ML060760535
| ML060760535 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 02/03/2006 |
| From: | Westinghouse |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| FOIA/PA-2010-0209 AEP-05-107, Rev 1 | |
| Download: ML060760535 (22) | |
Text
Westinghouse Non-Proprietary Class 3 Engineering Report for D. C. Cook Units 1 & 2 Turbine Trip without Reactor Trip Transient from the P-8 Setpoint Analysis AEP-05-107 Revision 1 NP-Attachment February 3, 2006 Westinghouse Electric Company LLC P.O. Box 355 Pittsburgh, PA 15230-0355 2006 Westinghouse Electric Company LLC All Rights Reserved Page 1 of 6 AEP-05-107 Revision 1 NP-Attachment
Introduction Westinghouse has performed an evaluation for the Reactor Protection System (RPS) modification to change the reactor trip on turbine trip interlock from permissive P-7 to permissive P-8 at D. C. Cook Units I and 2. This evaluation has been performed to determine the impact of a turbine trip without reactor trip on the potential actuation of a pressurizer Power Operated Relief Valve (PORV).
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 II-K.3.l0 (Reference 6).
In order to satisfy the NRC position, a best estimate plant specific analysis was performed to demonstrate that the pressurizer PORVs would not lift following the change in the reactor trip on turbine trip protection interlock from the P-7 to P-8 permissive setpoint. The current P-8 setpoint is [ ]"c % of rated power (Reference 2, Item Al 1). If the PORVs are challenged on a turbine trip from the existing P-8 setpoint, then the P-8 setpoint will be reduced until the pressurizer PORVs are not challenged.
Additionally, Control Systems Failure Mode Analyses were also performed to determine the effect of degraded control systems on the analyses results.
Key Assumptions and Inputs A best estimate analytical study was performed to determine the transient plant response to a turbine trip without reactor trip transient. The analysis was performed using a LOFTRAN computer code (Reference
- 1) model of the D. C. Cook Units 1 and 2. The LOFTRAN computer code was previously used with best estimate methodology for the D. C. Cook Units I and 2 uprate. This computer model simulates the overall thermal-hydraulic and nuclear response of the NSSS as well as the various control and protection systems.
Since the object of this study was primarily to determine the peak in pressurizer pressure following the initiation of the transient, assumptions were mad-that would contribute to a conservatively high prediction of pressurizer pressure. These assumptions are as follows:
- 1. This detailed analysis was performed for Unit 2; however, it is also applicable to Unit 1. The results of steam dump capacity analyses performed for Units I and 2 (References 3 and 4, respectively) were compared, and Unit 2 provided the more limiting margin (to the overtemperture/overpower reactor trip setpoints) than Unit 1. Also, Unit 2 operates at a higher pressurizer pressure than Unit 1. Therefore, it was assumed that Unit 2 would provide higher pressurizer pressure responses than Unit 1.
- 2. The turbine was tripped from [ ]c % ofthe full Nuclear Steam Supply System (NSSS) power,
[
]a.C MWt (Reference 2, Item Al) after [ ]"c seconds of steady state conditions; this includes [
]I c for conservatism on the current P-8 setpoint of [
]ac % of power (Reference 2, Item All). This [
]Bc conservatism is Westinghouse best estimate methodology.
- 3. The transient was analyzed for a range of the full power vessel average temperature (Tavg) from
[
]a cF to [
]asc 'F (Reference 2, Item A4). Note that the [
]I.c Tavg will result in the smallest temperature error for the steam dump system, and therefore, the largest plant heat up.
Also note that of the two steam pressures provided in Reference 2 ([
]ac psia versus [
]"
psia; Item A7), the [
]ac steam pressure ([
]' c psia) was modeled because a [
]axc steam pressure provides a greater pressurizer pressure response.
AEP*05-107 Revision 1 NI'-Attacflment Page 2 0? 6 AEP-05-107 Revision 1 NP-Attachnient Page 2 of 6
- 4. Best estimate Beginning-Of-Life (BOL) reactivity parameters for Unit 2 from Cycle 15 were used. BOL reactivity parameters have lower differential rod worth and the least negative moderator temperature coefficient and thus, using BOL parameters in the analysis yield
[
]ac results, which bound the full cycle of operation.
- 5. Initial primary and secondary side conditions are at [
]aC % of the rated power; these conditions, such as, Tavg and pressure, [
]aC (i.e. best estimate analysis).
- 6. The [
arC overall heat transfer coefficients from fuel to coolant were used, consistent with BOL conditions (Same as in Reference 3).
- 7. The Rod control system was in the automatic mode of control; the turbine trip transient is a load decrease, therefore the rods are inserted to mitigate the transient.
- 8.
The Pressurizer Pressure control system, Steam Dump control system (Turbine Trip Controller; Reference 2, Item B22) and Steam Generator (SG) Level control system were assumed operational and in the automatic mode of control. Steam dump to the atmosphere relief valves is not credited in the analysis. Since the analyses were best estimate, the parameter values including the control systems setpoints and PORV setpoints were assumed at [
]aC conditions
[
]aC
- 9. SG level control is modeled in the LOFTRAN code [
]-. The initial and nominal conditions were both specified [
]ac [
I.
.]a,c
- 10. 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 [
]c kW (Reference 2; Item D14, 15 and 16).
- 11. In order to develop the pressurizer spray capacity, it was assumed that the difference between Tavg and the cold leg temperature (Tcold) is linear with power level and the Reactor Coolant System (RCS) flow.
A majority of the fluid systems thermal-hydraulic input data was taken from the steam dump capacity analysis performed for the power uprate program (Reference 3). The nominal operating conditions (i.e.
steam flow, steam pressure, Tavg, pressurizer level) and the settings (i.e. the gains and time constants) for the Steam Dump, Pressurizer Pressure, Pressurizer Level, and Rod control systems were taken from Reference 2. The Steam Generator Level control settings were taken from Reference 3. The initial operating conditions had to be developed based on the nominal operating conditions specified in Reference 2. In addition, the steam dump capacity for 6 steam dump valves had to be developed based on a proportional estimation using a capacity defined in Reference 5 for 9 steam dump valves.
AEP-05-107 Revision 1 NP-Attachment Page 3 of 6
Description of Analysis and Evaluations A best estimate analysis for a turbine trip without reactor trip transient from the P-8 setpoint was performed to determine if the pressurizer PORVs are challenged. The turbine trip without a reactor trip transient was initialized from an initial power level of [ ]a c % ([
]a.c % power with [
]a c power uncertainty) with all normal control systems assumed operational. This best estimate analysis addresses the NRC position in NUREG-0737, Item II.K.3. I0 (Reference 6).
Additionally, a sensitivity study was performed to consider the effects of potential degraded control systems on the turbine trip without reactor trip transient.
The three control systems that primarily 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:
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.
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 begin to drive the rods into the core at maximum speed following the turbine trip.
Steam Dump Control System
- Failure of 1 Steam Dump Valve (I of 3 total valves failed) in Bank 1 Failure of I Steam Dump Valve (1 of 3 total valves failed) in Bank 2 Failure of I Steam Dump Valve (1 of 3 total valves failed) in each of Banks I & 2 These degradations reduce the total steam dump capacity, thus increasing the plant heat up.
Acceptance Criteria and Results The acceptance criterion for the turbine trip without a reactor trip transient from the P-8 setpoint is that the pressurizer PORVs are not challenged. The pressurizer PORV setpoint is 2350 psia (Reference 2, Item D17). Thus, if the pressurizer pressure is equal to or greater than 2350 psia, then the pressurizer PORVs would be challenged.
The best estimate analysis results for the turbine trip without the reactor trip transient from the P-8 setpoint are summarized in Tables 1 and 2. The analyses were performed at [
]C Tavg conditions (i.e.
based on a full load Tavg of [
]a"c 'F) with various RTD response times ([
]aC seconds) in order to provide flexibility to the applications of this analysis in case the actual RTD response time
([
Iac seconds; Reference 2, Item Al Op) is changed as a result of the future RTD Bypass Elimination project for D. C. Cook Units I and 2.
AEP-05-107 Revision 1 NP-Attachment Page 4of 6
Table 1 l
I I]
2,c
[
] arc Although, the response times of [
ra c seconds provide the same maximum pressurizer pressure, a response time of [
]ac seconds (Case 3 from Table 1) was used in the best estimate analysis for [
]Pc Tavg conditions (i.e. based on a full load Tavg of [
]3 c 'F). The results of this analysis are shown in Table 2.
Table 2 The "degraded" control system sensitivity analysis results for the turbine trip without the reactor trip transient from the P-8 setpoint are summarized in Table 3 for [
Iac Tavg conditions. Note that these degraded control system analyses were performed using the [
]" RTD response time of [
]ac seconds.
Table 3 1
F IC I~~~
I_________________________
__________________________________________l_
a,C l
Based on the results in Table 3, the PORVs would not be challenged with degradations assumed in the Pressurizer Pressure, Rod, and Steam Dump control systems.
AEP-05-107 Revision 1 NP-Attachmnent Page 5 of 6
Conclusions The turbine trip without reactor trip transient analysis from the current P-8 setpoint of [
]atC % power concluded that the pressurizer PORVs will not be challenged with best estimate simulation (i.e. all control systems performing as designed). Therefore, it is acceptable to change the reactor trip on turbine trip interlock from the P-7 to P-8 permissive setpoint.
Furthermore, the pressurizer PORVs will not be challenged during a turbine trip without reactor trip transient with degraded conditions in the Pressurizer Pressure control system (such as 50 % of nominal spray flow), the Rod control system (power mismatch failure) and the Steam Dump control system (failure of one steam dump valve in Bank 1, Bank 2 and Banks I and 2).
References
- 1.
WCAP-7907-P-A, "LOFTRAN Code Description," April 1984.
- 2.
AEP Correspondence DIT-B-03041-00 "Donald C. Cook Nuclear Plant Turbine Trip to Reactor Trip Permissive Change Analysis," T. E. Anderson, October 11, 2005.
- 3.
Westinghouse Letter, AEP-03-41, "American Electric Power Donald C. Cook Nuclear Plant Unit 2 Steam Dump/Margin to Trip Analysis Revision to Address Cycle 14 Fuel Axial Offset Deviation (AOD)," W. R. Rice, May 21, 2003.
- 4.
Westinghouse Letter, AEP-02-189, "American Electric Power Donald C. Cook Nuclear Plant Unit 1 Steam Dump/Margin to Trip Analysis," W. R. Rice, December 19, 2002.
- 5.
Westinghouse Calculation Note, CN-SEE-02-35, Revision 1, "D. C. Cook Unit 2 NSSS/BOP Interface Evaluation for the 2.0% Power Uprate Program," M. F. McGuire, February, 2003.
- 6.
NUREG-0737, "Clarification of TMI Action Plan Requirements," Item II.K.3.10, Proposed Anticipatory Trip Modification, October, 1980.
AEP-05-107 Revision 1 HP-Attachment Page 6 ct 6 AEP-05-107 Revision 1 HP-Attachment Page 6 cof 6 A to AEP:NRC:6331 DONALD C. COOK NUCLEAR PLANT UNIT I TECHNICAL SPECIFICATION PAGES MARKED TO SHOW CHANGES 3.3.1-13
RTS Instrumentation 3.3.1 Table 3.3.1-1 (page 3 of 6)
Reactor Trip System Instrumentation APPLICABLE MODES OR OTHER SPECIFIED REQUIRED SURVEILLANCE FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS ALLOWABLE VALUE
D(f)
SR 3.3.1.1 SR 3.3.1.11 SR 3.3.1.13 2 9.7%
Coincident with Steam Flow/Feedwater Flow Mismatch (per SG) 1,2 2
D(f)
SR 3.3.1.1 SR 3.3.1.11 SR 3.3.1.13 s 0.73E6 lb/hr steam flow at RTP
- 16. Turbine Trip
- a. Low Fluid Oil Pressure
- b. Turbine Stop Valve Closure (per train)
- 18. Reactor Trip System Interlocks 1
1,2 3
4 2 trains D
SR 3.3.1.13 SR 3.3.1.18 J1 SR 3.3.1.6 SR 3.3.1.19 2 750 psig 2 1% open NA
- a.
Intermediate Range Neutron Flux, P-6 2 (d), 3(), 4(), 5()
2 L
SR 3.3.1.14 SR 3.3.1.16 26E-11 amp
- b.
Low Power Reactor Trips Block, P-7
- c.
Power Range Neutron Flux, P-8
- d.
Power Range Neutron Flux, P-10 1
I 1 per train 4
L SR 3.3.1.5 L
SR 3.3.1.14 SR 3.3.1.16 NA s 31% RTP 1,2 4
L SR 3.3.1.14 SR 3.3.1.16 2 9% RTP and s 11% RTP
- e. Turbine First Stage Pressure, P-13 1
2 L
SR 3.3.1.1 SR 3.3.1.13 SR 3.3.1.16 s 37 psig (a) With Rod Control System capable of rod withdrawal or one or more rods not fully inserted.
(d) Below the P-6 (Intermediate Range Neutron Flux) interlock.
(c) Abovc the P 7 (Low Power Reactor Trips Block) intWreFG (f) Separate condition entry is allowed per SG for only 1 of the 4 total Reactor Trip System Instrumentation Function 15 channels inoperable on each SG (i.e., for only 1 of 2 SG Water Level - Low channels or 1 of 2 Steam Flow/Feedwater Flow Mismatch channels inoperable on each SG). Any combination of 2 or more inoperable Reactor Trip System Instrumentation Function 15 channels on any SG requires immediate entry into LCO 3.0.3.
ZI.5:V:.l.".i,5::ilZI.:ri4;w55fcn,.:2...'fi:51.45'..".
Cook Nuclear Plant Unit 1 3.3.1-13 Amendment No. 287
Attachment l'B to AEP:NRC:6331 DONALD C. COOKS NUCLEAR PLANT UNIT 2 TECHNICAL SPECIFICATION PAGES MARKED TO SHOW CHANGES 3.3.1-13
RTS Instrumentation 3.3.1 Table 3.3.1-1 (page 3 of 6)
Reactor Trip System Instrumentation APPLICABLE MODES OR OTHER SPECIFIED REQUIRED SURVEILLANCE FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS ALLOWABLE VALUE
D(f)
SR 3.3.1.1 SR 3.3.1.11 SR 3.3.1.13 D(D SR 3.3.1.1 SR 3.3.1.11 SR 3.3.1.13 225.0%
Coincident with Steam Flow/Feedwater Flow Mismatch (per SG) 1,2 2
s 1.56E6 lb/hr steam flow at RTP
- 16. Turbine Trip
- a. Low Fluid Oil Pressure b.. Turbine Stop Valve Closure (per train)
- 18. Reactor Trip System Interlocks i 1§hf 1(6 1,2 3
4 2 trains D
SR 3.3.1.6 SR 3.3.1.19 2 57 psig 2 1% open NA
- a.
Intermediate Range Neutron Flux, P-6 2, 3(a),4( ), 5(a) 2 L
SR 3.3.1.14 SR 3.3.1.16 2 6E-11 amp
- b. Low Power Reactor Trips Block, P-7
- c.
Power Range Neutron Flux, P-8 1
1 1 per train 4
L SR 3.3.1.5 L
SR 3.3.1.14 SR 3.3.1.16 NA s 31% RTP
- d.
Power Range Neutron Flux, P-10 1,2 4
L SR 3.3.1.14 SR 3.3.1.16 2 9% RTP and f 11% RTP
- e.
Turbine First Stage Pressure, P-13 1
2 L
SR 3.3.1.1 SR 3.3.1.13 SR 3.3.1.16 s 51 psig (a) With Rod Control System capable of rod withdrawal or one or more rods not fully inserted.
(d) Below the P-6 (Intermediate Range Neutron Flux) interlock.
(c) Above the P 7 (Low Power Reactor Trips Block) inter6ct (f) Separate condition entry is allowed per SG for only 1 of the 4 total Reactor Trip System Instrumentation Function 15 channels inoperable on each SG (i.e., for only 1 of 2 SG Water Level - Low channels or 1 of 2 Steam Flow/Feedwater Flow Mismatch channels inoperable on each SG). Any combination of 2 or more inoperable Reactor Trip System Instrumentation Function 15 channels on any SG requires immediate entry into LCO 3.0.3.
- tWNW, K
MiWRtie(
sa Cook Nuclear Plant Unit 2 3.3.1-13 Amendment No. 269 A to AEP:NRC:6331 DONALD C. COOK. NUCLEAR PLANT UNIT 1 TECHNICAL SPECIFICATION PAGES WITH THE PROPOSED CHANGES INCORPORATED 3.3.1-13
RTS Instrumentation 3.3.1 Table 3.3.1-1 (page 3 of 6)
Reactor Trip System Instrumentation APPLICABLE MODES OR OTHER SPECIFIED REQUIRED SURVEILLANCE FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS ALLOWABLE VALUE
- 15. SG Water Level - Low 1,2 2
D(f SR 3.3.1.1 2 9.7%
(per SG)
SR 3.3.1.11 SR 3.3.1.13 Coincident with Steam 1,2 2
D(0 SR 3.3.1.1 5 0.73E6 lb/hr Flow/Feedwater Flow SR 3.3.1.11 steam flow at RTP Mismatch (per SG)
SR 3.3.1.13
- 16. Turbine Trip
- a. Low Fluid Oil Pressure 1(h) 3 D
SR 3.3.1.13 2750 psig SR 3.3.1.18
- b.
Turbine Stop Valve 1(h) 4 D
SR 3.3.1.13
Ž1% open Closure (per train)
SR 3.3.1.18
- 17. Safety Injection (SI) Input 1,2 2 trains J
SR 3.3.1.6 NA from Engineered Safety SR 3.3.1.19 Feature Actuation System (ESFAS)
- 18. Reactor Trip System Interlocks
- a.
Intermediate Range 2(d) 3( a)
)4() 5(a) 2 L
SR 3.3.1.14 26E-11 amp Neutron Flux, P-6 SR 3.3.1.16
- b.
Low Power Reactor 1
1 per train L
SR 3.3.1.5 NA Trips Block, P-7
- c.
Power Range Neutron 1
4 L
SR 3.3.1.14 s 31% RTP Flux, P-8 SR 3.3.1.16
- d. Power Range Neutron 1,2 4
L SR 3.3.1.14 2 9% RTP and Flux, P-10 SR 3.3.1.16 5 11% RTP
- e. Turbine First Stage 1
2 L
SR 3.3.1.1 5 37 psig Pressure, P-13 SR 3.3.1.13 SR 3.3.1.16 I
I (a) With Rod Control System capable of rod withdrawal or one or more rods not fully inserted.
(d) Below the P-6 (Intermediate Range Neutron Flux) interlock.
(f) Separate condition entry is allowed per SG for only 1 of the 4 total Reactor Trip System Instrumentation Function 15 channels inoperable on each SG (i.e., for only 1 of 2 SG Water Level - Low channels or 1 of 2 Steam Flow/Feedwater Flow Mismatch channels inoperable on each SG). Any combination of 2 or more inoperable Reactor Trip System Instrumentation Function 15 channels on any SG requires immediate entry into LCO 3.0.3.
(h) Above the P-8 (Power Range Neutron Flux) Interlock.
I Cook Nuclear Plant Unit 1 3.3.1-13 Amendment No. I&,
B to AEP:NRC:6331 DONALD C. COOK NUCLEAR PLANT UNIT 2 TECHNICAL SPECIFICATION PAGES WITH THE PROPOSED CHANGES INCORPORATED 3.3.1-13
RTS Instrumentation 3.3.1 Table 3.3.1-1 (page 3 of 6)
Reactor Trip System Instrumentation APPLICABLE MODES OR OTHER SPECIFIED REQUIRED SURVEILLANCE FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS ALLOWABLE VALUE
- 15. SG Water Level - Low 1,2 2
D(t SR 3.3.1.1 2 25.0%
(per SG)
SR 3.3. 1.11 SR 3.3.1.13 Coincident with Steam 1,2 2
D(O SR 3.3.1.1
SR 3.3.1.13
- 16. Turbine Trip
- a. Low Fluid Oil Pressure 1(h) 3 D
SR 3.3.1.13 2 57 psig SR 3.3.1.18
- b.
TurbineStopValve 1(h) 4 D
SR 3.3.1.13 2 1% open Closure (per train)
SR 3.3.1.18
- 17. Safety Injection (SI) Input 1,2 2 trains J
SR 3.3.1.6 NA from Engineered Safety SR 3.3.1.19 Feature Actuation System (ESFAS)
- 18. Reactor Trip System Interlocks
- a.
Intermediate Range 2 (d), 3(a), 4(a) 5(a) 2 L
SR 3.3.1.14
Ž 6E-11 amp Neutron Flux, P-6 SR 3.3.1.16
- b.
Low Power Reactor 1
1 per train L
SR 3.3.1.5 NA Trips Block, P-7
- c.
Power Range Neutron 1
4 L
SR 3.3.1.14 s 31% RTP Flux, P-8 SR 3.3.1.16
- d.
Power Range Neutron 1,2 4
L SR 3.3.1.14 2 9% RTP and Flux, P-10 SR 3.3.1.16
- 11% RTP
- e. Turbine First Stage 1
2 L
SR 3.3.1.1 s 51 psig Pressure, P-13 SR 3.3.1.13 SR 3.3.1.16 (a) With Rod Control System capable of rod withdrawal or one or more rods not fully inserted.
(d) Below the P-6 (Intermediate Range Neutron Flux) interlock.
(f) Separate condition entry is allowed per SG for only 1 of the 4 total Reactor Trip System Instrumentation Function 15 channels inoperable on each SG (i.e., for only 1 of 2 SG Water Level - Low channels or 1 of 2 Steam Flow/Feedwater Flow Mismatch channels inoperable on each SG). Any combination of 2 or more inoperable Reactor Trip System Instrumentation Function 15 channels on any SG requires immediate entry into LCO 3.0.3.
(h) Above the P-8 (Power Range Neutron Flux) Interlock.
I Cook Nuclear Plant Unit 2 3.3.1-13 Amendment No. 26, to AEP:NRC:6331 APPLICATION FOR WITHHOLDING PROPRIETARY INFORMATION FROM PUBLIC DISCLOSURE
OWestinghouse Westinghouse Electric Company Nuclear Services P.O. Box 355 Pittsburgh, Pennsylvania 1 5230-0355 USA U.S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555-0001 Direct tel:
Direct fax:
e-mail:
(412) 3744419 (412) 374-4011 maurerbf(westinghouse.com Our ref: CAW-06-2098 February 3, 2006 APPLICATION FOR WITHHOLDING PROPRIETARY INFORMATION FROM PUBLIC DISCLOSURE
Subject:
AEP-05-107, Rev. I P-Attachment, "Engineering Report for D. C. Cook Units I & 2 Turbine Trip without Reactor Trip Transient from the P-8 Setpoint Analysis," (Proprietary)
The proprietary information for which withholding is being requested in the above-referenced report is further identified in Affidavit CAW-06-2098 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 American Electric Power.
Correspondence with respect to the proprietary aspects of the application for withholding or the Westinghouse affidavit should reference this letter, CAW-06-2098, and should be addressed to B. F. Maurer, Acting Manager, Regulatory Compliance and Plant Licensing, Westinghouse Electric Company LLC, P.O. Box 355, Pittsburgh, Pennsylvania 15230-0355.
Very truly yours, B. F. Maurer, Acting Manager Regulatory Compliance and Plant Licensing Enclosures cc: B. Benney L. Feizollahi A BNFL Group company
CAW-06-2098 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, Acting Manager Regulatory Compliance and Plant Licensing Sworn to and subscribed before me this day of 7
,2006 Notary Public Notarial Seal Sharon L. Fiori, Notary Public Monroeville Boro, Allegheny County My Commission Expires January 29,2007 Member. Pennsylvania Association Of Notaries
2 CAW-06-2098 (1)
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(2)
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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 AEP-05-107, Rev. I P-Attachment, "Engineering Report for D. C.
Cook Units I & 2 Turbine Trip without Reactor Trip Transient from the P-8 Setpoint Analysis," (Proprietary), dated January 31, 2006, being transmitted by the American Electric Power Company letter and Application for Withholding Proprietary Information from Public Disclosure, to the Document Control Desk. The proprietary information as submitted by Westinghouse for the Donald C. Cook Units I & 2 Engineering Report is for review and approval.
This information is part of that which will enable Westinghouse to:
(a) Assist the customer by providing an analysis to increase the turbine trip without reactor trip setpoints, as applicable for D. C. Cook Units 1 and 2.
(b) Show that there will be no additional challenges to the pressurizer PORVs as a result of the increase.
Further this information has substantial commercial value as follows:
5 CAW-06-2098 (a)
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 analyses 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.
The development of the technology described in part by the information 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 information, similar technical programs would have to be performed and a significant manpower effort, having the requisite talent and experience, would have to be expended.
Further the deponent sayeth not.
PROPRIETARY INFORMATION NOTICE Transmitted herewith are proprietary and/or non-proprietary versions of documents furnished to the NRC in connection with requests for generic and/or plant-specific review and approval.
In order to conform to the requirements of 10 CFR 2.390 of the Commission's regulations concerning the protection of proprietary information so submitted to the NRC, the information which is proprietary in the proprietary versions is contained within brackets, and where the proprietary information has been deleted in the non-proprietary versions, only the brackets remain (the information that was contained within the brackets in the proprietary versions having been deleted). The justification for claiming the information so designated as proprietary is indicated in both versions by means of lower case letters (a) through (f) located as a superscript immediately following the brackets enclosing each item of information being identified as proprietary or in the margin opposite such information. These lower case letters refer to the types of information Westinghouse customarily holds in confidence identified in Sections (4)(ii)(a) through (4)(ii)(f) of the affidavit accompanying this transmittal pursuant to 10 CFR 2.390(b)(1).
COPYRIGHT NOTICE The reports transmitted herewith each bear a Westinghouse copyright notice. The NRC is permitted to make the number of copies of the information contained in these reports which are necessary for its internal use in connection with generic and plant-specific reviews and approvals as well as the issuance, denial, amendment, transfer, renewal, modification, suspension, revocation, or violation of a license, permit, order, or regulation subject to the requirements of 10 CFR 2.390 regarding restrictions on public disclosure to the extent such information has been identified as proprietary by Westinghouse, copyright protection notwithstanding. With respect to the non-proprietary versions of these reports, the NRC is permitted to make the number of copies beyond those necessary for its internal use which are necessary in order to have one copy available for public viewing in the appropriate docket files in the public document room in Washington, DC and in local public document rooms as may be required by NRC regulations if the number of copies submitted is insufficient for this purpose. Copies made by the NRC must include the copyright notice in all instances and the proprietary notice if the original was identified as proprietary.