LR-N10-0163, Response to Request for Additional Information - License Amendment Request (H09-01) Supporting the Use of Co-60 Isotope Test Assemblies (Isotope Generation Pilot Project)

Response to Request for Additional Information - License Amendment Request (H09-01) Supporting the Use of Co-60 Isotope Test Assemblies (Isotope Generation Pilot Project)

ML101390314
Person / Time
Site:	Hope Creek
Issue date:	05/11/2010
From:	Jamila Perry Public Service Electric & Gas Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
LR-N10-0163
Download: ML101390314 (38)

Text

PSEG Nuclear LLC P.O. Box 236, Hancocks Bridge, New Jersey 08038-0236 0 PSEG Nuclear LLC MAY 1 12010 10 CFR 50.90 LR-N10-0163 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Hope Creek Generating Station Facility Operating License No. NPF-57 NRC Docket No. 50-354

Subject:

Response to Request for Additional Information - License Amendment Request (H09-01) Supporting the Use of Co-60 Isotope Test Assemblies (Isotope Generation Pilot Project)

References:

(1) Letter from PSEG to NRC, "License Amendment Request Supporting the Use of Co-60 Isotope Test Assemblies (Isotope Generation Pilot Project)," dated December 21, 2009 In Reference 1, PSEG Nuclear LLC (PSEG) submitted a license amendment request (H09-01) for the Hope Creek Generating Station (HCGS). Specifically, the proposed change would modify License Condition 2.B.(6) and create new License Conditions 1 .J and 2.B.(7) as part of a pilot program to irradiate Cobalt (Co)-59 targets to produce Co-60. In addition to the proposed license condition changes, the proposed change would also modify Technical Specification (TS) 5.3.1, "Fuel Assemblies," to describe the specific Isotope Test Assemblies (ITAs) being used.

The NRC provided PSEG a Request for Additional Information (RAI) on the license amendment request. The NRC RAI questions and the PSEG responses are provided in Attachment I to this letter, with the exception of RAI Questions 4, 5 and 6; the response to these questions will be provided in a subsequent letter, In addition, an Errata and Addendum (E&A) to NEDC-33529P (Attachment 3 to Reference 1) will be subsequently provided incorporating the changes discussed in the attached responses to RAI Questions 9 and 17.

Attachment 1 to this letter provides information which GEH considers to be proprietary. The proprietary information is identified by bracketed text. GEH requests that the proprietary information in Attachment 1 be withheld from public disclosure, in accordance with the requirements of 10 CFR 2.390, "Public inspections, exemptions, requests for withholding,"

paragraph (a)(4). A signed affidavit supporting this request is provided in Attachment 2 to this letter. Attachment 3 to this letter provides a nonproprietary version of Attachmentl, Attachments 4 and 5 to this letter provide calculations discussed in the response to RAI

\A I s OflzA (ec u-,L o

95-2168 REV. 7/99

Document Control Desk Page 2 LR-N10-0163 Question 19. Attachment 6 to this letter provides the 1 OCFR50.59 evaluation related to . Attachment 7 to this letter provides additional proposed changes to the HCGS TSs.,

PSEG has reviewed the information supporting a finding of no significant hazards consideration that was provided in Reference 1. The additional information provided in this submittal does not affect the bases for concluding that the proposed license amendment does not involve a significant hazards consideration. No new regulatory commitments are established by this submittal.

If you have any questions or require additional information, please do not hesitate to contact Mr.

Jeff Keenan at (856) 339-5429.

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

Executed on X k e/I

,'_uC)&

(Date)

Sincerely, John F-Perry Site Vice President Hope Creek Generating Station Attachments (7)

S. Collins, Regional Administrator - NRC Region I R. Ennis, Project Manager - USNRC NRC Senior Resident Inspector - Hope Creel<

P. Mulligan, ManagerIV, NJBNE Commitment Coordinator - HopeCreek PSEG Commitment Coordinator - Corporate

LR-N10-0163 Attachment 2 GE-Hitachi Affidavit for Withholding Portions of RAI Responses from Public Disclosure

GE-Hitachi Nuclear Energy Americas LLC AFFIDAVIT I, James F. Harrison state as follows:

(1) I am the Vice President, Fuel Licensing, Regulatory Affairs, GE-Hitachi Nuclear Energy Americas LLC ("GEH"), 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 Enclosure 1 of Global Nuclear Fuel-Americas, LLC letter, LRW-PSG-KT1-10-030, Lauren Watts to Don Notigan (Exelon Nuclear), entitled "Responses to Request for Additional Information 3, 7, 8, 9-13, 15, 17, 18, 20, and 21 Related to License Amendment Request to Modify Hope Creek Power Station Facility Operating License in Support of the Use of Isotope Test Assemblies," May 10, 2010.

GEH proprietary information in Enclosure 1, which is entitled "Responses to Request for Additional Information 3, 7, 8, 9-13, 15, 17, 18, 20, and 21", is identified

{*3)1by aA,,((",arin dotted underline inside double square brackets. ((LTh.is. _sentence is an exa..-ple......[]"Amarking at the beginning of a table, figure, or paragraph closed with a "))" marking at the end of the table, figure or paragraph is used to indicate that the entire content between the double brackets is proprietary. 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, GEH relies upon the exemption from disclosure set forth in the Freedom of Information Act ("FOIA"), 5 USC Sec. 552(b)(4), and the Trade 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, 975F2d871 (DC Cir. 1992), and Public Citizen Health Research Group v. FDA, 704F2d1280 (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 GEH's competitors without license from GEH 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; LRW-PSG-KT1-10-030 Enclosure 1 Affidavit Page 1 of 3

c. Information which reveals aspects of past, present, or future GEH customer-funded development plans and programs, resulting in potential products to GEH;

d. 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.

(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 GEH, 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 GEH, 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, are 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. Access to such documents within GEH is limited on a "need to know" basis.

(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 GEH 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), above, is classified as proprietary because it contains detailed results including the process and methodology for the design and analysis of the GE14i Isotope Test Assembly. The GE14i Isotope Test Assembly has been developed at a significant cost to GEH.

The development of the GE14i Isotope Test Assembly is derived from the extensive experience database that constitutes a major GEH asset.

LRW-PSG-KT1-10-030 Enclosure 1 Affidavit Page 2 of 3

(9) Public disclosure of the information sought to be withheld is likely to cause substantial harm to GEH's competitive position and foreclose or reduce the availability of profit-making opportunities. The information is part of GEH's comprehensive BWR safety and technology base, and its commercial 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 proce'ss. 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 GEH.

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.

GEH's competitive advantage will be lost if its competitors are able to use the results of the GEH experience to normalize or verify their own process or if they are able to claim an equivalent understanding by demonstrating that they can arrive at the same or similar conclusions.,

The value of this information to GEH would be lost if the information were digcl6sed to the public. Making such infornmation available to competitors without their having been required to undertake a similar expenditure of resources would unfairly provide competitors with a windfall, and deprive GEH of the opportunity to. exercise its competitive advantage to seek' an, adequate return on its large investment in developing these very valuable analytical tools.

I declare under penalty of perjury that the foregoing affidavit and the matters stated therein are true and correct to the best of my knowledge, information, and belief.

Executed on this 1 0 th day of May 2010.

James F. Harrison Vice President, Fuel Licensing, Regulatory Affairs GE-Hitachi Nuclear Energy Americas LLC LRW-PSG-KTI-10-030 Enclosure 1 Affidavit Page Cý 3 of 3

LR-N10-0163 Attachment 3 Additional Information Supporting the Request for a License Amendment to Modify HCGS Operating License in Support of the Use of Isotope Test Assemblies (Non-Proprietary)

LR-N10-0163 ADDITIONAL INFORMATION SUPPORTING PROPOSED LICENSE AMENDMENT USE OF ISOTOPE TEST ASEMBLIES4FOR COBALT-60 PRODUCTION HOPE CREEK GENERATING STATION DOCKET NO. 50-354 In reviewing the PSEG letter LR-N09-0290 (LAR H09-01) submittal dated December 21, 2009 (ADAMS No. ML093640193, Reference 1 of this attachment), related to a pilot program to irradiate Cobalt (Co)-59 targets to produce Co-60, for the Hope Creek Generating Station (HCGS), the Nuclear Regulatory Commission (NRC) staff has made a Request For Additional Information (RAI) in order to complete its review:

NRC RAI#1 In Table 1, "EquilibriumCobalt-60 Inventory," on page 14 of Attachment 7 to the application dated December 21, 2009 (Reference 1), the licensee uses the same values of neutron flux as that used in a similar table1 for Clinton Power Station (CPS). HCGS and CPS are boiling-water reactorswith different rated thermal power levels and number of fuel assemblies. Explain why the fluxes in Table I for the two reactors are the same. If the fluxes at the given exposure are different, please repeatthe calculations and modify Table 1.

RESPONSE TO RAI#1 The flux values provided in Table 1, "Equilibrium Cobalt-60 Inventory," on page 14 of (non-proprietary attachment) and Attachment 5 (proprietary attachment) of LAR H09-01 are identical to flux values provided for the CPS application because the values and the mathematical expression are generic. As stated in the HCGS response to RAI 9c: "The response to (c) is generic information that is of general interest to cobalt production."

NRC RAI#2 HCGS Technical Specification (TS) 5.3. 1, "FuelAssemblies," currently reads as follows:

The reactorcore shall contain 764 fuel assemblies and shall be limited to those assemblies which have been approved for use in BWRs.

The proposed amendment would revise TS 5.3. 1 to add the following:

A maximum of twelve GE14i Isotope Test Assemblies may be placed in non-limiting core regions, beginning with Reload 16 Cycle 17 core reload, with the purpose of obtaining surveillance data to verify that the GE14i 1 Reference page 21 of Attachment 3 to letter dated November 4, 2009, from Exelon to NRC (ADAMS Accession No. ML093100313).

1 of 31 LR-N10-0163 cobalt Isotope Test Assemblies perform satisfactorilyin service (priorto evaluatinga future license amendment for use of these design features on a production basis). Each GE14i assembly contains a small number of Zircaloy-2 clad isotope rods containing Cobalt-59. Cobalt-59 targets will transitioninto Cobalt-60 isotope targets during cycle irradiationof the assemblies.

(a) TS 5.3.1 lacks explicit information on the type of clad, type of fuel, type of materialof filler rods for potential substitution for fuel rods, approved methodology for fuel design analysis, and information on potential use of a limited number of test assemblies that may be placed in non-limiting locations.

Please propose further changes to TS 5.3.1 to address these issues. For example, see TS 4.2.1 of NUREG-1433, "StandardTechnical Specifications, General Electric Plants, BWR/4."

(b) In order to adequately describe the specific design of the ITAs which would be allowed to be inserted into the HCGS reactorplease add a sentence to the end of proposed TS 5.3. 1 similar to the following:

Specific details regardingthe design of the GE14i assemblies are contained in GE-Hitachireport NEDC-33529P, "Safety Analysis Report to Support Introduction of GE14i Isotope Test Assemblies (ITAs) in Hope Creek Generating Station," Revision 0, dated December 2009.

RESPONSE TO RAI#2 (a) Additional changes to TS 5.3.1, to align with NUREG-1433, will be added. See Attachment 7 of this submittal.

(b) The following sentence: "Details of the GE14i assemblies are contained in GE-Hitachi report NEDC-33529P, "Safety Analysis Report to Support Introduction of GE14i Isotope Test Assemblies (ITAs) in Hope Creek Generating Station," Revision 0, dated December 2009," will be added to TS 5.3.1. See Attachment 7 of this submittal.

NRC RAI#3 Page 4 of Attachment 1 to the applicationdated December 21, 2009 (Reference 1) indicates that "[tjhesecycle specific analyses will also ensure that the core loading has been designed such that the ITAs will not be the most limiting fuel assemblies at any time during the operating cycles, based on planned control rod patterns." Explain the relationshipbetween the "ITAs not being the most limiting assemblies"and the "plannedcontrol rod patterns."

RESPONSE TO RAI#3 The GESTAR Lead Test Assembly (LTA) process allows for the introduction of small quantities (less than approximately 2% of the total bundles in a core) of new fuel product designs in non-limiting reactor core locations without the need for full NRC review, evaluation and approval as long as the analysis of the LTAs uses approved methods and meets the approved criteria.

The GE14i design involves only a small change to a fully approved fuel design and utilizes 2 of 31 LR-N10-0163 previously licensed materials, bundle designs and analytical methods. Although the pilot project is not being licensed as an-LTA program and is, undergoing full NRC review, evaluation and approval, the-conservative design practice of introducing a quantity,,of less than 2% of the total bundles in a core into non,-imiting'core positionsis still being employed. This introductory approach is not required but is being utilized for an additional level of conservatism and to be consistent with precedent for introducing new fuel designs.

The placement of thetlsotope Test Assemblies (ITA) in the Hope Creek cycle 17 core was addressed with the normal GNF and PSEG core design processes; Specifically for control rod patterns, the procedures are designed tO ((

GNF has significant experience'with new fuel product line introduction and even' has experience introducing segmented fuel rods under the LTA provisions stated above. This LTA process has shown to be valuable in obtaining surveillance data to verify that a fuel bunrdle desigh'i perfb6rm~s' satisfactorily in service prior to implementation on a production basis.

NRC RAI#4. 5 and 6 The response to Questions 4, 5 and 6 will be provided'in a subsequent letter.

NRC RAI#7 In response to Clinton Power Station (CPS) RAI Number 10 on page 15 of Attachment 7 to the application dated December 21, 2009 (Reference 1), PSEG stated that "[t]he response to RAI 10(a)'is incorporated'ihtoSection 2. 1, 'New Design Features,and Section 4.6, Manufacturing Qudaity A~ssurance, of NEDC-33529P,Revision 0, '"Safety Analysis Report to Sup port Introduction of GE14i Isotope"Test Assemblies>.(I TA's) in Hope Creek Geherating-Station"."

These afoire-mentioned sections do not contain the Table 2, "C6'balt Targ'et Material Content" and Table 3, "Nickel Plating MaterialContent;," included in Exelon's response to RAi" 0 for CPS (Reference 3). These two tables list the Cobalt and Nickel coating material compositions of the cobalt pellets that were used for CPS. Please address whether these tabl6s'areappiicable to the Cobalt and Nickel coating for HCGS. If they are not applicable,provide new tables for the Cobalt and Nickel coating 'materialcomposition.

3 of 31 LR-N10-0163 RESPONSE TO RAI#7 Table 2, "Cobalt Target Material Content" and Table 3, "Nickel Plating Material Content,"

included in Exelon's response to RAI 10 for CPS (Reference 3) listing the Cobalt and Nickel coating material compositions of the cobalt targets that were used for CPS are also applicable to the Cobalt and Nickel plating for HCGS. The information is repeated here for completeness.

Cobalt Target Material Content Material %Content Nickel Plating Material Content Material  % Content

((__________ I NRC RAI#8 Provide a detailed engineering sketch of the cross sectional view of a Cobalt isotope rod showing the target placement rod (TPR), inner tube, and outer tube. The drawing should show diameters of the tubes, thicknesses of the walls of the tubes and sizes of gaps between the TPR, inner and outer tubes. This detailed diagram will enable the NRC staff to verify the licensee's thermal-mechanicalevaluation of the GE14i segmented rod and related confirmatory calculations.

4 of 31 LR-NIO-0163 RESPONSE TO RAI#8 Figure 1 below provides the, requested sketch.

FR Figure 2. Isotope Rod cross Section NRC RAI#9 Table 3-1 on page 33 of NEDC-33529P (Reference 2) lists a summary of methodologies and analysis codes applicable to the GE14i ITAs. Please add a column to this table that lists all references for each of the methodologies and the respective analysis codes with revision numbers. Also include the details of the references in the -Reference section of NEDC-33529P.

RESPONSE TO RAI#9 Table 3-1 on page 33 of NEDC-33529P (Reference 2) will be modified as follows:

5 of 31 LR-N10-0163 Table 3-2 Summary of GNF Methods Applicability to GE14i Methodology Analysis Version Supported Reference Code TGBLA 06 X Nuclear PANAC PANAC 06 11 X 3,20 Thermal ISCOR 09 X 21 Hydraulic Safety Limit GESAM 02 X 22, 23, 24 MCPR Transient ODYNM 10 X 25, 26, 27 Analyses TASC 03 X 28 ISCOR 09 X 21 PANAC 11 X 3,20 Stability ODYSY 05 X 29 TRACG 04 X 30 TASC 03 X 28 ATWS ODYNM 10 X 31 Thermal Menal GSTRM 07 X 32, 33 Mechanical LAMB 08 X 34 ECCS-LOCA TASC 03 X 28 SAFER 04 X 35 NEDC-33529P Section 6 Reference additions:

20. NEDE-30130-P-A, "Steady State Nuclear Methods," April 1985.

21. The ISCOR code is not approved by name. However, the SER supporting approval of NEDE-24011-P Rev. 0 by the May 12, 1978 letter from D. G. Eisenhut (NRC) to R.Gridley (GE) finds the models and methods acceptable, and mentions the use of a digital computer code. The referenced digital computer code is ISCOR. The use of ISCOR to provide core thermal-hydraulic information in reactor internal pressure differences, Transient, ATWS, Stability, and LOCA applications is consistent with the approved models and methods."

22. NEDC-32601 P-A, Methodology and Uncertainties for Safety Limit MCPR Evaluations, August 1999

23. NEDC-32694P, Power Distribution Uncertainties for Safety Limit MCPR Evaluation

24. NEDE-2401 1-P-A on Cycle Specific Safety Limit MCPR (TAC Nos. M97490, M99069 and M97491), March 11, 1999, Amendment 25.

25. NEDO-24154-A, Qualification of the One-Dimensional Core Transient Model (ODYN) for BWRs, Vol. 1, August 1986.

26. NEDO-24154-A, Qualification of the One-Dimensional Core Transient Model (ODYN) for BWRs, Vol. 2, August 1986.

27. NEDE-24154-P-A, Qualification of the One-Dimensional Core Transient Model (ODYN) for BWRs, Vol. 3, August 1986.

6 of 31 LR-N10-0163

28. NEDC-32084P-A Rev. 2, TASC-03A - A computer program for Transient Analysis of a Single Channel, July 2002.

29. NEDC-32992P-A,,ODYSY Application for Stability Licensing Calculations, July 2001.

30. NEDO-32465A, Reactor Stability Detect and Suppress Solutions Licensing Basis MethodoJpgy for Reload Application, August 1996.

31. NEDC-241"54-P-A, Suppl'enent 1 - Volume 4, Revision 1, Qualificati6h of the One-Dimensional Core Transient- Model for Boiling Water Reactors, February 2000.

32. MFN-036-85, Acceptance for Referencing of LicensingTopical Report NEDE-24011-P-A Amendment 7Ao Revision 6, "GE Standard.Applicati0n for Reactor Fuel Letter",

C.O. Thomas (NRC) to J. S. Charnley (GE), March1.1, )1985.

33. MFN-082-85, Letter, C. 0. Thomas (NRC) to J. S.- Charnley (GE), Aceptance For Referencing of LTR NEDE-24011-P-A-6, Amendment .10,"GE'Standard Application for Reacto'r Fuel," May ?8, 1985.

34. NEDE-20566-P-A, "General Electric Company Analytical Model for Loss-of-Coolant analysis in Accordance with 10CFR50 Appendix", Volumes 1-3, September 1986.

35. NEDE-23785-1-PA, Revision 1, "GESTR-LOCA and SAFER Models for the Evaluation of the Loss-of-Coolant Accident," Volumes II and III, October 1984.

NRC RAI#10 Explain in detail, with. assumptions, analysis, and calculations, why the licensee concludes that the GE14i ITAs will not have significant'impacton 'the in-core inhstrurientation-andcore monitoring system -of.the HCGS (as discussed in Section 3.2.33of NEDC-33529P (Reference 2)). Section 3.2.3 of NEDC-33529Pcontains-insufficient information to comp5lete arn effective and efficient .reviewb'6fthe. material.cited. In additi ;,provide an 'evaluationof the gamma radiation effects fro&4 the GE14i assemblies on other vessel internal components>. .

RESPONSE TO RAI#10 The major sources of gamnmas or photons in tle operating reactor are from the fission events and the, neutron captur events. Tolform the TIP instrumentation correlation constants (J-factors), g'amma sources-frbnm.each material region and signal contribution (attenuation) factors are used to deieirýine, the amount of gamma energy deposited in the Gamma TIP detector.

The rdleased energy from a neutron capture reaction in Co-5.9 is approximately 7.5 MeV per capture (mass defect) and approximately 2.5 MeV per decay from the subsequent decay of Co-60;to Ni-60. The total energy (approximately 10.0 MeV/event),fro'mthe neutron capture and the subsequent decay .of Co-60 is assumed to -occur at the time of.neutron captur'e. This assumption will 6ver-estimate the gamma energy from the cobalt material.early in' life but the error will reduce as the contribution from the decay of Co-60 increases. The assumption that the decay energy, is released at time zero, is consistent with the TGBLA assumption for all explicitly modeled fission product isotopeswith si.milar half-lifes.

The methodology for determining the gamma eneIrgy heatdeposition in the gamma detector incorporates the energy released from each nuclear reaction event (capture, fission, and decay), spatial location of the event; attenuation due to material between the event and the gamma detector, and the energy .deposition in the gamma detector configuration. This methodology has been used in BWRs since the mid 1980s.

7 of 31 LR-NI0-0163 To determine the contribution from the cobalt material in total Gamma TIP signal, an evaluation was performed with the energy released from the cobalt capture and decay defined as zero.

This demonstrates the error that would result if the gamma production in the cobalt rods were ignored. Due to the location and source strength of the cobalt isotope rods, the total gamma energy deposition in the gamma TIP detector from the cobalt material in four surrounding GE14i lattices is approximately [U )) or less. By including the cobalt gamma energy release model in the Gamma TIP detector signal correlation, the impact of the cobalt material on the accuracy of the Gamma TIP signal is reduced to a level significantly below (( )).

For the neutron in-core instrumentation (LPRMs), the in-core instrumentation signal-to-lattice power relationship is formed using the thermal detector J-factor. The thermal detector J-factor provides the relationship between the lattice power and the signal generated by the LPRM detector. The cobalt rods are explicitly modeled in the GE1 4i design and the impact of the cobalt neutron absorption is incorporated in the thermal J-factors and neutron flux predictions at the LPRM instrumentation location.

With the inclusion of the cobalt material effects in the lattice physics model, the perturbations on the instrumentation (gamma or neutron) signal from cobalt material are captured, and the adequacy of the in-core instrumentation is assured.

Additionally, the replacement of a fission material bearing fuel rod with a cobalt isotope rod will result in approximately a factor of 10 reduction in the gamma energy emitted from that rod location, as supported by NEDC-33529P, Section 4.4. The gamma energy from the fission material bearing rod is generated from prompt fission gammas, delayed fission gammas, and neutron capture gammas in actinides and fission products. Only after reactor shutdown and subsequent decay of short half life fission products and actinides in the ITA will the gamma from Co-60 decay become a significant contributor to the total gamma energy production. Therefore, the effects on the material characteristics of instruments or local vessel internals will be bounded by what is seen from a U0 2 fuel rod at that location.

NRC RAI#11 General Electric (GE) letter MFN 07-040 to the NRC dated January21, 2007 (ADAMS Accession No. ML072290203), provided an evaluation of potential non-conservatism in the GE Thermal-MechanicalMethodology, GSTRM. Please provide an evaluation of the impact of the information in MFN 07-040 on the adequacy of the use GSTRM model in the thermal-mechanical evaluation of the GE14i fuel bundle. This evaluation should containjustification for the use GSTRM methodology in the following areas of thermal-mechanicaldesign of GE14i:

" Internal pressure design

" Clad mechanical analyses

" Loss-of-coolant accident response

" Cladding strain analysis; and

" GSTRM calculated gap conductance that is used in the stability and transientanalyses.

8 of 31 LR-N10-0163 RESPONSE TO RAI#11 In MFN 07-040,GNF evaluated a potential non-conservatism in the GSTRM thermal-mechanical calculations. .Specifically, [

S] model on the;GSTRM fuel temperature, fuel design analyses, and downstream safety analyses have been evaluated. As, reported in the MFN 07-040 the.evaluated condition does. not constitute a reportablecondition per 10 CFR 21:- The NRC staffs evaluations of MFN 07-040 and associated supplements recommend an additional ((

)) for the GSTRM fuel rod internal pressure analyses to address ((

)) model in GSTRM (Reference -R-1):

As requestediin this RAI, the applicability of the GSTRM methodology tothe GE14i design.

analyses;, including the MFN .07-040.evaluatibn/conclusions and the NRC staff recommendations, has been evaluated and the following conclusions have been made.

LHGR limits-for the full length U0 2 rod, partial length U0 2 rod and Gadolinia containing rods have been updated to include the.additional (( . . ))for the GSTRM 'od inter.n al pressure.anaiyses (Reference R-2). G E 14i bundles for Hope Creek Generatihg Stati6n are designed with these revised LHGR limits and will be monitored in the core based on these revised LHGR limits.

-NRC staff recommended,(( .. , )) is-not applicable forthe. GE14i qob rods as~fuel failure due to excessive internal pressure is not a likely failure..mechanism for these -isotope, rods. ((

.. .~].an*d,also there is no fission gas ,release-from the cobaltitargets to increase the rod internal, pressure during, the' rraiati0 ."As- netrIesult, the rod-internal pressure -at, the- end, of life-is sig.lificantly., beo :.ther.,reactor.system-pressure and-thus the.fuel failure due to high rod internal pressure is not a likely~failure mechanism for these rods and no additional pressure design margin is required.

MFN 07-040 also demonstrated the applicability of GSTRM for the cladding mechanical analyses, loss-of-coolant accident response, cladding strain analyses and the 'gap conductances, generated by GSTRM for the transient and stability analyses. The cladding mechanical/strain analyses and the downstream safety analyses are ((

)) and thus the application of GSTRM with its conservative uncertainties treatment is'adequate for.those analyses.

NRC staffs evaluation of 'the MFN 07-040 also did not recommend any additional design margins for these calculations.

References:

[R-1] Appendix F of the NRC SER for NEDC-33173P, "Applicability of GE Methods to Expanded Operating Domains," July 21, 2009.

[R-2] Appendix C of the NEDC-32868P Revision 3, "GE14 Compliance With Amendment 22 of NEDE-2401 1-P-A (GESTAR II)" April 2009.

9 of 31 LR-N10-0163 NRC RAI#12 Provide a detailed description of the stability methodology mentioned in Section 3.2.6 of NEDC-33529P (Reference 2). The information contained in Section 3.2.6 is not sufficient for a full review of the methodology.

RESPONSE TO RAI#12 Detailed descriptions of the stability analysis and methodology mentioned in Section 3.2.6 of NEDC-33529P (Attachment 3 of Reference 1) are described in Section S.4 of GESTAR (Reference 2 in Attachment 3 of Reference 1).

Hope Creek Generating Station implements the Option III stability Long-Term Solution (LTS).

The plant and cycle-specific calculations required for the Option III stability LTS are described in Section S.4.1 of Reference R-2. The cycle-specific stability analyses are also described in the HCGS response to Clinton Power Station RAI Number 6 in Attachment 5 of Reference 1.

The approval status of the codes mentioned in Section 3.2.6 of Reference R-1 are summarized in the table below.

Cohmputer Code Vyersion or ~ NiC ~ Comments

~Revision Approved ~

ISCOR 09 Y(1) NEDE-24011-P Rev. 0 SER PANACEA 11 Y NEDE-30130P-A (2)

ODYSY 05 Y NEDC-32992P-A TRACG 04 N(3) NEDO-32465-A (1) The ISCOR code is not approved by name. However, the SER supporting approval of NEDE-24011P, Rev. 0, by the May 12, 1978 letter from D. G. Eisenhut (NRC) to R. Gridley (GE) finds the models and methods acceptable, and mentions the use of a digital computer code. The referenced digital computer code is ISCOR. The use of ISCOR to provide core thermal-hydraulic information in stability applications is consistent with the approved models and methods.

(2) The physics code PANACEA provides inputs to the transient code ODYN or to TRACG.

The improvements to PANACEA that were documented in NEDE-30130-P-A were incorporated into ODYN by way of Amendment 11 of GESTAR II (NEDE-24011-P-A). The use of TGBLA Version 06 and PANACEA Version 11 in this application was initiated following approval of Amendment 26 of GESTAR II by letter from A. A. Richards (NRC) to G. A. Watford (GE)

Subject:

"Amendment 26 to GE Licensing Topical Report NEDE-2401 1-P-A, GESTAR II Implementing Improved GE Steady-State Methods," (TAC NO. MA6481), November 10, 1999.

(3) TRACG02 has been approved in NEDO-32465-A by the US NRC for the stability DIVOM analysis. The current licensed thermal power (extended power uprate conditions) stability analysis is based on TRACG04, which has been shown to provide essentially the same or more conservative results in DIVOM applications as the previous version, TRACG02. The use of TRACG04 at HCGS was introduced with acceptance of License Amendment 163, implementing Ayerage Power Range Monitor, Rod Block Monitor Technical Specifications concurrent with Maximum Extended Load Line Limit Analysis at HCGS.

10 of 31 LR-N10-0163 NRC RAI#13 Section 4.5. 1 of NEDC-33529P (Reference 2) provides a brief qualitative assessment of the impact of GE14i ITAs on thermal-hydraulicinstabilityfor HCGS. During an audit performed by the NRC staff supporting the review of the proposed amendment, the staff was informed that a cycle-specific stability analysis will be performed for the up-coming cycle to determine the impact of GE14iITAs.on stability. Please provide details of the stability analysis.

RESPONSE TO RAI#13 The cycle-specific stability analysis will be performed for the upcoming cycle and provided to the NRC by OJly '8, 2b10. o NRC RAI#14 Section 4, "Licensing Evaluations"of NEDC-33529P(Reference 2) states that "[c]ycle-specific analyses.will be performed for HCGS Reload 16 Cycle 17 to establish fuel operatinglimits for.

th* lTA.s'that assýre corhmpliance with regulatorylimits." Provide the NRC staff with a summary of tfheI.CGS Relo6ad,1i6 CycVle 17, Supplemental Reload Liceosing Report.(SRLR) for review and v6rific.tion, "ftheresults of the cycle-specific analyses. This' re'prt'should.be similarto Global Nuclear Fuel report 0000-0099-4244-SRLR, Revision 0, "SupplementalReload Licensing Report for, d-inton'Power S.tat .Oion Unit I Reload 12 Cycle 13" attached toExelon's leiter RS 171 dated bernhbe-r 14,, .2009, for CPS (ADAMS Accession No.. ML093490375).

RESPONSE TO RAI#14 The HO"S Rkload 16 Cycle 17 Supplemental Reload Licensing Report (S.LR) will be provided to the Nt:RC by August 4, 2010.

NRC RAI#15 Section 4.2.1 of NEDC-33529P (Reference 2) states that "[t]he GE14i ITAs represent a small fraction of the total bundles in the core. As a result, their impact on the core.averagenuclear.

parametersis negligible. Furthermore,,the hydraulic characteristicsof GE14i ITAs are,similarto the GE14 bundles. Therefore, as in HCGS Cycle 16 (Reference 77), a cycle-specific A TWS

[fnticipated transientswithout scram],analysis is not requiredbecauseof the introduction of GE14iiTAs." . , -

a) Provide details of the disposition of the ATWS event at HCGS for Cycle 17 and justify that the A TWS acceptance criteriaas listed in Section 2. 14.2 of Reference 4 has been met.

b) What would be a minimum threshold number of ITAs in the HCGS core that would require the licensee,to perform a reanalysis of the A TWS event?

11 of 31 LR-N10-0163 RESPONSE TO RAI#15 (a) HCGS has margin to the ATWS acceptance criteria as shown in the cycle-independent calculation results documented in Reference R-1. This document is also referenced in NEDC-33529P. The GE14i ITA geometry and enrichment are similar to GE14. Therefore, the differences in nuclear characteristics of the GE14i bundle design will not be any greater than what is expected when transitioning to a different BWR fuel design. The impact of a core-wide

(( )) increase in ODYN void coefficient on the ATWS analysis has been assessed for reactor cores consisting of (( 1] fuel designs. The results documented in Tables 1 &

2 provide evidence that a (( )) increase in void coefficient has a (( )) on suppression pool temperature and (( )) change in peak vessel overpressure.

It is known that the plant response during an ATWS event is primarily affected by plant characteristics (SRV capacity, SLCS operating parameters, ATWS recirculation pump trip, etc). Minute changes in fuel design being loaded in small quantities (<2% batch fraction) does not impact the conclusions of Reference R-1. As such, ATWS is treated on a plant specific, cycle independent manner.

Table 1 Peak Vessel Pressure Void Coefficient Study Event and Description Exposure Peak Vessel, Pressure (MPa)

PRFO Base Case BOCO PRFO with (( )) void coefficient increase BOC PRFO Base Case EOC PRFO with (( )) void coefficient increase EOC Table 2 Peak Suppression Pool Temperature Void Coefficient Study Event and Description Exposure Peak Pool Temperature (0 C)

PRFO Base Case BOC ((

PRFO with (( )) void coefficient increase BOC PRFO Base Case EOC PRFO with (( )) void coefficient increase EOC ))

References R-1. GE Nuclear Energy, "Safety Analysis Report for Hope Creek Constant Pressure Power Uprate," NEDC-33076P, Revision 2, August 2006.

12 of 31 LR-N10-0163 (b) The proposed changes to TS 5.3.1 in LAR H09-01 specifically state: "A maximum oftwelve GE14i Isotope Test Assemblies may be placed in non-limiting core regions." As part of the ITA pilot program 12 is the maximum number of ITAs that will be placed in the HCGS. core.

The analysis provided in LAR H09-01. provides adequate technical justification for operation with 12 ITAs. Itis'known that the plant response durihg an ATWS ebvent is primarily affected by plant characteristics (SRV capacity, SLCS operating parameters, ATWS recirculation pumrp tip, etc). Mind ie changes in fuel design being ioaded in small quantities"(<2% batch fraction) does not impact the ATWS analysis cond1usions.

NRC RAI#1.6 Section 4.3,, "Evaluationof Design-Basis'Accidents," of NEDC-33529P (Reference2-)states:

The HCGS Design-BasisAccidents (DBAs). to be evaluated are,identified in Chapter 15.0 of the HCGS Updated Safety:Analysis Report (UFSAR).

The Control Rod Drop Accident (CRDA), Main!Steamline Break (MSLB) accidentoutside containment, Fuel Handling Accident (FHA); and Loss-of-Coolant Accident (LOCA) are licensed under 10 CFR 50.67, utilizing Alternate Source Term (A,ST), methodology per.Regulatory,Guide (RG)

Per RG 1. 183, 'Alternative Radiological Source Terms for-Evaluciating Design Basis Accidents at Nuclear Power Reactors," dated.July.2000 (ADAMS Accession No. ML003716792), Regulatory.Position C.1.3.32;"Rearnaly.yis Giuideline,"

The NRC staff does not expect a complete recalculaf6in of all facility radiologicalanalyses, but does expect licensees to evaluate all impactsof the proposed changes-nc to.qp'daie.l}*e affected aibljy'es and the, design bases appropriately. An analysis is considered to be affected if the proposed modification changes one or more assumptions or inputs used in that analysis such .that the-results, or the conclusions drawn on those results, .ar~enolonger valid.,

Also, RG. 1.183 Section B, "Discussion," states:

Although the LOCA is tyjbically the maximum credible accident, NRC staff experience in reviewing license applicationshas indicatedthe needto considerother-accident sequences of lesserconsequence but higher probabilityof occurrence.

Page 15.0-5 of Standard Review Plan (SRP) 15.0., "Introduction- Transient and Accident Analyses," Revision 3, dated March 2007 (ADAMS Accession No. ML070710376) states:

The reviewer-considersthe po~ssible case-variationsof AQOs [anticipated operationaloccurrences] and postulated accidents presented to verify that the licensee has identified the limiting cases.

The proposed change only provides an evaluation of the impact on the DBAs described above. Pleaseprovide an evaluation of the impact of the proposedchange on all 13 of 31 LR-NIO-0163 accidents in the design bases or include a justification why an evaluation of the impact is not needed. If an evaluation of other DBAs is provided, please provide the regulatory bases for the acceptance criteria(i.e., 10 CFR Part 100, 10 CFR Part 50.67) and any regulatoryguidance or SRPs used to make this determination.

RESPONSE TO RAI#16 A detailed explanation of all probable isotope rod failure modes is provided in Section 2.2 of NEDC-33529P (Attachment 3 to LR-N09-0290, LAR H09-01). Section 2.2 also describes key protective design features; the isotope rods will operate at a significantly lower heat generation rate compared to fuel rods, the isotope rods have a double layer of Zircaloy encapsulation before exposure of the nickel-plated cobalt targets, and the isotope rods have Zircaloy connections at all spacer locations. Section 2.2 provides a technical basis to conclude that isotope rods are not more vulnerable to common failure modes than normal fuel rods during operation. Section 2.3, Online Failure Detection, also provides the HCGS ability to measure changes in cobalt-60 activity and take appropriate response. The response to RAI#21 of this attachment provides further discussion on failure modes and cobalt detection.

Section 4.3 of NEDC-33529P, Evaluation of Design-Basis Accidents, identifies that the HCGS Design Basis Accidents (DBAs) to be evaluated are identified in Chapter 15 of the HCGS Updated Final Safety Analysis Report (UFSAR). The section states that Control Rod Drop Accident (CRDA), Main Steam Line Break (MSLB) accident outside containment, Fuel Handling Accident (FHA), and Loss-of-Coolant Accident (LOCA) are licensed under 10 CFR 50.67 utilizing Alternate Source Term (AST) methodology per Regulatory Guide (RG) 1.183. In addition to these four events, Chapter 15 of the HCGS UFSAR identifies five other events classified as limiting faults. These events are:

6. Reactor Recirculation Pump Shaft Seizure (UFSAR 15.3.3)

7. Reactor Recirculation Pump Shaft Break (UFSAR 15.3.4)

8. Instrument Line Break (UFSAR 15.6.2)

9. Feedwater Line Break - Outside Primary Containment (UFSAR 15.6.6)

10. Gaseous Radwaste Subsystem Leak or Failure (UFSAR 15.7. 1)

As discussed in the response to RAI#21, leakage of cobalt (including entire cobalt targets and/or cobalt particulate) from an isotope rod in an ITA is not a credible event during normal operations, transients or design basis accidents not involving fuel melt accidents (i.e., Loss of Coolant Accident and Control Rod Drop Accident). None of the additional five postulated events involve fuel failures or fuel melt; therefore, isotope rod failure or leakage is not credible during any of these events. Therefore, the radiological consequences for these five events are unchanged for a core operating with isotope test assemblies. The five events are described below.

1. Reactor Recirculation Pump Shaft Seizure (UFSAR 15.3.3)

The seizure of a reactor recirculation pump is a design basis accident that does not result in the failure of fuel. Since no fuel rod failures occur due to the recirculation pump shaft seizure, no GE14i isotope rod failures are postulated to occur, and the consequences of this event will be unchanged in operation with GE14i.

While the consequence of this transient does not result in fuel failure, it does result in the discharge of normal coolant activity to the suppression chamber via SRV operation. Since this 14 of 31 LR-N10-0163 activity is contained in the primary containment, there will be no exposures-to operating personnel. Because this transient does not-result in an uncontrolled release to the environment, the plant.operator-can choose to leave the activity in the primary containment or discharge it to the environment under controlled release conditions. If purging of the containment is chosen, the release will be in accordance with established technical specification limits.

2. Reactor Recirculation Pump Shaft Break (UFSAR 15.3.4)

This event is less severe than the Reactor Recirculation Pump Shaft Seizure event, and consequences of this event-are considered.to be .bounded by the shaft seizure event in HCGS UFSAR.i 5..3.3. .Since nofuel rod.failures occur dueto the recirculationpump shaft break event, no GE 14i isotope rod~failures-are postulated to occur, and-the consequences'of:.,this event will be unchanged:in operation with GE14i.

While the consequence' of this transient does notresult in fuel failure,' it does result in the discharge of normal coolant activity to the suppressionchamber via SRV operation. Since this activity is contained -in the-primary containment,-there will be no exposu.resý to operating.

personnel. Because this transient does not result in an uncontrolled release to the environment, the: pant operator can choose to. leavethe activity in the primary containment or discharge it to.

the enviroriment'under controlled release.conditions. If'purgingof.the containment is chosen; the release will be in accordance with~established'technical specification;limits..,

3. InstrLumejnt Line .Break (UFSAR 15.6.2)

The Instrument Line Break involvdsthd'ipdstUlationof asmall break ina steam or liquid line inside or outside containment but within a controlled release structure.

No fuel damage is associated with,,this,acdident..ZSince no fuel rod failures occurdue td.the instrument line break, no GE14i isbtbOdrod failures are. postulated to occur, and 'the -.

consequence of this event will be unchanged in operationwith GE14i. As-a-resultof depressurizing the Reactdr Coolant System, normal operating concentrations ofiiodine and, noble gases can be released 1incliudihd consideration of iodine spikihg. :.The analysis results indicate that offsite and control room doses are small fractions of 10 CFR 50.67 guidelines.

4. Feedwater Line Break -- Outside' Primary Containment (UFSAR,.15.6.6)-

To evaluate the pipe .breaksin a: large liquid prbcess line outside primary.containment, -the failure: ofa feedwater line,is assumed. The feedwateir line break outside primary containment results innofuel failures.:-Since-no fuel rod failures occur due to thefeedwater-line break outside.primary containment!, no GE14i isotope rod failures are postulated to 'occur, and the radiological consequences of this event will be unchanged in operation with GE1 4i.

Though there is no fuel damage as aconisequence of thisraccident, the activity in the. main condenser hotwell prior to occurrence of the break is released. The radiological release considerationis, primarily'one-of iodine'release-;i.:. Noble gas activity, in .the condensate is considered negligible. :The analysis results indicate. that offsite and control' room-doses are small fractions of 10 CFR 50.67 guidelines.

15 of 31 LR-N10-0163

5. Gaseous Radwaste Subsystem Leak or Failure (UFSAR 15.7.1)

The Gaseous Radwaste Subsystem Leak or Failure does not affect the nuclear fuel as there is no reactor core transient associated with this event. Since no fuel rod failures occur due to the gaseous radwaste subsystem failure, no GE14i isotope rod failures are postulated to occur, and the consequences of this event will be unchanged in operation with GE14i. Branch Technical Position 11-5 identifies that only radioactive noble gases (xenon and krypton) are to be considered to be released to the environment since the assumed transit time is long enough to permit major radioactive decay of oxygen and nitrogen isotopes. The branch technical position also identifies that particulates and radioiodines are assumed to be removed by pretreatment, gas separation, and intermediate radwaste treatment equipment. The analysis results indicate that offsite dose is a small fraction of 10 CRF 100 guidelines.

NRC RAI#17 The release fraction for Co-60 used in the design bases analyses assume that the Co-60 is in the fuel cladding and structuralmaterials. For the proposed change, the Co-60 available to be released during a DBA is not mixed with cladding and structuralmaterials, as considered for the RG 1. 183 release fractions, but is in high concentrationswithin the isotope rods. Pleasejustify why the DBA Co-60 release fraction used is applicable or conservative for the proposed isotope test assemblies. Please include any experimental data to justify the proposed release fraction.

RESPONSE TO RAI#17 The design of the Isotope Test Assemblies (ITAs) is such that the nickel-plated cobalt (Co) targets in the ITAs are isolated from the reactor environment by a double layer of Zircaloy encapsulation. Because there is no uranium fuel present in the cobalt isotope rods, the isotope rods have much lower heat generation than fuel rods. It is expected that the lower heat generation rate and double Zircaloy barrier features of cobalt isotope rods would justify the assumption that the fraction of cobalt released from the passive isotope rods during a design basis LOCA or CRDA would be equal to or less than the fraction of cobalt released from other passive materials present in the reactor core. However, no experimental data can be provided as further justification for this expectation. Therefore, the methodologies in sections 4.3.1 and 4.3.4 of NEDC-33529P have been updated (as shown below) to include analysis of potentially higher cobalt release fractions for CRDA and LOCA dose evaluations, respectively. The previously assumed release fraction of 0.0025, which is consistent with the recommended post-LOCA cobalt release fraction in RG 1.183, was (( )) and analyzed for CRDA and LOCA. For both accidents, assuming the ((

)) the dose impact of introducing 12 ITAs at HCGS remains negligible.

Updated Sections of NEDC-33529P:

4.3.1 Control Rod Drop Accident The HCGS licensing basis CRDA analyzed in Reference Al assumes a failure of 850 rods (8x8 fuel). The mass fraction of fuel in the damaged rods that reaches or exceeds the initiation temperature of fuel melting is estimated to be 0.77%. Fuel reaching melt conditions is assumed to release 100% of the noble gas inventory and 50% of the iodine inventory. [t 16 of 31 LR-N10-0163

)) Therefore, the licensing basis CRDA radiological analysis is not impacted by the introduction of 12 GE14i assemblies at HCGS.

As.described in Reference 9, comp.liance with licensing limits governing CRDA isassured thr6iugh,adherence t6.the Banked Position Withdrawal Sequence (BPWS). The associated any*sl have genenriallyddemor*strated large mar'gin to iicensihglimiits goye'rnig "

acdeptabl- e'tlpy' iisrtions. The BPWS analyses demonstrated that tha chtiacteristic control rod worh associated wih limiting rods ina BPWS sequence are lpow as compared to that required to challenge :the_ 280- cal/gm fuel design limit. The reactivity chtaracteristics of GE14i are sir ia~r to GEl4~thereforejthe introduction of 12 GE1.4i"assermblies at HCGS will have negligible effects on the existing CRDA margin. In addition t6-similar fuel reactivit5 characteristics, the. irn pact on the rod worths is constrained by other design factors such as shutdown margin and,in-sequence rod worths.

4.3.4 Loss-of-Coolant Accident (LOCA) the HCGS LOCA source term was previously evaluated in Reference A2. The impact of 12 GE14i assemblies on the HCGS licensing basis LOCA source term and radiological consequences was evaluated.

17 of 31 LR-NIO-0163 The introduction of 12 GE14i bundles at HCGS presents no significant impact on the AST LOCA source term.

6. References

9. GE Hitachi Nuclear Energy, "General Electric Standard Application for Reactor Fuel (Supplement for United States)," NEDE-2401 1-P-A-16-US, Revision 16, October 2007.

Al. Calculation H-1-CG-MDC-1 795, Revision 5, "Control Rod Drop Accident Radiological Consequences", June 2007.

A2. Calculation H-1-ZZ-MDC-1880, Revision 3, "Post-LOCA EAB, LPZ and CR Doses",

September 2009.

NRC RAI#18 Section 4.3.4, "Loss-of-CoolantAccident (LOCA)," of NEDC-33529P (Reference 2) states that the HCGS LOCA source term was previously evaluated for an extended power uprate (EPU).

The first sentence in the 2nd paragraphof this section makes a statement regardingone of the assumptions for the HCGS EPU LOCA source term. This statement appears inconsistentwith a calculation submitted by the licensee in support of the EPU license amendment review.

Specifically, the statement in NEDC-33529Pappears to be inconsistent with the isotopic core inventory information shown in Section 5.3.1.3 of PSEG Calculation Number H- 1-ZZ-MDC-1880, "Post-LOCA EAB, LPZ and CR Doses," Revision 21RO (ADAMS Accession No. ML063110185). Please resolve this apparentinconsistency and provide a revisedjustification for the impact of the proposed change on the LOCA analysis as necessary.

RESPONSE TO RAI#18 The analysis documented in NEDC-33529P Section 4.3.4 has been modified (see RAI#i7 response) to correctly consider the Co-60 present in the HCGS licensing basis post-LOCA radiological consequences evaluation source term. The revised analysis is consistent with the HCGS licensing basis methodology as documented in Calculation H-1-ZZ-MDC-1880 Revision

3. The conclusion that the introduction of 12 GE14i bundles at HCGS presents no significant impact on the licensing basis LOCA source term is still supported.

18 of 31 LR-NIO-0163 NRC RAI#19 Please provide enough information (i.e., design bases parameters,assumptions or methodologies) to replicate the dose results provided in NEDC-33529PSection 4.3.1, "Control Rod Drop Accident," and Section 4.3.4, "Loss-of-CoolantAccident (LOCA)," and provide the results of the calculationin rem Total Effective Dose Equivalent. If the only change is to add Co-60 to calculationnumber H- 1-CG-MDC-1795, Revision 4, "ControlRod Drop Accident RadiologicalConsequences," and H-1-ZZ-MDC- 1880, Revision 21RO, "Post-LOCA EAB, LPZ and CR Doses," please state this in your response.

If any design bases parameters,assumptions or methodologies (otherthan those provided in NEDC-33529P)were changed in the radiologicalDBA analyses used to support the proposed amendment change, please provide them. If there are many changes it would be helpful to compare and contrast them in a table. Also, please provide a justification for any changes.

The NRC staff has found that the efficiency of the review can be increasedby having the calculations available for review. In addition to providing any changes to the current licensing bases and justifications for these changes, the licensee is encouraged to provide above requested information (i.e. design bases parameters,assumptions or methodologies) by providing'the modified calculations-(LOCA and Contol Rod.Drop Accident) including their attachrhents., As an alternativ/e; the information may be provided in some other.format.

RESPONSE TO RAI#19 The radiological analyses examining the effect of introducing GE14i ITAs on licensing basis CRDA and post-LOCA doses have been updated (see RAI#1 7 response) to be consistent with HCGS calculations H-1-CG-MDC-1795 Revision 5 and H-1-ZZ-MDC-1880 Revision 3 and present results in rem Total Effective Dose Equivalent (TEDE). The alternative RADTRAD-analyses performed and their differences from these HCGS calculations are described in the revised sections 4.3.1 and 4.3.4 Of NEDC-33529P.,

The current revisions of Calculations HW1 -CG-MDC-1795 (Revision 5)2, "Control Rod Drop Accident Radiological Consequences," and H-I-ZZ-'MDC-1880 (Revision 3), "'Post-LOCA EAB, LPZ and CR:Doses,," areprovided as Attachments 4,and 5 of.this submittal. During an April, 6 auditsupporting the review of.this. LAR; the NRC asked for~clar'ification ontwo issues related to H-1-ZZ.MDC-*1880,; and subsequently asked for the 10CFR50:59 evaluation that was performed for H-1-ZZ-MDC-1880,- Revisidn.,3-. The two issues identified during the audit are discussed below; inbluding:discussion and comparison tables on parameters,,assumptions and methodologies in the current licensing basis; the 10CFR50.59 evaluation is provided as to this submittal.

Issue 1: For the HCGS LOCA dose calculation H-1-ZZ-MDC-1880, Revision 3, specific to the MSIV leakage, why did the doses, *o down from Revision 2 to Revision 3?

The HCGSfull scope AlternativerSource Term (AST) license amendment request, and.

subsequeht Amendment 134; dated October-3.; -2001, included an aerosol deposition modelfor the Mainý Steam Isolation Val~Ie(MSIV) leakage path based on the guidance in NRC document AEB 98-03 (Reference 5). The aerosol depbsition model that was subsequently included in H-1-ZZ-MDC-1 880, Revision 2, was developed using the following very conservative assumptions:

2 Revision 5 only corrected typographical errors; Revision 4 was previously docketed to support the HCGS

• EPU amendment request (ADAMS ML063110190) 19 of 31 LR-N10-0163

6. The Technical Specification MSIV leakage rates of 150 scfh (2.5 cfm) and 50 scfh (0.833 cfm) were modeled without reducing these leak rates to address post-LOCA primary containment pressure and temperature conditions. A 50% reduction in the MSIV leak rate was credited after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

7. One volume node for each release path - MSIV failed line and intact line - was modeled, with one aerosol removal efficiency per path. Although HCGS has seismically supported main steam lines beyond the outboard MSIVs, the piping upstream and downstream of the outboard valve was modeled as a single volume.

8. One aerosol settling velocity of 4 0 th percentile was used for both MSIV failed and intact lines, upstream and downstream of the outboard MSIVs.

Subsequent HCGS plant receiving its AST license amendment, the industry and NRC gained experience with, and an understanding of, aerosol deposition in the main steam lines following a LOCA. The NRC informed some AST license amendment applicants of a concern related to the modeling of lighter aerosol particles, which experience lesser gravitational deposition in the seismically supported lines beyond the outboard MSIVs. This concern was addressed in AST license amendments for the Peach Bottom (PB) plant. While this issue was not identified as an industry concern, PSEG NUCLEAR made the prudent decision to address the concern in H ZZ-MDC-1880, Revision 3, by updating the aerosol deposition model with respect to the latest regulatory developments (see discussion below). This resulted in some loss of dose margin.

In 2009, PSEG NUCLEAR initiated a revision to the HCGS LOCA analysis in H-1-ZZ-MDC-1880, Revision 2, to (1) allow for keeping the primary containment isolation valves (PCIVs) open for 120 seconds post LOCA, and (2) increase allowable Engineered Safety Feature (ESF) leakage from 1.0 to 2.85 gpm. H-1-ZZ-MDC-1880, Revision 3, was updated as follows:

8. Each piping segment upstream and downstream of the outboard MSIVs in the MSIV failed and intact lines were modeled as well mixed volumes. Two well mixed volumes for each MSIV release path is consistent with AEB 98-03 (See Table 1, Item 1).

9. The MSIV leakage in the release path was reduced based on the post-LOCA drywell and wetwell pressure and temperature, which significantly reduced the MSIV leakage and consequently reduced the resulting doses from the MSIV leakage paths (Table 1, Items 2 and 3).

10. The aerosol removal in each MSIV release path was divided between two well mixed volumes, which created two aerosol removal filters in a series configuration that reduced the aerosol released to the environment by factors of about 4 and 10 for the MSIV failed and intact lines, respectively (as shown in the computations provided in Table 1, Items 4 and 5). The MSIV leakage dose from these release paths were reduced proportionately.

The aerosol removal filter efficiencies were calculated using the horizontal projected area (diameter x length) of the main steam piping.

11. Hold-up times of 9.32 hrs and 29.52 hrs were credited for the MSIV failed and intact lines, respectively, in Revision 2. The hold-up time credit is not appropriate for the well mixed volumes; therefore, hold-up times are not credited in H-1-ZZ-MDC-1880, Revision 3, which is conservative with respect to radiological consequences (Table 1, Item 6).

20 of 31

Attachment 3 LR-NIO-0163

12. The maximum primary containment isolation valve (PCIV) isolation time was increased to 120 seconds (Table 1, Item 7). The open PCIVs present a release path to the environment for airborne containment activity due to the radionuclide inventory in the

.reactor coolant system liquid which is not considered, in H-1-ZZ-MDC-1880, Revision 2.

13. The allowable ESF leak rate was increased from 1.0 gpm to 2.85 gpm to facilitate acceptable results for-future plant maintenance surveillances(Table 1, Item 8).

14. The removal of the elemental iodine by wall deposition on wetted surfaces inside containment is-modeled in accordance with NUREG-0800, Standard Review Plan 6.5.2 (Table 1, Item 9).

Table 1 presents a comparison of all differences between H-1-ZZ-MDC-1 880, Revisions 2 and 3.

Table 1 Comparison of MSIV Leakage Modeling in H-1-ZZ-MDC-1880, Revision 2-versus Revision 3 Item Design Input H--ZZ--1)DC-1880. H-1-ZZ-1IIDC-1880 No. Information . .. Revision 2 .,Revision 3 I ,Well mixed volume Both release paths. - MSIV-failed Both release paths are model as and intact steam line releases are two well.,mixed volume nodes model as a single node volumes, based on AEB 98-03.

2 MSIV LeakageRate In MSIV Assumed 150 scfh for 0-24 hrs and Reduced based on the~post-LOCA I

Faiileýd Lie 5 schfi for", cfm). containmiht c24hs(2.5/1.25 presscre anad No reductidn in MISIV leaka-6 tmpmerture"

.. .. ,' 0.- 6,,and credited' frdreljost-.LOCA 0.223'cfiiifor 602, 2-24 and 24-conditidn. 72"0

, . r..sp ctiv"'el ' "'hic significantly reduced MSIV 3 -MSIV Leakage Rate In MSIV MSIV leakage-of 100 scfh:was' *- MSIV leakage bf,100 scflh was Intact Line .divided-beiween two intact.:MS. line allocated to one intact MS line -

- 50 scfhlline (0.8334 and 0.417 cfm 100 scfl/line with the leakage for.0-24 and 24-720.hr; . reduction based-on the post-LOCA respectively). No reduction in containment-pres sure and MSIV. leakage credited for drw*ell temperature.(0.539, 0.297, and post -LOCA condition. 0.149 cfm for 0-2, 2-24 and 24-720 hr, respectively), which significantly reduced MSIV leakage.

4 " Aerosol deposition efficiency - One aerosol removal efficieficy was Tw6 aorosolreovalefficiencies MSIV failed line calculated for both the MS1V failed (85i92%4forithie piingp segment and intact lines. The use of one 'betweenthe inboard and-outboard aerosol removal efficiency of MSIVs and-96r96% for the 98,32% for MSIV failed line segment beyond the .outboard resulted in 1.68%: of aerosols ,MSIV) were calculated for MSIV released to the environment, failed piping segments, which resulted in 0.43% of aerosol released to the environment. This reduced the aerosol release by about afactor of 4 (1.68/0.43 =

3.91).

21 of 31

Attachment 3 LR-N10-0163 Table 1 Comparison of MSIV Leakage Modeling in H-1-ZZ-MDC-1880, Revision 2 versus Revision 3 Item Design Input H-1-ZZ-MIDC-1880 H-1-ZZ-MDC-1880 No. Information Revision 2 Revision 3 5 Aerosol deposition efficiency - One aerosol removal efficiency of Two aerosol removal efficiencies MSIV intact line 99.46% was calculated for one (97.32% for the piping segment intact line well mixed volume node between the RPV nozzle and resulting a 0.54% aerosol released outboard MSIV and 97.95% for the environment, the segment beyond the outboard MSIV) were calculated for MSIV intact piping segments, which resulted in 0.055% of aerosol released to the environment. This reduced the aerosol release by about a factor of 10 (0.54%/0.055% = 9.82) 6 MSIV Leakage Holdup Time Hold up times of 9.32 hrs and 29.52 Holdup times are not credited for hrs were credited for the MSIV any MSIV leakage path, which is failed and intact lines respectively conservative with respect to radiological consequences..

7 Primary contaimnent isolation PCIVs not modeled as a release Primary containment isolation valves (PCIVs) path (i.e., the PCIV release path is valves (PCIVs) remain open for isolated prior to the onset of the 120 seconds.

AST gap release).

8 Allowable ESF Leakage Rate Allowable ESF leak rate of 1.0 gpm Allowable ESF leak rate increased (modeled as 2.0 gpm) to 2.85 gpm (modeled as 5.7 gppm).

This increased the ESF leakage dose by a factor of 2.85

__(= 2.85 gpm /1.0 gpm).

9 Elemental iodine removal by Not Credited Credited wetted surface deposition The combined effects of the above changes are such that the doses resulting from the MSIV leakage path are reduced substantially. This also demonstrates that the aerosol deposition model in the original AST license amendment based on Revision 2 to H-1 -ZZ-MDC-1880 was extremely conservative.

The net impact of the MSIV, PCIV and ESF leak rate changes was an increase in the control room dose from 4.16 to 4.17 Rem TEDE, a decrease in the Exclusion Area Boundary (EAB) dose from 3.10 to 1.43 Rem TEDE, and a decrease in the Low Population Zone (LPZ) dose from 0.696 to 0.548 Rem TEDE.

For comparison, Table 2 summarizes the differences between the EXELON Peach Bottom AST MSIV leakage model (as implemented in LOCA Analysis PM-1 077, Revision 1) and the HCGS MS!IV leakage model (as implemented in H-1-ZZ-MDC-1880, Revision 3).

22 of 31

Attachment 3 LR-N10-0163 Table 2 Comparison of Peach Bottom and Hope Creek MSIV'Leakage Aerosol Deposition Model

.Peach Bottom AST Analysis -"PM-1077,'Rev 1' Hope Creek AST Afialysis -'H-1-ZZ-MDC-1880, Rev 3

,MSIV Failed Line Intact Line MSIV Failed Line-- Intact Line Var.iable .RPVTfo Betveen . Beiween BetwNieen Between RPV To Between Between . Between Between Parameter Inboard. Inboard Outboard ,REN and Outboard Inboard Inboard Outboard RPV and Outboard MSIV and MSIV Outboard MSIV MSIV and MSIV and Outboard MSIV and Outboard and TSV MSIV and TSV Outboard TSV MSIV TSV MSIVs MSIVs Piping Ruptured -

Integrity ,

Assumed. .. RemainsIntact Remains Intact RutureditdRemains not credited Intact Assumed .

Aerosol Deposition Not Credited CreditedCredited Not Credited Piping - Not Volume Credited Credited . Credited Credited Dilution Drywell P/T Related.

MSIV Leak Not Applicable Credited N/A Credited Rate:

Reduction, Holdup Not Credited.

Time Not Credited Deposition.

Not Cred ited 40 Percentile* Not 50 30 50 30 D.itr VelDcity, .ot.Cr*edi ted Credited Percentile Percentile Percentile Percentile Elemental Not Ceie 5%

Iodine Not Credited

• Credited (50%) Credited Credited (50%)

Removal

• Peach Bottom LOCA analysis in PM- 1077, Rev 1 uses the 40 percentile aerosol deposition velocity in both MS1V failed and intact.lines. PM-1077, Rev 1, Appendix A documents the parametric study, which demonstrates that the results in the calculation using the 40 percentile aerosol deposition velocity is bounding for a lower deposition of the lighter aerosol particles in the piping downstream of the outboard MSiVs due to conservatism in the-calculation by neglrctin& tlýe aerosol and el&niefital iodine removal in the pipingsegment between the'inboardand outboardMSIVs inthe MSIV failed'line. "

Table 3 lists the input parameters associated with the AST methodology differences between the FPB. and HCGS'AST kLOCA lculations. In addition-to the Tables '2 and 3 AST' methodology differences, th-f 'm'6st sighifidaritmodeling differences between'the PB AST LOCA analysis and the HCGS ASTLOCA analysisare:

3) Containment Leakage,- The PB model has an initial higher containment leakage rate (O..7 -eight.%/day vs. 0.5 w eight %Iday)"

4) ESF Leakage - The PB model has a higher ESF leak rate (5 gpm vs. 2.85 gpm) 23 of 31 LR-NIO-0163 Table 3 Comparison of Design Input Related to Methodology Differences AST LOCA Calculations PM-1077, Revision 1 versus H-1-ZZ-AIDC-1880, Revision 3 Item Design Input H-1-ZZ-MDC-1880 PM-1077 No. Information Revision 3 Revision 1 1 Containment Elemental Iodine Standard Review Plan 6.5.2 Standard Review Plan 6.5.2 Removal by Wetted surface Area Model 2 Particulate (Aerosol) Powers' 10 percentile model Powers' 10 percentile model Deposition/Plateout Model 3 Total MSIV Leak Rate Through 250 scfh for <24 hrs @ 50.6 psig 360 scfh for <38 hrs @49.1 psig All Four Lines 125 scfh for > 24 hr @ 50.6 psig 180 scfi for > 38 hrs @ 49.1 psig 4 MSIV Leak Rate Through Line 150 scfh for <24 hrs @ 50.6 psig 205 scfh for < 38 hrs @ 49.1 psig With MSIV Failed 75 scfh for > 24 hrs @ 50.6 psig 102.5 scfi for >38 hrs @ 49.1 psig 5 MSIV Leak Rate Through First 100 scfi for <24 hrs @ 50.6 psig 155 scfh for < 38 hrs @ 49.1 psig Intact Line 50 scfh for >24 hrs @ 50.6 psig 77.5 scfh for > 38 hrs @ 49.1 psig 6 Maximum PCIV Closure 120.0 sec Instantaneously (Isolation) Time 7 Iodine Specific Activity 0.2 pCi/g DE 1-131 N/A 8 Noble Gas Specific Activity 100/tF aCi/g N/A 9 Maximum RCS Noble Gas Release Rates jaCi/sec KR-83M 3.40E+03 N/A KR-85M 6.1OE+03 KR-85 2.OOE+01 KR-87 2.OOE+04 KR-88 2.OOE+04 XE-131M 1.50E+01 XE-133M 2.90E+02 XE-133 8.20E+03 XE-135M 2.60E+04 XE-135 2.20E+04 XE- 138 8.90E+04 10 Maximum RCS Iodine Activity pCi/g 1-131 1.30E-02 N/A 1-132 1.20E-01 1-133 8.90E-02 1-134 2.40E-0 I 1-135 1.30E-01 24 of 31 LR-N1O-0163 -

Issue 2: For volumes after the outboard MSIV, what is the justification for using a 30th percentile deposition velocity?

Subsequent to HCGS receiving its AST license amendment, the industry and NRC gained experience with, and an understanding'of;l:aerosol deposition in the main steam lines following a LOCA. The NRC informed some AST: license amendment applicants: of a concern related to the modeling of light aeroso61 p"rticles which expenrienc' lessergi'ravitaftinal *deý0'6sitidh in the seismically supported lines beyond the outboard MSlVs. This concern:was addressed in many successful AST license amendments for the EXELON fleet. PSEG NUCLEAR made the prudent decision to address the. cohcern in H-.1-ZZ-MDC-1880, Revision 3, by updating the aerosol deposition model with respect to the latest regulatory developments. This resulted in some loss of dose margin.

The NRC staff concluded inAEB 98-03, page 11, that:

"Given the conservatism associated with-using a well-mixed model for the entire length of pipe and .a number of additidonal conse"rvatisms inherent in the piping deposition analysis, the use of a 10h Opercentilesettling velocity with a well-mixed model is not appropriate. Additional conservatisms include additional deposition by,thermophoresis, diffusiophoresis, and flow irregularities; ,additional deposition as a result of hy groscopicity; and a possible plugging of the leaking MSIV by'aerosols. Given the conservatism of the well-mixed assumption, we believe it,is acceptable to use median values (as compared to more conservative va~lues) for.deposition."'

Therefore, a 5 0 th percentile aerosol- sttlin'g velocity is used in main-steam piping upstream of the outboard MSIV, where the majorIity of heavier aerosol particle are expected"to be deposited.

The remaining lighter aerosol particles experience lesser gravitational deposition'in the piping beyond the outboard MSIV. This mechanisrsn is modeled using the 3 0 th percentile aberosol settling, which is a median value betweenh.the 1 0 th and 5 0 th percentile settling velocities. The use of a lower 3 0 th percentile settling velocity,.reduces the removal of the remaining lighter aerosols, and is conservative. The use of a lower settling velocity further increases the resulting doses.

The comparisons provided in Table 4 demonstrate that the aerosol deposition model used in H-1-ZZ-MDC-1 880, Revision 3 for the MSlIV leakage paths conservatively complies with the AEB 98-03 guidance.

25 of 31

Attachment 3 LR-NIO-0163 Table 4 Hope Creek MSIV Leakage Aerosol Deposition Model - Compliance With AEB 98-03 Methodology AEB 98-03 Hope Creek AST Analysis - H-1-ZZ-MDC-1880, Revision 3 MSIV Failed Line Intact Line MSIV Failed Line Intact Line Variable RPV To Between RPV To Between RPV To Between Between RPV To Between Between Parameter Inboard Inboard Inboard Inboard Inboard Inboard Outboard Inboard Inboard Outboard MSIV and MSIV and MSIV and MSIV and MSIV and MSIV and Outboard Outboard Outboard TSV Outboard TSV MSIVs MSIVs MSIVs MSIVs Ruptured Ruptured Piping - not Remains Remains Remains - not Remains Remains Remains Remains Remains Integrity credited Intact Intact Intact credited in Intact Intact Intact Intact Intact Assumed in analysis analysis Aerosol Not Not Credited Credited Credited Credited Credited Credited Credited Credited Credited Credited Deposition Piping Not Not Volume Diuin Not Credited Credited Credited Credited Not Credited Credited Credited Credited Credited Credited Dilution Holdup Not Not Not Not Not Not Not Not Not Not Time Credited Credited Credited Credited Credited Credited Credited Credited Credited Credited Deposition Used 40 and 50 Percentile velocity for Not 50 30 30 Distribution sensitivity study Credited Percentile Percentile 50 Percentile Percentile Elemental Not Credited Credited Credited Not Credited Credited Credited Credited Credited Iodine Remova Credited (50%) (50%) (50%) Credited (50%) (50%) (50%) (50%)

Removal (50%)

Perry (AEB 98-03 pilot plant) does not have the seismically supported main steam line beyond the outboard MSIV; therefore, unlike the Hope Creek Plant, the main steam line between the outboard MSIV and Turbine Stop Valve (TSV) is not modeled.

The 10CFR50.59 evaluation that was done supporting Revision 3 of Calculation H-1 -ZZ-MDC-1880 is provided as Attachment 6 to this submittal. While the 50.59 process/evaluation format is not designed to document the detail provided in the above discussion on the Revision 3 methodology changes, the evaluation does sufficiently describe and evaluate the Revision 3 methodology changes, and appropriately concludes that the calculation revision does not result in a departure from a method of evaluation that would require prior NRC approval.

In conclusion, the parameters, assumptions and methodologies used in the current licensing basis analysis are consistent with plant specific design inputs, NRC guidance, and industry applications and prior NRC approvals.

26 of 31 LR-N10-0163 NRC RAI#20 Section 4.3.2 of NEDC-33529P (Reference 2) states that "[t]he HCGS licensing basis MSLB analyzed in Section 15.6.4 of the HCG.S UFSAR [Steam System Piping Break Outside Containment]assumes no fuel damage occurs as a result of the eveht." Although the analysis assumes that no fuel rods are damaged, there is no explicit statement inr NEFDC-33529P regarding-the isotope rods. Confirm that no damage to the isotope rod occurs because of the event.

R E.SPOQN.SE TO RAI1#20 No daamage to cobalt isotope rods occurs due to a MSLB event at HCGS, and no cobalt is released fromrthe cobalt isotope rods. Cobalt isotope rods are significantly less likely to fail than fuel rods during operation, transients and design basis accidents not involving fuel melt (see discussion in RAI#21 response). Any event where no fuel damage is assured can safely use the assumption that-no isotope rod damage occurs.

NRC RAI#211 During circulation, the reactorcoolant acquires radioactivematerials due'to release of fission prb6dhcts from fuel leaks into the coolant anda:"tiva'tion'6fcorrosion prdduct.s in the ,eactor coolant. These radioactive-materials in the cbolant can plate out in the reactorc oolant system (RCS), and, at times, an -accumulation will break away to .spike. the, normal level of radioactivity'.

The release of coolant:duringa DBA could sehdxradioactivematerials into the environment. 'A limiting condition of operation (LCO) on the maximum allowable level of radioactivityin the i&eator coolant is #established, consistent w .ith 10 dFR 50.36(c)(2)'ii), Criterion 2, to einsure, in the-eveht of a release.of any radioactive material to the,environment during a DBA, radiation dosdý'are,maihnhined within the limits of4 0 CFR 100,;."Reactor-Site Criteria"and/or,1'0 CFR 50.6.7,.. "Accident Source. Term. '" The.limits.on.RCS specific activity-are also used for estbblishing 'standardizationiin radiatiof'shieldihg and plant personnel radiationprotection practices.

HCGS-,TS LCO 3.4.5 "ReactorCoolant;SystemrSp~ecific.,Activity, states,-that the primary coolant DQSE.;EQUlVALENTJ-131 ispecificactivity of the reactorcoolant shall,be less than or equal to 0.2 microcuriest.pergram,,(pgýi/gm) andJIess than,,orequaltol O0/tOIE pgCi/gmn Per the TS Definition 1.11, DOSEýEQUIVALENT -:13 1 is based,upon l-131., /- 1-32, j- 133, /-134, and /-135.

The.NRC-staff is concerned about whether the LCO adequately.,addressesa release of Co-60 into the RCS.

While no "fuel damage" is assumed for some DBA events, the burrent design-'ba-,sssfety analysts conservatively/assumes the fuel pins leak. Clarify whetlher the opera'tonhaldesignhlimrit for the isotope rods is no leakage. Since the TSs are derived frbm the safety analysis, describe how the TSs will ensure that the assumption of no Co-60 leakage from the Co-60 ITA's remains valid. Justify how LCO 3.4.5 remains able to ensure that 10 CFR 50.67 and 10 CFR 100 limits (as applicable), and radiationshielding and plant personnel radiationprotection design limits are

,met, or modify LCO 3.4.5 so that and these limits continue to be met after the proposedchange.

27 of 31 LR-N10-0163 RESPONSE TO RAI#21 The operational design limit for the isotope rods is no leakage. Furthermore, leakage of cobalt (including entire cobalt targets and/or cobalt particulate) from an isotope rod in an ITA is not a credible event during normal operations, transients or design basis accidents not involving fuel melt accidents (i.e., Loss of Coolant Accident and Control Rod Drop Accident). Based on regulatory guidance provided for fuel melt design basis accidents, it is conservatively assumed that cobalt (Co) isotope rods melt along with the fuel rods during a fuel melt design basis accident. The negligible impact of ITAs on CRDA and LOCA radiological consequences is addressed in the revised NEDC-33529P Sections 4.3.1 and 4.3.4, (See response to RAI#17)..

The isotope rod design, discussed in Section 2.1 of NEDC-33529P provides multiple features to prevent cobalt isotope rod failures. The main features that provide multiple levels of safety for the cobalt isotope rods are:

• The nickel-plated cobalt targets are encapsulated with two layers of Zircaloy-2 cladding

• The solid Zircaloy-2 connections between cobalt rod segments are located at each spacer location (debris fretting failures normally occur at spacer locations)

" The heat generation rate of a cobalt isotope rod is significantly less than a typical fuel rod GNF has experience with segmented rods in previous Lead Test Assembly programs.

Introduction of a small number of isotope rods into non-limiting locations in the core add to the argument that leakage of cobalt is not a credible event during normal or transient events.

The GE14i materials and bundle configuration were purposely selected to be the same as GE14; the design that.GNF has now deployed in approximately 26,000 bundles with over 10 years of successful operating experience. Of the over 70,000 rods in the HCGS core, only a small quantity will be cobalt bearing rods. The selection of the well-established bundle design for HCGS further reduces risk and performance uncertainty.

An explanation of isotope rod failures is provided Section 2.2 of NEDC-33529P. The failure mechanisms addressed include:

• Fuel handling accidents

• Manufacturing defects and assembly error

" Pellet cladding interaction

" Corrosion

" Primary hydriding 0 Cladding creep collapse

" Rod bow

" Unthreading of segments

• Stress

• Seismic and flow induced vibration

" Internal fret from inner capsule

" Spacer location fretting

• Mid-span fretting

• Failures during disassembly 28 of 31 LR-N10-0163 To further expand upon the failure modes discussed in NEDC-33529P, additional multiple levels of failure considerations are discussed below:

Targets being mechanically pulverized;iworn-out by fluid, flow, corroded or otherwise damaged while still inside the inner tube-or.dapsule6to compromise nickel coating and release -cobalt...

In addition to thefailure modes-required to compromise.theinner and outer cladding;'not being credible,-this failure scenario itself is not: credible for multiple reasons. Thenickel plating~of the targets is harder than all the Zircaloy-2 components that suriroundý it.:The nickel-would therefore not be the material to grind, or wear. Itiis more likely that the Zircaloy-2 tubing or canister grind or wear than the nickel. The coolant flow into an opening in the outer cladding and into an opening in the. inner cladding wduld not'have the necessary flow rate :to~cause anysignificant.

wearýof any internal isotope rod-components.

Additionally, there are no forces to excite the targets and sustain vibration or wear. Even considering the unlikely case that targets were to become excited, the-magnitude of the displacement of the isotope rod and, in turn,.(( )).

would be so small that damage to inner tubing is highly implausible.

Finally, nickel is chosen as plating or alloying material in many applications, including BWR alloys partly, because of its ability to withstand severe operating conditions involving corrosive environments. . - ,

Targets escaping segment assembly, througqha cladding .hole and being. mechanically pulverized to release cobalt.

L,4

• 4"

• " . ,' ;,4 , ,... , " ," . '

In addition .tojthe jailure modes reqqired to compromise-the inner-and.,outer: cladding npt being credible, this-fa"ilurescenario itself is also not-credible for multiple reasons. If.an .inner,tube were.to be compromise*.the (( - .....

)). Two layers of cladding would have to be breached at the exact-same axial and radial position and the breach would have to be greater than the size of a target for any targets to,escape.- After escape, the target would have to find a,mechanical:pulverizing mechanism against a material harder than nickel. This scenario is considered-highly,.

implausible.

Targets escaping segment assembly resulting from canister [f I.

and release of cobalt.

In addition to the failure modes required to compromise the inner and outer cladding not being credible, this failure scenario itself is also not credible for multiple reasons. In-the remote chance that full circumferential failure of the inner and outer cladding, occurred at the same location, on the same end of the same segment, the rod-to-rod and rod-to-channel spacing of the surrounding rods and/or fuel channel is too small to allow a ((

)) and release targets. ((

))

Regarding coolant flow into the opening after two full circumferential failures, as described above, the nickel plating of the targets is harder than all the Zircaloy-2 components that surround it. The nickel would therefore not be 'the material to experiencesignificant flow induced wear. Additionally, the coolant flow ((

)) would not have the necessary flow rate to cause any significant wear of any internal 29 Of 31 LR-NIO-0163 isotope rod components. The nickel plating of the targets would remain intact to prevent cobalt release into the coolant.

Even assuming (( )) and targets escaping from the segment, the targets would have to find a mechanical pulverizing mechanism against a material harder than nickel. These scenarios are considered highly implausible.

In summary, there are no plausible mechanisms for both the outer and inner cladding of an isotope rod to be compromised such that cobalt targets come in contact with the reactor coolant. If it is assumed that some unknown event were to occur such that the outer and inner cladding of the same rod segment (there are 9 independent rod segments in each cobalt isotope rod) were compromised, there is no plausible mechanism for cobalt targets to lose their nickel coating and release cobalt. The nickel-plating on the targets is harder than the Zircaloy-2 cladding materials surrounding them, so any wear associated with component interaction would be to the softer Zircaloy parts.

Additionally, combining any of these non-credible events such that the outer and inner cladding of the same segment were compromised there is no plausible mechanism to align the breach points to allow a cobalt target to escape or allow ((

)) to release targets to the coolant. Coolant flow is also not sufficient to negatively affect the plating on the targets.

Even adding these multiple levels of non-credible events, the segmented rod structure, with 9 individual double encapsulated containers, also ensures that the number of cobalt targets that can escape is limited to a small volume fraction of the targets in a single rod. This additional characteristic ensures that, in the event of multiple levels of failure that result in a single isotope rod segment failure, cobalt activity release is limited.

Traditional design basis analysis assumes some leakage of fuel rods, which is incorporated into technical specifications (TS) and is consistent with the design basis analyses. As described above, isotope rods have multiple layers of cladding and design features beyond a fuel rod's single layer of cladding and the isotope rods essentially act as a passive component in the operation of the bundle. Leakage of cobalt from an isotope rod is not a credible event during normal operations, transients or design basis accidents not involving fuel melt.

Fuel leakage is characterized by release of highly volatile gaseous fission products after failure of a single layer of cladding. Isotope rod leakage is characterized by the release of a low volatility metal (i.e., cobalt in target and/or particulate form) after the failure of an outer layer of cladding, an inner layer of cladding and compromising nickel plating. ((

1]

In summary, by design and definition, isotope rod failure is not credible and isotope rod leakage does not need to be incorporated into TS to remain consistent with traditional design basis analyses. If there is no fuel melt due to an accident, the source term available for release to the environment is based upon the activity in the RCS during normal operations. With Co-60 leakage not credible during design basis accidents not involving fuel melt, the proposed change will have no impact on the source term.

Although isotope' rod leakage is not a credible event during normal operations, transient and design basis accidents not involving fuel melt, it should be further noted that the existing TS surveillance requirements and periodic reactor coolant sampling detect Co-60 activity. HCGS 30 of 31 LR-NIO-0163 TS 3.4:5, "RCS Specific Activity," has, in addition to an equivalent 1-131 specific activity limit, a limit for RCS gross specific activity (E-bar). E-bar is defined in the TS as:

E-AVERAGE DISINTEGRATION ENERGY E shall be the average, weighted in proportion to the concentration of each radionuclide in the reactor~coolant at the. time, of sampling, of the sum of the average, beta and gamma energies per disintegration, in MeV,.for isotopes, with half lives:greater than 15-minutes, making.up at least 95% of the total non-iodine activity, in the coolant.

HC-GS-Surveillalne"Requirement (SR)4.4.5 requires'that, .in Mode 1ýi; 'radiochemical analysis for E-ba*rdetermn~ationshall: be performed at least, once per 6'months' (there is, also a -further requirelme'net:forE-ba-r that a sample tO-be *taken.afteiraaminimum of,2;EFPD and 20 days of.."

power operation have elapsed since reactor was last subcritical for 48-hours or longer).):- SR 4.4.5 also requires, in Modes 1, 2 and 3, a Gross Beta and Gamma Activity Determination every 72hours. Consequently,; the existing. LCO 3.4.5, remains adequate for ensuring dose, lirmits and radiation shielding and plant personnel radiation protection design, limits are met with'GE14i' .

ITAs installed.

References for Attachment 1 (unless uniquely identified in individual responses)

1. PSEG letter LR-N09-0290 to NRCQ "License Amendment Request Supporting.the Use of Co060 IsotopeTest-Assembliesý (Isotope Generation Pilot Project)," datedýDecemb-er 21, 2009 (ADAMS Package A&ccession No. ML093640193).

2 GE-Hitachi proprietary report NEDC-33529P, "Safety Analysis Report to Support Introduction of GE14i'lsotope&Test Assemblies-(ITAs) in Hope Creek Generating Station," Revision 0; dated December 2009 (Attachment 3 to Reference 1). A non-proprietary version of report (Attachment 4 to Reference1) is-included as part of ADAMS Accession No.ML093640199"

3. Exelon Nuclear letter RS-09-150 to NRC, "Additi6nal Information Supporting the Request for a License Amendment to Modify Clinton Power Station Facility Operating License in Support of the Use-of IsbtOpe Test Assemblies," dated November-4, 2009 (ADAMS-Accession No. ML093100313).

4. Global Nuclear Fuel proprietary report' NEDC-32868P, "GE14 Compliance with Amendment 22 of NEDE-2401 1-P-A (GESTAR II)," Revision 3, dated April 2009.

5. J. Schaperow et al., "Assessment of Radiological Consequences for the Perry Pilot Plant Application using the Revised (NUREG-1465) Source Term," U.S. Nuclear Regulatory Commission,.AEB 98-03, December 9, 1998 31 of 31