HNP-11-090, Response to Request for Additional Information Regarding TAC No. ME5409 (Harris-M5TM Cladding)

Response to Request for Additional Information Regarding TAC No. ME5409 (Harris-M5TM Cladding)

ML11286A254
Person / Time
Site:	Harris
Issue date:	10/06/2011
From:	Holbrook K Progress Energy Carolinas
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
HNP-11-090, TAC ME5409
Download: ML11286A254 (16)

Text

Progress Energy 10 CFR 50.90 October 6, 2011 Serial: HNP-11-090 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 Shearon Harris Nuclear Power Plant Unit No. 1 Docket No. 50-400 / Renewed License No. NPF-63

Subject:

Request for Additional Information Regarding TAC No. ME5409 (Harris-M5TM cladding)

Reference:

1. Letter from Chris Burton, Progress Energy Carolinas, to USNRC, "Application for Revision to Technical Specification 5.3.1 and Core Operating Limits Report References for M5 Cladding," HNP-10-124, dated January 13, 2011 (Adams M

T Ascension No. ML110250265)

2. Email from B. L. Mozafari, USNRC, to J. R. Caves, Progress Energy Carolinas, "Request for Additional Information Regarding TAC No. ME5409 (Harris-M5 cladding)," dated September 2, 2011.

Ladies and Gentlemen:

In Reference 1, Carolina Power & Light Company (CP&L), doing business as Progress Energy Carolinas, Inc., requested changes to the Technical Specifications (TS), Appendix A of Renewed Operating License No. NPF-63, for the Shearon Harris Nuclear Power Plant Unit No. 1 (HNP). The proposed changes would modify the HNP TS to permit the use of the AREVA fuel cladding alloy designated as M5 TM .

In Reference 2, the USNRC issued a request for additional information (RAI). The enclosure to this letter contains HNP's response to that RAI.

CP&L has concluded that the information provided in this response meets the intent of the original submittal (Reference 1) and does not impact the conclusions of the: 1) Technical Analysis, 2) No Significant Hazards Consideration under the standards set forth in 10 CFR 50.92(c), or 3) Environmental Consideration as provided in the original submittal.

This letter contains no new regulatory commitments.

Progress Energy Carolinas, Inc.

Harris Nuclear Plant P.O. Box 165 New Hill, NC 27562 FoA-0

U.S. Nuclear Regulatory Commission Page 2 HNP-11-090 If there are any questions or if additional information is needed, please contact Dave Corlett at 919-362-3137.

I declare under penalty of perjury that the foregoing is true and correct. Executed on / -- /1 Sincerely, Keith Holbrook Manager, Support Services Harris Nuclear Plant

Enclosure:

Response to Request for Additional Information Regarding TAC No. ME5409 (Harris-M5 M T

cladding) cc: Regional Administrator, USNRC/Region II Project Manager, Harris Nuclear Plant, USNRC/NRR Resident Inspector, Harris Nuclear Plant, USNRC Section Chief, NC Division of Environmental Health

HNP-11-090 Enclosure Shearon Harris Nuclear Power Plant / Unit No. 1 (HNP)

Docket No. 50-400 / Renewed License No. NPF-63 Application for Revision to Technical Specification 5.3.1 and Core Operating Limits Report References for M5 M T

Cladding Response to Request for Additional Information Regarding TAC No. ME5409 (Harris-M5 TM cladding)

HNP-11-090 Page 2 of 14 Enclosure Question 1 The Safety evaluation for Topical Report BAW-10240(P)(A) lists the following four conditions which Framatome (FANP) has accepted:

a) The corrosion limit, as predicted by the best-estimate model will remain below 100 microns for all locations of the fuel.

b) All of the conditions listed in the SEs for all FANP methodologies usedfor MSTM fuel analysis will continue to be met, except that the use of M5TM cladding in addition to Zircaloy-4 cladding is now approved.

c) All FANP methodologies will be used only within the rangefor which M51 data was acceptable and for which the verifications discussed in BA W-10240(P) or Reference 2 was performed.

d) The burnup limit for this approval is 62 GWd/MTU Explain in detail, how each of the above conditions has been implemented at HNP Unit 1.

Response 1:

AREVA is responsible for completing safety analysis, neutronic, thermal-hydraulic and fuel mechanical analysis per the NRC-approved methods listed in the HNP Core Operating Limits Report (COLR). BAW-10240 incorporates M5 TM material properties into a set of NRC-approved mechanical analysis, small break loss-of-coolant accident (SBLOCA), and non-loss-of-coolant accident methodologies.

a) The restriction that corrosion limit, as predicted by the best-estimate model, will remain below 100 microns for all locations of the fuel is implemented in AREVA design processes. A scoping study of M5TM implementation was performed and confirmed that the fuel mechanical limits can be satisfied for the HNP 17 x 17 HTP fuel design. This is implemented by AREVA, is subject to CP&L owner review and audit, and is subject to NRC audit.

b) AREVA implements the reload specific design. Conditions from approved Safety Evaluations are incorporated as restrictions in AREVA design procedures and guidelines that control the core reload designs provided to Harris Nuclear Plant. This is implemented by AREVA, is subject to CP&L owner review and audit, and is subject to NRC audit.

c) AREVA implements the reload specific design. Limitations to ensure FANP methodologies will be used only within the range for which M5TM data was acceptable and for which the verifications discussed in BAW-10240(P) or Reference 2 was performed are incorporated as restrictions in AREVA design procedures and guidelines that control the core reload designs provided to Harris Nuclear Plant. This is implemented by AREVA, is subject to CP&L owner review and audit, and is subject to NRC audit.

d) The burnup limitation is not a change for the 17 x 17 HTP fuel assembly. Burnup limits identified in approved methodologies are contained in HNP core functional requirements and AREVA design processes, which are currently limited to 62 GWd/MTU. This is implemented by AREVA, is subject to CP&L owner review and audit, and is subject to NRC audit.

HNP-11-090 Page 3 of 14 Enclosure Question 2 The safety evaluation for EMF-2310, Revision 0 topical report (Section 5.0), has restated one of the restrictions (Number 1) from the safety evaluation for ANF-89-151(P)(A) (Section 2.2, TER Conclusions) as stated below:

The stated applicationof the S-RELAP5 code is for the events listed above in Table 1. There are other computer codes and methodologies employed for evaluation of the events not listed in the table. For each licensing basis event analyzed, the applicantmust, as always, justify the methodology used whether by reference to S-RELAP5 or whatever methodology has been used.

Explain how this restriction is implemented for the upcoming cycle reload analyses when the transition to M5 M T

cladding occurs.

Response 2 Table 1 lists the events analyzed in the HNP Final Safety Analysis Report (FSAR). Table 1 is exclusive of radiological analyses that do not interface with transient analyses (e.g. Fuel Handling Accident, Waste Gas Decay Tank Rupture, etc.). Radiological analysis is performed in accordance with Alternate Source Term (AST) methodology of Regulatory Guide (RG) 1.183 as described in CP&L's submittal on AST. The NRC acceptance of CP&L's implementation of AST is contained in License Amendment 107. Radiological analyses are not impacted by the cladding material.

The Non-LOCA events that rely on methodologies outside of ANP-89-151 and EMF-2310 are Main Steam Line Break (MSLB), Ejection of a Full Length RCCA and Steam Generator Tube Rupture (SGTR). MSLB is analyzed using a separate AREVA methodology, but in the future could be analyzed using EMF-2310.

Rod Ejection is analyzed with an event specific methodology (XN-NF-78-44). SGTR is analyzed using Westinghouse methodology (WCAP-10698),

SGTR is analyzed for margin to overfill and offsite dose consequences. From the perspective of SAFDLs the SGTR is bounded by other more severe depressurization events. The change in the cladding material has negligible impact on the analysis of SGTR margin to overfill or offsite dose consequences.

In conclusion, NRC approved methodologies are used for accident analyses and those methods that do not directly interface with BAW-10240 are not impacted by the change in cladding material.

HNP-11-090 Page 4 of 14 Enclosure Question 3 RODEX2 fuel deformation and conductivity models were incorporated in S-RELAP5 for transient and accident analyses. Provide details of the methodology in the fuel model to evaluate fuel thermal conductivity as a function of burnup and temperature, considering all of the effects that take in the fuel during the irradiation in the reactor core.

Response 3 M

T The addition of BAW-10240 (P)(A) as a COLR methodology is specific to the incorporation of M5 material properties and correlations into other approved AREVA methodologies. BAW-10240(P)(A) does not deal with pellet (U0 2) material.

The RAI applies to burnup dependency on the pellet material. This fuel property is the subject of separate NRC approved methodologies, which are not related to the subject license amendment request. The NRC approved methodologies for RODEX2 is collectively listed in HNP Technical Specification Section 6.9.1.6.2.o, Core Operating Limits Report, Mechanical Design Methodologies. No additions or changes to the TS 6.9.1.6.2.o are being made as part of the subject license amendment request. A conservative penalty has been assessed for pellet material thermal conductivity degradation.

HNP-11-090 Page 5 of 14 Enclosure Question 4 Provide a list of Shearon Harris Unit 1 FSAR Chapter 15 Non-LOCA events that will be either analyzed or dispositioned for the upcoming fuel cycle when the licensee is planning to use M5 TM cladding. Also summarize the methodologies and codes used for the analyses.

Response 4 Table 1 provides the requested information for the upcoming fuel cycle 18. Table 1 does not include radiological events that are listed in HNP FSAR Chapter 15.0; those analyses do not interface with BAW-10240. Table 1 is a projection of the expected safety analysis activities that are currently in progress, but are not yet complete.

As described in ANP-89-151 and EMF-2310, non-LOCA events generally consist of two parts. The first part is a system transient. The second part is an assessment of the impact of cycle specific peaking factors on Minimum Departure from Nucleate Boiling Ratio (MDNBR) and Fuel Centerline Melt (FCM)

Specified Acceptable Fuel Design Limits (SAFDLs). Main Steamline Break is a special case in that an additional check is needed to confirm the reactivity conditions at limiting state points.

A number of events are dispositioned based on the characteristics of individual events versus other bounding events. The basis for these dispositions is contained in ANP-89-151 and EMF-2310. The dispositioned events have no associated system transient and do not have SAFDL evaluation on a cycle-by-cycle basis.

Within this framework, only one Non-LOCA system transient is being performed to support Cycle 18.

The affected system transient is being updated to recover MDNBR margin and the re-analysis is not related to the change in the cladding material. The impacted FSAR events are Uncontrolled Bank Withdrawal at Power (FSAR 15.4.2) and Withdrawal of a Single Full Length RCCA (FSAR 15.4.3.2). These events share a common system transient; the events differ by the peaking factors applied for the respective RCCA configurations.

The Inadvertent Boron Dilution Event (FSAR 15.4.6) is performed for every cycle. For Cycle 18 the methodology of EMF-2310 will be used. The analysis is performed for almost every reload campaign.

The changes in the core reactivity dominate the result and the change in the cladding material has negligible effects compared to differences in cycle design.

SBLOCA and LBLOCA are both being revised for Cycle 18. The applicable SBLOCA methodology (EMF-2328) was previously approved for the HNP docket. The LBLOCA is being re-analyzed using AREVA's EMF-2103 methodology; an application to add this COLR methodology is currently pending NRC review and approval.

The neutronics methods used to assess the power distribution and reactivity inputs for the MDNBR and FCM analysis and the thermal hydraulic methods used to assess compliance with SAFDLs are listed in Table 2.

HNP-11-090 Page 6 of 14 Enclosure Table 1 Listed in EMF-FSAR Event Name (ANS 2310 TER Cycle 18 activity Cycle 18 Comment Section category) Table 1? Methodology The event is included in the TER table for ANP-89-151 (P)(A). By the FW System nature of the event it is considered S

lfW i tem Nin scope for EMF-2310, even though 15.1.1 Malfunction Results that in a Decrease No (See Comment) Event 15.1.3 FSAR Bounded by ANP-89-151 not listed tbe u inoteapiaiiyo the EMF-2310 TER h in FW temperature (11) table, due to the applicability of the EMF-2310 method to "Increase in Heat Removal by the Secondary System" event grouping.

FW System System transient to be dispositioned Malfunction that Evaluate MDNBR for MS.

Results in an Increase and FCM SAFDL_

in FW flow (11)

Excessive Increase in Evaluate MDNBR System transient to be dispositioned 15.1.3 Secondary Steam Flow Yes ala MDNBR ANP-89-151 for M5.

(11) and FCM SAFDL Inadvertent Opening Event Bounded by Event bounded by FSAR 15.1.5 with 15.1.4 of SG relief or Safety Yes FSAR 15.1.3 (at ANP-89-151 no fuel failure after reactor trip Valve (11) power)

Evaluate MDNBR System transient not currently 15.1.5 Steam System Piping Yes and FCM SAFDL EMF-84-093 performed using EMF-2310. System Failures (IV) and reactivity transient to be dispositioned for MS.

interface

HNP-11-090 Page 7 of 14 Enclosure Table 1 (cont.)

FSAR Event Name (ANS Listed in EMF- Cycle 18 category) 2310 TER Cycle 18 activity Methodology Comment Section Table 1?

15.2.1 BWR event NA to N/A HNP 15.2.2 Loss of External Yes Event Bounded by ANP-89-151 Electrical Load (11) FSAR 15.2.3 15.2.3 Turbine Trip (11) Yes Evaluate MDNBR SFLANP-89-151 System transient to be dispositioned foM SAFDL for M5.

Inadvertent closure 15.2.4 of Main Steam Yes Event Bounded by ANP-89-151 Isolation Valves (11) FSAR 15.2.3 15.2.5 Loss of Condenser Yes Event Bounded by ANP-89-151 Vacuum (11) FSAR 15.2.3 Loss of Non-15.2.6 Emergency AC Power Yes Event Bounded by ANP-89-151 to the Station FSAR 15.3.2 Auxiliaries (11) 15.2.7 Loss of Normal Yes No SAFDL analysis ANP-89-151 System transient to be dispositioned Feedwater Flow (11) required. for M5.

15.2.8 Feedwater System Yes No SAFDL analysis ANP-89-151 System transient to be dispositioned Pipe Break (IV) required. for M5.

HNP-11-090 Page 8 of 14 Enclosure Table I (cont.)

FSAR SAi I

Event Eeteg Name (ANS (ANS TListed Litdiin EMF-2310 TERM-Cycle Cycle 18 activity Cycle 18 Comment Section category) Table 1? Methodology Event Bounded by Partial Loss of Forced FSAR 15.3.2 with 15.3.1 Reactor Coolant Flow Yes more restrictive ANP-89-151 (11) ANS II acceptance criteria Complete Loss of Evaluate to ANS II System transient to be dispositioned 15.3.2 Forced Reactor Yes MDNBR SAFDL ANP-89-151 for M5.

Coolant Flow (111)

Reactor Coolant Evaluate number of 15.3.3 Pump Shaft Seizure Yes fuel assemblies taexedDBANP-89-151 System transient to be dispositioned foM.

that exceed DNB for M5.

(Locked Rotor) (IV) criteria Reactor Coolant Event Bounded by 15.3.4 Pump Shaft Break Yes FSAR 15.3.3 ANP-89-151 (IV)

Uncontrolled Rod Cluster control Assembly Bank Evaluate MDNBR System transient to be dispositioned 15.4.1 Withdrawal from a Yes and FCM SAFDL ANP-89-151 for M5.

Subcritical or Low Power Startup Condition (II)

HNP-11-090 Page 9 of 14 Enclosure Table 1 (cont.)

FSAR Event Name (ANS Listed in EMF-2310 TER Cycle 18 activity Cycle 18 T Comment Section category) Table 1? Methodology Uncontrolled Rod Cluster Control Evaluate MDNBR New system transient to be 15.4.2 Assembly Bank WithdrAwal at Poe Yes EvlaeMNREMF-2310 Nwsse rnin ob and FCM SAFDL performed for Cycle 18.

Withdrawal at Power (11)

Dropped Full Length Evaluate MDNBR System transient to be dispositioned 15.4.3.1 RCCA or RCCA Bank Yes and FCM SAFDL ANP-89-151 for M5.

Withdrawal of a Evaluate number of 15.4.3.2 Single Full Length Yes fuel assemblies taexedDBEMF-2310 New system transient is performed foCyl18 RCCA (111) that exceed DNB for Cycle 18.

criteria and FCM No transient analysis performed.

15.4.3.3 Statically Misaligned Yes Evaluate MDNBR ANP-89-151 M5 properties included in M

T RCCA or Bank (11) and FCM SAFDL neutronic input.

Event bounded Startup of Inactive by FSAR 15.4.1 Reactor Coolant in Modes 3-5.

15.4.4 Pump at Incorrect Yes Event precluded ANP-89-151 Temperature by administrative control in Modes 1 and 2.

15.4.5 BWR event NA to N/A HNP

HNP-11-090 Page 10 of 14 Enclosure Table 1 (cont.)

ESAR Event Name (ANSFSAR Listed Evet310inNae ER (NS ycle28 ctiity EMF- ycl 28Comment Cycle 18 Cycle 18 activity Methodology C category) 2310 TER Section Table 1? 1 Chemical and Volume Control Mode 1 bounded System Malfunction by FSAR 15.4.2. No transient analysis performed.

1546 that Results in a Other modes M

T 15.4.6 Decrease ininthe t Yes Overed covered by cyclee EMF-2310 M5 properties reactivity included in calculations Boron Concentration specific reactivity in the Reactor calculations Coolant (11)

Inadvertent Loading Evaluate number of and Operation of a feasmbisNo transient analysis performed.

15.4.7 Fuel Assembly in an fuel assemblies ANP-89-151 M5 TM properties included in 154.oFel Assemlon that exceed DNB neutronic input.

Improper Position and FCM criteria Spectrum of Rod Evaluate number of Cluster Control fuel assemblies XN-NF-78-44 and System transient to be dispositioned 15.4.8 Assembly Ejection that exceed DNB ANP-89-151 for MS.

Accidents (IV) and FCM criteria 15.4.9 BWR event NA to N/A HNP

HNP-11-090 Page 11 of 14 Enclosure Table 1 (cont.)

FSAR Event Name (ANS Listed in EMF-2310 TER Cycle 18 activity I Cycle 18 Comment Section category) Table 1? Methodology Inadvertent Operation of the No impact on 15.5.1 Emergency Core Yes SAFDL as reactor Coolant System trip occurs at event During Power start.

Operation (11)

CVCS failures that Event bounded by 15.5.2 increase RCS Yes FSAR 15.5.1 and inventory 15.4.6 Inadvertent Opening 15.6.1 of Pressurizer Safety Yes Evaluate MDNBR ANP-89-151 System transient to be dispositioned or Power Operated SAFDL for M5.

Relief Valve (11)

Break or inOther Line Instrument Line Radiological analysis uses bounding eYes (see non-mechanistic release of RCS.

15.6.2 From RCPB that Ys(e comment) Analysis does not interface with penetrates BAW-10240 referenced methods.

Containment

HNP-11-090 Page 12 of 14 Enclosure Table 1 (cont.)

FSAR Section Event Name (ANS category) 2310 2TableTER 1? TCycle 8 activity Cycle 18 Methodology Comment Comment The methodology for SGTR overfill and SGTR offsite dose are not performed using AREVA 15.6.3 SGTR See comment Disposition WCAP-10698 methodology. The methodology employed was approved by NRC in the HNP License Amendment No.

107.

Steam Generator Tube Rupture (IV) See comment for Margin to Overfill line above.

(MTO) 15.6.4 BWR event NA to N/A HNP

HNP-11-090 Page 13 of 14 Enclosure Table 2 Technical Specification entry Report Title Application number The low power physics testing methodology for "rod swap" is contained in XN-75-27 Exxon Nuclear Neutronics this topical report. The 6.9.1.6.2.a Design Methods for Pressurized method is used to confirm (P)(A) Water Reactors RCCA reactivity measurements. This method is not used for core design.

ANF-RELAP Methodology for Non-LOCA methodology ANF Pressurized Water Reactors 151(P)(A) Analysis of Non-LOCA Chapter 15 Events This method provides mixed core penalties for MDNBR analyses. The switch to Application of Exxon Nuclear M5 is not considered a M

T 6.9.1.6.2.c XN-NF-82-21 Company PWR Thermal Margin "mixed core" for the Methodology to Mixed Core purposes of thermal Configurations hydraulic analyses since the flow channel geometry of the M5 TM and Zircaloy 4 assemblies is the same.

Method used to 6.9.1.6.2.d XN-75-32(P)(A) Computational Procedure of determination when Evaluation Fuel Rod Bowing additional MDNBR penalty applied.

6.9.1.6.2.e EMF-84-093 Steam Line Break Methodology MSLB methodology for PWR For cycle 18, EMF-2087 SEM/PWR-98: ECCS Evaluation methodology being replaced 6.9.1.6.2.f EMF-2087 Model for PWR LBLOCA with EMF-2103 based Applications methodology in separate license amendment.

HNP-11-090 Page 14 of 14 Enclosure Table 2 (cont.)

Technical Topical Report Specification entry Number Title Application number A Generic analysis of the Methodology employs ANF-6.9.1.6.2.g XN-NF-78-44 control Rod Ejection Transient RELAP for PWR PDC-3: Advanced Nuclear Fuels Provides neutronics inputs 6.9.1.6.2.h ANP-88-054 Corporation Power Distribution to safety analyses.

(P)(A) Control for Pressurized Water Reactors Method for statistical Statistical Setpoint /Transient application of DNB 6.9.1.6.2.i EMF9(A) 1 methodology for Westinghouse correlation and checks of Type Reactors OPAT and OTAT trip setpoints HTP: Departure from Nucleate DNB correlation for 6.9.1.6.2.j EMF 9) 1 Boiling Correlation for High evaluation of SAFDL j(P)(A) Thermal Performance Fuel Exxon Nuclear Company Not used with M5TM fuel.

6.9.1.6.2.k (NNF(A) 9 Evaluation Model EXEM PWR (P)(A) Small Break Model EMF Reactor Analysis System for Provides neutronic input to 029(P)(A) PWRs safety analyses.

SBLOCA SLC methodology ehdlg EMF-2328 PWR Small Break LOCA 6.9.1.6.2.m (M) 232 P Sall Brel compatible with M5 TM (P)(A) Evaluation Modelcldig cladding.

SRP Chapter 15 Non-LCOA Non-LOCA methodology.

6.9.1.6.2.n EMF-2310 Methodology for Pressurized Water Reactors