ML070920142

LTR-SCS-07-14-NP, Revision 0, Engineering Report, Seabrook, Turbine Trip Without Reactor Trip Transient from the P-9 Setpoint Analysis.

ML070920142
Person / Time
Site:	Seabrook
Issue date:	01/31/2007
From:	Westinghouse
To:	Office of Nuclear Reactor Regulation
References
LTR-SCS-07-14-NP, Rev 0
Download: ML070920142 (20)

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Attachment 5 Engineering Report Seabrook Nuclear Power Station Turbine Trip without Reactor Trip Transient from the P-9 Setpoint Analysis (Non-Proprietary),

WESTINGHOUSE NON-PROPRIETARY CLASS 3 LTR-SCS-07-14-NP Aftachment Engineering Report Seabrook Nuclear Power Station Turbine Trip without Reactor Trip Transient from the P-9 Setpoint Analysis Revision 0 January 2007 Westinghouse Electric Company LLC P.O. Box 355 Pittsburgh, PA 15230-0355

© 2007 Westinghouse Electric Company LLC All Rights Reserved Introduction Westinghouse has performed a best estimate plant specific analysis for Seabrook on the reactor trip without turbine trip P-9 permissive setpoint change from 20% to 45% of full power., Similar setpoint changes have been are approved by the Nuclear Regulatory Commission (NRC) for other utilities.

This study evaluated the impact of a turbine trip without reactor trip on the potential actuation of a pressurizer Power Operated Relief Valve (PORV). In addition, a sensitivity study was performed to consider the effects of degraded control systems.The Westinghouse design criterion is that turbine load rejections up to the maximum load rejection capability (i.e. the design basis) of the plant should not actuate a reactor trip if all control systems function properly as designed.

However, after the Three Mile Island (TMI) incident, the Nuclear Regulatory Commission (NRC) expressed concern regarding implementation of any plant features which could increase the probability of a stuck-open pressurizer PORV. The NRC position is addressed in NUREG-0737, Item ll-K.3.10 (Reference 1).In order to satisfy the NRC position, a best estimate plant specific analysis was performed to demonstrate that the pressurizer PORVs would not open following the change to this P-9 permissive setpoint from 20% to 45% of rated power. Based on all other Westinghouse licensing submittals, this best estimate analysis was performed without any instrument uncertainties.

Key Assumptions and Inputs The initial conditions and assumptions used in this analysis are unchanged from the current licensing basis for Seabrook.

The methods and computer codes are the same as used for the Stretch Power Uprate, and P-9 analyses for other Westinghouse plants. The methods have been reviewed and approved by the NRC as part of the licensing submittals for these plants.The analysis was performed using the lumped loop version of the NRC approved LOFTRAN computer code, (Reference 2), which simulates the overall thermal-hydraulic and nuclear response of the Nuclear Steam Supply System (NSSS) as well as the various control and protection systems.Consistent with the objective of this analysis, the transient conditions were simulated to determine the maximum pressurizer pressure following the initiation of the transient.

Therefore, the following key assumptions and inputs were utilized to ensure a conservatively high prediction of pressurizer pressure.1. The turbine trip transient was initiated after [ ]a.c seconds of steady state conditions.

The turbine was tripped from [ ]a.c of the full power (NSSS power = [ ]a,c MWt, Reference 4);this includes [ ]a'c power for conservatism on the P-9 setpoint of 45% of power (Reference 3). The [ ]a'c conservatism accounts for modeling limitations (i.e. not every control system is modeled in detailed) and to ensure the results are conservative with respect the actual plant responses.

The turbine load will be stepped down to [ ]a.c of load within [ ]a'c seconds. The transient is modeled as a step load decrease from [ ]a.c to [ ]a,, of load within [ ].c seconds.LTR-SCS-07-14-NP Attachment Page 2 of 10

2. The transient nominal conditions were based on a full power vessel average temperature (Tavg) of [ ]a*c OF, a full power feedwater temperature of [ ]a'c 0 F,and a nominal steam pressure of [ ]a.c psia and [ Steam Generator Tube Plugging (SGTP).a. Note that of the two high Tavg cases provided in Reference 4 ([ ]j*- psia versus]a,c psia), the [ ]a,c steam pressure ([ ]a.c psia) was modeled because a]ac level corresponds to a [ ],c steam pressure, which results in a higher pressurizer pressure.b. Note that of the two feedwater temperatures provided in Reference 4 ([ ]a'ac F versus [ ]ac OF), the [ ]ac feedwater temperature was chosen since it results in the highest steam/feedwater flow and therefore, the lowest steam dump capacity (in fraction of rated steam flow).3. Best estimate [ ]a c reactivity parameters from Cycle 11 were used (Reference 5). [ ]ac reactivity parameters have [ ]a-c differential rod worth and the I ]a,c moderator temperature coefficient and thus, using [ ]ac parameters in the analysis yields more conservative results, which bound the full cycle of operation.

4. Initial primary and secondary side conditions are at [ ]a.c of the rated power; these conditions, such as Tavg and pressure, do not account for any uncertainties (i.e. best estimate analysis).

5. The [ ]a'c overall heat transfer coefficients from fuel to coolant are used, consistent with [ ]a.c conditions.

6. Rod control is assumed operational and in the automatic mode of control. Since the turbine trip transient is a load decrease, the rods are automatically inserted to mitigate the transient.

However, it is assumed that once the nuclear power reaches about [ ]a'c % or below, the operator will place the control rods in manual to mitigate the transient to steady state no-load conditions.

This operator action (i.e. manual control of the rods) is not simulated in the LOFTRAN code; therefore, the results without manual control provide conservative responses.

7. The pressurizer pressure control system, steam dump control system (loss of load controller) and Steam Generator (SG) Level control system were assumed operational and in the automatic mode of control. The three primary control systems that act to mitigate this transient are the pressurizer pressure control system, rod control system, and steam dump control system 8. The Steam Generator (SG) level control system was not explicitly modeled because it is not credited for mitigating this transient.

Steam generator level control is indirectly modeled in the LOFTRAN code as the water mass in the steam generator.

The initial and nominal conditions were both specified as the mass at [ ]a~c of power (i.e. nominal conditions).

9. For the purposes of this analysis, it was assumed that all the pressurizer heaters (backup and proportional), were functioning properly, providing a total capacity of [ ]a'c kW (Reference 5). It should be noted that the peak pressurizer pressure occurs at the initial LTR-SCS-07-14-NP Attachment Page 3 of 10 stages of the transient; the heaters are not actuated during this time and as installed capacity will not have any impact on the peak pressurizer pressure.10. The pressurizer level program low and high setpoints were provided in Reference 5, as a function of no-load and full load Tavg, respectively.

It is assumed that the pressurizer level program varies linearly as a function of Tavg; this assumption is consistent with Westinghouse's pressurizer level program methodology.

11. The steam dump capacity is input to the code based on the nominal steam pressure.

The conservative steam dump capacity corresponding to [ ]a,c SGTP and low Tav.g was used in the analysis.

The LOFTRAN code compensates the steam dump capacity due to the steam pressure response during the turbine trip without reactor trip transient.

12. All control systems are functioning per design except for the degraded conditions that were analyzed.The LOFTRAN model developed in Reference 5 was used in this analysis.

The steam dump valve model was refined to model all four banks explicity and the initial conditions were revised to reflect the 45% power level.Analysis Description An analysis of the turbine trip without reactor trip transient from the P-9 setpoint was performed to determine if the pressurizer PORVs are challenged.

The analysis methods are the same as used for the Stretch Power Uprate, and the P-9 analyses for other Westinghouse Plants. The methods have been reviewed and approved by the NRC as part of the licensing submittals for these plants The turbine trip without a reactor trip transient was initialized from an initial power level of [ ]c with all normal control systems assumed operational.

This best estimate analysis addresses the NRC position in NUREG-0737, Item II.K.3.10 (Reference 1). Note that this analysis does not consider the effects of the anticipatory feature (i.e. the integral action) of the proportional-integral (PI) controller used for the pressurizer pressure control. An evaluation was performed for a similar plant; this evaluation indicated that the impact of the anticipatory feature is on the order of [ ]a,c psi, which is not considered to be significant relative to the overall analysis and conclusions.

Therefore, the same conclusion will be applied within.Additionally, a sensitivity study was performed to consider the effects of potential degraded control systems and to assess various P-9 setpoint values. The degraded cases listed below were analyzed independently.

The three control systems that act to mitigate this transient are the pressurizer pressure control system, rod control system, and steam dump control system. The degradations assumed for each control system are as follows: LTR-SCS-07-14-NP Attachment Page 4 of 10 Pressurizer Pressure Control System* 50% reduction in spray flow capacity (i.e., one spray valve fails to open).As the plant responds to the transient, the pressurizer pressure is increasing, and only one spray valve is functioning to relieve this increase in pressure.Rod Control System" Failure of the power mismatch channel." Failure of the power and temperature mismatch channel.The purpose of the power mismatch channel is to provide a fast signal to the rod control system during a rapid change in turbine load. If this signal is not present, then the rods are controlled by the Tavg error signal, which has a slower response and thus takes longer to drive the rods into the core at maximum speed following the turbine trip.Steam Dump Control System* Worst Credible Single Failure of I Condenser Steam Dump Valve in Bank 1 (1 of 3 total valves failed)* Failure of 1 Condenser Steam Dump Valve in Bank 2 (1 of 3 total valves failed)* Complete Failure of Bank 1 of the Condenser Steam Dump Valves (3 of 3 total valves failed)" Complete Failure of Bank 2 of the Condenser Steam Dump Valves (3 of 3 total valves failed)* Failure of I Condenser Steam Dump Valve in Banks 1, 2,3 and 4 (4 of 12 total valves failed)* Complete Failure of all Condenser Steam Dump Valves in all Banks (12 of 12 total valves failed)* Complete Failure of all Condenser Steam Dump Valves in all Banks with the Atmospheric Dump Valves Credited These degradations reduce or eliminate the steam dump capability, which results in a plant heatup.Acceptance Criteria The acceptance criterion for the turbine trip without a reactor trip best estimate transient from the P-9 setpoint is that the pressurizer PORVs are not challenged.

The pressurizer PORV setpoint is]a,c psia (Reference 6). Thus, if the pressurizer pressure is equal to or greater than [ ]a,c psia, then the pressurizer PORVs would be challenged.

For the degraded control system analyses, it is desired to satisfy the same acceptance criteria as the best estimate analysis; however, these analyses are to provide information regarding the sensitivities to the degraded systems. These degraded control system analyses are not required to satisfy the NRC position in NUREG-0737, Item I1.K.3.10 (Reference 1).The steam generator steam pressure results were further evaluated to determine if the SG safety valves were challenged during the transient.

The first safety valve setpoint is [ ]a C psig or]a,c psia (Reference 6). Thus, if the steam generator steam pressure is equal to or greater LTR-SCS-07-14-NP Attachment Page 5 of 10 than [ ]a~c psia, then the first safety valve would be challenged.

Results For the best estimate case at normal plant conditions with all control systems functioning per design the results were acceptable, See Table 1. The pressurizer PORVs and the steam generator safety valves were not challenged following the turbine trip without reactor trip transient.

The best estimate case modeled the current load rejection trip controller settings for the steam dump control system with Banks 1 & 2 and Banks 3 & 4 set to trip-open at [ ]ac OF and [ ]a,. °F, respectively, for low Tavg (Reference 6). The steam dump condenser valves modulate open with a proportional gain of [ ]a'c %/OF.The results of the degraded control systems analyses are summarized in Table 2. Note, the results in Tables 1 and 2 do not include the impact of the anticipatory feature which would increase the values by about [ ],,c psi. The results show that the pressurizer PORVs and the steam generator safety valves are not challenged except for the case with all steam dump banks failed.Challenging the pressurizer PORVs and the steam generator safety valves with all steam dumps failed is expected because there is no heat sink for the primary side under this scenario.

The steam dump system is considered highly reliable.

The steam dumps are used during plant startup.A failure of all steam dump banks would be detected prior to synchronization and power ascension.

The results presented for all steam dumps failed are consistent with the results for similar analyses from Westinghouse 4-loop plants. Based on similar analyses for other Westinghouse plants, the PORVs have been challenged with all steam dump banks failed and does not violate any design or licensing criteria., For the case with all steam dumps failed, steam will be released through the PORVs and overfilling the pressurizer will not occur. This Condition 2 event will not initiate a Condition 3.Thus, the requirements for the NRC position in NUREG-0737, Item II.K.3.10 (Reference

1) for the best-estimate (i.e. nominal) case are satisfied.

With all condenser steam dumps failed and crediting the atmospheric dump valves ([ ]a,c capacity), the results show the pressurizer PORVs and the steam generator safety valves are challenged.

For this transient, primary side heats up faster and the pressurizer PORVs open first, followed a few seconds later by the steam generator safety valves opening. The capacity of the atmospheric dump valves is not adequate to prevent the steam generator safety valves from opening.An analysis was also performed for the case with all steam dumps failed to determine the maximum power level at which the pressurizer power operated relief and steam generator safety valves are not challenged.

The maximum power level was determined to be [ ]a c 20%.Although there are no explicit uncertainty calculations required for technical specification permissives, the protection system methodology as defined in Reference 8 was applied to the determination of the Allowable value (AV) for permissives.

This approach is consistent with the recent technical specification changes for setpoints as part of the recent Seabrook Stretch Power Uprate program.LTR-SCS-07-14-NP Attachment Page 6 of 10 In Reference 8, the AV method is defined as the Nominal Trip Setpoint (NTS) +/- Rack Calibration Accuracy (RCA). The RCA is further defined as the calibration tolerance of the P-9 permissive bistable as controlled by the plant calibration procedures.

Westinghouse reviewed the plant procedures for the Nuclear Instrumentation System (NIS) which includes the P-9 permissive bistable actuation and determined that the calibration accuracy for the P-9 permissive bistable is l]ac Reactor Thermal Power (RTP). Therefore, the AV for the P9 permissive is defined as[ ]ac.LTR-SCS-07-14-NP Attachment Page 7 of 10 Table 1 Turbine Trip without Reactor Trip Analysis -Best Estimate Pressurizer Steam Pressure Pressure (psia) (psia)Maximum Maximum I Sa.c Table 2 Turbine Trip without Reactor Trip Analysis -Degraded Control Systems Pressurizer Steam Pressure Pressure (psia) (psia)Maximum Maximum-ja,c LTR-SCS-07-14-NP Attachment Page 8 of 10 Non-LOCA Safety Evaluation To increase the P-9 setpoint, an evaluation of the loss of load/turbine trip transients initiated from a nominal level of 45% must be performed.

The following evaluation demonstrates that the consequences of a turbine trip event from 45% power without the anticipatory reactor trip is bounded by the analyses presented in the Seabrook UFSAR for the loss of load/turbine trip and complete loss of reactor flow events (URSAR Sections 15.2.3 and 15.3.2).When a reactor trip on a turbine trip signal is blocked the safety analysis concern focuses on a turbine trip transient followed by a loss of reactor coolant flow. Reactor protection for the loss of reactor coolant flow event is based on the assumption that the loss of flow event is initiated while the plant is operating at normal operating conditions.

If a reactor trip due to an overpower, overtemperature or low pressure condition were to immediately precipitate a loss of reactor coolant flow event, the accident consequences could be more severe than those reported for the design basis analyses presented in USFSAR Chapter 15. If electrical power for the rector pumps is lost after [ ]a,c seconds following the reactor trip, the subsequent loss of reactor coolant flow is not considered serious because the reactor has already been shutdown.

At the Seabrook station, this is not a concern. With the electrical distribution configuration, a turbine trip initiates a generator circuit breaker trip, transferring unit auxiliary load to the preferred offsite power supply without any distribution bus transfer action and power to the reactor coolant pumps is maintained by reverse power through the offsite power supply without interruption.

The loss of load/turbine trip analyses for Seabrook station are presented in section 15.2.3 of the UFSAR. The analysis presents two cases. In one case, the behavior of the unit is evaluated to demonstrate the adequacy of the pressure relieving devices to limit the reactor coolant pressure to 100% of the design value. The pressurizer pressure control system is not credited in the analysis.The reactor is tripped from full power without pressure control on a high pressurizer pressure trip function approximately

[ ]a.c seconds after the start of the event. If the reactor were tripped from a power level just below the P-9 setpoint (45% power), the system pressure transient will be significantly less severe than that observed for the full power case. The reactor trip on high pressure would still occur but would be delayed approximately

[ ]a,c to [ ]a.c seconds beyond that in the full power case. Therefore, the full power analyses bounds those for part power.The second case analyzes the accident with respect to determining the minimum core Departure from Nucleate Boiling (DNB) ratio. In this analysis, the pressurizer pressure system is credited and is assumed to limit the reactor coolant pressure which minimizes the core DNB ratio. This second case also represents the limiting transient with respect to peak steam generator pressure because it usually results in a longer time to reactor trip. Both cases are analyzed from full power conditions with minimum reactivity feedback.

No credit is assumed for anticipatory reactor trip signal on turbine trip. The pressurizer pressure control system initiates a trip from full power on overtemperature Delta-T trip function approximately

[ ]a,c seconds after the initiation of the event.If this case was analyzed at a power level just below the P-9 setpoint (45% power), the reactor trip setpoint may not be reached. With the assumption of minimum reactivity feedback, the core power level will remain at the initial power level. This lower power level results in a substantial increase in margin to DNBR limit and results would be less severe than those observed in the full power case as documented in the UFSAR.Increasing the P-9 interlock setpoint from 20% to 45% will have no impact on the safety analyses for the inadvertent operation of emergency core cooling system during power operation.

The P-9 LTR-SCS-07-14-NP Attachment Page 9 of 10 interlock allows the anticipatory reactor trip on turbine trip to be blocked when power is below the P-9 setpoint.

It should be noted that none of the design basis analyses presented in Chapter 15 of the UFSAR simulate or take credit for the operation of the reactor trip on turbine trip protection function.

The P-9 interlock has no effect on a turbine trip signal generated on a reactor trip.Increasing the P-9 setpoint from 20% to 45% rated thermal power is acceptable with respect to non-LOCA design basis accident analyses.

The consequences of loss of load/turbine trip events below 50% power without anticipatory reactor trip on turbine trip are bounded by the full power analyses presented in the Seabrook Station UFSAR, Sections 15.2.3 and 15.3.2.Conclusions The turbine trip without reactor trip transient analysis from the current P-9 setpoint of 45% power concluded that the pressurizer and steam generator PORVs will not be challenged for the best estimate evaluation (i.e. all control systems performing as designed).

All degraded control systems senarios analyzed are acceptable except for the case with all steam dump banks failed.The pressurizer PORVs and steam generator safety valves are challanged in this case only.No changes to control system setpoints or parameters are required for the setpoint change from 20% to 45% of full power.References

1. NUREG-0737, "Clarification of TMI Action Plan Requirements," Item II.K.3.10, Proposed Anticipatory Trip Modification, October, 1980.[]ac 3. FPL Energy Contract No. 02207552, "P9 Setpoint Change," September 5, 2006.[4. ]~-5.6. FP 55837 Issue No. 023, NSSS Precautions Limitations and Setpoints for Seabrook Station, Revision 2, July 24, 2004.7. FPL Letter No CE-06-35, "l&C Inputs for P-9 Setpoint Change," September 22, 2006.[8. ]a,c LTR-SCS-07-14-NP Attachment Page 10 of 10 Attachment 6 Application for Withholding Proprietary Information from Public Disclosure

• Westinghouse U.S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555-0001 Westinghouse Electric Company Nuclear Services P.O. Box 355 Pittsburgh, Pennsylvania 15230-0355 USA (412) 374-4643 (412) 374-4011 greshaja@westinghouse.com Direct tel: Direct fax: e-mail: Our ret CAW-07-2232 January 24, 2007 APPLICATION FOR WITHHOLDING PROPRIETARY INFORMATION FROM PUBLIC DISCLOSURE

Subject:

LTR-SCS-07-14-P Attachment, January 2007, "Engineering Report Seabrook Nuclear Power Station Turbine Trip without Reactor Trip Transient from the P-9 Setpoint Analysis" (Proprietary)

Withholding the proprietary information in the above-referenced report is requested by Westinghouse Electric Company, LLC (owner of the information) as identified in the accompanying and attached Affidavit CAW-07-2232.

This affidavit 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 affidavit by Florida Power and Light Energy for the Seabrook Nuclear Power Station.Correspondence with respect to the proprietary aspects of the application for withholding or the Westinghouse affidavit should reference this letter, CAW-07-2232 and be addressed to J. A. Gresham, Manager, Regulatory Compliance and Plant Licensing, Westinghouse Electric Company LLC, P.O. Box 355, Pittsburgh, Pennsylvania 15230-0355.

Very truly yours, J. A. Gresham, Manager Regulatory Compliance and Plant Licensing Enclosures cc: Jon Thompson (NRC O-7E1A)

CAW-07-2232 bcc: J. A. Gresham IL R. Bastien, IL (Nivelles, Belgium)C. Brinkman, IL (Westinghouse Electric Co., 12300 Twinbrook Parkway, Suite 330, Rockville, MD 20852)RCPL Administrative Aide (ECE 4-7A) IL, 1A (letter and affidavit only)

CAW-07-2232 AFFIDAVIT COMMONWEALTH OF PENNSYLVANIA:

ss COUNTY OF ALLEGHENY:

Before me, the undersigned authority, personally appeared B. F. Maurer, 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: B. F. Maurer, Principal Engineer Regulatory Compliance and Plant Licensing Sworn to and subscribed before me this J day of January, 2006 Notary Public Nolarial SealI Sharon L. Fori, Notary Public Monrooville Boro, Allegheny County My Commission Exorr.c January 29,2007 Member. Penn)d;Vi6r'A 4;:fcitIun Or Notaries 2 CAW-07-2232 (1) 1 am Principal Engineer, Regulatory Compliance and Plant Licensing, in Nuclear Services, 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) 1 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" accompanying this Affidavit.

(3) 1 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 (b)(4) 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 compinnies, 3 CAW-07-2232 (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.

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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, any one component may be the key to the entire puzzle, thereby depriving Westinghouse of a competitive advantage.

4 CAW-07-2232 (e) Unrestricted disclosure would jeopardize the position of prominence of Westinghouse in the world market, and thereby give a market advantage to the competition of those 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 LTR-SCS-07-14-P Attachment, January 2007, "Engineering Report Seabrook Nuclear Power Station Turbine Trip without Reactor Trip Transient from the P-9 Setpoint Analysis" (Proprietary), being transmitted by the Florida Power and Light Energy letter and Application for Withholding Proprietary Information from Public Disclosure, to the Document Control Desk. The proprietary information as submitted by Westinghouse for the Seabrook Nuclear Power Station is expected to be applicable for other licensee submittals in response to certain NRC requirements for justification of increasing the turbine trip without reactor trip permissive (P-9) setpoint value.This information is part of that which will enable Westinghouse to: (a) Analyze turbine trip without reactor trip transient scenarios.(b) Draw specific conclusions and results for individual applications.

Further this information has substantial commercial value as follows: (a) Westinghouse plans to sell the use of similar information to its customers for purposes of analyzing turbine trip without reactor trip transient scenarios.

5 CAW-07-2232 (b) Westinghouse can sell support and defense of analyses related to turbine trip without reactor trip transient evaluations.(c) The information requested to be withheld reveals the distinguishing aspects of a methodology which was developed by Westinghouse.

Public disclosure of this proprietary information is likely to cause substantial harm to the competitive position of Westinghouse because it would enhance the ability of competitors to provide similar transient analyses and evaluations and licensing defense services for commercial power reactors without commensurate expenses.

Also, public disclosure of the information would enable others to use the information to meet NRC requirements for licensing documentation without purchasing the right to use the information.

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