RS-11-178, Additional Information Supporting Request for License Amendment Regarding Measurement Uncertainly Recapture Power Uprate
| ML113430811 | |
| Person / Time | |
|---|---|
| Site: | Byron, Braidwood  |
| Issue date: | 12/09/2011 |
| From: | Borton K Exelon Generation Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML113430815 | List: |
| References | |
| RS-11-178 | |
| Download: ML113430811 (37) | |
Text
Ex!!lotl G-eneratiol1 4300 Winfield Road Wammvllle, 60555 wwwexeloncorp,tol11 RS-11-178 December 9, 2011 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Braidwood Station, Units 1 and 2 Facility Operating License Nos. NPF-72 and NPF-77 NRC Docket Nos. STN 50-456 and STN 50-457 Byron Station, Units 1 and 2 Facility Operating License Nos. NPF-37 and NPF-66 NRC Docket Nos. STN 50-454 and STN 50-455 Generation 10 CFR 2.390 10 CFR 50.90
Subject:
Additional Information Supporting Request for License Amendment Regarding Measurement Uncertainty Recapture Power Uprate
References:
- 1. Letter from Craig Lambert (Exelon Generation Company, LLC) to u. S. NRC, "Request for License Amendment Regarding Measurement Uncertainty Recapture Power Uprate," dated June 23, 2011
- 2. Letter from N. J. DiFrancesco (U. S. NRC) to M. J. Pacilio (Exelon Generation Company, LLC), "Braidwood Station, Units 1 and 2 and Byron Station, Unit Nos. 1 and 2 - Request for Additional Information RE: Measurement Uncertainty Power Uprate Request (TAC NOS. ME6587, ME6588, 6589, AND ME6590),"
dated November 28, 2011 In Reference 1, Exelon Generation Company, LLC (EGC) requested an amendment to Facility Operating License Nos. NPF-72, NPF-77, NPF-37 and NPF-66 for Braidwood Station, Units 1 and 2, and Byron Station, Units 1 and 2, respectively. Specifically, the proposed changes revise the Operating License and Technical Specifications to implement an increase in rated thermal power of approximately 1.63% based on increased feedwater flow measurement accuracy. In Reference 2, the NRC requested additional information to support review of the proposed changes. In response to this request, EGC is providing the attached information.
EGC has reviewed the information supporting a finding of no significant hazards consideration and the environmental consideration provided to the NRC in Reference 1. The additional information provided in this submittal does not affect the bases for concluding that the proposed license amendment does not involve a significant hazards consideration. In addition, the
December 9, 2011 U.S. Nuclear Regulatory Commission Page 2 additional information provided in this submittal does not affect the bases for concluding that neither an environmental impact statement nor an environmental assessment needs to be prepared in connection with the proposed amendment.
There are no regulatory commitments contained in this letter.
In accordance with 10 CFR 2.390, *Public inspections, exemptions, requests for withholding,"
EGC requests withholding of Attachment 1. Attachment 1 contains information considered proprietary by Westinghouse Electric Corporation, the owner of this information. An affidavit from Westinghouse Electric Corporation supporting this request is included in Attachment 2. A non-proprietary version of Attachment 1 is provided in Attachment 3.
Should you have any questions concerning this letter, please contact leslie E. Holden at (630) 657-3316.
I declare under penalty of perjury that the foregoing is true and correct. Executed on the 9th day of December 2011.
Respectfully, Kevin F. Borton Manager, licensing - Power Uprate Attachments: Response to Request for Additional Information (Proprietary Version) Affidavit from Westinghouse Electric Corporation supporting withholding of information in Attachment 1 Response to Request for Additional Information {Non-Proprietary Version} Byron and Braidwood Stations Generator Capability Curves NSAl-07-11, Decay Heat Assumption in Steam Generator Tube Rupture Margin-to-Overfill Analysis Methodology cc:
NRC Regional Administrator, Region III NRC Senior Resident Inspector - Braidwood Station NRC Senior Resident Inspector - Byron Station Illinois Emergency Management Agency - Division of Nuclear Safety
Braidwood and Byron Stations Measurement Uncertainty Recapture License Amendment Request (MUR LAR)
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RAI)
December 9, 2011 ATTACHMENT 2 Westinghouse Electric Corporation Affidavit Supporting Withholding
(8 Westinghouse U.S. Nuclear Regulatory Commission Document Control Desk 11555 Rockville Pike Rockville, MD 20852 Westinghouse Electric Company Nuclear Services 1000 Westinghouse Drive Cranberry Township, Pennsylvania 16066 USA Direct tel: (412) 374-4643 Direct fax: (724) 720-0754 e-mail: greshaja@westinghouse.com Proj letter: CAE-ll-MUR-61 CAW-1l-3324 December 7,2011 APPLICATION FOR WITHHOLDING PROPRIETARY INFORMATION FROM PUBLIC DISCLOSURE
Subject:
CAE-l1-MUR-61 P-Attachment, "Byron and Braidwood Units 1 and 2 - Response to NRC Informal Request for Additional Information (RAJ) from the Nuclear Performance and Code Review Branch Related to the Measurement Uncertainty Recapture Power Uprate (MUR)
License Amendment Request (TAC Nos. ME6587, ME6588, ME6589, and ME6590)"
(Proprietary)
The proprietary information for which withholding is being requested in the above-referenced report is further identified in Affidavit CAW-11-3324 signed by the owner of the proprietary information, Westinghouse Electric Company LLC. The affidavit, which accompanies this letter, sets forth the basis on which the information may be withheld from public disclosure by the Commission and addresses with specificity the considerations listed in paragraph (b)(4) of 10 CFR Section 2.390 of the Commission's regulations.
Accordingly, this letter authorizes the utilization of the accompanying affidavit by Exelon Nuclear.
Correspondence with respect to the proprietary aspects of the application for withholding or the Westinghouse affidavit should reference this letter, CA W-l1-3324, and should be addressed to J. A. Gresham, Manager, Regulatory Compliance, Westinghouse Electric Company LLC, Suite 428, 1000 Westinghouse Drive, Cranberry Township, Pennsylvania 16066.
Enclosures 1~~~
J. A. Gresham, M~ger Regulatory Compliance
CAW-1l-3324 AFFIDAVIT COMMONWEALTH OF PENNSYLVANIA:
ss COUNTY OF BUTLER:
Before me, the undersigned authority, personally appeared T. Rodack, who, being by me duly sworn according to law, deposes and says that he is authorized to execute this Affidavit on behalf of Westinghouse Electric Company LLC (Westinghouse), and that the averments of fact set forth in this Affidavit are true and correct to the best of his knowledge, information, and belief:
Sworn to and subscribed before me this 7th day of December 2011 COMMONWEALTH OF PENNSYLVANIA Notarial Seal Cynthia Olesky, Notary Public Manor Bora, Westmoreland County CommISSIOn ExpIreS July 16, 2014 j PI:!nnsvIvanIa AssocIation of Notartes T. Rodack, Director Quality and Licensing Programs
2 CAW-11-3324 (1)
I am Director, Quality and Licensing Programs, in Nuclear Fuel, Westinghouse Electric Company LLC (Westinghouse), and as such, I have been specifically delegated the function of reviewing the proprietary information sought to be withheld from public disclosure in connection with nuclear power plant licensing and rule making proceedings, and am authorized to apply for its withholding on behalf of Westinghouse.
(2)
I am making this Affidavit in conformance with the provisions of 10 CFR Section 2.390 of the Commission's regulations and in conjunction with the Westinghouse Application for Withholding Proprietary Information from Public Disclosure accompanying this Affidavit.
(3)
I have personal knowledge of the criteria and procedures utilized by Westinghouse in designating information as a trade secret, privileged or as confidential commercial or financial information.
(4)
Pursuant to the provisions of paragraph (bX4) of Section 2.390 of the Commission's regulations, the following is furnished for consideration by the Commission in determining whether the information sought to be withheld from public disclosure should be withheld.
(i)
The information sought to be withheld from public disclosure is owned and has been held in confidence by Westinghouse.
(ii)
The information is of a type customarily held in confidence by Westinghouse and not customarily disclosed to the public. Westinghouse has a rational basis for determining the types of information customarily held in confidence by it and, in that connection, utilizes a system to determine when and whether to hold certain types of information in confidence. The application of that system and the substance of that system constitutes Westinghouse policy and provides the rational basis required.
Under that system, information is held in confidence if it falls in one or more of several types, the release of which might result in the loss of an existing or potential competitive advantage, as follows:
( a)
The information reveals the distinguishing aspects of a process (or component, structure, tool, method, etc.) where prevention of its use by any of Westinghouse's competitors without license from Westinghouse constitutes a competitive economic advantage over other companies.
3 CAW-1l-3324 (b)
It consists of supporting data, including test data, relative to a process (or component, structure, tool, method, etc.), the application of which data secures a competitive economic advantage, e.g., by optimization or improved marketability.
(c)
Its use 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 a similar product.
(d)
It reveals cost or price information, production capacities, budget levels, or commercial strategies of Westinghouse, its customers or suppliers.
(e)
It reveals aspects of past, present, or future Westinghouse or customer funded development plans and programs of potential commercial value to Westinghouse.
(f)
It contains patentable ideas, for which patent protection may be desirable.
There are sound policy reasons behind the Westinghouse system which include the following:
(a)
The use of such information by Westinghouse gives Westinghouse a competitive advantage over its competitors. It is, therefore, withheld from disclosure to protect the Westinghouse competitive position.
(b)
It is information that is marketable in many ways. The extent to which such information is available to competitors diminishes the Westinghouse ability to sell products and services involving the use of the information.
(c)
Use by our competitor would put Westinghouse at a competitive disadvantage by reducing his expenditure of resources at our expense.
(d)
Each component of proprietary information pertinent to a particular competitive advantage is potentially as valuable as the total competitive advantage. If competitors acquire components of proprietary information, anyone component
4 CAW-11-3324 may be the key to the entire puzzle, thereby depriving Westinghouse of a competitive advantage.
(e)
Unrestricted disclosure would jeopardize the position of prominence of Westinghouse in the world market, and thereby give a market advantage to the competition ofthose countries.
(f)
The Westinghouse capacity to invest corporate assets in research and development depends upon the success in obtaining and maintaining a competitive advantage.
(iii)
The information is being transmitted to the Commission in confidence and, under the provisions of 10 CFR Section 2.390, it is to be received in confidence by the Commission.
(iv)
The information sought to be protected is not available in public sources or available information has not been previously employed in the same original manner or method to the best of our knowledge and belief.
(v)
The proprietary information sought to be withheld in this submittal is that which is appropriately marked in CAE-ll-MUR-61 P-Attachment, "Byron and Braidwood Units 1 and 2 - Response to NRC Informal Request for Additional Information (RAI) from the Nuclear Performance and Code Review Branch Related to the Measurement Uncertainty Recapture Power Uprate (MUR) License Amendment Request (TAC Nos. ME6587, ME6588, ME6589, and ME6590)" (proprietary) for submittal to the Commission, being transmitted by Exelon Nuclear letter and Application for Withholding Proprietary Information from Public Disclosure, to the Document Control Desk. The proprietary information as submitted by Westinghouse for use by Byron Stations Units 1 and 2 and Braidwood Stations Units 1 and 2 is expected to be applicable for other licensee submittals in response to certain NRC requirements for Measurement Uncertainty Recapture submittals and may be used only for that purpose.
5 CAW-U-3324 This infonnation is part of that which will enable Westinghouse to:
(a)
Assist the customer in obtaining NRC review of the Byron Stations Units 1 and 2 and Braidwood Stations Units 1 and 2 Measurement Uncertainty Recapture Power Uprate Program.
(b)
Provide inputs for customer-specific calculations.
(c)
Provide licensing support for customer submittals.
Further this infonnation has substantial commercial value as follows:
(a)
Westinghouse plans to sell the use of this infonnation to its customers for the purpose of meeting NRC requirements for licensing documentation associated with measurement uncertainty recapture power uprate.
(b)
Westinghouse can sell support and defense of the technology to its customer in the licensing process.
( c)
The infonnation requested to be withheld reveals the distinguishing aspects of a methodology which was developed by Westinghouse.
Public disclosure of this proprietary infonnation is likely to cause substantial hann to the competitive position of Westinghouse because it would enhance the ability of competitors to provide similar calculations and licensing defense services for commercial power reactors without commensurate expenses. Also, public disclosure of the infonnation would enable others to use the infonnation to meet NRC requirements for licensing documentation without purchasing the right to use the infonnation.
The development of the technology described in part by the infonnation is the result of applying the results of many years of experience in an intensive Westinghouse effort and the expenditure of a considerable sum of money.
In order for competitors of Westinghouse to duplicate this infonnation, similar technical programs would have to be perfonned and a significant manpower effort, having the requisite talent and experience, would have to be expended.
Further the deponent sayeth not.
Braidwood and Byron Stations Measurement Uncertainty Recapture License Amendment Request (MUR LAR)
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RAI)
December 9, 2011 ATTACHMENT 3 RESPONSES TO REQUESTS FOR ADDITIONAL INFORMATION (NON-PROPRIETARY VERSION)
NRC Request 1 Braidwood/Byron Stations MUR LAR Response to RAI December 9, 2011, page 1 NON-PROPRIETARY Please identify the numerical values for the parameter uncertainties and biases that were used in the derivation of the departure for nucleate boiling ratio design limit of 1.19 for the ABB-NV correlation in accordance with the revised thermal design procedure and further justify that these values are acceptable for Byron and Braidwood.
Response
The parameter uncertainties and biases that were used in the derivation of the departure from nucleate boiling ratio (DNBR) design limit of 1.19 for the ABB-NV correlation are in accordance with WCAP-11397-P-A, "Revised Thermal Design Procedure," (Reference 1) and are provided in Table 1-1.
Table 1-1:
Uncertainties and Biases Used in the RTDP ABB-NV DNBR Limit for the ByronlBraidwood MUR-PU Safety Analyses Parameter Uncertainties and Biases Core Power
+/-O.5%(1)
+/-3.5%
Pressurizer Pressure
+/-43.0 psi RCS Average Temperature
+/-7.6°F
+1 SF (bias) (2)
Effective Core Flow Fraction
[
] a,c Nuclear Enthalpy Rise Hot Channel Factor, (FN LiH)
[
] a,c Fuel Fabrication Enthalpy Rise Engineering Hot Channel Factor, (FE LiH,1)
[
] a,c Subchannel Analysis Code
[
] a,c Transient Code
[
] a,c Note:
(1 ) A conservatively high power uncertainty was assumed to calculate the RTDP DNBR design limit.
The final power uncertainties for the Byron and Braidwood units are in the range of +/-O.334% to
+/-O.345%.
(2) Only the random portion (+/-7.6°F) of the RCS average temperature uncertainty is included in the RTDP DNBR design limit. The bias on the RCS average temperature is explicitly modeled in the safety analyses.
Consistent with the Revised Thermal Design Procedure (RTDP) methodology in WCAP-11397-P-A (Reference 1), biases are not included in the development of the RTDP DNBR limit. Only the random portion of the parameter uncertainties is included in the statistical combination of
Braidwood/Byron Stations MUR LAR Response to RAI December 9, 2011, page 2 NON-PROPRIETARY uncertainties for the RTDP DNBR limit. For the Byron and Braidwood Measurement Uncertainty Recapture Power Uprate (MUR-PU) DNB analyses, the adverse instrumentation bias associated with the reactor coolant temperatures is explicitly modeled in the safety analyses.
The only change to the Byron and Braidwood MUR-PU DNB analyses with RTDP parameter uncertainties from the pre-MUR-PU values is the reduced power measurement uncertainty associated with the use of the LEFM CheckPlus system to measure feedwater flow. As discussed in Attachments 8a through d of the June 23, 2011 MUR-PU License Amendment Request (LAR) submittal (Reference 2), the final power uncertainties for the Byron and Braidwood units are in the range of +/-0.334% to +/-0.345%. A conservatively high power uncertainty of +/-0.5% was assumed to calculate the RTDP DNBR design limit.
The remaining parameter uncertainties from the current RTDP DNB analyses are not affected by the increase in power level or the use the LEFM CheckPlus system and therefore continue to be applicable to Byron and Braidwood at MUR-PU conditions.
NRC Request 2 Section 11.3.2, "Auxiliary Equipment Design Transients" indicates that the only auxiliary equipment design transients impacted by the power uprate are those associated with the reactor coolant system hot and cold leg temperatures. It is further stated that the existing auxiliary equipment design transients are conservative and bounding for the power uprate.
Please discuss whether the analysis included changes in nitrogen-16 activity that would potentially affect letdown line decay time requirements.
Response
Auxiliary equipment design transients deal with Nuclear Steam Supply System (NSSS) temperature, pressure and flow. Activity and radiation levels are not included in detailed equipment design transients, including the auxiliary design transients. As such, N-16 is not associated with auxiliary equipment design transients.
The existing design basis reactor coolant activity is calculated at a core thermal power level of 3,658.3 MWt. Since the MUR PU core thermal power level is 3,645 MW the existing design basis reactor water fission product and activation product inventory remains bounding.
Even if a small increase in activity and associated radiation levels based on the increased core power level is assumed to occur in the letdown line, the impact of this small increase would be insignificant. Applying conservative assumptions, the radiation field would not be anticipated to increase by more than 0.2% from all sources of radiation in the letdown line flow at the point where the letdown line exits containment. The dose contribution from N-16 would decrease relatively rapidly from that point because of the short half-life (Le. approximately 7 seconds) of N-16. This small change in the N-16 dose contribution does not necessitate a change in the letdown line decay time requirements.
NRC Request 3 Section IV.1.E.iii, "Erosion/Corrosion Program," describes the flow accelerated corrosion (FAG) monitoring program. The FAG monitoring program includes the use of a predictive method to calculate the wall thinning of components susceptible to FAG. For the piping lines that have been recommended for FAG review, please provide a sample list of the components expected to experience the greatest increases in wear rates for MUR power uprate conditions. Include
Braidwood/Byron Stations MUR LAR Response to RAI December 9, 2011, page 3 NON-PROPRIETARY the initial wall thickness (nominal), current (measured) wall thickness, a comparison of the measured wall thickness to the thickness predicted by the GHEGWORKS SteamlFeedwater Application [SF A] FAG model, and predicted time to reach minimum allowable wall thickness.
Response
As indicated in Section IV.1.E.iii of the MUR-PU LAR submittal (Reference 2), the GHEGKWORKS SFA models will be updated to incorporate the changes associated with the MUR-PU. The changes will be incorporated into the inspection program as part of the implementation of the MUR-PU.
Exelon reviewed the components/systems that experienced the greatest wear rates based on the results of the stretch uprate performed at Byron and Braidwood Station in 2001. It is expected that the MUR-PU will have similar effects.
Table 3-1 below provides a sampling of 20 components that are expected to experience the greatest wear rates prior to and following the MUR-PU. The components selected are representative of the highest predicted wear rates from the GHEGKWORKS updates that were performed for the stretch power uprate for Byron Unit 1; the unit which historically has operated at the highest power level of the four Byron and Braidwood units.
These components are in the Extraction Steam (ES) reheat piping, Extraction Steam piping to the Feedwater heaters, Feedwater Pump (FWP) discharge piping, Heater Drain (HD) piping, and High Pressure Turbine (HPT) to Main Steam Reheater (MSR) piping.
Table 3-1 provides the component number and description, along with the predicted wear rate from the stretch power uprate, initial wall thickness (nominal), last inspection date, predicted wall thickness at the time of the indicated inspection date, current (measured) wall thickness based on the indicated inspection date, current next scheduled inspection interval (NSI), and projected NSI due to the MUR-PU. The NSI is selected such that the inspection occurs prior to the time the minimum wall thickness is predicted to be reached. The evaluations performed for the stretch uprate were used to develop a projected new NSI following the MUR-PU increase.
The NSI is given in terms of the outage number in which the inspection is required per the model; for example, the next Byron Unit 1 outage number is 18.
Table 3-1:
Component Line 1ES132 1 ECES05: CRH TEE TO FWH15 TEE 1ES100 1 ECES04: ES TEE TO FWH16 TEE 1 FW007 1ECFW01: FWP 1A TO HDR 1ES141 1 ECES05: RV TEE TO FWH15B 1ES117 1 ECES05: CRH TEE TO FWH15 TEE 1 ES086 1 ECES04: ES TEE TO FWH16 TEE 1FW011 1ECFW01: FWP 1A TO HDR 1 FW024 1ECFW01: FWP 1B TO HDR 1 FW035 1 ECFW01: FWP 1 C TO HDR 1 ES079A 1 ECES02: MSR 1A TEE TO MSR1A N 1 ES079B 1 ECES02: MSR 1 A TEE TO MSR 1 A S 1 ES079D 1 ECES03: MSR 1B TEE TO MSR1B S 1ES122 1 ECES05: CRH TEE TO FWH15 TEE 1 ES143B 1 ECES05: RV TEE TO FWH 15B 1HD085A 1 ECHD03: HDP 1 A TO HDR 1HD085B 1 ECHD03: HDP 1 B TO HDR 1HD091 1 ECHD03: HDP 1 C TO HDR 1ES147A 1 ECES05: RV TEE TO FWH 15A 1ES147B 1 ECES05: RV TEE TO FWH 15B 1 FW097A 1 ECFW03: SG HDR TO SG 1A TEE Acronyms:
CRH -
Cold Reheat FWH-Feedwater Heater ES -
Extraction Steam FWP-Feedwater Pump FW-Feedwater HD-Heater Drain NON-PROPRIETARY Braidwood/Byron Stations MUR LAR Response to RAI December 9,2011, page 4 Projected Wear Rates and Next Inspection Intervals Predicted Predicted Measured Current Projected Wear Rate-Nominal Wall at Wall at NSI NSlfor MUR Stretch Uprate Wall Inspection Inspection Inspection (Refueling (Refueling (mils/yr)
(in)
Date (in)
(in)
Outage#)
Outage #)
17.1 0.688 9/25/2000 0.420 0.588 38 35 16.4 0.688 9/11/2006 0.365 0.417 21 21 15.9 1.812 9/23/2000 1.267 1.932 21 21 15.7 0.688 3/7/2005 0.414 0.631 69 65 14.6 0.688 3/10/2005 0.434 0.774 25 24 14.0 0.688 3/9/2005 0.413 0.621 24 24 13.8 1.812 4/1/1999 1.452 1.709 21 21 13.8 1.812 4/2/1999 1.452 1.713 19 19 13.8 1.812 4/2/1999 1.452 1.722 20 20 11.5 0.594 3/7/2005 0.394 0.358 19 19 11.5 0.594 3/25/2008 0.311 0.374 24 24 11.5 0.594 3/16/2011 0.269 0.279 20 20 10.7 0.5 3/14/2002 0.314 0.497 90 81 10.6 0.688 9/25/2000 0.523 0.688 68 61 10.4 0.844 9/27/2000 0.608 0.825 42 40 10.4 0.844 9/26/2000 0.608 0.806 47 45 10.4 0.844 9/27/2000 0.608 0.771 45 43 10.4 0.5 4/13/1996 0.430 0.497 27 25 10.4 0.5 4/3/1999 0.377 0.502 41 38 10.3 1.219 12/4/1997 1.081 1.139 18 18 I
HDP -
Heater Drain Pump SG-Steam Generator HDR - Header RHT -
Reheat MSR - Moisture Separator Reheater RV-Relief Valve
NRC Request 4 Braidwood/Byron Stations MUR LAR Response to RAI December 9, 2011, page 5 NON-PROPRIETARY Section IV. 1.A. vi. 2.c, "Steam Generator [SG] Tube Bundle Integrity, Flow Induced Vibration and Wear, H*, and Chemistry, 11 states that changes in the key operating parameters associated with the MUR power uprate do not impact on the H* lengths or leakage factors for the Byron and Braidwood, Unit 2, Model 05 SGs. Please describe the key operating parameters affected by the MUR power uprate and discuss how they are bounded by the H* analysis to support the claim that H* lengths and leakage factors are not affected.
Response
The limiting H* length for the Byron and Braidwood Unit 2 steam generators in the most recent H* amendment (Reference 3) is a result of the main steam line break (MSLB) conditions. The key parameters at MSLB operating conditions that are used for determining H* lengths are provided in Table 4-1 and are not affected by the MUR-PU.
The leakage factors identified in Reference 3 were calculated using high T avg normal operating parameters conditions as initial conditions, there are no significant changes in the input conditions (pressure differential and temperature) between the conditions contained in Reference 3 and the revised conditions for the MUR-PU as shown on Table 4-1. The SG hot leg inlet temperature for the MUR-PU conditions is limited to 618.4 of by operating restrictions and the same SG inlet temperature restriction (618.4 OF) was used in Reference 3. The SG cold leg temperature decreases by 0.6 OF from 555.4 of to 554.8 of. Based on these temperatures and the slight decrease in pressure differential ratio subfactor that occurs during the postulated control rod ejection (CRE), locked rotor (LR) event, or feedwater line break (FLB) event due to an increased high Tavg normal operating parameter pressure differential at MUR-PU conditions, it is concluded that the leakage factors for CRE, LR, and FLB in Reference 3 remain bounding. High T avg normal operating parameter pressure differential is conservatively used in determining the CRE, LR and FLB leakage factors because its use results in a larger pressure differential subfactor versus using the low T avg normal operating parameter pressure differential.
Braidwood/Byron Stations MUR LAR Response to RAI December 9, 2011, page 6 NON-PROPRIETARY Table 4-1:
Key Operating Parameters - Current H* Analysis vs. Measurement Uncertainty Recapture Power Uprate Conditions H* Leakage Factors H* Length AOR MUR-PU H* Analysis H* Analysis MSLB (10% Plugging-(10% Plugging-Conditions(C)
Analysis Parameters High Tavg)
High Tavg)
Primary Side Pressure (psia) 2250 2250 2575 Secondary Pressure (psia) 923 914 15 Primary-to-Secondary Differential 1327 1336 2560 (psid)
SG Inlet/Outlet Temperature (OF) 618.4(a)/555.4 618.4(a)/554.8 297 Steam Temperature (OF)
(b)
(b) 212 Secondary Fluid Temperature (OF)
(b)
(b) 212 (a) Exelon Generation has limited the reactor vessel outlet temperature (SG inlet temperature) to less than or equal to 618.4°F.
(b) Parameter is not used in determining H* length or leakage factors.
(c) Parameters used to determine H* lengths are not impacted by MUR-PU NRC Request 5 Section IV. 1.A. vi.2.d, "Steam Generator Steam Drum Evaluation," states that increased fluid velocity at MUR power uprate conditions may increase current estimated degradation rates up to 25 percent for Byron, Unit No.2, and Braidwood, Unit 2. Additionally, it states that erosion-corrosion has previously been detected in several components of the Byron, Unit No.2, and Braidwood, Unit No.2 Model 05 SGs steam drum internals. Please provide examples of the most rapid degradation rates observed to date for steam drum internals and discuss whether a 25 percent increase in degradation rate would impact the inspection frequency andlor scope of these components.
Response
The Exelon steam generator inspection program is based on the requirements of NEI 97-06, "Steam Generator Program," Revision 3 and its referenced EPRI Steam Generator Management Program Guidelines, specifically, the EPRI PWR Steam Generator Examination Guidelines, Revision 7, and the EPRI Steam Generator Integrity Assessment Guidelines, Revision 3.
Steam drum degradation at Braidwood Unit 2 and Byron Unit 2 was first observed in 2005.
Since 2005, visual inspections of the steam drum internals at both plants have been performed during selected subsequent outages.
At Braidwood Unit 2, steam drum inspections were performed and ultrasonic (UT) thickness measurements were collected during outages in 2005, 2006, 2008, 2009, and 2011. At Byron Unit 2, steam drum inspections were performed and UT thickness measurements were
Braidwood/Byron Stations MUR LAR Response to RAI December 9, 2011, page 7 NON-PROPRIETARY collected during outages in 2005,2007, and 2011. Based on a study of all of the UT thickness measurements for Braidwood Unit 2 and Byron Unit 2, it was concluded that a 0.036 inch/cycle degradation rate represents an accurate (yet conservative) rate of degradation. The Byron Unit 2 Fall 2011 steam drum inspections were performed after the MUR-PU LAR submittal (Reference 2). An engineering evaluation was performed to determine the next inspection interval of the degraded steam drum components and used a 0.045 inch/cycle degradation rate, which is the 0.036 inch/cycle rate with a 25% increase for the MUR-PU.
Historical inspection results, based on the UT thickness measurements, determined the maximum observed one-cycle direct material loss at Braidwood Unit 2 was 0.062 inch. This material loss occurred on the swirl vane blade of a primary moisture separator. The maximum observed one-cycle direct material loss at Byron Unit 2 was 0.056 inch. This material loss occurred on the swirl vane blade of a primary moisture separator.
The 0.062 inch/cycle degradation rate is a maximum observed historical rate at one specific location. Typical rates of steam drum degradation observed at both Braidwood Unit 2 and Byron Unit 2 are considerably less. The largest degradation rate determined during the most recent Byron Unit 2 outage steam drum inspection (Fall 2011) was 0.021 inch/cycle over the last three cycles. Therefore, there is considerable margin to the post MUR-PU degradation rate of 0.045 inch/cycle that was used in the engineering evaluation.
The increase in degradation rate has been incorporated into the steam generator inspection program. The inspection scope and frequencies for steam generator steam drums remains unchanged. However, Exelon will continue to evaluate the impact of any potential accelerated degradation rates in the steam drum and will adjust inspection frequency based on the results of these inspections.
NRC Request 6 Section IV. 1.A. vi. 2.f, "Steam Generator Loose Parts," states that some existing objects in the Byron, Unit No.2, SGs have caused wear on the tubing during past cycles, however, these objects are termed, "unknown objects or inaccessible objects," since the support plate locations are difficult to access. It is also stated that the inspection criteria for these objects will remain unaffected as a result of the MUR power uprate. Please describe the inspection frequency, type of eddy-current technique used, and extent to which secondary-side inspections are performed to disposition these "unknown objects or inaccessible objects." Also, describe how it was determined that the inspection criteria for these objects can remain the same under MUR power uprate conditions.
Response
Steam generator eddy current inspections are performed in accordance with Technical Specification 5.5.9, "Steam Generator (SG) Program," and the EPRI PWR Steam Generator Examination Guidelines, Revision 7. These routine inspections have been successful at identifying foreign object wear indications and possible loose part (PLP) indications. The indications are identified during routine bobbin coil inspections and confirmed with Plus-Point probes. At a given inspection, the tubes surrounding the reported wear and/or PLP indications, as well as tubes surrounding each tube previously plugged for foreign object wear or PLP indication, are inspected at the same elevation as the indication using Plus-Point probes to verify the extent of the indication and to monitor for migration of the loose part during subsequent operating cycles. Preventive plugging and stabilization of every tube with a foreign
Braidwood/Byron Stations MUR LAR Response to RAI December 9,2011, page 8 NON-PROPRIETARY object wear indication in the inaccessible Tube Support Plate (TSP) region or a PLP indication is performed.
Subsequent to the MUR-PU LAR submittal (Reference 2), steam generator inspections at Byron Unit 2 in September/October 2011 (Le., refueling outage B2R16) included Plus-Point probe inspections in a two-tube "box" around each foreign object wear indication, PLP indication, or each tube previously plugged for foreign object wear or PLP indication in the upper TSPs.
These inspections included all indications identified during previous outages and the Fall 2011 refueling outage (Le., B2R 16). The purpose of these inspections was to determine the extent of the foreign object wear, to verify the tubes that were left in service had no detectable degradation at the elevation where the foreign objects were identified, and to confirm that the objects which caused previously reported wear indications had not migrated.
Exelon also has performed a quantitative evaluation of foreign objects at the upper TSPs. This work consisted of the following activities:
- A detailed history review of previous eddy current data to determine if flaw indications were present during previous outages,
- Application of a geometric (axial and circumferential length extent) sizing technique for upper bundle wear scars that enabled more accurate determination of the true flaw geometry than compared to industry methods,
- Projection of a maximum wear scar size for a two-cycle inspection interval,
- Robust structural evaluation which considered the wear as either volumetric or cracking degradation morphologies to help to ensure the structural and leakage integrity is ensured for all plant conditions,
- An assessment of identified foreign object wear indications at the upper TSPs through Byron Unit 2 refueling outage 16 (Le., B2R16) for two cycles of operation, and
- The potential impact resulting from the implementation of the MUR-PU.
No subsequent secondary side inspections of these objects are currently planned since all of the foreign objects are at locations inaccessible by the inspection equipment.
The potential impact of the MUR-PU was assessed by comparing the local thermal-hydraulic conditions at the flaw locations for both pre-and post-MUR-PU conditions. In general, flaw locations on the hot leg side below TSP 8, which is the uppermost elevation of the preheater, had a slight decrease in localized dynamic pressure (pV2). Flaw locations in the preheater or above TSP 8 had a slight increase in dynamic pressure. Conservative scaling factors were then developed to predict the potential impact of the MUR-PU. Since each foreign object wear indication in the upper TSPs has been plugged and stabilized on detection, the only way active tubes in the region can be affected is if the foreign object migrates during the operating period.
Based on historical trends since 1995 (Le., B2R05 inspections) and the analytical work described above, it was concluded that the foreign objects identified in the upper TSPs were acceptable for a two-cycle steam generator inspection interval at pre-and post-MUR-PU conditions.
NRC Request 7 In the LAR, Attachment 5,Section V.1.B.i, "Alternate AC (AAC) Power Source," the licensee stated that the AAC power source has sufficient capacity to operate systems necessary for coping with a station blackout (S80) event for the required 4-hour coping duration.
Braidwood/Byron Stations MUR LAR Response to RAI December 9,2011, page 9 NON-PROPRIETARY Provide a summary of the load study versus AAC power source capacity to validate the above statement under MUR conditions.
Response
As stated in Section V.1.B.i of Attachment 5 of the MUR-PU LAR submittal (Reference 2), the electrical loads that changed as a result of the MUR-PU are not fed from the EDG system.
There are no increases to the emergency buses loads supported by the EDGs. Therefore, MUR-PU has no impact on the Alternate AC Power Source or its ability to support the systems necessary for coping with an SBO event.
The worst-case loading for a diesel-generator during an SBO event for any of the Byron or Braidwood Units occurs at Byron and the loading is 5722 KW, which is well below the 2000-hour rating of the diesel-generator of 5935 KW.
NRC Request 8 In the LAR, Attachment 5,Section V.1.B.iii, "Condensate Storage Tank Inventory," the licensee stated that the condensate storage tank provides adequate inventory for decay heat removal following a SBO event at MUR power uprate conditions.
Provide a discussion of condensate storage tank inventory required under normal operating conditions versus the required inventory for the SBO duration at MUR power uprate conditions.
Response
The minimum Condensate Storage Tank (CST) inventory per Technical Specifications Section 3.7.6, "Condensate Storage Tank (CST)," and the associated Technical Specification Bases Section B 3.7.6 is 212,000 gallons for each unit. Using NUMARC 87-00, "Guidelines and Technical Bases for NUMARC Initiatives Addressing Station Blackout at Light Water Reactors,"
Section 7.2.1, "Condensate Inventory for Decay Heat Removal," methodology, the necessary condensate inventory per unit for an SBO event is 81,231 gallons of CST water under MUR-PU conditions. The Station's minimum CST inventory of 212,000 gallons is well above the SBO required water inventory.
NRC Request 9 In the LAR, Attachment 5,Section V.1.B.iv, "Class 1 E Battery Capacity," the licensee stated that the Class 1 E batteries have sufficient capacity to provide adequate power for safe shutdown loads.
Provide a discussion of capacity margins available in the Class 1 E batteries under MUR conditions.
Response
As stated in Section V.1.B.iv of Attachment 5 of the MUR-PU LAR submittal (Reference 2), the MUR-PU does not affect any DC powered indication, control, or protection equipment.
Therefore, MUR-PU has no impact on the loading for the Class 1 E batteries and the batteries remain acceptable at MUR-PU conditions.
The capacity margins for the Class 1 E batteries are provided in the Table 9-1. These margins do not change as a result of MUR-PU conditions.
Braidwood/Byron Stations MUR LAR Response to RAI December 9, 2011, page 10 NON-PROPRIETARY Note the battery remaining capacity margins are in addition to the factors that are included in the battery sizing calculation (Le., 1.05% for design margin, 1.25% for aging and 1.11 % for temperature correction).
Table 9-1:
Battery Remaining Capacity Margin BATTERY Braidwood Byron 111 (1DC01E) 2.7%
1.8%
112 (1DC02E) 6.1%
7.6%
211 (2DC01 E) 4.7%
4.7%
212 (2DC02E) 8.3%
8.7%
NRC Request 10 In the LAR, Attachment 5,Section V.1.C, "Environmental Qualification (EQ) of Electrical Equipment," the licensee stated that they conducted an evaluation and concluded that the power uprate will not impact the equipment qualification.
Provide a discussion/summary of temperature, pressure, and radiation levels/profiles to demonstrate that adequate margins remain with respect to EQ of electrical equipment in accordance with IEEE Std. 323-1974, under the worst case accident conditions at MUR power uprate conditions.
Response
Radiation:
The total integrated dose used for determining electrical equipment environmental qualification (EQ) did not change for Byron and Braidwood Station at MUR-PU conditions and therefore remains bounding. The current integrated dose includes a 10% margin per IEEE-323, "IEEE Standard for Qualifying Class 1 E Equipment for Nuclear Power Generating Stations,"
requirements. The current radiological environmental parameters for post-accident dose contribution to the total integrated doses had been analyzed with respect to a power level of
~ 102%, which bounds the MUR-PU power level.
Pressure:
The long-term LOCA mass and energy (M&E) release and the Main Steam Line Break (MSLB)
M&E were reanalyzed for the Byron and Braidwood Stations MUR-PU. The resultant containment pressures that would potentially affect the EQ profiles are discussed in sections 111.15.5 and 111.16.5 of the MUR-LAR submittal (Reference 2).
For the reanalyzed LOCA M&E cases, the resultant maximum containment pressure of 42.6 psig and 38.26 psig, respectively for Units 1 and 2, are less than the peak containment pressures for the current licensing bases of 42.77 psig and 38.36 psig for Units 1 and 2, respectively. A comparison of the reanalysis peak containment pressures for MUR-PU to those from the current licensing bases is presented in Table 111.15-3 of the MUR LAR submittal (Reference 2). Figures 111.15-1,111.15-3,111.15-5 and 111.15-7 of the MUR LAR submittal (Reference 2) provide the resultant containment pressure responses for LOCA cases reanalyzed for MUR-PU conditions.
Braidwood/Byron Stations MUR LAR Response to RAI December 9,2011, page 11 NON-PROPRIETARY For reanalyzed MSLB M&E peak containment pressure case on each unit, the resultant maximum containment pressures of 34.6 psig and 35.9 psig 1, respectively for Units 1 and 2, are less than the peak containment pressures for the current licensing bases of 39.3 psig and 38.3 psig for Units 1and 2, respectively. Figures 111.16-1 and 111.16-3 of the MUR LAR submittal (Reference 2) provide the resultant containment pressure responses for the MSLB cases reanalyzed for MUR-PU conditions.
The current peak pressure input for EO is 50 psig. This input is unchanged for MUR-PU conditions. The EO pressure profile is shown on Figure 10-1.
Temperature:
The electrical equipment EO temperature profile is a composite curve that bounds the results from both the MSLB M&E release inside containment and the LOCA M&E release analyses.
As discussed above, the long-term LOCA M&E release and the MSLB M&E were reanalyzed for the MUR-PU. The resultant containment temperatures that would potentially affect the EO profiles are discussed in sections 111.15.5 and 111.16.5 of the MUR-LAR submittal (Reference 2).
For the reanalyzed LOCA M&E the resultant maximum containment temperature of 264.24 of and 257.4 of, respectively for Unit 1 and Unit 2, are less than the peak containment temperatures for the current licensing bases of 264.5 of and 257.57 of for Units 1and 2, respectively. A comparison of the reanalysis peak containment temperatures for MUR-PU to those from the existing licensing bases is presented in Table 111.15-3 of the MUR-PU LAR submittal (Reference 2). Figures 111.15-2,111.15-4,111.15-6 and 111.15-8 of the MUR-PU LAR submittal (Reference 2) provide the resultant containment temperature responses for the LOCA M&E cases reanalyzed for MUR-PU conditions. The LOCA temperature curves are also replicated on Figure 10-1.
For the reanalyzed MSLB M&E peak containment temperature case on each unit, the resultant maximum containment air temperature increased by 0.6°F to 333.6°F for Unit 1; the previous maximum containment temperature of 330.8°F remains bounding for Unit 2. Figures 111.16-2 and 111.16-4 of the MUR-PU LAR submittal (Reference 2) provide the resultant containment temperature responses for the MSLB M&E cases reanalyzed for MUR-PU conditions. The composite MSLB temperature curves are also replicated on Figure 10-1.
The updated Containment EO temperature profile is shown on Figure 10-1.
The effect of the revised containment conditions on the environmental qualification of electrical equipment in containment was evaluated. The evaluation included consideration of margins in peak temperature and Post Accident Operating Time (PAOT) in accordance with the recommendations of IEEE-323-1974 (e.g., 15 of peak temperature and 10% in PAOT). This is consistent with the current EO program and industry practice. The smallest margin was 2.9%
and is associated with the PAOT. The evaluation concluded that all equipment is qualified with sufficient margin. Occurrences of reduced margin are qualitatively justified based on separate test data, equipment functional requirements or thermal lag.
Evaluation results show that all equipment in containment within the scope of the EO program remains qualified.
1 Section 111.16.5 of the MUR LAR submittal (Reference 2) contained a typographical error that identified the Unit 2 value as 31.4 psig.
NON-PROPRIETARY Braidwood/Byron Stations MUR LAR Response to RAI December 9,2011, page 12 Figure 10-1: Containment EQ Temperature and Pressure Profile 360 I 150 340*
320 I
~
I 1
300
/
280
/;-~~.. \\ -i t
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---:--~~.
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~
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~
NRC Request 11 Braidwood/Byron Stations MUR LAR Response to RAI December 9, 2011, page 13 NON-PROPRIETARY In the LAR, Attachment 5,Section V.1.F.v, "Reserve Station Service Transformers," the licensee stated that the plant operation at power uprate conditions has no effect on loss of voltage or degraded grid voltage protection schemes, and motor starting scenarios.
Provide steady state voltages at 4. 16 kV safety-related buses under worst case design basis conditions, before and after MUR conditions, in a tabular form.
Response
Table 11-1 provides the steady-state voltages at the 4.16 kV safety-related buses before and after MUR-PU conditions. These voltages are based on worst-case design basis conditions (Le., post-LOCA conditions with the minimum required voltage being available in the switchyards). This table shows that 4.16 kV safety-related bus voltages do not change as a result of MUR-PU.
The 6.9 kV system load changes required for MUR-PU are not safety-related and have no impact on 4.16 kV safety-related bus voltages.
Table 11-1:
4.16 kV Safety-Related Worst Case Bus Voltages Braidwood Byron Current Current (Pre-MUR)
Post-MUR (Pre-MUR)
Post-MUR Bus Number Voltage (V)
Voltage (V)
Voltage (V)
Voltage (V) 141 4069.2 4069.2 3946.5 3946.5 142 4041.4 4041.4 3929.2 3929.2 241 4064.5 4064.5 3923.8 3923.8 242 4075.1 4075.1 3969.3 3969.3 NRC Request 12 In the LAR, Attachment 5,Section V.1.F.i, "Main Generator," the licensee stated that the generator is operated within the Capability Curve.
Provide the Capability Curve to verify the MW [megawatt] and MVAR [megawatt volt amperes reactive] capability of the main generator.
Response
Copies of the Byron and Braidwood Generator Capability Curves are provided in Attachment 4.
NRC Request 13 In the LAR, Attachment 5,Section V.1.E, "Onsite Power Systems," the licensee stated that 'The LEFM [Leading Edge Flow Meter] CheckPlus System is being installed as an MUR power uprate device, however, no changes to the 120 V design loading will occur."
Braidwood/Byron Stations MUR LAR Response to RAI December 9, 2011, page 14 NON-PROPRIETARY Provide a discussion on the auxiliary power requirement for the Cameron LEFM CheckPlus System, such as direct current or AC power requirements, and its loading impact, if any, on the associated safety-related or nonsafety-related buses.
Response
For each Byron and Braidwood Unit there is one LEFM electronics cabinet with two CPUs and two AlC units. The power to the LEFM cabinet is supplied from non-safety related motor control centers (MCCs). Each CPU power feed requirement is half of the fully configured power (600 VA /2 CPUs = 300 VA). Each LEFM electronics cabinet AlC units will be powered via an outlet in the LEFM cabinet that is fed from an independent 120 VAC distribution panel.
Listed below are the electrical power requirements for each of the LEFM electronics cabinets:
LEFM Cabinet CPUs Nominal Power:
Nominal Voltage:
600 VA fully configured (300 VA per CPU) 120 VAC, single phase 50/60 Hz Steady State voltage range: +/- 10% nominal Transient voltage range:
+/- 20% for less than 10 ms LEFM Cabinet AlC Units Rated Current:
12.5 A (each AlC Unit)
Rated Voltage:
115 VAC +/-10%, single phase 50160 Hz Spare breakers from non-safety related MCCs are used to provide power to each of the CPUs in the LEFM electronics cabinet and the cabinet AlC units. Table 13-1 provides the existing and proposed loading on each of these MCCs.
Table 13-1:
MCC Loading Byron Braidwood MCC Unit1 Unit2 Unit 1 Unit2 Existing 1 (2)33Y1
<9KVA
<8 KVA
< 20 KVA
<11 KVA 1 (2)33V4
<9KVA
<19 KVA
<7.2 KVA
<17 KVA Addition 1.8 KVA 1.8 KVA 1.8 KVA 1.8 KVA The load addition to each of the 22.5 KVA distribution transformers on each of the MCCs is acceptable.
An evaluation also determined that current loading levels under MUR power uprate conditions have no impact on the 4.16 kV buses existing capability. The 120/208 VAC distribution panels for MCC's 1 (2)33Y1 and 1 (2)33V4 are modeled as fully loaded (Le. 22.5 KVA). Therefore, the load additions to these panels do not impact the modeled MCC bus loading or require a change to the load analysis calculations. Therefore, there is an insignificant change in the margin of the on-site electrical power systems.
The 125Vdc system loads are not impacted by the addition of the LEFM.
NRC Request 14 In the LAR, Attachment 5,Section V.1.E, "Onsite Power System," the licensee identified some load changes in the 6.9 kV system.
Braidwood/Byron Stations MUR LAR Response to RAI December 9,2011, page 15 NON-PROPRIETARY Provide a discussion that as a result of some load changes in 6.9 kV system, there are no adverse impacts on any 6.9 kV switchgear equipment such as short-circuit ratings, protective relay settings etc.
Response
The MUR-PU results in some load changes in the 6.9 kV system. Table 14-1 identifies the affected loads and the worst-case changes as a result of MUR-PU.
These changes result in a maximum change and percentage change to an individual bus load flow of 66 KVA and < 0.5% respectively. The remaining loading margin after these changes for any of the 6.9 kV buses is a minimum of 36%.
Table 14-1 shows that the new break horsepowers (BHPs) at MUR-PU conditions are still within the nameplate ratings for the affected motors. This means the existing motors are adequate and will not be altered. Since the motors are not being physically altered, the existing fault current contribution by the pump motors will not change as a result of MUR-PU. Therefore, the load changes for MUR-PU does not change the short-circuit current on the 6.9 kV system and has no adverse impact on the 6.9 kV switchgear equipment short-circuit rating.
Also, protective relay settings are not impacted since the change in BHP is still within the nameplate ratings for the motors and the motors will not change. The condensate / condensate booster pump motor, Reactor Coolant Pump motor and the heater drain pump motor breaker protective relays have been set to protect the existing motors.
Based on the above, there are no adverse impacts on any 6.9 kV switchgear equipment.
Table 14-1:
MUR-PU 6.9 kV LOAD CHANGES MOTOR CURRENT MUR-PU NAMEPLATES LOAD BHP BHP RATING (HP)
Condensate/ Condensate Booster 3201 3236 3500 (CD/CB) Pump Motor Duty Heater Drain (HD) Pump Motor Duty 1901 1888 2250 Reactor Coolant Pump (RCP) Motor Duty 6981 6986 7000 NRC Request 15 Please provide the additional references associated with the Steam Generator Tube Rupture Analysis. These references are needed to understand the treatment of potential non-conservative assumption in WCAP-10698-P-A and its Supplement 1 as discussed in a (ADAMS Accession No. ML111790038).
NSAL-07-11, "Decay Heat Assumption in Steam Generator Tube Rupture Margin-to-Overfill Analysis Methodology, " November 2007.
WCAP-16948-P, "Clarifications for the Westinghouse Steam Generator Tube Rupture Margin to Overfill Analysis Methodology, " December 2008.
Response
Braidwood/Byron Stations MUR LAR Response to RAI December 9, 2011, page 16 NON-PROPRIETARY A copy of NSAL-07 -11 is provided in Attachment 5. Westinghouse Electric Company has provided a copy of WCAP-16948-P in Reference 4.
References Braidwood/Byron Stations MUR LAR Response to RAI December 9, 2011, page 17 NON-PROPRIETARY
- 1) WCAP-11397-P-A, "Revised Thermal Design Procedure," Friedland, A. J. and Ray, S.,
April 1989.
- 2) Letter from Craig Lambert (Exelon Generation Company, LLC) to u. S. NRC, "Request for License Amendment Regarding Measurement Uncertainty Recapture Power Uprate,"
dated June 23, 2011
- 3) Letter from N. J. DiFrancesco (USNRC) to M. J. Padllo (Exelon), "Braidwood Station, Units 1 and 2 and Byron Station, Units Nos. 1 and 2 - Issuance of Amendment RE:
Changes to Technical Specification Sections 5.5.9, 'Steam Generator (SG) Program' and 5.5.9 'Steam generator (SG) Tube Inspection Report.' (TAC Nos. ME5198, ME5199, ME5200, and ME5201 )," April 13, 2011 [Accession No. ML110840580]
- 4) Letter from J.A. Gresham (Westinghouse) to USNRC, LTR-NRC-11-22, "Submittal of WCAP-16948-P, Revision 0, 'Clarifications for the Westinghouse Steam Generator Tube Rupture Margin to Overfill Analysis Methodology' (Proprietary)", dated November 17, 2011
Braidwood and Byron Stations Measurement Uncertainty Recapture License Amendment Request (MUR LAR)
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RAI)
December 9, 2011 ATTACHMENT 4 BYRON AND BRAIDWOOD STATIONS GENERATOR CAPABILITY CURVES
Braidwood and Byron Stations Measurement Uncertainty Recapture License Amendment Request (MUR LAR)
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RAI)
December 9, 2011 1000 800 600 400 200 a
-200
-400
-600
-800 GENERATOR CAPABILITY CURVES AND UNDEREXCITATION LIMITER SEmNGS Byron Units 1 or 2 Regulator: In-Servlce
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Braidwood and Byron Stations Measurement Uncertainty Recapture license Amendment Request (MUR LAR)
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RAI)
December 9, 2011 Braidwood Station Unit 1 Gallelalor~ Curves and Underexcitalioo Urititer SeIIings RegulaIor: in Service MEGAWATTS - Q2800 Notes; 1. Not to be: used with wbge regulator out of service.
- 2. Based on 10., maruirL
- 3. Maximum exciler amps = 101.
- 4. Maximum stator amps = 31431.
Braidwood and Byron Stations Measurement Uncertainty Recapture License Amendment Request (MUR LAR)
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RAI)
December 9, 2011 Braidwood Station Unit2 Ga......
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- 3. Maximum exciter amps = un_
- 4. Maximum stator amps = 31431.
(1D5'l1i)
Braidwood and Byron Stations Measurement Uncertainty Recapture License Amendment Request (MUR LAR)
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RAI)
December 9, 2011 ATTACHMENT 5 NSAL-07 Decay Heat Assumption in Steam Generator Tube Rupture Margin-to-Overfill Analysis Methodology
Nuclear Safety.westinghouse
============~AdwsoryLeff~
This is a notification of a recently identified potential safety issue pertaining to basic components supplied by Westinghouse.
This information is being provided so that you can conduct a review of this issue to determine if any action is required.
P.O. Box 355, Pittsburgh, PA 15230
Subject:
Decay Heat Assumption in Steam Generator Tube Rupture Margin-to-Overfill Analysis Methodology Basic Component: Steam Generator Tube Rupture Analysis Affected Plants: See page 4 Substantial Safety Hazard or Failure to Comply Pursuant to 10 CFR 21.21 (a)
Transfer of Information Pursuant to 10 CFR 21.21 (b)
Advisory Information Pursuant to 10 CFR 21.21 (d)(2)
References:
See page 5
SUMMARY
Number: NSAL-07-11 Date:
11115/2007 Yes D No 181 N/A D Yes D Yes D A generic methodology for the steam generator tube rupture (SGTR) margin-to-ruptured SG overfill analysis was developed by a subgroup of the Westinghouse Owners Group (WOG) and is documented in WCAP-I 0698-P-A (Reference I). The methodology included the conclusion that higher decay heat was conservative with respect to margin to overfill, based on sensitivities performed for a reference plant.
Recently, a customer identified an issue with the Reference I decay heat conclusion. The customer determined that lower decay heat is conservative for their plants' SGTR margin-to-overfill analyses.
Westinghouse confirmed the customer's results and determined that lower decay heat could be conservative for some plants analyzed with the WCAP-l 0698 method.
In the development of the Reference 1 methodology, Westinghouse evaluated the dose consequences of SG overfill following a SGTR using best-estimate assumptions. This evaluation was documented in WCAP-II002 (Reference 2). The results were acceptable on a best-estimate basis and accepted by the Nuclear Regulatory Commission (NRC) although they noted the work could not be used in a plant's licensing basis since it does not utilize conservative (FSAR Chapter 15) methodology. Although the issue described in this NSAL may result in the design basis SGTR analysis predicting SG overfill, dose consequences are bounded by WCAP-Il 002. Since the dose consequences resulting from a SGTR event that includes SG overfill are acceptable on a best-estimate basis, this issue is not a substantial safety hazard pursuant to 10 CFR 21.21(a).
Additional information, if required, may be obtained from Sean Kinnas, (412) 374-4640 Originator:(s)
Approved:
J. T. Crane J. A. Gresham, Manager Regulatory Compliance and Plant Licensing Regulatory Compliance and Plant Licensing S. T. Kinnas Containment and Radiological Analysis Electronically approved records are authenticated in the Electronic Document Management System
ISSUE DESCRIPTION NSAL-07-11 Page 2 of5 As a consequence of the January 1982 Ginna SGTR event, the NRC questioned assumptions about the analyses presented in licensees' safety analysis. The NRC required some plants to address the issues raised by the Ginna event. In response, a WOG subgroup developed a generic SGTR methodology, WCAP-I0698-P-A, to serve as the framework for plant specific analyses. Participants in the WOG subgroup could then use the WCAP's generic methodology to perform their licensing basis SGTR analysis, or contract a vendor to perform the analysis. The WOG subgroup consisted of Shearon Harris, Byron and Braidwood, Catawba, Beaver Valley Unit 2, South Texas, Millstone Unit 3, Diablo Canyon, Ginna, Vogtle, Watts Bar, Comanche Peak and Seabrook.
WCAP-I0698-P-A examined the competing effects of decay heat on the margin-to-SG overfill following a SGTR. Showing margin-to-ruptured SG overfill demonstrates that potential consequences of water releases do not have to be considered. For the margin-to-overfill analysis, higher decay heat yields a benefit by increasing steam releases from the ruptured SG, but results in a penalty from a longer cooldown and a conservatively delayed break flow termination. Conversely, lower decay heat yields a penalty by reducing steam releases from the ruptured SG, but results in a benefit from a shorter cooldown and earlier break flow termination. WCAP-l 0698-P-A concluded that higher decay heat was conservative for the SGTR margin-to-overfill analysis, and therefore 120% of the 1971 ANS decay heat curve should be used in future SGTR margin-to-overfill analyses.
Recently, it was determined by a licensee that lower decay heat was conservative for their plants' margin to overfill analyses, resulting in a loss of steam generator margin to overfill. Westinghouse confirmed the lower decay heat could be conservative for some plants' margin to overfill analyses, while others will not be adversely affected.
This issue is only known to be applicable to those plants utilizing the WCAP-I 0698 method for their SGTR licensing basis. Combustion Engineering (CE) plants and Westinghouse plants that do not use the WCAP-I0698 method are not impacted by this issue.
TECHNICAL EVALUATION Although SG overfill could result in increased radiological consequences, there is a reasonable basis to conclude that the applicable off site dose guidelines would continue to be met. This is based on best-estimate dose evaluations performed as part of the generic SGTR methodology and presented to the NRC at that time as detailed below.
WCAP-II002 (Reference 2) presents an analysis of the consequences of SG overfill resulting from a SGTR. The analysis includes consideration of the failure of a safety valve on the ruptured SG following overfill and the resulting continued releases from the ruptured SG. The analysis modeled the recovery actions to cool down and depressurize the ruptured SG to cold shutdown. WCAP-II002 presents a dose calculation for the SGTR with overfill using realistic or best-estimate assumptions and concludes that the resulting doses are within the guideline values of 10 CFR 100.
The NRC reviewed WCAP-Il 002 and presented the conclusions of the review together with those of the review ofWCAP-10698-P-A in their evaluation dated March 30, 1987. This evaluation is included in WCAP-I0698-P-A. The conclusion of the evaluation ofWCAP-11002 includes the following statements, essentially accepting the conclusion that the dose guidelines would be met with best-estimate assumptions but not allowing its use for licensing basis analyses:
"The report concludes that the consequences of SGTR overfill are acceptable based on 'best estimate' off site dose calculations and the waterhammer evaluation as discussed above. The
NSAL-07 *11 Page 30f5 staff concludes that the analytical approach utilized in the report is technically sound with respect to providing best estimate" information regarding the effects of SOTR overfill.
However, this report should not be used as a licensing basis document since it does not utilize conservative (FSAR Chapter 15) methodology."
The above demonstrates that the radiological consequences of SO overfill were evaluated during the development of the SOTR methodology. The consequences were determined to be within 10 CFR 100 guidelines using best estimate assumptions which were found acceptable to the NRC for evaluation purposes. It is reasonable to extend this discussion for plants licensed to the dose guidelines of 10 CFR 50.67 (Alternate Source Term), since the transition from thyroid to total effective dose equivalent (TEDE) doses generally results in increased margin to the limits.
Both WCAP-I 0698 and WCAP-11 002 analyzed a single reference plant that was judged to apply generically. However, the recent discovery concerning the decay heat assumption in the generic methodology (WCAP-10698) raises the concern of whether the best estimate WCAP-11002 remains generically applicable. It is the Westinghouse engineering judgment that WCAP-11 002 continues to be applicable and may be used for operability determinations and other non-licensing basis applications that allow for the use of best-estimate assumptions. The basis for this judgment is presented below.
The method employed in the development ofWCAP-ll 002 was to force the reference plant from WCAP-10698 to overfill. The assumed operator action times were extended until SO overfill occurred and water was released through a safety valve on the ruptured SO. Once water went through the safety valve, a consequential failure of that safety valve was assumed. Two scenarios with this consequential failure were examined: the safety valve was assumed to either fail full-open or to fail partially open. Both scenarios result in an uncontrolled depressurization of the ruptured steam generator with continued break flow and atmospheric releases until cold shutdown conditions are reached.
With respect to decay heat, once overfill occurs, a higher decay heat is more limiting. After overfill occurs, higher decay heat results in a longer time to reach cold shutdown conditions and a later break flow termination, leading to more limiting dose consequences. Thus, the decay heat assumption used for the WCAP-II002 reference plant does not affect its continued applicability to demonstrate that dose limits are not challenged by a SOTR on a best-estimate basis.
For the SOTR thermal and hydraulic analysis for input to the radiological consequences analysis discussed in Reference 3 (modeling a failed-open PORV on the ruptured SO), there are no competing effects with respect to decay heat. Higher decay results in increased steam releases from the ruptured SO and a longer cooldown, leading to a later break flow termination. These effects are conservative for the SOTR radiological consequences calculation, and thus, lower decay heat does not need to be considered for the SOTR thermal and hydraulic analysis for input to the radiological consequences analysis performed following Reference 3.
AFFECTED PLANTS This issue is only applicable to plants utilizing the WCAP-l 0698 method for their SOTR margin-to-overfill analysis.
The WOO subgroup consisted of Shearon Harris, Byron and Braidwood, Catawba, Beaver Valley Unit 2, South Texas, Millstone Unit 3, Diablo Canyon, Oinna, Vogtle, Watts Bar, Comanche Peak and Seabrook.
Since these plants sponsored WCAP-l 0698-P-A, it is likely their plant-specific tube rupture analysis is based on the generic methodology, and they should review their assumptions with respect to decay heat.
NSAL-07 -11 Page 4 of5 All other utilities who have a SGTR margin to overfill analysis, whether licensing basis or supplemental, based on the WCAP-l 0698 method, should review their assumptions with respect to decay heat.
Therefore, the potentially affected plants will include the WOG subgroup participants and those plants that have utilized the WCAP-I0698 methodology. Based on a review of Westinghouse engineering records, the table below lists those plants where Westinghouse has performed a SGTR analysis per WCAP-l0698, and those plants in the WOG subgroup.
PLANTS THAT POTENTIALLY USE WCAP-I0698 METHODOLGY WOG SGTR Subgroup Beaver Valley 2 Braidwood 1&2 Byron 1&2 Catawba 1&2 Comanche Peak 1&2 Diablo Canyon 1&2 Ginna Millstone 3 Seabrook 1 Shearon Harris 1 South Texas 1&2 VogUe 1&2 Watts Bar 1 Other Plants Using WCAP-I0698 Beaver Valley 1 D. C. Cook 1&2 Farley 1&2 Indian Point 2&3 Kewaunee Point Beach 1 &2 Turkey Point 3&4 V.c. Summer Angra Beznau 1 Ringhals 2&3 CE plants and Westinghouse plants that do not use the WCAP-I0698 methodology are not impacted by this issue.
SAFETY SIGNIFICANCE The issue may result in the design basis SGTR analysis predicting SG overfill for a given plant, but the dose consequences are bounded by the WOG work discussed above (Reference 2) which demonstrates there is no safety concern based on best estimate analyses.
NRC AWARENESS Westinghouse has not formally notified the NRC.
RECOMMENDED ACTIONS
- 1. It is the Westinghouse engineering judgment that WCAP-11002 continues to be applicable and may be used for operability determinations and other non-licensing basis applications that allow for the use of best-estimate assumptions
- 2. Westinghouse will present a project authorization to the Analysis Subcommittee of the Pressurized Water Reactor Owners Group (PWROG) in December 2007 to develop a resolution plan for this issue. In addition to decay heat, other input assumptions with known competing effects will be examined. Once that program is complete, plants should examine the effects of the revised assumptions on their SGTR margin-to-overfill analyses.
- 3. The proposed PWROG program will investigate input parameters with competing effects, including decay heat. If a plant does not participate in the proposed PWROG program, it is
NSAL"()7-11 Page 5 of5 Westinghouse's recommendation that the SGTR analysis use decay heat based on full power operation. Early into an operating cycle, the decay power rapidly (on the order of hours) approaches an equilibrium value. It is not anticipated that a SGTR would occur during this time period since the likelihood of a SGTR occurring during this short period is low. A similar justification is contained within WCAP-l 0698 and has been accepted by the NRC. Hence, while lower decay heat has been shown to be conservative for some plants' SGTR margin-to-overfill analyses, only decay heat based on full power operation needs to be considered.
- 4. As noted above, the SGTR thermal and hydraulic analysis that provides input to the radiological consequences analysis (as discussed in Reference 3) is not impacted by this issue.
REFERENCES
- 1. WCAP-10698-P-A, "SGTR Analysis Methodology to Determine the Margin to Steam Generator Overfill," August 1987.
- 2. WCAP-ll002, "Evaluation of Steam Generator Overfill Due to a Steam Generator Tube Rupture Accident," February 1986.
- 3. Supplement 1 to WCAP-10698-P-A, "Evaluation of Off site Radiation Doses for Steam Generator Tube Rupture Accident," March 1986.
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