October 3, 1988, allows licensees to include the cycle specific parameters (i.e.,
MCPR, LHGR, and APLHGR etc.) in the core operating limits report (COLR), provided the changes in the parameters are determined using an NRC-approved methodology and consistent with all applicable limits of the analysis of record. Please affirm that the COLR will provide the limits and penalties to compensate for the inoperability of the MTBS and included the NRC-approved topical reports with applicable revision number.
Page 5 of 6

Attachment I Additional Information Related to License Amendment Request to Revise Technical Specifications Limiting Condition for Operation 3.7.6, "Main Turbine Bypass System" Exelon Generation Company. LLC (EGC) Response to NRC Request No. 2 As discussed in the September 23, 2010, Exelon Generation Company, LLC (EGC) request (i.e., Reference), all computer codes utilized to determine the limits for the main turbine bypass system out of service fall under the General Electric Standard Application for Reactor Fuel (GESTAR) (i.e., NEDE-2401 1-P-A) by reference. The GESTAR has been previously reviewed and approved for use by the NRC. Revision 16 of the GESTAR is currently the NRC-approved methodology referenced in the Clinton Power Station (CPS) Core Operating Limits Report (COLR). Moreover, EGC is obligated to provide the cycle-specific operating limits in accordance with Technical Specifications (TS) Section 5.6.5, "Core Operating Limits Report (COLR)." These limits will include the new operating limits for the main turbine bypass system out-of-service as discussed in the Reference markups of TS Sections 3.7.6, "Main Turbine Bypass," and TS 5.6.5. As discussed in the EGC response to NRC Request 1.a above, NRC-approved methodology will be used to determine the limits, and the applicable analyses will meet the limits of the analysis of record.
Lastly, should the September 23, 2010, proposed changes receive NRC approval, the References section of the CPS COLR will be updated to contain a reference to the Global Nuclear Fuel analysis that supports operation with the main turbine bypass system out of service.

==Reference:==

Letter from J. L. Hansen (Exelon Generation Company, LLC (EGC)) to U. S. NRC, "License Amendment Request to Revise Technical Specifications Limiting Condition for Operation 3.7.6,
'Main Turbine Bypass System,"' dated September 23, 2010 Page 6 of 6

Attachment 2, Enclosure 1, Contains Proprietary Information.
Withhold from Public Disclosure Under 10 CFR 2.390(a)(4).
When separated from Attachment 2, Enclosure 1, this cover letter is decontrolled.
Attachment 2 Global Nuclear Fuel CFL-EXN -LH 1-11-021 GNF Response to NRC RAI 1 for Clinton Power Station (CPS)
Turbine Bypass Valve Out -of-Service (TBVOOS) dated April 19, 2011 Attachment 2, Enclosure 1, Contains Proprietary Information.
Withhold from Public Disclosure Under 10 CFR 2.390(a)(4).
When separated from Attachment 2, Enclosure 1, this cover letter is decontrolled.

Proprietary Notice This letter transmits proprietary information in accordance with 10CFR2.390. Upon the removal of Enclosure 1 the balance of the letter may be considered non-proprietary.
Global Nuclear Fuel A Joie Venn, of GE, Toshiba, & Hitachi Charles F. Lamb                                                 Global Nuclear Fuel - Americas, LLC Customer Project Manager                                       Castle Hayne Road, Wilmington, NC 28402 (910) 819-5613, charlesf.lamb@ge.com CFL-EXN-LH1-11-021 April 19, 2011 Hossein Youssefnia Exelon Nuclear 4300 Winfield Way Warrenville, IL 60555

==Subject:==
GNF Response to NRC RAI 1 for Clinton Power Station (CPS) Turbine Bypass Valve Out-of-Service (TBVOOS)

==References:==
: 1. "Contract between Amergen and Global Nuclear Fuel for the supply of Nuclear Fuel and Related Components and Services for Unit 1 of the Clinton Power Station,"
January 12, 2001, as amended ("Fuel Contract").

==Dear Mr. Youssefnia:==

This letter transmits the Global Nuclear Fuel (GNF) response to the Nuclear Regulatory Commission (NRC) Requests for Additional Information (RAIs). contains the GNF response to RAI 1. Please note that Enclosure 1 contains proprietary information of the type that GNF maintains in confidence and withholds from public disclosure. The information has been handled and classified as proprietary to GNF as indicated in the enclosed affidavit. The affidavit contained in Enclosure 3 identifies that the information contained in Enclosure 1 has been handled and classified as proprietary to GNF. GNF hereby requests that the information in Enclosure 1 be withheld from public disclosure in accordance with the provisions of 10 CFR 2.390 and 9.17. Enclosure 2 is a non-proprietary version of .
GNF requests that any transmittal of this proprietary information to the NRC be accompanied by the enclosed affidavit and proprietary notice. In order to maintain the applicability of the affidavit and to meet the requirements of 10 CFR 2.390, the transmittal to the NRC should:
: 1) Faithfully reproduce the proprietary information,
: 2) Preserve the proprietary annotations, and
: 3) Include the words similar to "GNF Proprietary Information" at the top of first page and each page containing the proprietary information.

Further, 10 CFR 2.390 requires that the proprietary information be incorporated, as far as possible, into separate paper.         Therefore, Enclosure 1 hereto contains the proprietary information, and the non-proprietary and redacted information is provided in Enclosure 2.
Based on past discussions with the NRC, GNF has been encouraged to request its customers to provide a paragraph similar to the following in the customer letters transmitting proprietary information to the NRC in order to clearly indicate the proprietary nature of the information and to document the source of the proprietary information as indicated in the GNF affidavit.
     "The enclosed RAI responses contain proprietary information as defined by 10 CFR 2.390.
GNF, as the owner of the proprietary information, has executed the enclosed affidavit, which identify that the enclosed proprietary information has been handled and classified as proprietary, is customarily held in confidence, and has been withheld from public disclosure.
The proprietary information was provided to Exelon Nuclear in a GNF transmittal that is referenced by the affidavit. The proprietary information has been faithfully reproduced in the enclosed RAI responses such that the affidavit remains applicable. GNF hereby requests that the enclosed proprietary information be withheld from public disclosure in accordance with the provisions of 10 CFR 2.390 and 9.17. A non-proprietary version of the RAI responses also is provided."
The verified inputs associated with these RAI responses are included eDRF Section 0000-0132-4798 R0.           A signed copy of this letter is included in eDRF Section 0000-0132-4799 R0. If you have any questions, please do not hesitate to contact me.
Sincerely, Charles F. Lamb Customer Project Manager Enclosures
: 1. Response to NRC RAI 1 for Clinton Power Station TBVOOS Submittal - GNF Proprietary Information - Class III (Confidential)
: 2. Response to NRC RAI 1 for Clinton Power Station TBVOOS Submittal - Non-Proprietary Information - Class I (Public)
: 3. Affidavit for Enclosure 1

ENCLOSURE 2 CFL-EXN-LH 1-11-021 Response to NRC RAI 1 for Clinton Power Station TBVOOS Submittal Non-Proprietary Information-Class I (Public)
INFORMATION NOTICE This is a non-proprietary version of CFL-EXN-LH1-11-021        Enclosure 1, which has the proprietary information removed. Portions of the document that have been removed are indicated by white space inside open and closed bracket as shown here ((     )).

CFL-EXN-LH1-11-021                                    Non-Proprietary Information - Class I (Public)                                                                              Page 1 of 11 REQUEST FOR ADDITIONAL INFORMATION REGARDING REVISIONS TO TECHNICAL SPECIFICATIONS FOR MAIN TURBINE BYPASS SYSTEM CLINTON POWER STATION, UNIT NO. 1 DOCKET NO. 50-461 In reviewing the Exelon Generation Company's (Exelon's) submittal dated September 23, 2010, related to a request to revise Technical Specifications [TS] Limiting Condition for Operation 3.7.6, "Main Turbine Bypass System [MTBS]", for the Clinton Power Station (Clinton),
Unit No.1, the Nuclear Regulatory Commission (NRC) staff has determined that the following information is needed in order to complete its review:
RAI 1: Section 5 of Attachment 5 to the licensee's letter of September 23, 2010, discusses the results of the analysis of the following events: (1) the feedwater controller failure, (2) rod withdrawal error, (3) loss of feedwater heating, and (4) slow recirculation flow increase.
: a. Discuss the methods and computer codes used in the analysis for each of the above referred events, verify that the methods and codes used were previously approved by the NRC, and address compliance with restrictions and conditions specified in the NRC safety evaluation report approving the methods and codes.
: b. Provide information to show the interfaces (inputs and outputs) of PANAC, ISCOR, ODYN, and TASC used in the calculation of the MCPR and LHGR for the analysis of the feedwater controllers failure events.
: c. Discuss the plant parameters considered in the analysis, identify the major input initial conditions and the worst single-failure used in the analysis, discuss the bases used to select the numerical parameters and demonstrate that the numerical values with consideration of the uncertainties and fluctuations around the nominal values are conservative, resulting in minimum margins to the Minimum Critical Power Ratio (MCPR) and Linear Heat Generation Rate (LHGR) safety limits.

CFL-EXN-LH1-11-021                                  Non-Proprietary Information - Class I (Public)
Enclosure 2                                                                          Page 2 of 11 RESPONSE TO RAI 1.a:
Feedwater Controller Failure The Feedwater Controller Failure (FWCF) event is performed with what is commonly called the ODYN methodology. ODYN is the key transient analysis code for dynamic fast transient events.
The set of codes and their approval basis is described in the table below. The set of codes (PANACEA, ISCOR, ODYN, and TASC) applied for this methodology is also known as GEMINI methods. Refer to the response for RAI 1. b for a description of the interfaces among these codes.
NRC-Approved Codes          Reference PANACEA                     S. Richards (NRC) to G. Watford (GE), "Amendment 26 to GE Licensing Topical Report NEDE-24011-P-A, 'GESTAR II' -
Implementing Improved GE Steady-State Methods (TAC No.
MA6481)," MFN 99-035, November 10, 1999.
ISCOR                       The NRC Safety Evaluation supporting approval of NEDE-2401 1 P Revision 0 by the May 12, 1978 letter from D. G. Eisenhut (NRC) to R. Gridley (GE) finds the models and methods acceptable, and mentions the use of a digital computer code. The referenced digital computer code is ISCOR. The use of ISCOR to provide core thermal-hydraulic information in transient applications is consistent with the approved models and methods.
ODYN                        Licensing Topical Report, "Qualification of the One-Dimensional Core Transient Model (ODYN) for Boiling Water Reactors,"
NEDO-24154-A, Volumes 1 - 3, August 1986, NEDC-24154P-A Supplement 1, Volume 4, February 2000.
TASC                        "TASC-03A: A Computer Program for Transient Analysis of a Single Channel," NEDC-32084P-A, Revision 2, July 2002.
The following provides the limitations from the NRC Safety Evaluation for the latest approval of ODYN (NEDC-24154P-A Supplement 1, Volume 4, February 2000):
: 1. The downcomer level must remain above the jet pump suction, and no prolonged level in the active channel is allowed.
: 2. The duration of the simulation after the upper plenum subcools should be limited.

CFL-EXN-LH1-11-021                                  Non-Proprietary Information - Class I (Public)                                                                           Page 3 of 11
: 3. The mass in the separators should not remain zero and, therefore, the code is restricted to applications where the water level remains at or above the top of active fuel plus 5 feet.
: 4. The code is not presently qualified to perform stability calculations.
: 5. No lower plenum voiding is allowed.
All of these code limitations are met for the FWCF transient. Although there were no specific limitations identified for the other codes for Anticipated Operational Occurrence (AOO) application, the use of the codes is within the capabilities of PANACEA, ODYN, ISCOR and TASC.
Rod Withdrawal Error and Loss of Feedwater Heating The Rod Withdrawal Error and Loss of Feedwater Heating events are performed with the PANACEA three dimensional core simulator code. The PANACEA code and its approval basis are described in the table below.
NRC-Approved Codes          Reference PANACEA                    S. Richards (NRC) to G. Watford (GE), "Amendment 26 to GE Licensing Topical Report NEDE-24011-P-A, 'GESTAR II' -
Implementing Improved GE Steady-State Methods (TAC No.
MA6481)," MFN 99-035, November 10, 1999.
There are no specific limitations applied to PANACEA as a result of the NRC Safety Evaluation.
These events are performed within the capabilities of the PANACEA code.
Slow Flow Runout The slow flow runout event is analyzed with the PANACEA three dimensional core simulator code and the ISCOR code. The PANACEA and ISCOR codes and their approval bases are described in the table below.

CFL-EXN-LH1-11-021                                  Non-Proprietary Information - Class I (Public)                                                                           Page 4 of 11 NRC-Approved Codes          Reference PANACEA                     S. Richards (NRC) to G. Watford (GE), "Amendment 26 to GE Licensing Topical Report NEDE-24011-P-A, 'GESTAR II' -
Implementing Improved GE Steady-State Methods (TAC No.
MA6481)," MFN 99-035, November 10, 1999.
ISCOR                       The NRC Safety Evaluation supporting approval of NEDE-24011P Revision 0 by the May 12, 1978 letter from D. G. Eisenhut (NRC) to R. Gridley (GE) finds the models and methods acceptable, and mentions the use of a digital computer code. The referenced digital computer code is ISCOR. The use of ISCOR to provide core thermal-hydraulic information in transient applications is consistent with the approved models and methods.
There are no specific limitations applied to PANACEA or ISCOR as a result of the NRC Safety Evaluation. These events are performed within the capabilities of the PANACEA and ISCOR codes.

CFL-EXN-LH1-11-021                                   Non-Proprietary Information - Class I (Public)
Enclosure 2                                                                           Page 5 of 11 RESPONSE TO RAI 1.b:
The PANACEA, ISCOR, ODYN, and TASC transient analysis methodology as it is used in the boiling water reactor (BWR) safety analysis process is shown in Figure 1-1 of "TASC-03A A Computer Program for Transient Analysis of a Single Channel," NEDC-32084P-A, Revision 2, July 2002. The transient analysis methodology is used to perform the required transient analyses which result in establishing the operating limit minimum critical power ratio (OLMCPR),
demonstrating conformance to the reactor dome pressure vessel safety limit, and satisfying the fuel rod plastic strain and centerline melt limits. The primary parts of the transient analysis process include: (1) the lattice nuclear design methodology (TGBLA); (2) the three-dimensional BWR simulator (PANACEA); (3) the one-dimensional transient analysis model (ODYN); (4) the steady-state hydraulics and hot channel analysis methodology (ISCOR); (5) the transient critical power calculational methodology (TASC); (6) the fuel rod thermal-mechanical design methodology (GSTRM); and (7) the critical quality - boiling length correlation (GEXL).
The transient analysis process begins with the use of the lattice physics methods to develop the two-dimensional nuclear libraries which are required as input to the three-dimensional BWR simulator. To perform the required analyses, the lattice physics methods require a fuel assembly description and data from the cross section library. The lattice physics methods also provide the local peaking patterns used to determine the R-factors for use in the GEXL correlation.
The three-dimensional simulator (PANACEA) is used to define the core state and three-dimensional nuclear parameters used as input to the one-dimensional transient analysis model. In addition to the inputs from the lattice physics methods, the three-dimensional simulator requires the reference core loading pattern, core operating state as inputs, and the steady-state thermal-hydraulic loss coefficients. These loss coefficients are developed using the steady-state thermal-hydraulics methodology (ISCOR) and are derived from fuel assembly specific pressure drop data as a function of power and flow. With the GEXL correlation as input, the three-dimensional simulator is used to predict the anticipated MCPR throughout the operating cycle and can also be used in the analysis of slow transients to determine the change in critical power ratio (ACPR) for these events.
ODYN is used to determine the peak transient pressure, the transient change in power, and the transient heat flux and thermal -hydraulic parameter changes required as input to both the hot channel analysis and to the transient critical power methodology (TASC). The transient peak pressure is used to demonstrate conformance to the reactor pressure vessel safety limit, which is based on the reactor pressure vessel design pressure. The transient change in power is used to demonstrate conformance to the fuel rod plastic strain and centerline melt limits.
Compliance with these limits confirms the acceptability of the LHGR limits. The fuel rod plastic strain and centerline melt limits are developed from the fuel rod design analyses using the fuel rod thermal -mechanical methodology and are based on the fuel physical parameters inputs.
The one-dimensional transient analysis methodology collapses the three-dimensional nuclear

CFL-EXN-LH1-11-021                                    Non-Proprietary Information - Class I (Public)                                                                             Page 6 of 11 parameters from 3D PANACEA state point to one dimension for use in the ODYN 1D kinetics model and requires the plant configuration and performance parameters as inputs through the transient base deck.
The hot channel analysis (TASC) is performed by using the predicted flow distribution in the core during the transient to establish the flow to the limiting channels of each type in the core to be analyzed using the transient critical power methodology. The hot channel analysis is based on the inputs for the core-wide transient parameter changes provided by the one-dimensional transient analysis model (ODYN). The hot channel analysis requires the GEXL correlation, axial surface heat flux shape, local peaking factors, R-factor, initial critical power ratio (ICPR), and number of fuel assemblies of each type in the core and assembly description for each fuel type as input. The transient critical power calculational methodology is used to calculate the OCPR from the ICPR assumed as an initial condition for the transient being evaluated. This defines the minimum CPR during the transient. The transient critical power calculational methodology requires the GEXL correlation, hot channel hydraulic description, and the ICPR as input.

CFL-EXN-LH1-11-021                                   Non-Proprietary Information - Class I (Public)                                                                             Page 7 of 11 RESPONSE TO RAI 1.c:
Feedwater Controller Failure The tables below describe the plant parameter and initial condition considerations for the feedwater controller failure event. The plant parameters and initial conditions are not revised or changed for the Turbine Bypass Valve Out-of-Service (TBVOOS) analyses. In other words, the only difference in the analysis for the TBVOOS is the designated complete failure of the turbine bypass system. There are no changes to single failure assumptions. As described in the UFSAR, due to redundancy in the trip system, single failures are not expected to result in a more severe event than analyzed.
Plant Parameter                              Considerations Maximum Feedwater Runout Flow                The maximum feedwater (FW) runout capacity the plant can attain is used. ((
                                                                                  )) Maximizing the runout flow increases the subcooling and power increase.
High Level Scram Setpoint                    The maximum analytical limit is used for conservatism. The maximum value extends the period of increased subcooling.
Turbine Bypass Opening Performance           This event results in a high level turbine trip resulting in turbine stop valve closure. The turbine bypass opening characteristics are important because the opening characteristics (opening time and capacity) can affect the pressurization of the reactor. For this analysis with the turbine bypass system Out-of-Service (OOS), there is no credit for any bypass flow.
Turbine Stop Valve Closure Time              The fastest turbine stop valve closure time is specified for the analysis for conservatism.
Scram Speed                                  Limits are provided for both the Option B (measured performance basis) and Option A (tech spec scram speed) scram speeds. ((
                                                                       ))

CFL-EXN-LH 1-11-021              Non-Proprietary Information - Class I (Public)                                                         Page 8 of 11 Initial Conditions      Considerations Power Level             Initial power levels are considered from rated power down to the minimum power level for thermal limits monitoring. ((
1]
Core Flow              The maximum and minimum licensed core flow at rated power are considered in the analysis.
Feedwater Temperature  If the plant is operating with reduced feedwater temperature due to feedwater heater out of service or end of cycle feedwater temperature reduction, the amount of subcooling will increase resulting in a larger power increase. Both normal and reduced feedwater temperatures are considered in the analysis.
Water Level            The water level change from the initial to the high level scram setpoint affects the duration of the subcooling portion of the feedwater controller failure. The initial level is set at the low level alarm as the initial condition to maximize inlet subcooling for conservatism.
Core Exposure          All rods out End-of-Cycle (EOC) exposure used as this is a conservative state point for scram reactivity.

CFL-EXN-LH1-11-021                                   Non-Proprietary Information - Class I (Public)                                                                             Page 9 of 11 Loss of Feedwater Heating The tables below describe the plant parameter and initial condition considerations for the loss of feedwater heating event. There are no changes to single failure assumptions. As described in the UFSAR, due to redundancy in the trip system, single failures are not expected to result in a more severe event than analyzed.
Plant Parameter                     Considerations Loss of Feedwater Heating AT       The change in power and subcooling is primarily driven by this AT. A maximum value is selected to maximize the power increase for conservatism.
Thermal Power Scram Setpoint        The increase in subcooling from the feedwater temperature decrease results in core power levels reaching the high thermal power scram. This scram is not credited in the analysis for conservatism, which maximizes the ACPR and LHGR response.
Initial Conditions                  Considerations Power Level / Core Flow             The licensed power and minimum licensed core flow at rated power are used in the analysis. This initial core flow condition sets a higher initial void fraction to maximize the void reactivity and power increase.
Feedwater Temperature              The rated initial condition is used as this condition has the highest level of FW heating and would have the largest temperature drop.
Core Exposure                       Multiple core exposures are evaluated to capture the effect the core exposure has on reactivity feedback.

CFL-EXN-LH1-11-021                                  Non-Proprietary Information - Class I (Public)                                                                          Page 10 of 11 Rod Withdrawal Error The tables below describe the plant parameter and initial condition considerations for the rod withdrawal error event. There are no changes to single failure assumptions. As described in the UFSAR, due to redundancy in the rod control and information system, single failures are not expected to result in a more severe event than analyzed.
Plant Parameter                     Considerations Rod Withdrawal Limiter Setpoint     The rod withdrawal limiter setpoint dictates the distance a rod can withdraw. The system enforces 1 foot above 70%
power and 2 feet below 70% power consistent with the analysis basis.
Initial Conditions                  Considerations Rod Patterns                        The rod withdrawal error is primarily a local event, affecting the local conditions in the core near the selected error rod. ((
11 Core Exposure ll

CFL-EXN-LH1-11-021                                 Non-Proprietary Information - Class I (Public)                                                                         Page 11 of 11 Slow Flow Runout The tables below describe the plant parameter and initial condition considerations for the slow flow runout event. The slow flow runout is not a defined UFSAR event. ((
                                                                                 )) There are no changes to single failure assumptions.
Plant Parameter                    Considerations Flow Runout Slope                  A conservative ((                                          ))
is used in the MCPR slow flow runout calculations. The steady-state thermal hydraulics model ISCOR is used to perform the MCPR calculations. ((
                                                   ))
Maximum Flow Runout                Core flow is assumed to start from various offrated flow conditions with changes from the initial condition to the maximum capability for the plant. This maximizes the changes in core flow and power, and as a result maximizes OCPR and LHGR changes.
Initial Conditions                 Considerations Core Flow                          Various initial core flow rates are considered at conservative rod lines to determine the effect on the OCPR and change in LHGR from the flow increase to the maximum licensed core flow.
Core Exposure                      For the LHGR calculation, ((
11

ENCLOSURE3 CFL-EXN-LH 1-11-021 Affidavit

Global Nuclear Fuel - Americas AFFIDAVIT 7 Andrew A. Lingenfelter, state as follows:
(1) I am Vice President, Fuel Engineering, Global Nuclear Fuel - Americas, LLC ("GNF-A"),
and 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 Enclosure I of GNF's letter, CFL-EXN-LH1-11-021, C. Lamb (GNF-A) to H. Youssefnia (Exelon Nuclear), entitled "GNF Response to NRC RAI 1 for Clinton Power Station (CPS) Turbine Bypass Valve Out-of-Service (TBVOOS) TBVOOS Submittal," dated April 19, 2011.                        GNF-A proprietary information in Enclosure 1, which is entitled "Response to NRC RAI 1 for Clinton Power Station TBVOOS Submittal," is identified by a dotted underline inside double square brackets. ((Thus senitence is_ail.exammipke.'j')) A "((" marking at the beginning of a table, figure, or paragraph closed with a "))" marking at the end of the table, figure or paragraph is used to indicate that the entire content between the double brackets is proprietary. In each case, the superscript notation f3' refers to Paragraph (3) of this affidavit, which 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, GNF-A 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 qualify 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, 975F2d871 (DC Cir. 1992), and Public Citizen Health Research Group v. FDA, 704F2d1280 (DC Cir. 1983).
(4) Some examples of categories of information which 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 GNF-A's competitors without license from GNF-A constitutes a competitive economic advantage over other companies;
: b. Information which, if used by a competitor, would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing of a similar product;
: c. Information which reveals aspects of past, present, or future GNF-A customer-funded development plans and programs, resulting in potential products to GNF-A; CFL-EXN-LH 1-11-021                                                            Affidavit Page 1 of 3
: d. Information which discloses patentable subject matter for which it may be desirable to obtain patent protection.
The information sought to be withheld is considered to be proprietary for the reasons set forth in paragraphs (4)a. and (4)b. above.
(5)  To address 10 CFR 2.390 (b) (4), the information sought to be withheld is being submitted to NRC in confidence. The information is of a sort customarily held in confidence by GNF-A, 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 GNF-A, no public disclosure has been made, and it is not available in public sources. All disclosures to third parties including any required transmittals to NRC, have been made, or must be made, pursuant to regulatory provisions or proprietary agreements which provide for maintenance of the information in confidence. Its initial designation as proprietary information, and the subsequent steps taken to prevent its unauthorized disclosure, are as set forth in paragraphs (6) and (7) following.
(6) Initial approval of proprietary treatment of a document is made by the manager of the originating component, the person most likely to be acquainted with the value and sensitivity of the information in relation to industry knowledge, or subject to the terms under which it was licensed to GNF-A. Access to such documents within GNF-A is limited on 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, by the manager of the cognizant marketing function (or his delegate), and by the Legal Operation, for technical content, competitive effect, and determination of the accuracy of the proprietary designation. Disclosures outside GNF-A 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 agreements.
(8) The information identified in paragraph (2) is classified as proprietary because it contains details of GNF-A' s fuel design and licensing methodology. The development of this methodology, along with the testing, development and approval was achieved at a significant cost to GNF-A or its licensor.
The development of the fuel design and licensing methodology along with the interpretation and application of the analytical results is derived from an extensive experience database that constitutes a major GNF-A asset.
CFL-EXN-LH 1- 11-021                                                         Affidavit Page 2 of 3

(9)  Public disclosure of the information sought to be withheld is likely to cause substantial harm to GNF-A's competitive position and foreclose or reduce the availability of profit-making opportunities. The information is part of GNF-A'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.
The research, development, engineering, analytical, and NRC review costs comprise a substantial investment of time and money by GNF-A.
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.
GNF-A's competitive advantage will be lost if its competitors are able to use the results of the GNF-A 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 GNF-A 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 GNF-A 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 19th day of April 2011.
Andrew A. Lingenfelter Vice President, Fuel Engineering Global Nuclear Fuel - Americas, LLC CFL-EXN-LH1-11-021                                                               Affidavit Page 3 of 3}}