ML021370484

Extended Power Uprate - Presentation to the Advisory Committee on Reactor Safeguards

ML021370484
Person / Time
Site:	Brunswick
Issue date:	05/02/2002
From:	Carolina Power & Light Co
To:	Advisory Committee on Reactor Safeguards
References
Download: ML021370484 (84)

Text

Presentation to the Advisory Committee On Reactor Safeguards May 2, 2002 S*CP&L AProgress Energy Company

Bob Kitchen EPU Project Manager SjCP&L A Progress Energy Company 2

Overview Implementation Approach 170 Current BNP Operation P 105% Uprate Completed P 2-Year Operating Cycle

"* EPU Request Is a 15% Increase P 120% From Original Licensed Thermal Power (OLTP)

P No Reactor Pressure Increase

"*Two Step Uprate Required o 1st Power Uprate - 112 to 115% (OLTP)

P 2nd Power Uprate= 120% (OLTP)

• • CP&L 3

Overview Parameter Comparison I

I I I Parameter OLUP CLTP' EPIU, Reactor Thermal Power 2436 2558 2923 (MWt)

Reactor Steam Flow 10.470 11.089 12.781 (Mlb/hr)

Reactor Dome Pressure 1020 1045 1045 (psia)

CP&L 4

Overview Required Safety Significant Modifications o Standby Liquid Control (SLC)

P Increase amount of Boron injected P Change required to meet cold shutdown o Unit Trip Load Shed o Reduce Challenge to Offsite Power P Ensures Adequate Post-Unit Trip Voltage o Selected Load Shed of Large Balance-of-Plant Motors S

5

Overview Phase 1 - Balance-of-Plant Modifications I* Replace High Pressure Turbine P Modify Electro-Hydraulic Controls (EHC)

"*Replace Reactor Feedwater Pump Turbines

"*Replace Feedwater Heaters

"*Improve Grid Stability P Power System Stabilizer P Out-of-Step Protection

"*Improve Generator Isophase Cooling S

6

Overview Phase 2- Balance-of-Plant Modifications l Replace Main Transformers

. Replace Feedwater Heaters o Upgrade Moistu re Separator Reheaters CP&L 7

Overview Impact on Plant Margins I Plant Modifications to Maintain Operational Margins Plant,Ch-,ange Margin Impact SLC Boron Increase Improve SLC Margin Require 1 Pump Versus 2 Stability Option III Improve Operating Margin No Change to Safety Margin Improve Operator Interface, Power Range Instrumentation Improve Maintenance, Address Obsolescence Upgrade Condensate Maintain Standby Pumps Power System Stabilizer Upgrade Grid Stability SCP&L 8

Overview Control of Phase 1 Operation re Determination of Phase 1 Operating Level P Uprate Monitoring and Testing P Establishment of Interim Operating Power 9 Supporting Analysis

• Procedural Controls sCP&L 9

Overview Exceptions to ELTR Thermal-Hydraulic Stability

• ECCS-LOCA

"* Reactor Transients

"* Large Transient Testing Exceptions Consistent With Previously Approved EPU Submittals Q CP&L 10

Overview Unique Aspects

1. Implement Actions to Enhance Grid Stability

"*Maintain Hydrogen Water Chemistry

"*Manage Cycle Energy Requirements o 2-Year Fuel Cycle P 97% Capacity SCP&L 11

Tom Dresser Lead Engineer - Nuclear Fuels S*CP&L A Progress Energy Company 12

Core Considerations Equilibrium Cycle Design Targets Re(quired Changes

"* Cycle Energy 1(OxlO GE14 Bundles

"*Thermal Margin

• Higher Enrichment

"* Discharge Exposure

• U*arger Reloads

• Hot Excess Reactivity

• S LC Boron

"* SLC Margin eTIhermal Limits

"* Cold Shutdown MIonitoring Threshold Margin CP&L 13

Core Considerations BNP Unit 1 Cycle 14 I o Design Goals Met

. Current Methods and Margin Expectations eNoSLO Modification

. Very Fla at Radial Power Distribution o SLMCPR Change Adequate Margin Demonstrated for BNP Initial EPU Through Equilibrium EPU Cores CP&L

~ 14

Core Considerations ATWS - Analysis Results Eventf OLTP EPU Parameter Conditions Results Results Limit, Peak Vessel Bottom PRFO/BOC 1372 1487 < 1500 Pressure (psig)

Peak Suppression Pool MSIVC/EOC 194.8 195.5 < 207.7 Temperature (OF)

Peak Containment MSIVC/EOC 12.7 12.9 < 62 Pressure (psig)

Peak Cladding PRFO/EOC 1449 1309 < 2200 Temperature (OF)

CP&L 15

• D (0

-a

Blane Wilton Supervisor - Reactor Systems SC P&L A Progress Energy Company 21

Reactor Vessel & Internals Effects & Impacts e Pressure/Temperature (PT) Curves P Current PT Curves Approved for Use With EPU Through March 2003 o Will Submit New Curves With Updated Fluence Methodology Per RG 1.190 in June 2002 o Fluence o Fluence Increase Not Directly Proportional to Power Increase - Due to Core Configuration S

22

Reactor Vessel & Internals Effects & Impacts

• Embrittlement o The Upper Shelf Energy (USE) Equivalent Margins Analysis (EMA) Values for the Reactor Vessel Beltline Materials Remain Within the Limits of 10 CFR 50 Appendix G for 32 EFPY of EPU Operation 9 Fatigue o The ASME Code Fatigue Limits Are Met Through EOL + 20 Years for All Vessel Components for EPU Conditions SCP&L 23

Mark Grantham Superintendent - EPU Design CP&L A Progress Energy Company 24

Containment Response Key Results - DBA LOCA I I UFSAR, Current, Current Para met.,hods eter'L"/im MethocI: Methods Acceptance,;'it Parameter ,r(102% CLTP) (102% CLIP') (102%, EPU) , ln Drywell Pressure (psig) 40.9 44.2 46.4 62 Drywell Air Space 286.7 290.4 293 340 Temperature (OF)

Wetwell Pressure (psig) 14.0 30.5 31.1 62 Suppression Pool 189.4 197.9 207.7 220 Temperature (OF)

CP&L 25

Containment Response Net Positive Suction Head e Currently Committed to Safety Guide 1 P No Credit for Containment Overpressure 9 EPU Short Term NPSH P No Credit for Containment Overpressure Required

• EPU Long Term NPSH P Required Overpressure 3.1 psig P Requested Overpressure 5.0 psig o Available Overpressure 11.3 psig SCP&L 26

Dan Pappone LOCA Process Lead - General Electric CP&L AProgress Energy Company 27

Feedwater/Recirculation Line Break Dynamic Loads on Vessel Rreak FIclwhI B

2 CP&L 28

Feedwater/Recirculation Line Break Acoustic and Flow-Induced Loads

• Recirculation Line Break Limiting o Subcooled Region o Narrow Space Between Vessel Wall and Shroud o Single Large Break Area for Pressure Wave Source 9 Loads Evaluated for P Jet Pump P Core Shroud o Shroud Support (Acoustic Loads Only)

S 29

Feedwater/Recirculation Line Break EPU Approach e Evaluate Change in Mass, Energy Release From Break i Reduce Conservatism in EPU Break Flow Analysis to Stay Within Original Hydrodynamic Load Definition If Possible 9 Evaluate Increase in Hydrodynamic Load Definition, Consistent With Design Basis to Accommodate EPU o Use Margin Between Calculated and Allowable Stresses o Reduce Conservatism in Stress Calculations, Where Applicable SCP&L 30

Feedwater/Recirculation Line Break Analysis Results

• BNP EPU Results o Components in the Core Flow/Steam Flow Path Saw Increase in Stresses i Refined Analysis Performed for Core Shroud, Shroud Support, and Jet Pump Diffuser o All EPU Stresses on RPV Internals Are Within Allowable Limits S

31

Dan Pappone LOCA Process Lead - General Electric S\CP&L A Progress Energy Company 32

Appendix R - PCT IEvaluation Assumptions

. Loss-of-Offsite-Power

. FW Ramps Down to Zero in 5 Seconds

"*High Pressure System Not Credited for Vessel Makeup

"*Nominal Power 9 Nominal Decay Heat e 3 SRVs Start Blowdown at 40 Min utes

• 1 Low Pressure Coolant Injection Pump Available SCP&L 33

Appendix R - PCT Evaluation Conservatisms & Results i eConsE.rvatisms o 90% of SRV Capacity for Blowdown o ECCS Performance (Minimum Flow Rates As Functions of Vessel Pressure, Maximum Valve Stroke Times, etc.)

o High Pressure Injection Not Available for Makeup

• Results o 1468 0 F Versus 1500'F Limit CP&L 34

Bob Kitchen EPU Project Manager CP&L A Progress Energy Company 35

Operational Aspects Summary of Operator Impacts

"*EPU Test Plan per ELTR1 o Exception - Large Transient Testing

"*Operational Changes

) New Approach to Instability P More Power Reductions P Small Reduction in Operator Response Time CP&L 36

Operational Aspects Probabilistic Safety Analysis - Results "1. No Change in System Success Criteria

"* No New Accident Sequences Identified

"* No Significant Impact Due to Procedural Changes

"* No Significant Impact Due to Hardware Changes (e.g., Replaced "In-Kind" With Like Equipment)

"*Slight Decrease in Time Available for Four Operator Actions SCP&L 37

Operational Aspects Probabilistic Safety Analysis - Results I

CDF, LERF ,,

(delta) (dielta),

Base, 2.55E-5 4.27E-6 EPU 2.59E-5 4.46E-6

(+1.6%) (+4.5%)

With'SLC 2.32E-5 3.07E-6 Modif ication, (-9%) (-28%)

CP&L 38

Core Considerations Thermal-Hydraulic Stability IEPo I Option III Stability Solution

"*Exception to ELTR

"*GEý Recommended Validation of Setpoints

"*Impact of EPU P Met All Limits o Current Methods and Margin Expectations o No Change in Regions CP&L 17

Core Considerations ECCS LOCA - Impact of EPU I GE9Prop I Crt er CLTP EPU Li t Nominal PCT (TF) 1052 1049 N/A Appendix K PCT (°F) 1537 1530 N/A Upper Bound PCT(TF) < 1490 < 1487 < 1600 Licensing Basis PCT (TF) < 1560 < 1557 < 2200 Maximum Local Oxidation < 1% < 1% < 17%

Core-Wide Metal-Water<0.1%<0.1%<1%

Reaction Coolable Geometry, Maintained Maintained Maintained Long-Term Cooling_ __I-_I CP&L 18

Core Considerations Transients Overview I EPo -

ELTR Exception

• EPU Analysis P Overpressurization o Loss of Feedwater e Reload Analysis o Load Reject Without Bypass CP&L 19

Core Considerations Transient Analysis Results I GEPoritr I I Limitting Event CLTP`'6 EPU Limit Overpressu rization (psig Peak Vessel Pressure)

Loss of Feedwater Acceptable 7.26 > TAF (Feet Above TAF)

Load Reject Without Bypass 1.52 1.51 N/A (OLMCPR)

CP&L 20

BRUNSWICK STEAM ELECTRIC PLANT Units I & 2 Extended Power Uprate Brenda Mozafari NRR Senior Project Manager Division of Licensing Project Management May 2, 2002 1

Overview

• BWR4/Mark I

• 20 percent power uprate from OLRTP

• Constant reactor dome operating pressure

• 5 percent stretch uprate approved Nov 1996

• 2 part implementation (7% and 8%)

• BOP modifications

• GE14 fuel 2

Application Mostly follows ELTR1 and ELTR 2 Exceptions to ELTR1 and ELTR 2 are identified Non-risk-informed submittal

"*Experience from previous EPU reviews 3

SCOPE OF REVIEW

"*REACTOR CORE AND FUEL PERFORMANCE

"*REACTOR COOLANT SYSTEM

"*CONTAINMENT ANALYSES

"*ECCS / LOCA EVALUATION

"*SPECIAL EVENTS/LIMITING OPERATIONAL TRANSIENTS

"*RADIOLOGICAL CONSEQUENCES

"*HARDWARE CAPABILITIES/SYSTEMS AND COMPONENTS

• VESSEL/ NSSS PIPING

"*INSTRUMENTATION AND CONTROLS

"*ELECTRICAL POWER AND EQUIPMENT QUALIFICATION

"*HUMAN PERFORMANCE (OPERATOR RESPONSE TIME)

" PSA 4

BRUNSWICK UNITS 1 & 2 EXTENDED POWER UPRATE ACRS Meeting May 2, 2002 Reactor Systems Branch BWRs and Fuel Performance Section Ralph Caruso: Section Chief Zena Abdullahi: Lead Reviewer 5

Scope and Depth of Staff Reviews Code Applicability Staff review scope Experience from previous uprates Guidance documents (SRP, RG, topical reports)

, Other ongoing licensing reviews Operating experience and reports Knowledge/experience of reviewers Codes continue to be used within limits of applicability - no new phenomena

"*No new codes being used - no Maine-Yankee scenario

"*Staff has recent experience reviewing some of these codes SAFER/GESTR ODYSY TRAC-G 6

Scope and Depth of Staff Reviews Assurance of Appropriate Code Usage

• Vendors use codes in rigorous fashion

, Limits described in topical reports Methods described in detail and performed in accordance with written procedures Licensees audit calculations SAppendix B Quality Assurance Inputs maintained in controlled database

"*Staff has performed audits of selected BWR calculations and is comfortable that they are being done properly

"*Staff will continue to perform audits of BWR calculations 7

Review Agenda Example of Review Focus A-WS Standby Liquid Control Relief Valve Margin Evaluation N License Condition Conclusion 8

Example of Review Focus ATWS PUSAR peak vessel pressure of 1492 psig (1500 psig allowable)

Due to low margin, staff performed more in-depth review Staff determined limiting Unit not analyzed for (PRFO),

"*Unit 2 with larger bypass capacity (80.6 % v.s. 20.26% EPU steam flow)

"*ATWS analysis based on Unit 1 (limiting for LOOP and MSIVC)

CP&L reanalyzed

, PRFO ATWS event based on Unit 2 bypass capacity.

Reanalyzed PRFO event based on plant-specific data SYielded lower peak vessel pressure of 1487 psig 9

Example of Review Focus Standby Liquid Control No SLC relief valve margin evaluation in PUSAR

"*Initial BSEP SLC relief valve evaluation resulted in negative margin.

m.Evaluation based on GE data.

CP&L re-evaluated the SLC relief valve margin based on

, Predicted dome pressure Two pump system losses based on plant-specific tests 0 Original system losses based on 1984 GE evaluation Plant-specific elevation head calculation

"*Staff concluded o margin low but acceptable

"*CP&L acknowledges low margin plans to modify the SLC system to improve the margin Will replace the relief valves to obtain 50 psid margin 10

SLC License Condition No SLC Amendment Request Submitted SLC shutdown capability

"*Phase I, Unit 1, Cycle 14 o 660 ppm sufficient to meet cold shutdown concentration.

"*Phase II "o Increase boron (660 ppm to 720 ppm) required to meet cold shutdown "o Due to core design changes (i.e. larger batch fractions, higher enrichment in the GE14 fuel, driven by EPU and energy requirement to meet cycle length)

License Condition Requires amendment request 2 Changes to TS 3.i .7,"Standby Liquid Control (SLC) System" 0 6 months before implementation Single pump/squib valve commitment 0 Will start both pumps to avoid changing EPGs and operator retraining I1

Conclusions Licensing analyses are based on NRC-approved methods, codes and acceptance criteria.

BSEP EPU SAR is consistent with NRC-accepted guidelines and generic analysis for evaluating the impact of the extended power uprate on safe operation of the plant.

Deviations were presented to the Committee during the Clinton and the CPPU topical report meetings. (NEDC-32989P and NEDC-33004P)

The staff finds that CP&L provided sufficient evidence to support approvable for operation of BSEP Units 1 and 2 at the proposed power level of 2923 MWt.

12

Appendix R Evaluation

.10 CFR 50 Appendix R requires that tne train of systems necessary to achieve & maintain safe shutdown be maintained free of fire damage

• Brunswick used 1500 oF PCT as the fuel design limit for satisfying Appendix R

• The staff finds the use of 1500 oF PCT as the limit for a fire event to be acceptable 13

BRUNSWICK UNITS 1 & 2 EXTENDED POWER UPRATE ACRS Meeting May 2, 2002 MECHANICAL AND CIVIL ENGINEERING BRANCH CIVIL AND ENGINEERING MECHANICS SECTION Kamal Manoly: Section Chief May 2, 2002 14

Feedwater And Recirc Line Break Effects on Reactor Internals

"*In General, Loads Resulting From Pipe Break Dependent on Line Pressure and Temperature; But, Not Flow Rate Increase Due To Power Uprate Slight Increases in RIPDs For Some Internal Components as a Result of EPU Acoustic and Flow-Induced Loads Evaluated For Governing Recirc Line Break Blowdown Loads Governed by Main Steam Line Break Calculated Stresses and CUFs For Various Internal Components Either Bounded by Original Analysis or Show Slighit Increase; However, Within Allowable Limits of ASME Code,Section III 15

OVERALL CONCLUSIONS Analyses are based on NRC-approved analytical methods and codes On-site audit confirmed compliance with staff app roved methodology EPU SAR is consistent with NRC-accepted guidelines and generic evaluations Thermal limits and the applicable safety analyses would be reanalyzed or reconfirmed using NRC approved core reload analyses methodology 16

Applicability of Revised Fission Product Source Term (NUREG-1465) for High Burnup and MOX Fuels Presentation to the Advisory Committee on Reactor Safeguards Jason Schaperow Safety Margins and Systems Analysis Branch Division of Systems Analysis and Regulatory Effectiveness Office of Nuclear Regulatory Research May 2, 2002

Overview Objective Assess applicability of revised source term to high burnup and MOX fuels Approach Hold a series of expert panel meetings, including experts who developed basis for the revised source term Results Experts suggested source term values for high burnup and MOX fuels, identified issues, and recommended research 2

Revised Source Term Source term is the fission product release into containment atmosphere which is available for release to the environment RES published revised source term (aka alternative source term) in NUREG-1465 in 1995 More realistic than earlier TID-14844 source term Aerosol except for 5% of iodine which is vapor Four-phase release: gap, early in-vessel, ex-vessel, late in-vessel A few differences in release timing and magnitude between PWR and BWR (main difference is release timing for I) 3

Regulatory Applications Gap and early in-vessel phases of revised source term used for LOCA design basis accident analyses Exclusion Area Boundary, Low Population Zone, and control room doses containment isolation valve closure time (start time of gap release) integrated dose used to qualify equipment in containment post accident shielding, sampling, and access hydrogen generated by radiolytic decomposition of water All four phases of source term may be used for severe accident risk assessment 4

Regulatory Applications (cont.)

Revised source term being implemented voluntarily because of safety and cost benefits License amendments issued: Perry Fort Calhoun Grand Gulf Three Mile Island 1 Indian Point 2 Hope Creek Duane Arnold Surry 1 & 2 Crystal River Applications under review: Oyster Creek Brunswick Columbia (WNP2)

Oconee 1, 2, & 3 Kewaunee 5

Approach Held a series of expert panel meetings (Sep 2001 - Feb 2002)

Panel members were requested to judge applicability of each aspect of the revised source term, and if judged not applicable, to propose alternative As part of this effort, panel members...

considered recent data from international tests discussed physical phenomena affecting source term for high burnup and MOX fuels identified and prioritized source term research 6

Approach (cont.)

Panel of International Experts Bernard Clement (IPSN, France)

James Gieseke (consultant)

Thomas Kress (consultant)

David Leaver (Polestar Applied Technology)

Dana Powers (Sandia National Laboratories)

Others Principal Investigator: Mohsen Khatib-Rahbar (Energy Research)

Panel Facilitator: Brent Boyack (Los Alamos National Laboratory)

Consultant: Hossein Nourbakhsh (Energy and Environmental Science) 7

Applicability of Revised Source Term for High Burnup Fuel Panel assessment based on:

Maximum assembly burnup of 75 Gwd/t Core average burnup of 50 GWd/t Zirlo cladding (PWR), Zircaloy cladding (BWR)

Low pressure scenario (minimizes RCS retention) 8

Table 3.1 PWR Releases Into Containment (High Bumup Fuel)

Gap Release Early In-Vessel Ex-Vessel Late In-Vessel Duration (Hours) 0.4 (0.5)' 1.4 (1.3) 2.0 (2.0) 10.0 (10.0)

Noble Gases 0.05; 0.07; 0.07; 0.07 (0.05) 0.63; 0.63; 0.63; 0.65 (0.95) 0.3 (0) 0 (0)

Halogens 0.05 (0.05) 0.35 (0.35) 0.25 (0.25) 0.2 (0.1)

Alkali Metals 0.05 (0.05) 0.25 (0.25) 0.35 (0.35) 0.1 (0.1)

Tellurium group 0.005 (0) 0.10; 0.30; 0.30; 0.30 (0.05) 0.40 (0.25) 0.20 (0.005)

Barium, Strontium 0 (0) 0.02 (0.02) 0.1 (0.1) 0 (0)

Noble Metals (0) (0.0025) (0.0025) (0) 2 0; 0; 0.05; 0.05; TR Mo, Tc 0 0.15; 0.2; 0.2; 0.2; 0.7TR 0.02; 0.02; 0.2; 0.2; TR Ru, Rh, Pd 0 0.0025; 0.0025; 0.01; 0.01; 0.0025; 0.02; 0.02; 0.02; TR 0.01; 0.01; 0.01; 0.10; TR 0.2TR Cerium group (0) (0.0005) (0.005) (0)

Ce 0 0.0002; 0.0005; 0.01; 0.01; 0.005; 0.005; 0.01; 0.01; TR 0 0.02TR Pu, Zr 0 0.0001; 0.0005; 0.001; 0.002; 0.005; 0.005; 0.01; 0.01; TR 0 0.002TR Np 0 0.001; 0.01; 0.01; 0.01; 0.005; 0.005; 0.01; 0.01; TR 0 0.02TRI___I Lanthanides (one group) 0; 0; 0; 0; 0 (0) 0.0005; 0.002; 0.01 (0.0002) 0.005; 0.01; 0.01 (0.005) 0(0)

La, Eu, Pr, Nb 0.0002; 0.02TR 0.005; TR Y, Nd, Am, Cm 0.0002; 0.002TR 0.005; TR Nb 0.002; 0.002TR 0.005; TR Pm, Sm 0.0002; 0.002TR 0.005; TR I I The numbers in parenthesis are those from NUREG-1465, Accident Source Terms for PWR Light-Water Nuclear Power Plants (Table 3.13).

2 TR = total release. The practice in France is to assign all releases following the gap release phase to the early in-vessel phase.

Energy Research, Inc. 17 ERI/NRC 02-202 (DRAFT)

Results of Panel Assessment for High Burnup Fuel Physical/chemical forms expected to be applicable.

Only small changes in release-phase duration and release fraction expected.

Burnup-independent issues identified based on recent tests potential for enhanced Te release continued uncertainty in releases of noble metals, Ce, La groups recent data suggests subdividing noble metals, Ce, La groups Related issues BWR power uprates BWR fuel design 10

Tellurium Release Revised source term specifies early in-vessel Te release of 0.05 ORNL tests indicate Te gets sequestered in the Sn in Zircaloy cladding and not released until high fraction of cladding is oxidized More recent French tests (VERCORS, PHEBUS-FP) indicate that Te release could be similar to I (i.e., 0.30)

For PWRs, this was a contentious issue among panel members.

For BWRs, panel members specified release fractions similar to revised source term BWR zircaloy fuel channels tend to limit cladding oxidation 11

Other Source Term Issues Related to High Burnup BWR Power Uprates One expert saw no basis for significant effect on fission product release Another expert stated that flux-profile flattening associated with power uprates could increase the release rate for the outer assemblies.

BWR Fuel Design NUREG-1465 specifies a different source term for a BWR than a PWR Characteristics of more recent BWR fuel rod designs are closer to PWR fuel rod characteristics (e.g., pellet diameter, cladding thickness)

Panel indicated that similar rod designs tend to result in similar source terms.

12

Applicability of Revised Source Term for MOX Fuel Panel assessment based on:

Using MOX in PWR (about 1/2 of core)

Typical MOX assembly burnup of 42 GWd/t M5 cladding Low pressure scenario (minimizes RCS retention) 13

Results of Panel Assessment for MOX Fuel Physical/chemical forms expected to be applicable.

Only small changes in release-phase duration and noble gas, I, and Cs release fractions expected.

Same Te issue as for high burnup fuel.

Some of the experts did not recommend release fractions for Ba/Sr, noble metals, cerium, and lanthanum groups, because of the lack of test data.

Only data was a VERCORS test result for Cs with an arbitrary scale on the y-axis 14

Panel-Recommended Research High Priority Research Validate severe-accident analysis codes against recent source term tests Investigate in-vessel core degradation following vessel failure (air ingress)

Acquire any available data on fission product releases for high burnup and MOX fuels Perform fission product release tests for high burnup fuels using modern cladding designs (Zirlo and M5)

Perform revaporization tests Panel also recommended several medium and low priority source term research efforts.

15

Status Panel members recently provided comments on draft panel report.

Final panel report to be issued by June 2002.

Results of expert panel assessment to be used to help address reactor safety issues applications for high burnup and MOX fuels severe accident risk assessment other applications (e.g., vulnerability assessement) 16

High Burnup Fuel Research and Regulatory Issues Ralph Caruso Office of Nuclear Reactor Regulation Ralph Meyer Office of Nuclear Regulatory Research May 2, 2002

Background

• NRR User Need Request - October 4, 1993

• Updates to NRC Fuel Performance Models

• Revise Models for Stored Energy during LOCA, and Evaluate Impact on LOCA Analyses

• Evaluate Fuel Failure Thresholds for Normal Operation and RIA

• Commission Memorandum - November 25, 1996

• Low Enthalpy Fuel Failures

• Incomplete Control Rod Insertion

• Commission Memorandum - July 15, 1997

• Adequacy Assessment of Regulatory Guidelines and Licensing Criteria for High Burnup Fuel

• High Enrichment

• Spent Fuel Storage and Transportation 2

P..

Background (continued)

• Research Information Letter No. 174 March 3, 1997

• Proposed Changes to RIA Criteria

- Vendors and NRC calculations Below Proposed Limits

- Uncertainties may be Large

• Agency Program Plan for High Burnup Fuel - July 6, 1998

• Cladding Integrity and Fuel Design Limits

• Control Rod Insertion Problems

• Criteria and Analyses for RIA, LOCA, ATWS

• Update Fuel Rod and Neutronic Codes

• Source Term and Core Melt Progression

• Transportation and Dry Storage

• High Enrichments

• NRC Confirmatory Research for Burnup <62 GWD/MTU

• Industry Responsibility for Criteria/Data/Models for Burnup >62 GWD/MTU 3

3.

RIA Regulatory Criteria

"* General Design Criterion 28 "The reactivity control systems shall be designed with appropriate limits on the potential amount and rate of reactivity increase to assure that the effect of postulated reactivity accidents can neither (1) result in damage to the reactor coolant pressure boundary greater than limited local yielding nor (2) sufficiently disturb the core, its support structures or other reactor pressure vessel internals to impair significantly the capability to cool the core."

"* Standard Review Plan Section 4.2 - Coolable Geometry

"... [retaining a] rod-bundle geometry with adequate coolant channels to permit removal of decay heat."

"* Standard Review Plan Section 4.2 - Violent Expulsion of Fuel

" ... To meet the guidelines of Regulatory Guide 1.77 as it relates to preventing wide spread fragmentation and dispersal of the fuel and avoiding the generation of pressure pulses in the primary system of a PWR, a radially averaged enthalpy limit of 280 cal/g should be observed. This 280 cal/g limit should also be used in BWRs."

4

Revisions to Regulatory Criteria

• Current Criteria Challenged by Experience from CABRI and NSRR

• High Burnup Fuel Cladding Failed at Unexpectedly Low Enthalpy 0 Highly Corroded 0 Fabrication/Preconditioning Issues (REP/Nal)

• Interim Criteria Described in July 6, 1998 Commission Memorandum Based on RIL 174 0 Oxide Spalling: None Allowed 0 Cladding Failure: 100 callg (enthalpy increase)

S Coolability: 280 cal/g (enthalpy limit) for <30 GWD/MTU No Cladding Failure for > 30 GWD/MTU I

5

Current Situation

"* Detailed Analyses Performed by Vendors and by NRC 3-D Neutronics instead of 1-D

• Limiting Fuel (30GWD/MTU) Enthalpy Values less than 100 callg

- Therefore, No Fuel Cladding Failures Expected

• Radiological Consequences well within 10CFR100

• New Methods Under Review

"* New Fuel Cladding Materials

• Much Lower Corrosion

• No Spallation at Current Burnup

"* PWR Operational Practices

• All Rods Out at Full Power - No Control Rods to Eject

• Hot Zero Power Cases Analyzed

• LWR Rod Worths are Small

• High Burnup Fuel has Limited Reactivity

"* No Evidence From CABRI or NSRR of Molten Fuel Ejection 6

NRR User Needs

"* Proposed User Needs are Evaluated against Four Criteria

• Maintain Safety

• Improve Efficiency and Effectiveness

• Reduce Unnecessary Regulatory Burden

• Improve Public Confidence

"* No Licensing Action Under Review which Requires the Results of RES High Burnup Program 0 Current Plants meet Interim Criteria

"* Extended Burnup Criteria are Responsibility of Industry

• EPRI Topical Report on RIA Criteria is First of Several Expected Reports

• NRR/NMSS/RES Technical Advisory Group on Fuels will Provide Input to Help Identify Need for Research Information in Reviews

"* Agency Prioritization Process is Under Review

• Research "Needs" Process will be Revised in Integrated Fashion

• Revision will consider how parts of the High Burnup Fuel Program Plan fit into Agency Strategic Plan 7

Research Activites on High-burnup Fuel Agency Program Plan 0 1993 User Need memo of limited scope (codes and reactivity transients) 0 1994 memo to Commissioners on reactivity transients and high-burnup 0 1998 User Need from NMSS on dry storage and transportation 0 1998 Agency Program Plan (Commission paper)

• identified 4 reactor issues for confirmatory research

• said why okay to wait 3-5 years for resolution

• outlined research that would do the job 0 Research expanded to cover dry storage as well as reactor transients 0 Research activities broadened to include ZIRLO and M5 0 All experimental programs have industry cooperation PWR Rod Ejection 0 Regulatory Guide 1.77 needs revision of criteria at high burnups 0 1997 Research Information Letter (RIL-1 74) with 100 cal/g interim criterion 0 1999 Research Information Letter (RIL-177) on spalling 0 Heightened interest because of corrosion (Bulletins 2001-01, 2002-01) 0 Confirmatory assessment in 2-3 years with minimum effort 8

t LOCA 0 50.46 criteria and ECCS models were derived for unirradiated Zircaloy 0 Validation for unirradiated ZIRLO and M5 to be confirmed 0 Adequacy of corrosion adjustment for high burnup fuel to be confirmed 0 Significant hot-cell work left to complete confirmatory assessment 0 This research supports the performance-based approach to 50.46 revision BWR ATWS Instability

"* Higher risk than the postulated rod drop accident that is usually analyzed

"* Cladding temperature appears to be more important than mechanical loads

"* Improvement in analytical methods needed to complete confirmatory assessment

"* Value of related future testing being considered Analysis

"* Needed for resolution in all areas

"* Improvements in steady-state code, completed in 1997

"* Additional transient code development needed to complete resolution in above areas 9

I.

Description of Research on High-burnup Fuel Research on PWR Rod Ejection

"* Cabri (France) test reactor (NRC is a paying participant)

"* NSRR (Japan) test reactor (no cost to NRC)

"* IGR and BIGR (Russia) test reactors (small cost to NRC)

"* ANL measurement of mechanical properties (small part of larger program)

"* PNNL and BNL code analysis (small part of larger programs)

"* RES staff effort to document confirmatory assessment Research on LOCA

"* ANL LOCA tests and mechanical properties measurements in hot cell

"* Halden (Norway) LOCA tests in test reactor (small part of larger program)

"* JAERI (Japan) unirradiated test data (no cost to NRC)

"* Kurchatov (Russia) test in hot cell (small cost to NRC)

"* PNNL code analysis (small part of larger program)

"* RES staff effort on 50.46 and Reg. Guide 10

I.

Research on BWR ATWS

"* STUK and VTT (Finland) analysis with VTT-NRC codes (no cost to NRC)

"* Halden (Norway) test reactor data (small part of larger program)

"* PSU heat-transfer data

"* PNNL code analysis (small part of larger program)

"* RES staff effort on code analysis Research on Analysis

"* PNNL code development, analysis (FRAPTRAN, FRAPCON, MATPRO)

"* Halden (Norway) test reactor data

"* RES staff effort on code development and analysis Research on Dry Cask Storage

• ANL hot cell tests on creep and other mechanical properties

"* INEEL storage data

"* RES staff effort to assess regulatory requirements 11

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03 40 20 amm 0

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Slide 17 BROOKHAVEN NAII()NAL IA[IORATORY