RC-15-0025, License Amendment Request - LAR-12-04269, License Basis Changes in Steam Generator Tube Rupture Analysis, Response to Request for Additional Information

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License Amendment Request - LAR-12-04269, License Basis Changes in Steam Generator Tube Rupture Analysis, Response to Request for Additional Information
ML15055A143
Person / Time
Site: Summer South Carolina Electric & Gas Company icon.png
Issue date: 02/12/2015
From: Gatlin T
South Carolina Electric & Gas Co
To: Shawn Williams
Document Control Desk, Office of Nuclear Reactor Regulation
References
CR-12-04269, LAR-12-04269, RC-15-0025
Download: ML15055A143 (33)


Text

. Proprietary Information Thomas D. Gatlin Withhold Under 10 CFR 2.390 Vice President,Nuclear Operation 803.345.4342 February 12, 2015 A SCANA COMPANY RC-1 5-0025 U.S. Nuclear Regulatory Commission Document Control Desk Washington, D.C. 20555-0001 Attn: S. A. Williams

Dear Sir / Madam:

Subject:

VIRGIL C. SUMMER NUCLEAR STATION (VCSNS) UNIT 1 DOCKET NO. 50-395 OPERATING LICENSE NO. NPF-12 LICENSE AMENDMENT'REQUEST - LAR-12-04269 LICENSE BASIS CHANGES IN STEAM GENERATOR TUBE RUPTURE ANALYSIS RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION

References:

1. SCE&G Letter from Thomas D. Gatlin to NRC Document Control Desk, License Amendment Request - LAR-12-04269, "License Basis Changes in Steam Generator Tube Rupture Analysis," dated August 27, 2014

[MLI 4245A408]

2. NRC Letter from Shawn A. Williams to Thomas D. Gatlin, "Virgil C. Summer Nuclear Station, Unit No. 1 - Request for Additional Information Regarding License Basis Changes in Steam Generator Tube Rupture Analysis (TAC No.

MF4699)," dated October 1, 2014 [ML14268A096]

3. SCE&G Letter from Thomas D. Gatlin to NRC Document Control Desk, License Amendment Request - LAR-12-04269, "License Basis Changes in Steam Generator Tube Rupture Analysis Response to Request for Additional Information," dated October 31, 2014 [ML14308A075]
4. NRC Letter from Shawn A. Williams to Thomas D. Gatlin, "Virgil C. Summer Nuclear Station, Unit No. 1 - Request for Additional Information Regarding License Basis Changes in Steam Generator Tube Rupture Analysis (TAC No.

MF4699)," dated January 9, 2015 [ML14339A030]

-South Carolina Electric & Gas Company (SCE&G), acting for itself and as agent for South Carolina Public Service Authority pursuant to 10 CFR 50.90, submitted License Amendment Request (LAR)-per Reference 1 concerning license basis changes in the steam generator tube rupture analysis. NRC review of this request determined that additional information was required-and a request for additional information (RAI) was issued per Reference 4. This submittal's attachment contains SCE&G's response to the RAls dated January 9, 2015. In addition, SCE&G is providing its response to RAI No. 5 per the October 31, 2014 response submittal (Reference 3) to the NRC's request for additional information dated October 1, 2014 (Reference 2).

Enclosure 1 (CD) transmitted herewith contains Proprietary Information.

When separated from Enclosure 1, this document is decategorized. ~QDI Virgil C.Summer Station

  • Post Office Box 88- Jenkinsville, SC. 29065 . F(803) 345-5209

Document Control Desk CR- 2-04269 RC-15-0025 Page 2 of 2 contains one copy of compact disk (CD) entitled, "RADTRAD 3.03 Input and Output Files for V.C. Summer Unit 1 Steam Generator Tube Rupture Analysis," January 2015 (Westinghouse Proprietary). In addition, the Westinghouse Application for Withholding Proprietary Information from Public Disclosure CAW-15-4079, accompanying Affidavit, Proprietary Information Notice, and Copy Right Notice are included in Enclosure 2 of this submittal.

As the CD contains information proprietary to Westinghouse Electric Company LLC, it is supported by an Affidavit signed by Westinghouse, the owner of the information. The 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 Section 2.390 of the Commission's regulations. Accordingly, it is respectfully requested that the information which is proprietary to Westinghouse be withheld from public disclosure in accordance with 10 CFR Section 2.390 of the Commission's regulations.

Correspondence with respect to the copyright or proprietary aspects of the items listed in the above two paragraphs or the supporting Westinghouse Affidavit should reference CAW-15-4079 and should be addressed to James A. Gresham, Manager, Regulatory Compliance, Westinghouse Electric Company, 1000 Westinghouse Drive, Building 3 Suite 310, Cranberry Township, Pennsylvania 16066.

This letter contains no regulatory commitments.

If you have any questions regarding this submittal, please contact Mr. Bruce L. Thompson at (803) 931-5042.

I certify under penalty of perjury that the foregoing is correct and true.

)-)2 - 2e2/

Executed on Thomas D. Gatlin TS/TDG/wm

Attachment:

VCSNS Response to Request for Additional Information

Enclosures:

1. Class 2 RADTRAD 3.03 Input and Output Files for V.C. Summer Unit 1 SGTR Analysis Compact Disk (Proprietary Information)
2. Application for Withholding, Accompanying Affidavit, Proprietary Information Notice, and Copy Right Notice c: K. B. Marsh NRC Resident Inspector S. A. Byrne S. E. Jenkins J. B. Archie Paulette Ledbetter N. S. Carns K. M. Sutton J. H. Hamilton NSRC J. W. Williams RTS (CR-12-04269)

W. M. Cherry File (813.20)

V. M. McCree PRSF (RC-15-0025)

S. A. Williams

Document Control Desk CR-12-04269 Attachment RC-15-0025 Page 1 of 23 VIRGIL C. SUMMER NUCLEAR STATION (VCSNS) UNIT 1 DOCKET NO. 50-395 OPERATING LICENSE NO. NPF-12 ATTACHMENT VCSNS RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION

Document Control Desk CR-1 2-04269 Attachment RC-1 5-0025 Page 2 of 23 The U.S. Nuclear Regulatory Commission (NRC) staff reviewed the Virgil C. Summer Nuclear Station License Amendment Request (LAR), LAR-12-04269, dated August 27, 2014, concerning the license basis changes required due to an update to the Steam Generator Tube Rupture (SGTR) analysis. The NRC staff has determined that the following request for additional information (RAI) is required to complete its review.

RAI No. 5 (response to October 1. 2014 RAI)

In Attachment I of the licensee's submittal, it is stated that certain equipment can be locally controlled if needed, e.g. the SG PORVs, compressors XAC-3A/B, XAC-4A/B, and XAC-12, and the EFW flow control valves. Have the environmental conditions at those local control stations been confirmed as benign for plant operators during an SGTR event, especially regarding heat, humidity, steam, and radiation?

SCE&G Response The SGTR event is a loss of primary coolant from the Reactor Coolant System (RCS) to the secondary side of the ruptured steam generator. The event is assumed to occur at full power with the reactor coolant contaminated with fission products based on conservative assumptions and limits set by the Technical Specifications. The flow of radioactive reactor coolant results in contamination of the secondary systems. Operator response to the SGTR is primarily completed within the Main Control Room. Should control not be available from within the Main Control Room, alternate local operator actions are covered within the procedures to affect the appropriate SGTR response.

By the nature of the event, the SGTR does not result in a steam release within the confines of any buildings where these alternate local operator actions may take place. No steam driven increases in local pressure, temperature or humidity will occur. Potential degraded environmental conditions will be limited to temperature (i.e., due to loss of normal HVAC) and local radiation.

Increases in temperature will be driven by heat loss from contained fluids through the pipe insulation or operating equipment. The temperature increases would only occur in the event normal HVAC systems are lost. The HVAC systems will not be lost as a consequence of the SGTR, but may be lost due to a loss of offsite power.

Temperature rise upon loss of HVAC is covered by Station Blackout habitability for the steam generator (SG) power operated relief valves (PORVs) and the emergency feedwater (EFW) flow control valves. The operator actions for cooling the plant and maintaining SG levels are consistent for both events. The Station Blackout habitability submitted to the NRC (Ref. 1) concluded the temperature rise at these locations is acceptable for the multiple operator entries which may be needed to manipulate valves.

Based on studies completed and submitted for Station Blackout, the SG PORV and EFW flow control valves will remain accessible for SGTR with respect to temperature.

Document Control Desk CR-12-04269 Attachment RC-15-0025 Page 3 of 23

  • The alternate local operator actions for the compressors (XAC-3AIB, XAC-4A/B and XAC-12) differ in that multiple entries over the course of plant cooldown are not necessary. These compressor actions are early event single entries to attempt local start of equipment. With no steam releases in the compressor areas, near term temperature increases are not of concern.

" The diesel driven air compressor (XAC-14) is located at ground level outdoors, on the opposite side of the Turbine Building from the SG PORV vent stacks. Temperature excursions at the diesel air compressor are not of concern.

  • The instrument air compressors (XAC-3A/B) are located on the 412 foot elevation in a Turbine Building subfloor (FSAR Figure 1.2-18). These are in a large high bay (20 feet +

ceiling). After the SGTR with reactor trip and turbine trip, the heat sources in the Turbine Building are limited. With no direct steam release, the access to XAC-3A/B for early event alternate actions will not be challenged.

  • The RB air compressors (XAC-4A/B) are located on the 463 foot elevation in the Auxiliary Building (FSAR Figure 1.2-6). The compressors are located in a hallway location nearthe personnel hatch and separated from steam line heat sources. With no direct steam release, the access to XAC-4A/B for early event alternate actions will not be challenged.
  • The standby air compressor (XAC-12) is located on the 485 foot elevation in the Auxiliary Building (FSAR Figure 1.2-8). The compressor is located in a hallway location with no steam line heat sources on the 485 foot elevation. With no direct steam release, the access to XAC-12 for early event alternate actions will not be challenged.

For the SGTR, increases in local radiation within a building would only occur in those areas with piping or components containing contaminated secondary system fluid. The instrument air compressors in the Auxiliary Building (XAC-4A/B and XAC-12) and the Turbine Building (XAC-3A/B) would experience no increase in local radiation. For those locations with steam lines, the radiation dose would be based on the concentration of radioisotopes in the secondary coolant.

when assumed to be at or near Technical Specification limits) can result in higher than normal operation radiation fields. For the expected operator mission time for alternate local operation, the dose was conservatively calculated to be less than 1 millirem assuming activity levels consistent with the licensing basis SGTR.

" For the SG PORV in the ruptured loop, the secondary coolant activity would be higher than those on the intact loops since it would also include the reactor coolant flow (i.e.,

break flow) into the ruptured SG. Although the radiation field will be higher, the mission time for the PORV on the ruptured loop is shorter since only valve closure needs to be accomplished. For the expected operator mission time, the dose was conservatively calculated to be less than 650 millirem assuming activity levels consistent with the licensing basis SGTR.

Document Control Desk CR-12-04269 Attachment RC-15-0025 Page 4 of 23

  • For the EFW flow control valve operation, there are 4-inch steam lines going to the turbine driven EFW pump (one from Loop B and a second from Loop C) in the area where the valves are located. For the expected operator mission time for alternate local operation, the dose was conservatively calculated to be less than 50 millirem assuming activity level consistent with the licensing basis SGTR.
  • The diesel driven air compressor is located outside of the Auxiliary Service Building.

With the discharge from the SG PORV or safety valve vent stacks, there is a potential for dose when the operator completes a local action to start the compressor. For the operator mission time to start the diesel air compressor, the dose was calculated to be less than 50 millirem assuming activity releases consistent with the licensing basis SGTR.

For each of these cases, the conservatively calculated dose is within the NUREG-0737, Section ll.B.2 whole body limit of 5 Rem.

The alternate local operator actions for the SGTR response are thus all within accessible areas with acceptable environmental conditions.

Responses to RAIs dated January 9, 2015 RAI No. 1 In application dated August 27, 2014, it stated:

New SGTR dose analyses are also performed by Westinghouse using, consistent with the Virgil C. Summer Licensing Basis, the RADTRAD computer program following the guidance of Regulatory Guide 1.183.

Dose consequences are provided for the mass transfer input from the current licensing basis and new transient analyses.

During the review process the NRC staff noticed what appear to be some differences between the current license basis SGTR analysis parameters and the proposed SGTR analysis parameters that are not discussed in SCE&G's LAR. Therefore, the NRC staff is requesting that a table be provided for the SGTR analysis that lists: (1) The parameters used in the SGTR dose analysis, (2) the current approved SGTR modeled value for each parameter listed in VCSNS Updated Final Safety Analysis Report, (3) the new modeled value for each parameter in the revised SGTR analysis, and (4) a description of the technical basis that explains why the new modeled value changed.

(NOTE: Tables 9 through 14 in the LAR already provide items (1) and (3) above and could be revised to include items (2) and (4)).

Document Control Desk CR-12-04269 Attachment RC-15-0025 Page 5 of 23 SCE&G Response Tables 9-14 have been updated to include parameter listing for both the Licensing Amendment Request (LAR) analysis and the current analysis of record (AOR) which is reflected in the VCSNS Updated Final Safety Analysis Report. The AOR values correspond to those reviewed (ML102160020 with corrections per ML13051A372) in support of VCSNS's Alternative Source Term Amendment No. 183. Table changes include:

" Table 9 has been updated to include two additional entries: one for Analysis Code and one for ruptured flow flashing fraction.

  • Table 10 was split into Table 10A and 1OB to accommodate the AOR entries.
  • A comment column has been added to Tables 9 and 13 to highlight and clarify difference between the LAR and AOR values.

As indicated in the updated Tables, the LAR and AOR parameters and methods differ in the following general area:

  • Activity Release Calculation Methods
  • Initial RCS Activities
  • Initial Secondary Coolant Activities
  • Treatment of Pre-Trip Steam Release
  • Treatment of Pre-Trip and Post-Trip break flow flashing fractions
  • Primary and Secondary System Coolant Mass
  • Low population zone x/Q The majority of the differences involve user inputs. For most, the values used in the LAR analyses are more conservative than the values used in the AOR. For the less conservative differences, most of the LAR values reflect the removal of excessive conservatism. Overall, the LAR parameters and methods follow the Alternative Source Term methodology outlined in Regulatory Guide 1.183 for a SGTR and, as shown in Table 14, and their application results in an overall increase in the calculated doses when compared to the current AOR. Large margins continue to exist relative to the regulatory limits.

Additional details on the specific differences between the LAR and AOR parameters and methods are provided in the updated tables.

Document Control Desk CR-12-04269 Attachment RC-15-0025 Page 6 of 23 Table 9 Summary of Parameters Used in Evaluating the Radiological Consequences of a Steam Generator Tube Rupture LAR Value AOR Value Comment Analysis Code RADTRAD RADTRAD For the AOR, the released activity was calculated for Version 3.03 Version 3.03 defined intervals (e.g., 0 - 30 min, 0 - 2 hrs, etc.) using simplifying assumptions and then input to RADTRAD for the calculation of the resulting offsite and control room doses.

For the LAR analysis, RADTRAD was used to calculate the activity released as well as the resulting offsite and control room doses.

Both applications used RG-1.183 methodology. The revised approach for the LAR was chosen because it is a more rigorous application and is consistent with standard Westinghouse practices which make use of the RADTRAD codes capabilities.

RCS Iodine Activity

1. Pre-Accident Spike Primary coolant iodine activities Primary coolant iodine activities based The factor of 60 is applied in both the LAR analysis based on 60 ltCi/gm DE 1-131. These on 60 pICi/gm DE 1-131. These are 60 and AOR. The numerical differences shown in are 60 times the values given in Table times the values given in Table 1OA Table 1OA result from a revised treatment of the 10Awhich are based on 1.0 pCi/gm which are based on 1.0 gCi/gm source terms with defects in fuel producing 1% of core DE 1-131. DE 1-131. power (i.e., commonly referred to as 1% failed fuel).

See comment for the Accident-Initiated Spike.

Document Control Desk CR-12-04269 Attachment RC-15-0025 Page 7 of 23 Table 9 (continued)

Summary of Parameters Used in Evaluating the Radiological Consequences of a Steam Generator Tube Rupture LAR Value AOR Value Comment

2. Accident-Initiated Initial primary coolant iodine activities Initial primary coolant iodine activities The numerical differences shown in Table 1OA for the Spike based on 1.0 VCi/gm DE 1-131. The based on 1.0 [tCi/gm DE 1-131. The DE 1-131 values result from the revised treatment of iodine appearance rates for the iodine appearance rates for the the source terms for 1% failed fuel. The AOR values accident-initiated spike are 335 times accident-initiated spike are 335 times are derived from source terms based on a letdown flow the equilibrium appearance rates the equilibrium appearance rates which of 143 gpm with 1% failed fuel whereas the LAR which are given in Table 10A. The are given in Table 10A. The spike values are derived from the worst-case combination for spike continues until 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> from the continues until 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> from the start of minimum (60 gpm) and maximum (120 gpm) letdown start of the event, the event, with 1% failed fuel. The LAR approach achieves a more consistent treatment of all nuclides considered (see noble gas activity below) with letdown within its design range of 60 -120 gpm. As indicated in Table 10A, the changes result in a small increase in all Iodine activity levels except 1-131, which remained the same.

For both applications, the iodine appearance rates assumed a conservative letdown flow of 143 gpm. As shown in Table 10A, the iodine appearance rate for 1-131 is higher in the LAR analysis, and the differences in the AOR and LAR values for the other Iodine nuclides are small. These differences would have a minor impact on the calculated doses.

Document Control Desk CR-12-04269 Attachment RC-15-0025 Page 8 of 23 Table 9 (continued)

Summary of Parameters Used in Evaluating the Radiological Consequences of a Steam Generator Tube Rupture LAR Value AOR Value Comment Noble Gas Activity Primary coolant noble gas activities Primary coolant noble gas activities The AOR values are derived from source terms based based on operation with defects in based on operation with defects in fuel on 1% failed fuel assuming minimum letdown at fuel producing 1% of core power. producing 1% of core power with 60 gpm. Based on the assumption that the ratio of Values are given in Table 10A. reduction. Values are given in Table radioiodines to the other radionuclides in the reactor 1OA. coolant is a constant, a reduction factor of 3.84 was applied, where 3.84 corresponds to the equivalent DE 1-131 value for the 1% failed fuel source terms.

The LAR values are based on the worst-case combination of source terms for minimum (60 gpm) and maximum (120 gpm) letdown assuming 1% failed fuel. Since no reduction is applied, the LAR approach is more conservative.

RCS Bromine and Alkali Primary coolant Bromine and alkali Primary coolant noble Bromine and The AOR values are derived from source terms based Metal Activity metal activities based on operation alkali metal activities based on on 1% failed fuel assuming minimum letdown at with defects in fuel producing 1% of operation with defects in fuel producing 60 gpm. A reduction factor of 3.84 was applied, where core power. Values are given in Table 1% of core power with reduction. Table 3.84 corresponds to the equivalent DE 1-131 value for 1 OA. Table 1OA values correspond to 1OA values correspond to the initial 1% failed fuel. For the pre-accident spike, the reduced the initial activities assumed in the activities assumed in the accident Bromine values are conservatively increased by a accident initiated iodine spike initiated iodine spike analysis. For the factor of 60.

analysis. For the pre-accident spike, pre-accident spike, the RCS Bromine The LAR values are based on the worst-case the RCS Bromine activities are activities are increased by a factor of combination of source terms for minimum (60 gpm) increased by a factor of 60. 60. and maximum (120 gpm) letdown assuming 1% failed fuel without reduction. The LAR modeling is more conservative than the AOR approach.

Document Control Desk CR-1 2-04269 Attachment RC-1 5-0025 Page 9 of 23 Table 9 (continued)

Summary of Parameters Used in Evaluating the Radiological Consequences of a Steam Generator Tube Rupture LAR Value AOR Value Comment Secondary Coolant Initial secondary coolant iodine Initial secondary coolant activity The AOR approach for the pre-accident spike is very System Activity activity based on 0.1 p.Ci/gm DE 1-131 corresponds to the equilibrium activity conservative; that is, the resulting iodine activities (i.e, 1/ 1 0 th of the RCS activity within the secondary coolant for a correspond to approximately 1.94 pCi/gm DE 1-131 concentrations for 1 j.Ci/gm 1-gpm primary to secondary leak with which is well above the Tech Spec limit of 0.1 pCi/gm DE 1-131). SG halogen and alkali SG blowdown credited. For the pre- DE 1-131. Except for 1-134, the iodine activities in the metal activity concentrations are accident spike, the equilibrium values LAR analysis are lower. The values used, however, assumed to be 1/10th of the RCS are calculated assuming reactor coolant are set at the Tech Spec limit which represents an activity concentrations, reflecting the activities for 60 pCi/gm DE 1-131. For upper limit for normal plant operation. As shown in ratio of the Technical Specifications the accident initiated spike, the Table 1OB, the LAR values for the other radio-nuclides limits on DE 1-131. No noble gases equilibrium values are calculated are more conservative.

are contained in the secondary assuming reactor coolant activities for Except for 1-131, the LAR activities for the accident coolant. Values are given in Table 1% failed fuel with adjustment to reflect initiated spike are more conservative. Consistent with 10B. the Tech Spec limit of 0.1 pCi/gm the pre-accident spike approach, the radio-iodines are DE 1-131. No noble gases are set at the Tech Spec limit which represents an upper contained in the secondary coolant. limit for normal plant operation. The difference in the nuclide specific contributions to DE 1-131 would have a minor impact on the calculated doses.

Iodine Chemical 97% elemental, 3% organic, no 97% elemental, 3% organic, no No change.

Fractions particulates (no impact on analysis particulates (no impact on analysis since filter efficiencies credited are the since filter efficiencies credited are the same for all forms of iodine), same for all forms of iodine).

Document Control Desk CR-12-04269 Attachment RC-1 5-0025 Page 10 of 23 Table 9 (continued)

Summary of Parameters Used in Evaluating the Radiological Consequences of a Steam Generator Tube Rupture LAR Value AOR Value Comment RCS Mass Hand calc: Constant at 3.4E5 Ibm Hand calc: Constant at 3.9E5 Ibm The AOR modeling using RADTRAD is based on user Transient: Constant at 3.4E5 Ibm Transient: N/A input activity releases. The base 1% failed fuel source (Table 8) terms and iodine concentrations resulting from the concurrent iodine spike were calculated using an RCS mass of 1.8E8 gm (3.97E5 Ibm).

The LAR modeling uses RADTRAD to calculate the RCS activities during the accident initiated spike.

Use of a lower RCS mass is conservative since it maximizes the concentration buildup.

Intact SGs Mass Hand calc: Constant at 2.12E5 ibm Hand calc: Constant at 2.26E5 Ibm The AOR modeling using RADTRAD is based on user Transient: Constant at 1.9E5 Ibm Transient: N/A input activity releases. The secondary coolant source (Table 8) terms were calculated using an SG mass of 1.13E5 Ibm/SG or 2.26E5 Ibm for the two intact SGs. Dilution of the concentration of the activity leaking into the secondary mass (due to mixing with that mass) was not considered in the AOR.

The LAR modeling uses RADTRAD to calculate the concentration in the secondary as the event progresses and resulting activity release. Use of a lower SG mass is conservative since it maximizes the concentration buildup.

Document Control Desk CR-1 2-04269 Attachment RC-15-0025 Page 11 of 23 Table 9 (continued)

Summary of Parameters Used in Evaluating the Radiological Consequences of a Steam Generator Tube Rupture LAR Value AOR Value Comment Ruptured SG Mass Hand calc: Constant at 1.06E5 Ibm Hand calc: Constant at 1.13E5 Ibm The AOR modeling using RADTRAD is based on user Transient: Constant at 9.6E4 Ibm Transient: N/A input activity releases. The secondary coolant source (Table 8) terms were calculated using an SG mass of 1.1 3E5 Ibm/SG. Dilution of the concentration of the activity transferred by break flow into the secondary mass (due to mixing with that mass) was not considered in the AOR.

The LAR modeling uses RADTRAD to calculate the concentration in the secondary as the event progresses and resulting activity release. Use of a lower SG mass is conservative since it maximizes the concentration buildup.

Transient Timing Reactor Trip, SI, Hand calc: 385 sec Hand calc: 385 sec The AOR does not explicitly model pre-trip release via LOOP Transient: 76 sec (Table 5) Transient: N/A condenser off-gas. The activity release from the pre-trip break flow is, however, accounted for in the overall 0 - 30 minute release.

Break Flow Flashing Hand calc: 1800 sec Hand calc: 1800 sec No change for hand calculation.

Stops Transient: 2067 sec (Table 5) Transient: N/A Break Flow Hand calc: 1800 sec Hand calc: 1800 sec No change for hand calculation.

Terminates Transient: 3411 sec (Table 5) Transient: N/A

Document Control Desk CR-12-04269 Attachment RC-15-0025 Page 12 of 23 Table 9 (continued)

Summary of Parameters Used in Evaluating the Radiological Consequences of a Steam Generator Tube Rupture LAR Value AOR Value Comment Intact SG Releases 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> No change.

Terminated Ruptured SG Transient Release Data Rupture Flow" Hand calc:19,400 Ibm Hand calc: 19,400 Ibm No change for hand calculation.

Pre-Trip Transient: 3600 Ibm Transient: N/A (Table 6 and Figure 6*)

Post-Trip Until Hand calc: 73,500 Ibm Hand calc: 73,500 Ibm No change for hand calculation.

Flashing Stops Transient: 94,240 Ibm Transient: N/A (Table 6 and Figure 6*)

Post-Trip After Hand calc: 0 Ibm Hand calc: 0 Ibm No change for hand calculation.

Flashing Stops Transient: 37,880 Ibm Transient: N/A (Table 6 and Figure 6*)

Flashing Fraction Hand calc: 0.2 pre-trip and 0.125 Hand calc: 0.14 For the AOL, the pre and post-trip activity release from post trip Transient: N/A break flow is treated the same. A mass weighted Transient: as calculated (Figure 8) average flashing fraction of 0.14 is utilized with a total break flow of 92,900 Ibm. The total flashed rupture flow considered is 13,006 Ibm, which is essentially the same as the total pre and post trip release assumed in the LAR (3880 + 9188 = 13068 Ibm). Overall, the LAR provides a more rigorous treatment of the pre and post-trip activity release.

Document Control Desk CR-12-04269 Attachment RC-15-0025 Page 13 of 23 Table 9 (continued)

Summary of Parameters Used in Evaluating the Radiological Consequences of a Steam Generator Tube Rupture LAR Value AOR Value Comment Flashed Rupture Flow Hand calc: 3,880 Ibm Hand calc: See comment See Flashing Fraction comment.

Pre-Trip Transient: 610 Ibm Transient: N/A (Table 6 and Figure 9*)

Post-Trip Hand calc: 9,188 Ibm Hand calc: See comment See Flashing Fraction comment.

Transient: 5,120 Ibm Transient: N/A (Table 6 and Figure 9*)

Steam Releases 1,310 Ibm/sec N/A The AOR does not explicitly model pre-trip release.

Pre-Trip The activity release from the pre-trip break flow is, however, accounted for in the overall 0 - 30 minute release.

Pre-trip steam release from the intact SGs is explicitly accounted for in the LAR analysis. An iodine partition factor of 0.01 is assumed.

Post-Trip Until Hand calc: 56,800 Ibm post trip Hand calc: 56,800 Ibm post trip No change for hand calculation.

Break Flow Stops Transient: 73,740 Ibm Transient: N/A (Table 7 and Figure 7*)

Intact SGs Transient Release Data***

Primary to Secondary 1 gpm (8.34 Ibm/min) 1 gpm (8.34 Ibm/min) No Change Leakage

Document Control Desk CR-12-04269 Attachment RC-15-0025 Page 14 of 23 Table 9 (continued)

Summary of Parameters Used in Evaluating the Radiological Consequences of a Steam Generator Tube Rupture LAR Value AOR Value Comment Steam Releases 2,620 Ibm/sec N/A The AOR does not explicitly model pre-trip release.

Pre-Trip The activity release from the pre-trip intact SG leakage is, however, accounted for in the overall 0 - 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> release.

Pre-trip steam release from the intact SGs is explicitly accounted for in the LAR analysis. An iodine partition factor of 0.01 is assumed.

Post-Trip Until Hand calc: 381,400 Ibm Hand calc: 381,400 Ibm No change in hand calc values.

2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Transient: 348,300 Ibm until break Transient: N/A flow termination (Table 7 and Figure 7*), 256,400 from break flow termination until 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 2 to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 924,900 Ibm 924,900 Ibm No Change 8 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 1,200,000 Ibm 1,200,000 Ibm No Change Iodine Partition Coefficients Condenser Not modeled Not modeled No change Steam Release from 100 100 No change SGs

Document Control Desk CR-12-04269 Attachment RC-15-0025 Page 15 of 23 Table 9 (continued)

Summary of Parameters Used in Evaluating the Radiological Consequences of a Steam Generator Tube Rupture LAR Value AOR Value Comment Flashed Break Flow 1.0 1.0 No change Release from Ruptured SG Atmospheric Dispersion Values are given in Table 11 Values are given in Table 11 The LAR incorporates a 0 - 2 hr X/Q for the LPZ dose Factors which was not considered in the AOR. This additional XJQ was calculated using the same inputs and methods used for the other LPZ values shown in Table 11. Its use is conservative.

Breathing Rate No change.

EAB 3.5E-4 m3/sec for all time intervals 3.5E-4 m3/sec for all time intervals LPZ 3.5E-4 m3/sec until 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 3.5E-4 m3/sec until 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 1.8E-4 m3/sec from 8 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 1.8E-4 m3/sec from 8 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Control Room Modeling See Table 13 See Table 13 The control room modeling inputs shown in Table 13 for the LAR analysis and AOR are the same.

"*A The analysis conservatively added 10% to the mass transfer data presented in this figure.

"**"r

/ This is the total flow through the break and includes the flashed flow listed separately.

Data listed in total for both intact SGs.

Document Control Desk CR-12-04269 Attachment RC-15-0025 Page 16 of 23 Table 10A Specific Activities in the Primary Coolant and Associated Iodine Appearance Rates RCS Concentration RCS Concentration Based on Nuclide Based on 1.0 pCi/gm Rape Rate Operation with Defects in Fuel DE 1-131 Producing 1% of Core Power (pCi/gm) (pCi/gm)

AOR [AR AOR LAR' AOR' LAR (No Reduction) (With Reduction) 1-131 0.782 0.782 7.21E-3 6.84E-3 -

1-132 1.268 0.808 3.33E-2 3.59E-2 -

1-133 1.268 1.200 1.39E-2 1.40E-2 -

1-134 0.279 0.156 1.53E-2 1.50E-2 -

1-135 0.862 0.625 1.29E-2 1.29E-2 -

Br-83 - - - - 0.089' 0.02321 Br-84 - - - - 0.0421 0.01091 Cs-134 - - - - 4.4 1.15 Cs-136 - - - - 4.5 1.17 Cs-137 - - - - 2.1 0.547 Cs-1 38 - - - - 0.97 0.253 Rb-88 - - - - 6.5 0.99 Kr-85m - - - - 1.8 0.469 Kr-85 - - - - 7.9 1.98 Kr-87 - - - - 1.1 0.287 Kr-88 - 3.2 0.834 Xe-131m -- 3.4 0.599 Xe-1 33m -- 19.0 4.95 Xe-1 33 _ _- 290 75.5 Xe-135m -- 0.52 0.135 Xe-1 35 - 8.6 2.24 Xe-138 - 0.64 0.167

1. For the Pre-Accident Spike, these initial activities are increased by a factor of 60.
2. The AOR values are scaled down by a reduction factor of 3.84 which corresponds to the equivalent DEl-1 31 value for the 1% failed fuel source term.

Document Control Desk CR-12-04269 Attachment RC-15-0025 Page 17 of 23 Table IOB Specific Activities In Secondary Coolant Pre-Accident Iodine Accident Initiated Spike Nuclide Spike (pCi/gm) (pCi/gm)

LAR AOR LAR AOR 1-131 0.0782 1.6790 0.0782 0.086538 1-132 0.1268 0.2130 0.1268 6.40E-05 1-133 0.1268 1.4980 0.1268 0.012649 1-134 0.0279 0.0166 0.0279 8.97E-7 1-135 0.0862 0.3931 0.0862 0.000571 Br-83 0.0089 0.00625 0.0089 0.000322 Br-84 0.0042 0.00072 0.0042 3.7E-5 Cs-134 0.44 0.04493 0.44 0.139 Cs-136 0.45 0.04345 0.45 0.134 Cs-137 0.21 0.02146 0.21 0.066 Cs-1 38 0.097 0.00028 0.097 0.000866 Rb-88 0.65 0.00061 0.65 0.0019

Document Control Desk CR-12-04269 Attachment RC-15-0025 Page 18 of 23 Table 11 Atmospheric Dispersion Factors Exclusion Area Low Population Zone Control Room Time Period Boundary ec/lion) (seC/rnt

)

(sec/m 3 ) se (se LAR AOR LAR AOR LAR AOR 0 to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 1.24E-4 1.24E-4 5.06E-5 - 1.51 E-3 1.51 E-3 2 to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 1.24E-4 1.24E-4 2.42E-5 2.42E-5 1.17E-3 1.17E-3 0 to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> - 2.42E-5 -

8 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 1.24E-4 1.24E-4 1.68E-5 1.68E-5 5.75E-4 5.75E-4

Document Control Desk CR-12-04269 Attachment RC-1 5-0025 Page 19 of 23 Table 12 Dose Conversion Factors Reference 11 EDE Dose Conversion Factor - Rfrne1 Nuclide Conesin Cloudshine CEDE Dose Conversion Factor - Inhaled (Sv/Bq)

(Sv-m 3/Bq-sec)

LAR AOR LAR AOR 1-131 1.82E-14 1.82E-14 8.89E-09 8.89E-09 1-132 1.12E-13 1.12E-13 1.03E-10 1.03E-10 1-133 2.94E-14 2.94E-14 1.58E-09 1.58E-09 1-134 1.30E-13 1.30E-13 3.55E-11 3.55E-11 1-135 8.294E-14 1 8.294E-14 1 3.32E-10 3.32E-10 Br-83 3.82E-16 3.82E-16 2.41E-11 2.41E-11 Br-84 9.41E-14 9.41 E-14 2.61E-11 2.61E-11 Cs-134 7.57E-14 7.57E-14 1.25E-08 1.25E-08 Cs-136 1.06E-13 1.06E-13 1.98E-09 1.98E-09 Cs-137 2.88E-14 2 2.725E-14 3 8.63E-09 8.63E-09 Cs-138 1.21 E-13 1.21 E-13 2.74E-11 2.74E-11 Rb-88 3.36E-14 3.36E-14 2.26E-11 2.26E-11 Kr-85m 7.48E-15 7.48E-15 Kr-85 1.19E-16 1.19E-16 Kr-87 4.12E-14 4.12E-14 Kr-88 1.02E-13 1.02E-13 Xe-131 m 3.89E-16 3.89E-16 _ _ _

Xe-133m 1.37E-15 1.37E-15 Xe-133 1.56E-15 1.56E-15 Xe-135m 2.04E-14 2.04E-14 Xe-135 1.19E-14 1.19E-14 Xe-138 5.77E-14 5.77E-14 _ _ _

1 The value listed for 1-135 is the DCF for 1-135 plus the DCF for its daughter product Xe-1 35m adjusted by the branching fraction of 0.154.

2 The LAR value listed for CS-1 37 is the DCF for its daughter product Ba-137m.

3 The AOR value listed for CS-1 37 is the DCF for CS-1 37 plus the DCF for its daughter product Ba-1 37m adjusted by the branching fraction of 0.946.

Document Control Desk CR-12-04269 Attachment RC-1 5-0025 Page 20 of 23 Table 13 Control Room Modeling LAR AOR Comment Transition from Normal Mode The CR ventilation emergency mode The CR ventilation emergency mode Ventilation, to Emergency Mode is initiated at 0.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> after the start is initiated at 0.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> after the start No Change of the accident.* of the accident.

CR Volume 226,040 ft3 226,040 ft3 No Change CR Unfiltered In-Leakage Normal Mode 217 cfm 217 cfm No Change Emergency Mode 243 cfm 243 cfm CR Unfiltered Makeup Flow Normal Mode 1,291 cfm 1,291 cfm No Change Emergency Mode 0 cfm 0 cfm CR Filtered Makeup Flow Normal Mode 0 cfm 0 cfm No Change Emergency Mode 1,265 cfm 1,265 cfm CR Filtered Recirculation Flow Normal Mode 0 cfm 0 cfm No Change Emergency Mode 19,125 cfm 19,125 cfm CR Filter Efficiency 95% for elemental and organic iodines 95% for elemental and organic iodines No Change and particulates and particulates CR X/Q Values are given in Table 11 Values are given in Table 11 No Change CR Breathing Rate 3.5E-4 m 3 /sec 3.5E-4 m 3/sec No Change CR Occupancy Factors 1.0 for the first day 1.0 for the first day 0.6 from 1 to 4 days 0.6 from 1 to 4 days No Change 0.4 after 4 days 0.4 after 4 days

  • The conservative time has been retained to be consistent with the approved analysis (LAR Reference 1). In addition, sensitivity calculations were performed crediting the emergency mode of operation 60 seconds after the SI signal.

Document Control Desk CR-12-04269 Attachment RC-15-0025 Page 21 of 23 Table 14 SGTR Radiological Consequences Analysis Results Hand Calc TEDE Dose Transient TEDE Dose TEDE Dose Location (rem) (rem) Limit (rem)

Scenario LAR* AOR LAR* AOR Exclusion Area Boundary (0 to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />) 0.68 0.63 0.34 - 25 Pre- Low Population Zone (0 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />) 0.28 0.13 0.14 - 25 Accident Control Room (0 to 30 days)

Iodine CR ventilation emergency mode is initiated at:

Spike 30 minutes 1.30 1.18 0.53 -

SI Actuation + 60 seconds 0.37 - 0.11 -

Exclusion Area Boundary (0 to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />) 0.31 0.22 0.18 - 2.5 Low Population Zone (0 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />) 0.14 0.05 0.088 - 2.5 Accident-Initiated Control Room (0 to 30 days)

Iodine CR ventilation emergency mode is initiated at:

Spike 30 minutes 0.50 0.37 0.22 5 SI Actuation + 60 seconds 0.13 - 0.06 Approximately 5% margin was added to the total calculated dose to arrive at the dose reported. This margin is included to allow flexibility in addressing minor impacts on the dose analysis without requiring a reanalysis or changes in the reported dose.

Document Control Desk CR- 2-04269 Attachment RC-14-0178 Page 22 of 23 RAI No. 2 Provide the RADTRAD input and output files for the new SGTR analysis.

SCE&G Response The requested RADTRAD input and output files are provided on the enclosed CD. The RADTRAD files are Westinghouse proprietary information. Therefore, it is requested that they be withheld from public disclosure in accordance with the following application and accompanying affidavit (enclosed):

Westinghouse Letter, CAW-1 5-4079, "APPLICATION FOR WITHHOLDING PROPRIETARY INFORMATION FROM PUBLIC DISCLOSURE", dated 01/19/2015.

RAI No. 3 As presented in LAR 12-04269, LOFTTR2 was used, per the methodology of WCAP-10698-P, to perform SGTR transient analyses for VCSNS. Does the version of LOFTTR2 used in the analysis presented in LAR 12-04269 contain any code modifications?

SCE&G Response The V. C. Summer steam generator tube rupture (SGTR) transient calculations presented in the license amendment request (LAR) were performed using the LOFTTR2 computer program.

LOFTTR2 is an updated version of the LOFTTR1 program which was developed as part of the approved SGTR methodology presented in WCAP-1 0698-P-A. During the methodology development, the LOFTTR1 program was modified to accommodate steam generator overfill and the revised program was designated as LOFTTR2. LOFTTR2 was used for the evaluation of the consequences of overfill presented in WCAP-1 1002. The NRC evaluation of WCAP-11002 is included in WCAP-10698-P-A. The LOFTTR2 program is therefore considered to be approved by the NRC. LOFTTR2 has been used for numerous SGTR transient calculations approved by the NRC since 1987, including the three precedents listed in the LAR (ML110450159 and ML111170513; ML11293A359 and ML11293A365; and ML112521289).

The LOFTTR2 program was identical to the LOFTTR1 program, with the exception that the LOFTTR2 program included the additional capability to represent the transition from two regions (steam and water) on the secondary side to a single water region if overfill occurs, and the transition back to two regions again depending upon the calculated secondary conditions.

Westinghouse did not maintain the LOFTTR1 program since LOFTTR2 included all of the LOFTTR1 functionality.

LOFTTR2 is maintained by Westinghouse following its Quality Assurance procedures for software control, which include requirements for configuration control of changes, installation testing on different computers or operating systems, and software problem reporting and error correction. The LOFTTR2 code has not had substantive changes since its use as part of the WCAP-10698-P-A methodology outlined above. The LOFTTR2 version used for the V. C.

Summer analysis (LOFTTR2 Version 2.01) is the same as that used for the three precedents

Document Control Desk CR-12-04269 Attachment RC-14-0178 Page 23 of 23 listed in the LAR (ML110450159 and ML111170513; ML11293A359 and ML11293A365; and ML112521289).

RAI No. 4 The referenced Steam Generator Tube analysis methodology (WCAP-10698) contains a single failure analysis relative to margin to overfill. When a new methodology is referenced by an applicant, exceptions to portions of the methodology typically require a technical basis. What is the technical basis for not considering a single failure?

SCE&G Response LAR-12-04269 supports a change to the Virgil C. Summer Nuclear Station (VCSNS) licensing basis to incorporate supplemental analyses of a SGTR accident which explicitly models operator responses and quantified their impact on the potential for steam generator overfill and offsite and control room dose. The new transient calculations supplement, as opposed to replace, the VCSNS licensing basis analysis by demonstrating margin to steam generator overfill and providing input to a dose analysis that confirms the licensing basis mass transfer input is conservative.

As outlined in Section 3.1.4 of LAR-12-04269, the supplemental SGTR transient analyses were performed by Westinghouse using the LOFTTR2 computer program following the methodology developed in WCAP-10698-P-A and its Supplement with the exception that a single failure is not considered. The basis for not assuming a single failure is twofold. First, as stated in Section 3.1.2 of LAR-1 2-04269, VCSNS current licensing basis analysis does not consider the effects of a postulated single failure. This general approach was established during the initial licensing process and has been an inherent assumption in a number of previously approved LARs for VCSNS supporting fuel transitions, steam generator replacement and power uprate. Therefore, a no single failure assumption is consistent with VCSNS's current licensing basis analysis, which will be maintained to provide mass transfer inputs to the dose analysis. Secondly, as discussed in Section 4.2 of the LAR-12-04269, precedence supports the approach. VCSNS is using the same general approach applied in previously approved license amendments for Point Beach Nuclear Plant Units 1 and 2, Turkey Point Units 3 and 4, and Prairie Island Nuclear Generating Plant Units 1 and 2. The applicable references for the license amendments are provided in Section 4.2 of LAR-12-04269. In these approved license amendment requests, the SGTR analysis for input to dose maintained the licensing basis 30-minute hand calculation and was supplemented by (1) a calculation to show the 30-minute hand calculation remains conservative for dose compared to a calculation modeling actual operator actions and (2) a margin to overfill analysis that followed WCAP-1 0698 but with no single failure (consistent with the plants' original licensing bases).

Document Control Desk CR-1 2-04269 RC-14-0178 Page 1 of 8 VIRGIL C. SUMMER NUCLEAR STATION (VCSNS) UNIT 1 DOCKET NO. 50-395 OPERATING LICENSE NO. NPF-12 ENCLOSURE 2 APPLICATION FOR WITHHOLDING, ACCOMPANYING AFFIDAVIT, PROPRIETARY INFORMATION NOTICE, AND COPYRIGHT NOTICE

Westinghouse Electric Company

( ) Westinghouse Engineering, Equipment and Major Projects 1000 Westinghouse Drive, Building 3 Cranberry Township, Pennsylvania 16066 USA U.S. Nuclear Regulatory Commission Direct tel: (412) 374-4643 Document Control Desk Direct fax: (724) 940-8560 11555 Rockville Pike e-mail: greshaja@westinghouse.com Rockville, MD 20852 Proj letter:

CAW-15-4079 January 19, 2015 APPLICATION FOR WITHHOLDING PROPRIETARY INFORMATION FROM PUBLIC DISCLOSURE

Subject:

CD entitled "RADTRAD 3.03 Input and Output Files for V.C. Summer Unit 1 Steam Generator Tube Rupture Analysis," January 2015 (Proprietary)

The proprietary information for which withholding is being requested in the above-referenced documents is further identified in Affidavit CAW-15-4079 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.

The subject documents were prepared and classified as Westinghouse Proprietary Class 2. Westinghouse requests that the documents be considered proprietary in their entirety. As such, non-proprietary versions will not be issued.

Accordingly, this letter authorizes the utilization of the accompanying Affidavit by South Carolina Electric and Gas.

Correspondence with respect to the proprietary aspects of the Application for Withholding or the Westinghouse Affidavit should reference CAW-15-4079, and should be addressed to James A. Gresham, Manager, Regulatory Compliance, Westinghouse Electric Company, 1000 Westinghouse Drive, Building 3 Suite 310, Cranberry Township, Pennsylvania 16066.

Very truly yours, James A. Gresham, Manager Regulatory Compliance

CAW-15-4079 January 19, 2015 AFFIDAVIT COMMONWEALTH OF PENNSYLVANIA:

ss COUNTY OF BUTLER:

I, James A. Gresham, am 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 my knowledge, information, and belief.

Jmes A. Gresham, Manager Regulatory Compliance

2 CAW-15-4079 (1) I am Manager, Regulatory Compliance, 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) 1am 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 (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 constitute Westinghouse policy and provide 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

3 CAW-1 5-4079 Westinghouse's competitors without license from Westinghouse constitutes a competitive economic advantage over other companies.

(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.

(iii) 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.

4 CAW-1 5-4079 (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.

(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.

(iv) 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.

(v) 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.

(vi) The proprietary information sought to be withheld in this submittal is that which is appropriately marked in CD entitled "RADTRAD 3.03 Input and Output Files for V.C.

Summer Unit I Steam Generator Tube Rupture Analysis," January 2015 (Proprietary),

being transmitted by South Carolina Electric and Gas letter and Application for Withholding Proprietary Information from Public Disclosure, to the Document Control Desk. The proprietary information as submitted by Westinghouse is that associated with Seam Generator Tube Rupture Analysis, and may be used only for that purpose.

(a) This information is part of that which will enable Westinghouse to:

(i) Provide input to the Nuclear Regulatory Commission for review of the V.C. Summer Unit 1 steam generator tube rupture dose analyses.

5 CAW-15-4079 (ii) Provide customer specific calculations.

(b) Further this information has substantial commercial value as follows:

(i) Westinghouse plans to sell the use of the information to its customers for the purpose of performing Steam Generator Tube Rupture analyses.

(ii) Westinghouse can sell support and defense of the technology to its customer in the licensing process.

(iii) 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 the proprietary versions of documents furnished to the NRC associated with Seam Generator Tube Rupture Analysis, and may be used only for that purpose. The documents are to be considered proprietary in their entirety.

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.