LR-N16-0099, Hope Creek Generating Station License Amendment Request - Safety Limit Minimum Critical Power Ratio Change

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Hope Creek Generating Station License Amendment Request - Safety Limit Minimum Critical Power Ratio Change
ML16181A193
Person / Time
Site: Hope Creek PSEG icon.png
Issue date: 06/08/2016
From: Davison P J
Public Service Enterprise Group
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
LAR H16-03, LR-N16-0099
Download: ML16181A193 (19)
v • d • e


Text

{{#Wiki_filter:Enclosure 1 Contains Proprietary Information to be Withheld from Public Disclosure Pursuant to 10 CFR 2.390 LR-N16-0099 LAR H16-03 PSEG Nuclear LLC P.O. Box 236, Hancocks Blidge, NJ 08038-0236 JUN 08 2016 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-0001 Hope Creek Generating Station Renewed Facility Operating License No. NPF-57 NRC Docket No. 50-354 PSEG NudearlLC 10 CFR 50.90

Subject:

License Amendment Request -Safety Limit Minimum Critical Power Ratio Change In accordance with 10 CFR 50.90, PSEG Nuclear LLC (PSEG) hereby requests an amendment to Renewed Facility Operating License No. NPF-57 for Hope Creek Generating Station (HCGS). In accord.ance with 10 CFR 50.91 (b)(1 ), a copy of this request for amendment has been sent to the State of New Jersey.

  • The proposed license amendment request (LAR) modifies Technical Specifications (TS) Section 2.1, Safety Limits. Specifically, this change incorporates a revised Safety Limit Minimum Critical Power Ratio (SLMCPR) for single recirculation loop operation (SLO) due to the cycle specific analysis performed by Global Nuclear Fuel (GNF) for HCGS Cycle 21. The proposed change has been evaluated in accordance with 10 CFR 50.91(a)(1), using the criteria in 10 CFR 50.92(c), and it has been determined that this request involves no significant hazards considerations.

There are no regulatory commitments contained in this letter. There are three attachments and three enclosures to this letter. Attachment 1 provides an evaluation supporting the proposed changes. The marked-up TS page, with the proposed changes indicated, is provided in Attachment

2. Attachment 3 provides, for information only, proposed changes to the TS Bases. Enclosure 1, GNF Report 003N5734-R1-P, specifies the required SLMCPRs for HCGS Cycle 21. Enclosure 1 contains information proprietary to. GNF. GNF requests that the document be withheld from public disclosure in accordance with JUN 08 2016 Page 2 LR-N16-0099 Enclosure 1 Contains Proprietary Information to be Withheld from Public Disclosure Pursuant to 10 CFR 2.390 10 CFR 50.90 10 CFR 2.390(a)(4).

Enclosure 2 contains a non-proprietary version of the GNF Report, 003N5734-R1-NP. An affidavit supporting this request is contained in Enclosure

3. These proposed changes have been reviewed by the Plant Operations Review Committee.

PSEG requests NRC approval of the proposed LAR by October 14, 2016 to support the HCGS refueling outage In Fall 2016 (Reload 20). Once approved, the amendment shall be implemented prior to startup from the refueling outage. If you have any questions or require additional information, please contact Mr. Lee Marabella at (856) 339-1208. I declare-under penalty of perjury that the foregoing is true and correct. Executed on :f VJ\e.. d.,.o \' (Date) Respectfully, Paul J. Davison Site Vice President Hope Creek Generating Station Attachments:

1. Request for Changes to Technical Specifications
2. Technical Specification Pages with Proposed Changes 3. Technical Specification Bases Pages with Proposed Changes (For Information Only)

Enclosures:

1. Proprietary Version of GNF Report 003N5734-R1-P
2. Non-Proprietary Version of GNF Report 003N5'734-R1-NP
3. GNF Affidavit in Support of Request to Withhold Information cc: Administrator, Region I, NRG Project Manager, NRC NRC Senior Resident Inspector, Hope Creek Mr. P. Mulligan, Chief, NJBNE Mr. L. Marabella, Corporate Commitment Tracking Coordinator Mr. T. MacEwen, Hope Creek Commitment Tracking Coordinator !

LR-N16-0099 Attachment 1 Request for Changes to Technical Specifications i ',* LR-N 16-0099 Attachment 1 LAR H16-03 HOPE CREEK GENERATING STATION RENEWED FACILITY OPERATING LICENSE NO. NPF-57 DOCKET NO. 50-354 License Amendment Request-Safety Limit Minimum Critical Power Ratio Change Table of Contents

1.0 DESCRIPTION

......................................................................... ,. ............................. 2 2.0 PROPOSE:D CHANGE ...................................................................... .................... 2 3 .. 0 BACKGROUND .................................................................................................. 2 4.0 TECHNICAL ANALYSIS ............................................................................

........ 2 5.0 REGULATORY ANALYSIS ...................................................................................

3 5.1 No Significant Hazards Consideration ......................................................... 3 5.2 Applicable Regulatory Requirements and Criteria ........................................ 4 6.0 ENVIRONMENTAL CONSIDERATION .............................................................. 5

7.0 REFERENCES

................................................................................................... 5 ', Page 1of5 r . LR-N16-0099 Attachment 1

1.0 DESCRIPTION

LAR H16-03 This evaluation supports a request to amend Renewed Facility Operating License No. NPF-57 for Hope Creek Generating Station (HCGS). The proposed change modifies Technical Specifications (TS) Section 2.1, Safety Limits. Specifically, this change incorporates a revised Safety Limit Minimum Critical Power Ratio (SLMCPR) for single recirculation loop operable (SLO) due to the cycle specific analysis performed by Global Nuclear Fuels (GNF) for HCGS Cycle 21. 2.0 PROPOSED CHANGE The proposed change involves revising the SLMCPR contained in TS Section 2.1 for SLO. The SLMCPR for SLO is being changed from <:: 1.10 to :<'! 1.11. The SLMCPR value for two recirculation loop operation (TLO) remains unchanged at:<'! 1.08. Marked up TS page 2-1 showing the requested change is provided in Attachment

2. Proposed changes to the TS Bases are provided in Attachment 3 of this submittal for information only. Changes to the affected TS Bases pages will be incorporated in accordance with TS 6.15, "Technical Specifications (TS) Bases Control Program." -3.0 BACKGROUND The SLMCPR analysis establishes SLMCPR values that ensure at least 99.9% of all fuel rods in the core do not experience transition boiling during normal operation and analyzed transients.

The SLMCPRs are re-evaluated for each reload using NRG-approved methodology to incorporate plant and cycle specific parameters for the current core design. As such, the calculated SLMCPR values may change on a cycle specific basis. The proposed change involves revising the SLMCPR contained in TS Section 2.1 for SLO due to the cycle specific analysis performed by GNF for HCGS Cycle 21. Operation in accordance with the revised SLMCPR for SLO will continue to preserve the existing margin to transition boiling and therefore protect the integrity .of the fuel cladding barrier. * . 4.0 TECHNICAL ANALYSIS The proposed TS change revises the SLMCPR contained in TS Section 2.1 for SLO due to the cycle specific analysis performed by GNF for HCGS Cycle 21. The SLMCPR for TLO remains unchanged. The SLMCPRs are calculated using NRG-approved methodology described in "General Electric Standard Application for Reactor Fuel (GEST AR 11)," NEDE-24011-P-A, Revision 22 (Reference 1 ). Information supporting the cycle specific SLMCPRs is included in Enclosure

1. That enclosure summarizes the methodology, inputs, and results for the calculated SLMCPRs. Page 2 of 5 ! i 1-! ;

LR-N 16-0099 Attachment 1

  • LAR H16-03 The SLMCPR analysis establishes SLMCPR values that ensure at least 99.9% of all fuel rods in the core do not experience transition boiling during normal operation and analyzed transients.

The SLMCPRs are calculated to include cycle specific parameters and in general, are dominated by two key parameters: (1) flatness of the core bundle-by-bundle MCPR distribution, and (2) flatness of the bundle pin-by-pin power/R-factor distribution. In addition, the uncertainty in the MCPR boiling correlation (GEXL critical power uncertainty) varies by fuel product line. Because the fresh fuel bundles generally dominate the SLMCPR calculation, a change in product line can potentially change the calculated SLMCPRs. HCGS Cycle 21 will be the first reload of GNF2 fuel. The effects of the GNF2 correlation uncertainty are included in the cycle specific analysis contained in Enclosure

1. The SLMCPRs documented in Enclosure 1 do not contain an adder for Extended Power Uprate (EPU) operation.

The NRC's safety evaluation of NEDC-33173P, "Applicability of GE Methods to Expanded Operating Domains," initially imposed a 0.02 adder to the cycle-specific SLMCPR values for EPU operation. The NRC subsequently determined additional data and analysis provided by General Electric-Hitachi (GEH) justified the original uncertainties used in GE H's methods and approved the removal of the adder for EPU operation. The exclusion of an adder for HCGS Cycle 21 is consistent with the NRG Safety Evaluation Report (SER) for NEDC-33173P (Reference 2). No plant hardware or operational changes are required with this proposed change. 5.0 REGULATORY ANALYSIS 5.1 No Significant Hazards Consideration In accordance with 10 CFR 50.90, PSEG Nuclear LLC (PSEG) hereby requests an amendment to Renewed Facility Operating License No. NPF-57 for Hope Creek Generating Station (HCGS). The proposed change modifies Technical Specifications (TS) Section 2.1, Safety Limits. Specifically, this change incorporates a revised Safety Limit Minimum Critical Power Ratio (SLMCPR) for single recfrculation loop operation (SLO) due to the cycle specific analysis for HCGS Cycle 21. PSEG has evaluated whether or not a Significant Hazards Consideration is involved by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below: 1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The required SLMCPRs for HCGS Cycle 21 are calculated using NRG-approved methodology. The SLMCPR values, contained in TS Section 2.1, Safety Limits, ensure at least 99.9% of all fuel rods in the core do not experience transition boiling during normal operation and analyzed transients, preserving fuel cladding Integrity. The proposed change to the SLMCPR value for SLO ensures this criterion continues to be met, and therefore does not increase the probability Page 3 of 5 i ,. ..;_ ' LR-N16-0099 Attachment 1 LAR H16-03 or consequences of an accident previously evaluated. In addition, no plant hardware or operational changes are required with this proposed change. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No. The required SLMCPRs for HCGS Cycle 2.1 are calculated using NRG-approved methodology. The SLMCPR values, contained in TS Section 2.1, ensure at least 99.9% of all fuel rods in the core do not experience transition boiling during normal operation and analyzed transients. The proposed change to the SLMCPR value for SLO does not involve any plant hardware or operational changes and does not create any new precursors to an accident. Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any previously evaluated. '* 3. Does the proposed amendment involve a significant reduction in a margin of safety? Response: No. The required SLMCPRs for HCGS Cycle 21 are calculated usin*g NRG-approved methodology. The SLMCPR values, contained in TS Section 2.1, ensure at least 99.9% of all fuel rods 1n the core do not experience transition boiling during normal operation and analyzed transients, pres-erving fuel cladding integrity. The revised SLM CPR value for SLO ensures this criterion continues to be met. In acjdition, the proposed change to the SLMCPR for SLO does not adversely affect the design basis function or performance of a structure, system, or component as described in the HCGS UFSAR. I, Therefore, thE;l proposed amendment does not involve a significant reduction in a margin of safety. --Based upon the above, PSEG Nuclear concludes that the proposed amendment presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" ls justified. 5.2 Applicable Regulatory Requirements and Criteria 10 CFR 50.36 Technical Specifications 10 CFR 50.36, "Technical specifications" identifies the requirements for the Technical Specification categories for operating power plants: (1) Safety limits, limiting safety system settings, and limiting control settings, (2) Limiting conditions for operation, (3) Surveillance requirements, { 4) Design features, (5) Administrative controls, (6) Decommissioning, (7) Initial notification, and (8) Written Reports. Specifically, 10 CFR 50.36(c)(1)(i)(A) states: Safety limits for nuclear reactors are limits upon important process variables that are found to be necessary Page 4 of 5 \ LR-N16-0099 Attachment 1 LAR H16-03 to reasonably protect the integrity of certain of the physical barriers that guard against the uncontrolled release of radioactivity. The SLMCPR values, contained in TS Section 2.1, Safety Limits, ensure at least 99.9% of all fuel rods In the core do not experience transition boiling during normal operation and analyzed transients, preserving fuel cladding integrity. The proposed change to the SLM CPR value for SLO ensures this criterion continues to be met. . In conclusion, based on the considerations discussed above, (1) there is a reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted ln compliance with the NRC's regulations, and (3). the issuance of the amendment .will not be inimical to the common defense and security or to the health and safety of the public. 6.0 ENVIRONMENTAL CONSIDERATION A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located Within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility r;:riterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

7.0 REFERENCES

1. Global Nuclear Fuel, "General Electric Standard Application for Reactor Fuel (GESTAR-11)", NEDE-24011-P-A, 22, November 2015. 2. Final Safety Evaluation for GE-Hitachi Nuctear Energy Americas (GEH) Topical Report NEDC-33173P, Revision 2 and Supplement 2, Parts 1-3, "Analysis of Gamma Scan Data and Removal of Safety Limit Critical Power Ratio (SLMCPR} Margin -Non-Proprietary," March 15, 2012 (ADAMS Accession No. ML 113340473).

Page 5 of 5 ' .. LR-N16-0099 Attachment 2 Technical Specification Pages with Proposed Changes LR-N16-0099 Attachment 2 LAR H16-03 TECHNICAL SPECIFICATION PAGES WITH PROPOSED CHANGES The following Technical Specification for Renewed Facility Operating License No. NPF-57 is affected by this change request: Technical Specification 2.1.2 2-1 \

  • 2. 0 SAFE'lt' LIMITS AND LlMITING SAFETY SYSTmli'l SE'!'TINGS

........ ......

2. 1 SAE'E'.l.'Y LIMITS THJGRM/JJ, POWJi!R, Low Pre,'ilsure oi;: Low Flow :2, 1.1 THERMAT" l?OWER shall not exceed 24% of RA'rE:D 'l'HERMAL l?OWEiR with the :i:*eactor vessel steam dol\ie pressure le:ss than 785 psig or core flow less than 10% of rated flow. OPERA'l'IONAL CONDITIONS 1 and 2. With THERMAL POWER exceeding 24% of RATED THEa\MAL POWER and the reactor v<assel steam dome pressure l.ess than 785 psig br core fl.ow less than 1f% o.e rated flow, be in at least HOT SHUTDOWN within 2 hours and oomply with the requirements of Specification
6. 7 .1, '.l'HERMAL J?tWEm, High !?res.sure a_p.d High Flow, 2 .1. 2 With reactor steam dome p.ressure greater than 785 psig and oore flow greater than 10% Of rated flow;. The MJ'.NIMUM CRl.TlCAL POWER RATIO (MCl?R) shall be 1. 08 for two :ri;icirculation loop operation and shall be <'! 4.-rl:-B-for single reoirculation loop operation. JTTil APl?I1ICABILIT\:'.:

Ol?E!RATIONAJ" CONDITIONS l and 2, ll..JJJ M:.'.£1Ql'l

With reactor steam dome pressure greater than 7B5 psig and care flow greater than of rate.d flow and the MCPR below the values fol;' the fuel stated in LCO 2 .1. 2 1 be in at least HOT SHUTDOWN with1n 2 hours and comply with the requirements of Specification
6. 7* .1. gmACTOR COOl1At'l'l 1 BY.S'.l'lil/:1.

PRESSURE 2. 1. 3 The .i::eaotor coolant systern pressure, as measursd in the reactor vessel steam dome, shall not eJ<ceed 1325 ps.ig. Al?l?LICABILlTY: OPERA.TIONAL CONDITIONS l, 2 1 3 and 4, Wi.th the .!:'eaotor coolant system pressure, as measured iP the reactor vessel steam dome, above 1325 psig, be in at least HOT SH!JTPOWN with reactor coolant system pres$ure less than o:r. equal to 1325 psig within 2 hours and comply with the reqL1i.rements of Specification 6.7.1

  • HOPE Amendment No. "

LR-N16-0099 Attachment 3 Technical Specification Bases Pages with Proposed Changes (For Information Only) LR-N16-0099 Attachment 3 LAR H16-03 TECHNICAL SPECIFICATION BASES PAGES WITH PROPOSED CHANGES The following Technical Specification Bases for Renewed Facility Operating License No. NPF-57 is affected by this change request: Technical Specification Bases -2.0 B 2-1 I , e 2, 1 SAFmTY .LIMITS BASES

  • 2. 0 INT!\ODOCT:i:ON The fuel Cliidding, reaotor and primal:'y system piping are the principal bar:ders to thtOi release of radioactive mate:dals to the env.:l.rons.

Safety Limits are estabJ.ished to p:r.ohct the integrity of the*se barriers during normal plant operations and a.ntioipated transients, The fuel claddinq integrity Safety Limit is set such that rio fUel damage is calculated to ocour if the limit is not violated. Because .fuel damage is not direat.1.y observable, a .approach is used to establish a Sa.fety Limit that the MCl?R is <:: l. 08 for two recirculat.ion loop operation and for single recirculation loop operation. These 111Cl?R values represent a conservative* margin relative to the conditions required to maintain fuel cladding integd.ty, The fuel cladding is one of the ph,Ys-idal barriers whioh separate the radioactive materials from the enviroI'l.IJ, The integrity or this clad<l.ing barrier i$ r*elated to its relative freedom from pel:'forations ox"

  • oraoking.

Although some cotrosion or use related cracking may occur during. the life of the cladding, fission product migration from thi$ source is incrementalJ.y ournulati ve and cont:l.nuously measurable. E'uel a.ladding per.forations, however, can result from thermal stresses which occur from reactor operation significal'ltly above desi'gn oon.ditions and the Limiting Sa.fety System SetUngs, While fission proctuot migration .from oladding perforatbn is just as measurable as that from use related cracking, the thermally caused cJ.adding perforations signal a threshold beyond which stiJ.l great:er thermal stresses may cause gross rather than incremental cladding deterioration. Therefore, the fuel claddinq Sa.faty Limit is defined with a marg.tn to the cond-itions which would produce onset of transition boiling 1 MCPR of 1. O, These conditions represent a signi.ficant departt1re from the condition intended by desi!n for planned operation, 2.1.1 Low Pressure or Low Flow The use of the critical power co:r.relations are not valid for all critical power oalculation.s performed at l:'eduoed prE;lsl\lUres below 785 psig or oore flows less than l.0% of rate* flow. Therefore, the fuel cladding integrity Safety Limit is established by other means. This is done by establishing a Hmiting condHio:n on core Tl!ERMAr, POWER with the following basis, Since the preesui:*e 41.rop in the bypasl'l region :i.l'l essentially all elevation head, the core pressu:iie drop at low power and flows will always. be greater than 4. 5 psi, Analyse*s show that with a bundle flow of 28 x 10 3 lbs/hr, bundle prsssure drop is nearly Of bundle power and has a

  • value of 3. 5 psi. 'l'hus, the bundle flow with a 4. 5 psi driving head will be greater than 28 x 10 3 lbs/hr, Ful.l .;ioale ATLAS test data taken at pressures from 14, 7 psia to BOO psia in4lioate that the fl.tel assembJ,y critical power at this flow ia appro.ximately 3.35 MWt, With the design peaking factors, this *aorreeponds to a THERMAL P.OWElR of more than 50'% of* RATED TH&RMAI,, !?OWJi;L\, ThUs 1 a 'f.HERMAL POWER 1.imi t of 2.4 o:f RATED THERMAL POWER for ;i.*eacto.i:

pressure below 785 is conservative. HOPE CREllilI< >'XX. Amendment No, .... I l:"SEG J'.SSL\sd) LR-N16-0099 Enclosure 1 Contains Proprietary Information to be Withheld from Public Disclosure Pursuant to 10 CFR 2.390 Enclosure 1 Proprietary Version of.GNF Report 003N5734-R1-P r ,j ! : I I LR-N16-0099 Enclosure 3 GNF Affidavit in Support of Request to Withhold Information Global Nuclear Fuel -Americas AFFIDAVIT I, Lukas Trosman, state as follows: (1) I am Engineering Manager, Reload Design and Analysis, Global Nuclear Fuel -Americas, LLC (GNF-A), and have been delegated the function of reviewing the information described in paragraph (2) which is sought to be withheld, and have been authorized to apply for its withholding. (2) The information sought to be withheld is contained in GNP proprietary report GNF-003N5734-Rl-P, "GNP Additional Information Regarding the Requested Changes to the Technical Specification SLMCPR Hope Creek Cycle 21," dated May 2016. GNP proprietary information in GNP-003N5734-Rl-P is identified by a dotted underline inside double square brackets. [[[hl§ .. __ GNP proprietary information in some tables is identified with double square brackets before and after the object. In each case, the superscript notation {3} refers to Paragraph (3) of this affidavit, which provides the basis for the proprietary determination. (3) In making this application for withholding of proprietary information of which it is the owner or licensee, GNP-A relies upon the exemption from disclosure set forth in the Freedom of Infonnation Act ("FOIA"), 5 USC Sec. 552(b)(4), and theTrade Secrets Act, 18 USC Sec. 1905, and NRC regulations 10 CFR 9.17(a)(4), and 2.390(a)(4) for "trade secrets" (Exemption 4). The material for which exemption from disclosure is here sought also qualify under the narrower definition of "trade secret", within the meanings assigned to those terms for purposes of FOIA Exemption 4 in, respectively, Critical Mass Energy Project v. Nuclear Regulatory Commission, 975 F2d 871 (DC Cir. 1992), and Public Citizen Health Research Group v. FDA, 704 F2d 1280 (DC Cir. 1983). (4) Some examples of categories of information which fit into the definition of proprietary information are: a. Information that discloses a process, method, or apparatus, including supporting data and analyses, where prevention of its use by. GNP-A's competitors without license from GNP-A constitutes a competitive economic advantage over other companies;

b. Information which, if used by a competitor, would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing of a similar product; c. Information which reveals aspects of past, present, or future GNP-A customer-funded development plans and programs, resulting in potential products to GNP-A; cl. Information which discloses patentable subject matter for which it may be desirable to obtain patent protection.
  • The information sought to be withheld is considered to be proprietary for the reasons set forth in paragraphs (4)a. and (4)b. above. GNF-003N5734-Rl-P Affidavit Page 1 of 3 (5) To address 10 CFR 2.390 (b) (4), the information sought to be withheld is being submitted to NRC in confidence.

The information is of a sort customarily held in confidence by GNF-A, and is in fact so held. The information sought to be withheld has, to the best of my knowledge and belief, consistently been held in confidence by GNF-A, no public disclosure has been made, and it is not available in public sources. All disclosures to third parties including any required transmittals to NRC, have been made, or must be made, pursuant to regulatory provisions or proprietary agreements which provide for maintenance of the information in confidence. Its initial designation as proprietary information, and the subsequent steps taken to prevent its unauthorized disclosure, as set forth in paragraphs (6) and (7) following. ( 6) Initial approval of proprietary treatment of a document is made by the manager of the originating component, the person most likely to be acquainted with the value and sensitivity of the information in relation to industry knowledge, or subject to the terms under which it was licensed to GNF-A. (7) The procedure for approval of external release of such a document typically requires review by the staff manager, project manager, principal scientist or other equivalent authority, by the manager of the cognizant marketing function (or his delegate), and by the Legal Operation, for technical content, competitive effect, and determination of the accuracy of the proprietary designation. Disclosures outside GNF-A are limited to regulatory bodies, customers, and potential customers, and their agents, suppliers, and licensees, and others with a legitimate need for the information, and then only in accordance with appropriate regulatory provisions or proprietary agreements. (8) The information identified in paragraph (2) is classified as proprietary because it contains details of GNF-A's fuel design and licensing methodology. The development of this methodology, along with the testing, development and approval was achieved at a significant cost to GNF-A. The development of the fuel design and licensing methodology along with the interpretation and application of the analytical results is derived from an extensive experience database that constitutes a major GNF-A asset. (9) Public disclosure of the information sought to be withheld is likely to cause substantial harm to GNF-A's competitive position and foreclose or reduce the availability of profit-making opportunities. The information is part of GNF-A's comprehensive BWR safety and technology base, and its co1mnerCial value extends beyond the original development cost. The value of the technology base goes beyond the extensive physical database and analytical methodology and includes development of the expertise to determine and apply the appropriate evaluation process. In addition, the technology base includes the value derived from providing analyses done with NRC-approved methods. The research, development, engineering, analytical, and NRC review costs comprise a substantial investment of time and money by GNF-A. The precise value of the expertise to devise an evaluation process and apply the correct analytical methodology is difficult to quantify, but it clearly is substantial. GNF-003N5734-Rl-P Affidavit Page 2 of 3 i I + I I ! GNF-A's competitive advantage will be lost if its competitors are able to use the results of the GNF-A experience to normalize or verify their own process or if they are able to claim an equivalent understanding by' delilonstrating that they can arrive at the same or similar conclusions. The value of this information to GNF-A would be lost if the information were disclosed to the public. Making such information available to competitors without their having been required to undertake a similar expenditure of resources would unfairly provide competitors with a windfall, and deprive GNF-A of the opportunity to exercise its competitive advantage to seek an adequate return on its large investment -in-developing arid obtaining these very valuable analytical tools. I declare under penalty of perjury that the foregoing is true and correct. Executed on this 17th day of May 2016. GNF-003N5734-Rl-P Lukas Trosman Engineering Manager, Reload Design and Analysis Global Nuclear Fuel -Americas, LLC 3901 Castle Hayne Road Wilmington, NC 28401 lukas.trosman@ge.com Affidavit Page 3 of 3}}

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