ML20210B789

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Summary of 990630 Meeting with EPRI in Rockville MD Re Staff Review of EPRI RETRAN-3 Transients & Accidents Code.List of Meeting Attendees & Presentations Matls Used by Licensee Encl
ML20210B789
Person / Time
Issue date: 07/13/1999
From: Joshua Wilson
NRC (Affiliation Not Assigned)
To: Carpenter C
NRC (Affiliation Not Assigned)
References
PROJECT-669 NUDOCS 9907230234
Download: ML20210B789 (37)
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UNITED STATES l

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NUCLEAR REGULATORY COMMISSION l

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July 13,1999 MEMORANDUM TO: Cynthia A. Carpenter, Chief Generic issues, Environmental, Financial, and Rulemaking Branch Division of Regulatory improvement Programs, NRR FROM:

James H. Wilson, Senior Project Manager q-Generic Issues, Environmental, Financial, and Rulemaking Branch Division of Regulatory Improvement Programs, NRR

SUBJECT:

SUMMARY

OF MEETING HELD ON MARCH 2 AND 3, WITH EPRI CONCERNING THE STAFF'S REVIEW OF RETRAN-3D On June 30,1999, representatives of Electric Power Research Institute (EPRI) with the staff of the Nuclear Regulatory Commission (NRC) at the NRC's offices in Rockville, Maryland. The purpose of the meeting was to discuss the staff's review of EPRI's RETRAN-3 transients and accidents code. Attachment 1 provides a list of meeting attendees and their affiliations. provides the presentation materials used by EPRI at the meeting.

Discussion items Concerns of Dr. Wallis: Dr. Wallis, Chairman of the ACRS Thermal-Hydraulics Subcommittee, has stated that the momentum equation used by EPRI is not correct. EPRI does not understand what is wrong with the momentum equation as formulated in the RETRAN-3D documentation. EPRI refers to text books by Tong, Bergies, et al., and Lightfoot, et al., as having similar equations. EPRI would like the staff to indicate where the model is incorrect. The staff will prepare a request for additionalinformation (RAl) to address Dr.

Wallis' concems in this regard.

New Models: The five equation model revision is seen as cleaning up deficiencies in limited usage rather than errors. Code releases typically occur at two-year intervals. The material updated is for BWR applications where the five equation model is used in the core and four equstions elsewhere. The updates are to respond to cases run by Paul Sherer Institute that are beyond the initialintended applications of the code. Primary changes are made in interfacial mass transfer and flow regime maps.

EPRI has made a number of additional model changes, including incorporation of Purdue numerics, similar to PARCS code, as another kinetics option. The user will have choice of using a numeric method similar to PARCS or ARROTTA.

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. l SPERT Tests: The staff indicated the excellent prediction of the SPERT Hot Standby Case 81 and Full Power Test 86 obtained with the NESTLE code. EPRI has been asked to perform a similar prediction using *, RETRAN-3D code. EPRI stated that it cannot perform work for which it does not have. ;lity funding.- The staff will prepare an RAI formally requesting an EPRI RETRAN-3D calculation of the SPERT tests.

Status of New Materials: Draft of Volumes One and Three change pages have been made available to the staff for preliminary review Volume Four analyses of cases affected by five equation model changes and kinetics model additions will be made available when ready.

Final versions of the documentation are expected to be submitted by August 1999. The build of MOD 002b of the code is expected this week. That version will be very close to the

' MOD 003.0 version expected this summer, i

PIRT: EPRI has prepared a listing of important phenomena similar to a PIRT for the new applications to which RETRAN-3D is to be applied. The new applications are:

PWR Steam Line Break PWR Rod Ejection BWR Control Rod Drop BWR instability PWR ATWS (Full and Partial)

BWR ATWS (Full and Partial)

Shutdown Operation The listed phenomena are not ranked but are claimed to represent all of the high-importance phenomena. The staff notes that a large number of the assessments indicated in the tables are based on code-to-code comparisons, rather than code-to-data comparisons.

Manual Volume 5, User Guidelines: EPRI will not publish a complete User Guideline volume since the operational occurrences are not new and are covered by the RETRAN-02 Volume 5.

Only the new guideline material pertaining to 3-D kinetics and BWR stability will be published as a supplement to the existing volume. The staff notes that since this code is very user-dependent, a thorougn and complete user guideline specification is extremely important.

Staff Audit of Training: The staff will send two members to audit the RETRAN-3D training class scheduled for August 1999, in Idaho Falls. The strong influence of the user on the code results make it important to ensure adequate training of code users is done. This will also be discussed in the RETRAN-3D safety evaluation report.

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3-Schedule: A full ACRS meeting is scheduled for July 14,1999, at which Dr. Wallis will report on the review. EPRIintends to formally submit additional documentation by August 1999. The staff will conduct an audit of the RETRAN-3D training course August 1999. The staff intends to issue its final safety evaluation before the end of the calendar year.

Project No. 669 Attachments: As stated Distribution:

Central Files Public RGEB r/f RCaruso RLandry TUlses UShoop DMatthews SNewberry JHWilson BZalcman JDonohew J

rmiel JStaudenmeier d "g#M,lh a ll) 4 ljIh A l DOCUMENT NAME: g:\\jhwi\\meetsum.630

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C:q{lXB,gNV C:RGEB OFFICE RGEB q SC:RGEB JHWils[

JWhil CCarpenk NAME BZaleman DATE 7/[/99 7/ t'l/99 7//3/99 7$\\/99 OFFICIAL RECORD COPY

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Schedule: A full ACRS meeting is scheduled for July 14,1999, at which Dr. Wallis will report on the review. EPRI intends to formally submit additional documentation by August 1999. The staff will conduct an audit of the RETRAN-3D training course August 1999. The staff intends to issue its final safety evaluation be' ore the end of the calendar year.

Project No. 669 Attachments: As stated l

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-Project No. 669 Electric Power Research institute Mr. Kurt Yeager President and CEO Electric Power Research Institute 3412 Hillview Avenue Palo Alto, CA 94303 Robin Jones Vice Presidet and Chief Nuclear Officer Electric Power Research Institute 3412 Hillview Avenue Palo Alto, CA 94303 Mr. Raymond C. Torok Project Manager, Nuclear Power Group Electric Power Rose Mr. Kurt Yeager

. President and CEO Electric Power Research Institute.

3412 Hillview Avanue Palo Alto, CA 94303 Mr. Gary L. Vine Senior Washington Representative Electric Power Research Institute 2000 L Street, N.W., Suite 805 2

Washington, DC 20036 Mr. Bindi Chexal Electric Power Research Institute Post Office Box 10412

~ Palo Alto, CA 94303 I

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p LIST OF ATTENDEES AT MEETING WITH EPRI HELD IN ROCKVILLE, MARYLAND ON JUNE 29,1999

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EPRI RETRAN 3D REVIEW NAME AFFILIATION P. Boehnert NRC 4

W. Jensen NRC R. Landry NRC U.Shoop_

NRC J. Stauder.ineier NRC A. Ulses NRC L.Agee EPRI M. Paulsen EPRI j

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I EPali Meeting with NRC on Review of RETRAN-3D Presented by Lance Agee June 29-30,1999 h

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l Safety & Rellability Assessment Topics For Discussion

  • Purpose of Meeting
  • Introductory Comments
  • Features Needed for Analysis of Interest
  • Information Flow Between Physics and RETRAN
  • Licensing Analysis Requirements
  • Validation of RETRAN Safety & Reliability Assessment.

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r-EPRii Purpose of Meeting Focus of Review

  • Uses of RETRAN-3D for RETRAN or Applications
  • Closure of RETRAN-02 SER ltems R==nive current i=====

Volume 5 Guidelines Phenomena identification & Ranking Table Separate Effects Validation ACRS Comments Understanding of EPRI Code Development Process Define Future Interaction:

iSafety & Reliability Assessment

% %.. 3 emu EPRIi Currently RETRAN-02 SER Limitations Twenty RETRAN-02 SER limitations addressed by additional validation or supporting justification.

Six RETRAN-02 SER limitations addressed by New models in RETRAN-3D

- Three-Dimensional Kinetics

- Thermal Non-Equilibrium

- Non Condensable Gas Acknowledge that twelve items not changed.

. Safety & Reliability Assessment

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9 EPRii BWR & PWR Events Requiring New RETRANModels BWR Accidents & Special Events New Models Criteria

  • ATWS 3-D K Sup. Pool Tom.

' Stability MOC,3-D K Decay Ratio < 1.0 Th/Non-Eq PWR Condition IV & Snacial Events

  • Rod Ejection 3-D K PFE, PCT
  • Steam Line Break) 3-D K DNBR, CP, Dose
  • ATWS 3-D K Other
  • Mid Loop Operations Non Con Other iSafety & Reliability Assessment

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Proposed Resolution of lasues RETRAN Volume 5 Guidelines Existing guidelines adequate for all anticipated events.

Limited scope supplement to addressed:

- Method of Characteristics

-Three Dimensional Core Model

- Thermodynamic Non-Equilibrium

- Non-Condensable Gas

- Gap Conductance Phenomena identification & Ranking Table Address New Applications Current thoughts on structure gfety & Reliability Assessment env m ins.

EPRii D

i Least We Forget "We find a fundamental fault even with the exiting body of regulations.

A preoccupation developed with such large-break accidents as did the attitude that if the could be controlled, we need not worry about the analysis of 'less important' accidents"

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Report of the President' Commission on f

The Accident at Three Mile Island John G. G Kemeny, Chairman The Need For Change: The Legacy of TMI October 1979, Washington DC

, Safety & Reliability Assessment 3

EPRii Recommendations to the AIF Policy Committee on follow-up to the Three Mile Island Accident Existing safety and licensing analysis practice tends to concentrate on the limiting, or ' worst-case' events, with particular emphasis on hypothetical accidents to the neglect of higher frequency, lower consequence events.

Principle Recommendations of the subcommittee are:

- Close the loop between the plant safety anti performance analysis and actual operational experience, using more realistic analyses

- Use systematic engineering tools to extend preent analyses in scope, duration and events considered to provide a broader, more realistic base for operating procedures and reliability of systems.

Safety & Reliability Assessment'

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l EPRIi Features Needed for Analysis of Interest Purposeof Anseysee Thermal Neutronoce Feed Control Steady Sound Compon Hydraulice

& Power Beck System State Cond.

Models

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Deve6op Procedures and 3/4 Eqn Pt or 1 D Ins All Dom Dom Sons Traun Operators Evaluate Pland enodonte 3 or 4 Eqn Pt or 1 D Dom AA Dom Dom Dom

& AbnormalBehavior Review Plant 3/4 Eqn PTort D 6 ens Key Dom Dom Dom Moetcatsons bM i Zn~~

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,a BWR 4 Eqn 1-D Dom Key Dom Dom Dom PWR 3 or 4 Eqn PT Sene Key Dom Dom Sons Rod DrogwHod Eremort 4 Een 3-D Dom sne bars ans ins gg Lomed Rotor 3 Eqn PT ins ins Gene Ine

, one 52 PWR Steam LJne Broek 6 Eqn Pt or 3 D Dom tru Sens Gene ins j

Large preak LOCA 6 Een PT ins ins ins Sens ins f

ATWS S Egn 10 Dom Key Dom Dom Dom D

BWR Core 6tatWrty 5 Eon 10 or 3-D Dom ano Dom Dom Dom Key e includse trips and controllers scWee for e parescular event Dom. Domi re lui se. nouns.re me iu.e s a eu w ieve isafety & Reliability Assessment e,. _,,.

EPRii Characteristics of Operational Transients General Features

  • Control system action
  • Equipment responses (i.e., SCRAM, Valve Motion, etc.)

Important Modelina Features (Generally not considered in larae-break LOCA)

  • Must properly describe all applicable systems Hydraulic network Control system Component characteristics Feedback parameters
  • Correct steady-state imperative
  • Realistic boundary conditions essential Safety & Reliability Assessment i

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,EPRil Characteristics of Operational Transients (cont'd.)

Much Greater Demand on Modeler

  • Must understand plant system
  • Must know which plant equipment involved in transient Must understand computer code iSafety & Reliability Assessment senvuam ism ii EPRii 3

EG&G Predicted and Measured Pressurizer Level)

Loft Loss-of-Feedwater Test L6-5 i

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l EPRii Changes to EG&G's RETRAN-02 Loft L6-5 Model Added leak model to secondary Matched test initial conditions Used best estimate of decay heat Used non-equilibrium pressurizer model w

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EPRICode Lattice Phys. ics Linkage CPM-3 Cross sections & Physics Data u

System Analysis SS Core Simulation RETRAN-3D CORETRAN cross secuans Core Follow

& Physics Data Chapter 15 Core Design vents w a undw DNBR of CPR condiuons System Conditions

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EPRIj Steady State Physics Events & Quasi-Steady Transients

  • Rod Withdrawal Error
  • Fuel Loading Error

- Mislocated Fuel Assembly

- Rotated Fuel Assembly

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  • Kt s parameters must be in previously acceptable range
  • Applicable to PWR using RETRAN with PT kinetics
  • BWR over pressure events evaluated with 1-D kinetics RETRAN u,,y,io,

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Acceptable Range of a Key Safety Parameter An,y,;,

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Some Physics Safety Parameters Reactivity Terms

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  • Doppler coefficient a Moderator temperature coefficient

Controlitems

  • Rod withdrawal reactivity insertion rate
  • Scram reactivity curve Other issues
  • Kinetics information; delayed fraction and neutron lifetime
  • Peak linear heat rate as function of axial position
  • Pin census information a Umiting axlal shapes
  • Ex-core detector response functions StyVfLJAd5 S trWR 1?_

j Safety & Reliability Assessment EPRTi Plant System Analysis

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Output steady state Conditione System Geometric Description j

metrumentation system Deecription f

Protection System Description

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EPRii Chapter 15 Events Frequency Of Occurrence

__ BWR: 33 wal ;n.ti RETRAN not=='laak8= (N/A1 to 5

  • Accidents; Some Fuel Damage, cool-able geometry required (7,3 NA)
  • Special Events (ATWS, Stability); (6,2 NA)

PWR: 30 total events. RETRAN not applicabi.1st2 t

  • Incidents of Moderate Frequency (F-il); No fuel damage-ON8R Umit (22)
  • infrequent incidents (F-ill); Fuel damage must not preclude startup after outage (2,1 NA)
  • Umiting Faults (F-IV); Some Fuel Damage, cool-able geometry required (4,1NA)
  • Special Events (2)

-ATWS Mid-loop Operations iSafety & Reliability a==amament

' 2PRii BWR Accidents & Soecial Events )

BWR Accidents New Models Criteria Control Rod Drop Accident 3-D K PFE,Dosa Main Steamline Break No Dose Instrument une Break No Dose Loss of Coolant Accident NA Refueling Accident NA Recirculation Pump Selzure No MCPR, Dose Assembly Loading Error NA BWR Special Events Summary Shutdown Margin Demonstration NA Standby Uquid Control System NA Shutdown From Outside the CR No No Specific Umits Overpressure Protection Analysis No RV Pressure Stability MOC,3-D K Decay Ratio < t.0 TWS-3-D K Sup. Pool Tom.

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EPRij PWR Condition III, IV & Soecial Events PWR Condition ill Events New Models Criteria

  • Complete loss RC flow No DNBR
  • Fuelloading error NA DNBR,FDL PWR Condition IV Events
  • RC pump shaft seizure No DNBR, PSY, PCT
  • Feedwater system pipe break No PSY, Others
  • Rod Election 3-D K PFE, PCT
  • Steam Line Break) 3-D K DNBR, CP, Dose PWR Special Events
  • ATWS 3-D K Other
  • Mid-Loop Operations Non Con Other iSafety & Reliability Assessment Guidance Available i

l General Descriotion

  • EPRI NP-4498 The Reactor Analysis Support Package (RASP)

-Volume 2: BWR Event Analysis Guidelines

-Volume 3: PWR Event Analysis Guidelines RETRAN Related

  • EPRI NP-1850 RETRAN-02 Computer Code Manuel

-Volume 5 Modeling Guidelines LSafety & Reliability Assessment'

EPRii REG GUIDES for Operationhl Transient Evaluation

1. Information to be presented as a function of time.

(a) Neutron power

- (b) Averege and maximum heet flus.

(c) Reactor coolant system pressure (d) Thermallimits (MDNBR or MCPR)

(e) Coolant inlet temperature (f) Core average and hot channel coolant cult temperature (g) Maulmum fuel centerline temperature, cladding temperature, or fuel enthalpy

2. Evaluation of offects of chemical, thermal, Irradiation, mechanical and hydraulic interactions on fuel and cladding.
3. Defir.ition of the specified ecceptable fuel design limit (sAFDL).
4. Demonstration that the SAFDL le not exceedeo by calculating the minimum DNBR (CHFR or CPR for BWRs) and the maximum fuel temperature or a corresponding lineer heet generation rate.
5. Evaluation of fuel and cladding to ensure no rod perforetion.

iSafety & Reliability Assessment

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RASP Event Guideline Format BWR Event Guidelines PWR Event Guidelines

1) Event Description
1) Description
2) Phenomena
2) Key input Parameters
3) System Considerations
3) Key Calculated Parameters
4) Component
4) Assumptions Performance
5) Available Margin Characteristics
6) Modeling Phenomena
5) Integration of Codes and Analysis
7) Systems to be Considered
6) Utilization of Analysis
8) Integration of Codes and Analysis
7) Precautions
9) Utilization of Analysis
10) Useful Suggestions
11) Difference in Vendor Analysis
12) Demonstration Calculations

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EPRij Assessment of RETRAN l
  • NP-7450 RETRAN-3D-A Program for Transient Thermal-Hydraulle Analysis of Complex Fluid Flow Systems, Rev 4, Sept 1998

- Separate Effects Test (18)

- System Effects Test (10)

- RETRAN-02 & RETRAN 3D Comparisons (17) + TMl(12)

- Comparison to Plant Data (19)

  • RETRAN International Conferences Proceedings
  • NP-5840 7/1988 " Qualification of RETRAN[-02] for Simulator Application", July 1988 (80)
Safety & Reliability Assessment l

M EPRii Separate Effects Analyses in NP-7450 PRESSURE DROP CRITICAL FLOW

  • Ferrell-McGee Pressure Drop Data Fauske Cntical Flow Experiments
  • Non-condensable Gas Pressure Marviken Critical Flow Experiments

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HEA TRANSFER FRIGG-2 Void Fraction Data

  • Bennett, Hewitt, Kearsey, and FRIGG-4 Void Fraction Data Keeys Round Tube Data
  • Bennett, Collier, Pratt, and Thomton One-Foot GE Level Swell Test l

Annulus Data 1004 3 E chrock Grossman Round Tube 1

S Four-Foot GE Level Swell Test Data 5801-15

  • Condensation Analytic Solution ORNL THTF Void Profile Test
  • Condensation in the Presence of DECAY HEAT MODEL DESCRIPTION I

Non-condensables l

  • Non-condensable Convection Heat NATURAL CIRCULATION Purdue Thermosyphon Test Transfer Data
  • Method Of Characteristics Solution
Safety & Reliability Assessmont en-mea n

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EPRii System Effects Analysis in NP-7450

  • PRESSURIZER TESTS

- SHIPPINGPORT Loss-of-Load Tests

- MIT Pressurizer Tests

  • RETRAN-30 ANALYSIS OF LOFT SMALL BREAK

- LP-SB-1

- LP-SB-2

  • MULTfDIMENSIONAL KINETICS REFERENCE

- HERMITE REA Comparison

- NEACRP Benchmark

- PWR REA Comparison with ARROTTA

- PWR SLB Comparison with ARROTTA i

- NEACRP Benchmark

-TMI-1 Rod Election Accident Analyses

' Safety & Reliability Assessment' 9NV4 JAG tisen 77 EPRii RETRAN-02 and RETRAN-3D Comparisons in NP-7450 PWR Comnarlsons BWR ComnarigDjlt

_ ANO-2 Turbine Trip

  • Susquehanna Feedwater Calvert Cliffs Steem Line Break Controller Failure PWil Loss of Flow Susquehanna Feedwater Heater Failure

. Prairie Island SG Tube Rupture

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Controller Failure

. TMI Loss Of Feedwater Cofrentes MSIV Trip

. Almaraz Turbine Trip BWR ATWS

. KEPCO KNU 1 SG Tube Rupture KNU-2 Loss Of Normal River Band Two a

Recirculation Pump Trip Feedwater Yonggwang 1 Turbine Trip Peach Bottom Turbine Trip Safety & Reliability Assessment 9MV4 JAW 91MI6 PS

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All TMl Chapter 14 Compari.ons with both RETRAN-3D and RETRAN-02 Transients Analysis

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in NP-7450

  • Start-up Accident Comparisons with RETRAN 3D l

CORETRAN and TRAC

  • Moderator Dilution
  • Rod Election Calculation
  • Pump Start-up

- Hot Zero Power

  • Locked Rotor

- Hot Full Power

  • Dropped Rod

- Beginning of Cycle

  • Loss of Coolant Flow

- End of Cycle

  • Foodwater Line Break
  • SLB Calculation
  • Station Blackout
  • Loss of Foodwater Accident
  • Loss of Feedwater Safety & Reliability Assessment serm.we,= r.

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, EPRii RETRAN-3D Comparisons to Measured Data Performed by Utilities in NP-7450 KORI(PWR)

Lacuna Verde (BWR)

Nuclear Unit 1 Loss Of All

  • Generator Load Rejection Offsite Power q

N0 clear Unit 4 Large Load.

. MSIV Closure Cofrentes (BWR)

Reduction Test Nuclear Unit 2 Multiple Failure

  • HPCS Injection Event
  • FW Controller Failure Comanche Peak (PWR)
  • Level Setpoint Change Steam Electric Station Load
  • Feedweter Pump Trip Rejection
  • Recirculation Pump Low-BWR-5 Comparisons Speed Transfer Pressure Setpoint Change

' Level Setpoint Change

  • Generator Load Rejection with All MSIV Closure Partial Bypass Failure Load Rejection with Bypass Single MSIV Closura m.
Safety & Reliability Assessment f an.wm ii..

t EPRif Summary of BWR Analysis in NP-5840 Event

  1. of Analyses Peachbottom Tests 6

Pressurization Transients 7

Recirculation Pump Coastdown 4

Controller Setpoint Changes 3

Loss of Coolant Accidents 3

Miscellaneous Transients 5

Total 30 iSafety & Reliability Assessment.

EPRI, Summary of PWR Analysis in NP-5840 i

Event

  1. of Analyses Loss of Load Transients 11 Excess Heat Removal Transients 10 Loss of Feedwater Transients 4

Loss of Flow Transients 6

Loss of Coolant accidents 12 Miscellaneous Transients 7

Total 50 Safety & Reliability Assessment i

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EPRii Schematic of Boiling Curve l

~ Regions of Primary Interest

<y l Thom, et al l Nucleale 6415ng TRm81Gon.2Wng 9

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Correlation Basis Dittus-Boelter Thom, et. al.

Chexal-Lellouche Radiators Tubes or Annull Rod Bundles llr 2

L Distus. F. W.. and Boelter, L M K.," Heat Transfer in Automobile Rachators of the Tubular Type". Umversity of Cahfonus Pubhcanons in Engineenng. 2,443-41,19M Thom, J. R. S.. Walker. W. M., Fallon. T. A.. and Reissng, G F. S., "Boihng in Subcooled Water Dunns Flow Up Heated Tubes or Annub", Proc. Ins:n. Mech. Engrs.. I 80 pt 3c. 226246.1966 Chesal. V. K., tellouche, G S., Horowitz, ). H:alam J., and oh, S.,"The Chessi4ellouche Void Frachon Correlauon for Genershzed Apphcanons", NSAC-139.1991.

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i USNRC Presentation June 29,1999 Mark P. Paulsen CSA u{$$

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r RETRAN Development Process RETRAN-3D Development Since the Mid-1980s

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Code Evolution New Models Errors were Corrected Design Review was Conducted 1995-1996 MOD 001f Used for Design Review New Code Version Released Following Review - MOD 002.0 Formal QA implemented Code Releases Under Direction of the Maintenance Group Typically Two Years Model Development and Code Maintenance Work Continued ISR RETRAN Development Process (Cont'd)

NRC Review was initiated Development Activities were in Progress MOD 002.0 Formal Code Available for Review Summary of the Error Reporting and Status of the Code RETRAN Web Site Lists all Reported Errors l

mCSR l

New Development Areas Three-Dimensionan Kinetics Channel Model Simp! fy User input for Core Geonietry from CORETRAN vla CORETRAN Data interface (CDI) File Pren; ressor GUI Available New C.oss.Section Model Tabu ar Form Base Plus Change All Nodes Purdue Numerics Design Review items Four-Equation Model EOS for Noncondensable/ Water Mixtures Heat Transfer Linearization mCSR New Development Areas (Cont'd)

New/ Enhanced Modes Five-Equation interfacial Mass Transfer Steady-State Flow Split Calculation Second Order Transfer Function Purdue Numerics (Thermal-Hydraulics) mCSR

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l Five-Equation Model Originally Developed to Treat Subcooled Boiling in BWR's Simple Void Collapse Model Typically Applied in Core Region Validation Performed Frigg Void Data Turbine Trip Transients BWR Stability Analyses u _@.u f

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Five-Equation Model (Cont'd)

User Experience for BWR Applications Pressure Increase - Pressure Coefficient Problem Dependent Sensitive to the Pressurization Rate and Time-Step Size Limitations When Applied Outside of Core PSI Applied Model to Wider Variety of Problems Rapid Depressurizations Full Systems PWRs - Steam Generators Applied Beyond Intended Application mCSR 1

r Five-Equation Model (Cont'd) i Model Development and Assessment initiated with PSI Improve the Robustness Extend Range of Application

- Preliminary Results 9* Int. RETRAN Meeting Work Completed Fall of 1998 G

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Five-Equation Model (Cont'd) include in next Code Release - MOD 003.0 Formal QA Revise Documentation 9" International RETRAN Meeting Paper Report of Work in Progress Rapid Depressurization Transients Under Predicted Vapor Production j

BWR Pressurization Used Successfully in past Analyses Model Dependent Coefficient Lacked Robustness Interphase Mass Transfer Model Limitation or Deficiency Limitation vs Error mCSR

Five-Equation Model Use Not Required for Current RETRAN-02 Uses Limited Use for New Applications BWR Stability 1

New Application Areas Used in a Small Fraction of Analyses Interphase Mass Transfer Model Enhanced 4

Five-Equation Model not Replaced Misunderstanding Unfortunate mCSR New Review Material Status Draft of Volumes 1 Provided New Models Documented Error Corrections Draft of Volume 3 Provided New Models Documented Error Corrections Volume 4 Re-analysis Five-Equation 3D Kinetics Validation 9* International RETRAN Meeting Papers Code Availability QA Review Complete MOD 002b Build this Week Minor Changes for MOD 003.0 u %.CO P

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May 26 ACRS Meeting issues Five-Equation Model SPERT Analysis Assessment mCSR r

Five-Equation Model Misunderstanding Relative to Five-Equation Revisions Indicated Five-Equation Model Being Replaced Wholesale Interfacial Mass Transfer Model Change (Model improvement)

Improve the Robustriess Extend Range Model Can Be Applied Revision Does not Correct Errors Error Correction Made to Five-Equation with Noncondensables f(Q L %sm#I 11

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l Five-Equation Model (Cont'd)

Five-Equation Model Features and Change Status l

Featura Changes

- Basic Field Equations No Changes Made - None Planned

- Wall Heat Transfer No Changes Made - None Planned

- Wall Mass Transfer No Changes Made - None Planned

- Interphase Mass Exchange Revised Model

- Pressure Search Water Only Model Unchanged Water /NCG Model-Revised for Trouble Report 167 NCG Pressure Search Error Corrected m(_$.

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Five-Equation Model(Cont'd)

Statement at May 26 ACRS Meeting about Overspecification (RAI Item 3.9)

..They Came Back and Said Oh, Well, We're Revising That Whole Five-equatior' Model Anyway Because There Are Problems with it."

RAI Item 3.9 Response

!!!ustrated the Actual Code Steady-State Failed to Converge as Indicated Overspecification is a Common User-Input Error Related to Steady-State Solution of Mixture Momentum No Mention That the Model Was Being Revised Overspecification User Error Not Unique to Five-Equation Model More than One Set of Valid ICS & BCS l

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R SPERT Analysis April 20,1999, Conference Call with NRC Staff '

Followup Review Scope Sent 4/29/99 Generic Use of RETRAN-3D in "RETRAN-02 Mode"- Orgs with Approved Methodologies Lifting Restrictions in the RETRAN-02 SER New RETRAN-3D Models and Capabilities New Areas of Application Benchmarking of 3D Core Models to SPERT Test Data ACRS Testimony "So We're Stressing That We Do Recommend Very Strongly That the Applicant Begin Looking at the SPERT Data and so Some Similar Calculations."

mCSR F

SPERT Analysis (Cont'd)

Two Options for PWR Rod Ejection (REA) Review Code-to-Code Comparisons Application-Specific Justification Require a PWR REA Methodology Submittal Generic PWR REA Uncertainty Based Comparisons with SPERT Cata e

No Additional Justification for PWR REA Would Be Required.

Historical Code-to-Code Comparisons Preferred Ouestion Applicability of Neutronics Model Formulation to SPERT Possibility Excessive Uncertainty Considerable Technical Risk Schedule and Cost impact ARROTTA Core Model Approved by the NRC for Duke Power's PWRs NRC Should Accept RETRAN-3D for PWR REA Based on this Work L

v t.

l c-1 Inadequate Assessment RETRAN-02 Analyses Already Addressed NEW Applications Phenomena identified Assessment Reviewed i

l mCSR r

New RETRAN-3D Applications and important Phenomena New Application Areas PWR Steam Line Break Accident PWR Rod Ejection Accident BWR Control Rod Drop Accident BWR instability PWR ATWS (Full and Partial)

BWR A1WS (Full and Partial Shutdown Operation Draft Document for Phenomena identification by Application Area Comments on Assessment Avaliabie or Needed mCSR

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important Phenomena g.

(Draft)

PWR Simase use Break AccWont Assosoment Documenteelon U4em Stearn Generaturs Heat Treneler Srnter to Chap.15 Tron.

Vol 4 Level Smet SenestMty stumes Unty Sutwistus Entainment SonettMty stuees Utmy Sutweetets OnsMhrough Steam Generators Heat Tranater Sander to Chap.15 Tron.

Vd 4 IUtgity Sutzvetets Vold Deteueen Semiter to Chap.15 Tren.

Vol 4 IUnty Sutztuttets Loner Pienue Moran0 SensitMty stuessaast dets Utmy Sutweetsds

- Gem Reponed Power increase Code-to code comparson Vol. 4 Control Rod Worth Code-to code comportson Vol.4 Moderstar Temperature Codo40.cxxto cornpenson Vol. 4 Fuel Temperature Code 4> code compenson Vd 4 Local Voutng Code 40. code comparson Vol 4 o e Moderow wesiina S.anevity su es veny Sutwiews

- _ _ _ : TH Eflects Sensevtty stuees Vety Sutwesents Decay heet ANSI Std. Examples Vol.4 Pressunnar Oumurge Operatonal & Startup Date Vol. 4 IUsity Sutwrusses CSR r

important Phenomena (Draft) (Cont'd)

Asses. ment Docum.nt.non PWR ned am AccWont GIES Reponal Power increase (3D)

Code 4> code corrperson Vd 4 Contal Rod Worth Code-et> code comportson Vol 4 FuelTemperatura Code 4> code comportson Vd 4 Moderstar Temperature Code-t> code comportson Vol 4 Drect Moderator Hestrig Sensevtty stumes Utsly SutwnNtels Aseseement Documentation SWR Convol Rod Droc Accident GdES Reponal Power increase (3D)

Code-tc> code compenson Vol 4. add PSI Contal Rod Worth Code 4> code compenson Vol 4. add PSI Fuel Temperature Code 4> code comparson vol 4. add PSI Moderator Temperature Code 4> code compenson Vol d. edd PSI FRIGG Testa Vol 4 Local Voseng Drect Moderstar HeatMG Sonsttrvity stuees Utmy Submittels mCSR i

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r important Phenomena l

(Draft)(Cont'd)

BWR inetsbety Assesement Documentation Core Wide Power Oscannons Peach Bonom Tests Vad 4 Vermont Yansee Tests Vol. 4 Recenal Power OscAlstons Lelbstadt WPPS LaSase Cokentes Vous Feedback Peach Boeorn TT Tests Vol 4 edd 10 Kin.

Vosd Dettuton FRGG tests Vol 4 Drect Moderstar Heatm0 Senesvity stuees Ulsty Sutrruttais Densey Weve Osceshons FRIGG tests INET (PSI) tests Void Dettuton FRIGG tesis Vol. 4 & Vou! Book Seglo and Two-Phase Pressure Drop FRIGG tests Vol. 4 Hydraunes Channel Flow Dateuten CodNs> Code Utility Submutais Bypassieskape Flow Code-tc> Code Ushty Submittats vouf Genersion FRIGG Voul Compensons Vol 4 Re-Looo Centeugal Pump Perfornance Startup tests isensitn4ty stuees Vol 4 Jet Pump Performance Startup tests IsentAmly studes Vol.4 j

mCSR r

important Phenomena (Draft) (Cont'd)

PWR ATWS (Full and Partial)

Assessment Documentation QER Power REA and SLB Voi 4 Non4ymnpric poser shape (parhat ATWS)

REA and SLB Vol. 4 Dopprer Feedback REA and SLB Vol 4 Moderstar Temperature Feedback REA and SLB Vol 4 Baron Conconraton Feedbaca Sensibvity Studies Ubbty Submittain Rod Wor 1h REA and SLB Vol 4 Drect Moderstar >+eatmg SensitNety Studes UUiny SubmitWs Heat frontier Smier to Chapt 15 Trans Vol. 4 Demy Heat ANSI Std Examp6es Vol 4 Prasauntmhree Lme Choked Row Sander to Chapt.15 Trans.

Vol 4 U tube Steam Generators Heat Trarecer Sanilar to Chapt.15 Trans.

Vol 4 Once throuah ham Canaragg3 Meat Transfer Sannar to Chapt 15 Trans Vol 4 Bo'tirt irlaRChQn System ummg Sensevny Stuees Utlhty Submittels Transport Sensevity Stueen Utdrty submrttats J

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Important Phenomena (Draft)(Cont'd)

BWR ATWS { fun and Pernal)

Asessement Documente#en Com ' Power Amet Redsteuton (Fue)

Peach SotomTT Test Vol. 4 -add 10 lun 3 D Redeweuton (Perhal) -

Rod Drop Vol. 4 add PSI Doppler Feedback Rod Drop Vol 4.edd PSI Voed Feedteck Rod Drop Vol. 4. add PSI Baron Concentreton Feeu4 Sonnevtly Shados UWty Subnetels Heat Treneser sammer to Chapt 15 Trans.

Vol 4 Decay Heat ANSISid Exongiles Vol 4 Void Genershon FRIGG Tests Vol 4 Channel Flow Disttouton CodeeCode Ulbty SubmlReis Sysses/Leekage Flow Codee Code utsty Submmuus Weser Lewi Noen0Mel.Open.Sen States Uebly Submaines NeeselCirculeton -

Sep. Eflects / Sen. 3luess Vol. 4 / UWey SubmNiels unser -

Weser Levet NoengMe Opsi.Sen.Sauees Ulsty Submntus Notaal circulaton Sep. E8ects i Sen. Seuses Vol 4 iUWry Submetes councemer Weser Level NoengMe. Open. Sen. Sautes UWty Submittels Natural Crcunhon Sep. Eflects / Sen, Stuees Vol 4 (UWry Submdbois Enron innocean Svetem n==ng S.nsmay Saa es UWir submes.

Transport Sensavity Stuees UWty Submsness Retsf Velves chosung Seperses Eflecs &

Vol 4 Chant is Trans.

mCSR 1

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Important Phenomena i

(Draft) (Cont'd)

Assesernent DocumenteWon Shuldeurn Operseen M

Preuuru Drop and heet Vol.4 Low Pressure

' Nonconcensab6e gas gassent transfer eseessments Vol 4 i

GGE8 ANSI Sid. Esempies Vol 4 Decay heat PreO4F Heat Transeer Vol 4 Heat Transfer Vced Genershon FRIGG Voed Comportsons Vol4 NoengMas. Opn Sen. Steens Utsty Subnettels Weeer Level Natural Circuleton Gep. Effects i Sen. Studes Vol 4 iUWIr/ Submusst Unner PlenumfStancInsDes NodenCMe. Opn Sen. Stuees utsty Submntets Woest Leven Nenarsi ceculaton Sep EnoctsiSen.Studes Vol 4 IUWry Subminst i

i Nodmewd.Open.Sen Stuess UtiNty Submittels Water Levet Notarel Caculomon Sep. f sects 1 Sen. Stuees Vol 4 iUumty Submnists l

8 *'

Steam Genereinra Noncond ordy heet Wenefer Vol.4 Heat Trenefer wth Noncondensatm Condensaten wlNoncond Vol 4 ReAus Condensabon mCSR

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