ML12076A141

Summary of Meeting Summary with Carolina Power & Light to Discuss Realistic Large Break Loss-of-Coolant

ML12076A141
Person / Time
Site:	Harris
Issue date:	04/05/2012
From:	Billoch-Colon A Plant Licensing Branch II
To:
Billoch-Colon, Araceli
Shared Package
ML12076A140	List: ML12076A141 ML12076A142 ML12076A165
References
TAC ME6999
Download: ML12076A141 (56)

Text

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"/+" ~o.e' April 5, 2012 LICENSEE: Carolina Power & Light FACILITIES: Shearon Harris Nuclear Plant, Unit 1

SUMMARY

OF JANUARY 11, 2012, MEETING WITH CAROLINA POWER &

LIGHT TO DISCUSS REALISTIC LARGE BREAK LOSS-OF-COOLANT ACCIDENT (TAC NO. ME6999)

On January 11,2012, the U.S. Nuclear Regulatory Commission (NRC) staff conducted a closed meeting with Carolina Power and Light (the licensee) at NRC Headquarters, 11555 Rockville Pike, One White Flint North, Rockville, Maryland. The purpose of the meeting was to continue the discussion on the NRC staff concerns related to the realistic large break loss-of-coolant accident (LOCA) methodology amendment request for Shearon Harris Nuclear Plant, Unit 1 (HNP). Proprietary information was discussed during the meeting so it was closed to the public.

To facilitate discussion, the NRC provided suggested discussion topics (Enclosure 2). The licensee presented non-proprietary version of a slide presentation (Enclosure 3).

DISCUSSION In a letter dated August 22,2011 (Agencywide Documents Access and Management System Accession No. ML11238A077), the licensee submitted an amendment request. The proposed request would revise the HNP technical specifications (TSs) to add a plant-specific methodology that implements AREVA's NRC-approved topical report, EMF-2103(P)(A), 'Realistic Large Break LOCA Methodology for Pressurized Water Reactors, Revision 0' and add EMF-2103(P)(A),

'Realistic Large Break LOCA Methodology for Pressurized Water Reactors;' Revision 2 or higher upon approval of the specific revision by the NRC.

The licensee described the basis and intent of the submittal. This amendment request supports the on~oing efforts by the licensee to allow the use of the AREVA fuel cladding alloy designated as M5 M. The licensee indicated that the approval of the use of the realistic LOCA methodology topical reports supports a migration away from the existing legacy methodologies to a more updated methodology. These activities are in preparation for the use of the new clad during the HNP Cycle 18 refueling outage, currently scheduled for early in the second quarter of 2012.

During the meeting, the licensee and the NRC staff continued the discussion of the December 11, 2011, closed meeting topics. These topics covered the droplet shattering model and its impact on heat transfer, the packing factors, the swell/rupture data considerations, and the application of the Sugimoto/Murao correlation. The licensee responded many of the NRC staffs concerns during the discussion. These responses are to be submitted to the NRC staff formally by the licensee. Minor changes were made to the discussion topics. No commitments or regulatory decisions were made by the NRC staff during the meeting.

-2 Please direct any inquiries to me at 301-415-3302.

Sincerely, Araceli T. Billoch Colon, Project Manager Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-400

Enclosures:

1. List of Attendees

2. Discussion Topics

3. Licensee's Presentation cc w/encls: Distribution via Listserv

LIST OF ATTENDEES JANUARY 11, 2012, CLOSED MEETING WITH CAROLINA POWER & LIGHT COMPANY SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 REALISTIC LARGE BREAK LOCA LICENSE AMENDMENT REQUEST U. S. NUCLEAR REGULATORY COMMISSION Douglas Broaddus Anthony Mendiola Araceli T. Billoch Col6n Benjamin Parks Eva Brown Leonard Ward Carolina Power & Light Company John Caves Dean Tibbitts Dave Corlett Mike Blom AREVA Bob Baxter Bert Dunn Nithian Nithianandan Mireille Cortes Gayle Elliot Enclosure 1

ENCLOSURE 2 SHEARON HARRIS NUCLEAR POWER PLANT DOCKET NO. 50-400 DISCUSSION TOPICS FOR REALISTIC LARGE BREAK LOCA LICENSE AMENDMENT REQUEST

SUGGESTED DISCUSSION ITEMS LARGE BREAK LOSS-OF-COOLANT-ACCIDENT SUBMITTAL SHEARON HARRIS NUCLEAR POWER PLANT, UNIT NO.1 PROGRESS ENERGY DOCKET NO. 50-400

1. For droplet break up model, show the drop sizes produced by the model for several low reflood rate data. Present the clad, vapor temperature and total heat transfer coefficient versus time at the measured axial locations. Show the heat transfer coefficient for all of the components comprising the dispersed flow film boiling (DFFB) heat transfer, including the interfacial heat transfer coefficient.

2. Since RELAP5 is one-dimensional the vapor temperature and droplets are distributed evenly across the hot channel. The code computed cross-section average quantities appears to fail to properly capture the very high temperature gradient in the vapor phase boundary layer near the wall so that the distribution of the evaporating water droplets playa fundamental role in the heat transfer process. In particular, interfacial heat transfer is over predicted. This appears to be a major limitation for all one-dimensional codes. Test data shows that the channel is three-dimensional with accumulation of drops in the central region and a highly superheated region near the walls. Modeling this multi-dimensional behavior leads to a substantial reduction in the interfacial heat transfer and limiting of the droplet de-superheating to the central core and not the highly superheated layer near the walls.

Explain what adjustments are made to the DFFB model components to overcome this major discrepancy. That is, the sink temperature is not the average channel temperature for computing single phase heat transfer, an interfacial heat transfer between the drops and the vapor is control by the lower vapor temperature in the central core where the drops reside.

3. Due to the simplified one-dimensional averaging of thermodynamic quantities in RELAP5 and the limited data, it is difficult to quantify all of the component contributions to DFFB.

a. Address how the magnitude of the droplet contribution is verified in the RELAP5 model.

-2

b. Without detailed knowledge of the magnitude of all of the components to DFFB, validation of this model against reflood data may result in including other phenomena/effects that are not pertinent to the heat transfer benefits from the droplet break up model. Explain and justify the magnitude of the impact on DFFB heat transfer with this new model.

c. Describe the interfacial heat transfer model and the impact on interfacial heat transfer coefficient with the new droplet model. In comparing the DFFB against data with the new droplet model, show all of the contributions to the total heat transfer coefficient versus time at the peak clad temperature (PCT) location.

4. The packing fraction of 50 percent does not appear to capture all of the test data.

Packing fraction as a function of burst strain varies in the range 52 to 80 percent based on data from Broughton, J. M, 1981, "[Power Burst Facility] PBF [Ioss-of-coolant accident] LOCA Test Series, Test LOC-3 and LOC-5 Fuel Behavior Report,"

NUREG/CR-2073. The Nuclear Energy Agency (NEA) Organization for Economic and Co-operation and Development (OECD) Nuclear Fuel Behaviour in Loss-of-coolant Accident (LOCA) Conditions State-of-the-art Report identifies 55.5 and 61.5 percent fill fraction for the FR2 reactor test E2. Values for the high burnup fuel in IFA-650.4 are expected to be higher than 70 percent, consistent with the bounds for PBF/LOCA gamma scanning and micrographies and FR-2. (See Grandjean, C "IRSN Calculation of the IFA-650.4 and .5 LOCA Tests ISRN, Cadadache, Fr. EHPG Meeting, Storefjell, March 12-15, 2007 meeting).

Show the impact on PCT for fill fractions up to and including 80 percent. Please also describe how the fill fraction is sampled.

5. Address whether the use of a nominal decay heat curve has ever been applied to decay heat test data over the range of applicability to show that this approach captures all decay heat conditions. The discussion should also address the uncertainty in generating this nominal curve and demonstrate that use of the nominal curve does not capture the decay heat for the first two seconds. Provide a multiplier which appropriate captures the decay heat behavior during this first two seconds of the curve.

ENCLOSURE 3 SHEARON HARRIS NUCLEAR POWER PLANT DOCKET NO. 50-400 LICENSEE'S PRESENTATION FOR REALISTIC LARGE BREAK LOCA LICENSE AMENDMENT REQUEST

• Introductions

• Overview of Status

• Response to Individual questions from December 13th Meeting

• Schedule

• Concluding Remarks Harris M5 RLBLOCA questions - NRC Meeting p.2 ~ Progress Energy

Progress Energy

• Mike Blom

• John Caves

• Dave Corlett

• Dean Tibbitts AREVA

• Bob Baxter

• Bert Dunn

• Mireille Cortes

• Nithian Nithianandan

• Gayle Elliott (IQ~ Harris M5 RLBLOCA questions - NRC Meeting

p. 3 ~ Progress Energy

Cycle 18 Feed Assembly Zircaloy 4 M5 Clad LBLOCA Method EMF-2087 ANP-3011 based on EMF-2103 PCT 2081 OF 1919 OF Transient 7% <3%

Oxidation Safety Analysis 2958 2958 Core Power (MW)

~ Harris M5 RLBLOCA questions - NRC Meeting p.4 ~ Progress Energy

Appendix K 2958 Zircaloy 4 2081 7.0 EMF-2103 Best 2993 Zircaloy4 1887 2.02 Rev 0 Estimate EMF-2103, i Best 2958 1930 11.95 Rev 0 + Estimate Trans Pkg ANP-3011 Best 2958 M5 1919 2.94 Estimate Harris M5 RLBLOCA questions - NRC Meeting p.5 L Progress Energy

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• Additional Sensitivity Cases Run (Response to Question 1)

• Discussions of Questions 2 to 6 refined and responses prepared

• ~ Harris M5 RLBLOCA questions - NRC Meeting p.6 ~ Progress Energy

• Characterize droplet shattering model without modeling the fuel relocation

• Show the sensitivity of the fuel relocation to fuel relocation packing factor

• Consider including a range of packing factors 30 - 80%

• Utilize the Harris Nuclear Plant (HNP) limiting-peak clad temperature (PCT) case

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Harris M5 RLBLOCA questions - NRC Meeting

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Question #1 Expand Sensitivity Study Table 1: 0.5 Packing Fraction Cases with Hot Assembly Rupture Harris M5 RLBLOCA questions - NRC Meeting p.9 ~ Progress Energy

Question #1 Expand Sensitivity cont'd

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Table 2: 0.6 Packing Fraction Cases with Hot Assembly Rupture

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Question #1 Expanded Sensitivity cont'd

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Table 3: 0.7 Packing Fraction Cases with Hot Assembly Rupture

• ~ Harris M5 RLBLOCA questions - NRC Meeting

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• ~ Harris M5 RLBLOCA questions - NRC Meeting p.13 ~ Progress Energy

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Question #1 Maximum Packing Factor [PF] Considerations PBF - Power Burst Facility

• Only 1 gamma scanning data point at 80%

packing fraction and it is at 30% strain

• Data results are in question due to material movement during the handling of the test rods

• Micrographies measurement is more accurate compared to gamma scanning

• PBF micrographic data shows 70% PF for rupture strain below 50%

• AREVA estimates 45% PF for strains near 70%

~ Harris M5 RLBLOCA questions - NRC Meeting

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Question #1 Maximum PF Considerations conl'd Packing Fraction vs. Rupture Strain Note: The balloon Balloon filling rate (0/0) by relocated fragments filling rate is interpreted as packing 90'

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Harris M5 RlBlOCA questions - NRC Meeting p.16 ~ Progress Energy

FR-2 (E-5) 61.5% I Test maximized rupture strain and had very small rupture opening Halden IFA- 53% 92 GWd 650.4 Halden IFA- ,r Not measured, but observed -- 90 GWd 650.9 as similar to 650.4 (N EA report)

Halden Test No relocation (no strain) 60GWd IFA-650.10

~ Harris M5 RLBLOCA questions - NRC Meeting p.17 ~ Progress Energy

• Studsvik Tests - Provide results for burn-ups at 70 GWd, show no fuel in the ruptured region as it was all lost out of the rupture

• KfK/FR2 Tests - Only rod E5 showed a PF of 61.5%. During the swelling and rupture of the rod, the cladding expanded to make a seal around the 10 of the container tube, thus invalidating any other results Harris M5 RLBLOCA questions - NRC Meeting p.18 ~ Progress Energy

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• 50%, as in ANP-3011 P, is an appropriate PF based on the Halden test results, which are the most reliable information available

• 70% PF is an upper limit (based on older PBF data) for the strain range seen in the HNP plant cases

• Sensitivity study concludes base case is conservative and no bias is proposed (with droplet shattering)

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Harris M5 RLBLOCA questions - NRC Meeting p.22 ~ Progress Energy

Question #2 Droplet Shattering Model Address whether droplet shattering is calculated on all flow blockage (non-vertical) surfaces in the S-RELAP5 calculation. If not, provide the flow blockage surfaces which are assumed to cause droplet shattering.

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Harris M5 RLBLOCA questions - NRC Meeting p.23 ~ Progress Energy

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Page 122 of ANP-3011 (P), "Harris Nuclear Plant Unit 1 Realistic Large Break LOCA Analysis" states, In the present model, the rupture blockage ratio [which is correlated to the number of droplets to yield a maximum atomization factor], 8, is taken from the swelling and rupture correlation.

• Address whether droplet shattering is calculated only against the hot pin rupture, or the additional flow blockage areas (i.e., balloon/burst regions, spacer grids, etc.) assumed to be present upstream of the hot pin rupture location.

• If the additional flow blockage areas are not based on pre-transient core geometry, discuss how the locations and sizes of flow blockages are distributed.

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Address how well the S-RELAP5 modification predicted the data.

Harris M5 RLBLOCA questions - NRC Meeting p.29 ~ Progress Energy

Question #4 Benchmark to SR Data cont'd

• An Excel spreadsheet that contained blockage Test 61607 data has been placed in PGN FTP site for NRC to retrieve for further evaluation

• Figures 6-14 and 6-15 in ANP-3011 P show S-RELAP5 conservatively predicted PCT compared to blockage test data for the FLECHT-SEASET and REBEKA-6 tests

• Additional FLECHT-SEASET test benchmarks, 61509 and 61607, are presented on the next two slides

• COBRA-TF (NUREG/CR-4166) modeled all expected multi dimensional flow phenomena at rupture location, including flow diversion

• The figures show that S-RELAP5 conservatively predicts cladding response compared to the COBRA-TF results

.~ Harris M5 RLBLOCA questions - NRC Meeting p.30 ~ Progress Energy

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Question #6 Droplet shattering impacts on Heat Transfer Explain how droplet shattering model incorporates the following droplet-dependent heat transfer effects:

• Inter-phase heat transfer

• Fluid-structural interactions including cladding, balloon, and spacer heat transfer to coolant

• Validate heat transfer modeling for these separate effects

~ Harris M5 RLBLOCA questions - NRC Meeting p.35 ~ Progress Energy

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~ Harris M5 RLBLOCA questions - NRC Meeting p.41 ~ Progress Energy

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• Benchmarking demonstrates that sensitivity study model provides conservative predictions "Ir.p- . ~~

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Harris M5 RlBlOCA questions - NRC Meeting p.42 ~ Progress Energy

• Best estimate to upper bound packing factor is 50% to 70%

• The response of PCT over a range of PF shows a smooth response to changes

• AREVA's sensitivity method has been successfully benchmarked against applicable research

• AREV A's treatment of the SRR phenomenon is applied in a conservative manner within ANP-3011

• A PCT of 1919 from base case is therefore defensible

~ Harris M5 RLBLOCA questions - NRC Meeting ~

p.43 ~ Progress Energy

Discussion of questions received on January 10th , 2012.

~ Harris M5 RlBlOCA questions - NRC Meeting p.44 L Progress Energy

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• Progress Energy will docket the responses

• Projected docket date is January 20, 2012

• Target date for new round of questions to be docketed is the week of February 13th

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Harris M5 RLBLOCA questions - NRC Meeting p.45 ~ Progress Energy

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Pkg ML12076A140, Summary: ML12076A141, Discussion Topics: ML12076A14 Licensee's Presentation Slides: ML12076A165 OFFICE LPL2-2/PM LPL2-2/BC NAME ABiliochCol6n DBroaddus FSa 16n DATE 03/22112 03/22/12 04/05/12