ML20212K310

From kanterella
Jump to navigation Jump to search
Summary of 860708 Meeting W/Util in Bethesda,Md Re Topical Reload Program for Plant.List of Attendees,Proposed Meeting Agenda & Slide Presentation Encl
ML20212K310
Person / Time
Site: Peach Bottom  Constellation icon.png
Issue date: 08/13/1986
From: Gears G
Office of Nuclear Reactor Regulation
To:
Office of Nuclear Reactor Regulation
References
NUDOCS 8608200434
Download: ML20212K310 (95)
v • d • e


Text

~

-i August 13, 1986

{[

\

-Docket Nos. 50-277

nd 50-278

' LICENSEE: Phila &1phia Electric Company FACILITY: Peach Bottom Atomic Power Station, Units 2 and 3

SUBJECT:

JULY 8, 1986 MEETING WITH THE PHILADELPHIA ELECTRIC COMPANY (PEco)

TO DISCUSS TOPICAL RELOAD PROGRAM FOR PEACH BOTTOM, UNITS 2 AND 3 On July 8,1986, a meeting was held at the NRC headquarters in Bethesda, Maryland to. discuss the Philadelphia Electric Company's (the licensee or PECo)

Topical Reload Licensing Program. Attachment 1 is the list of individuals

~that attended the meeting. Attachment 2 is the proposed meeting agenda offered by the licensee. Attachment 3 is a hand-out of the slides presented by the licensee at the meeting. The following is a sumary of the significant items discussed and the actions, if any, taken or proposed.

1.0 Presentation of Licensee's Reload Program for Peach Bottom The licensee's presentation is attached as Attachment 3. PECo is in the process of developing reload licensing capability in-house. PECo's management is committed to this in-house approach for the following reasons:

1. Better plant specific analysis via in-house reloads vs. vendor approach
2. Provides a good understanding to the plant by doing in-house reloads
3. Better control of operating margins.
4. More cost effective in the long run due to multiple units.

An open dialog was maintained during the licensee's presentation with the NRR staff focusing in on the licensee's proposed application and verification of computer codes as part of the proposed reload licensing program.

Discussions also included what type of data /informatian would be needed by the NRR staff in the licensee's proposed submittals in order for the staff to provide an expedited review as requested by the licer.see.

2.0 Schedules for Method Report Submittals

- After the licensee's general' presentation on the content of the proposed submittals, identified-as Method Reports by PECo, discussions on schedules followed. The licensee's-proposed schedule for submittal of its Method Reports to the NRC is contained in Attachment 3. The NRR staff indicated that it would try to accommodate these schedules but this would require timely assistance from the licensee as well as high quality submittals. The NRR

~

8608200434 860813 PDR ADOCK 05000277 p PDR A

staff and licensee agree that a status / schedule meeting should be held shortly after receipt by NRR of the final Methods Report (3/87) in order to discuss the status of the NRR review and to provide the licensee with sufficient time to develop alternatives for the Unit 3 reload, if required. This meeting would most likely be held in the Bethesda, MD area during the month of June (1987). Additional technical meetings will be required upon the staff's review of the licensee's Methods Report.

h Gerald E. Gears, Project Manager BWR Project Directorate #2 Division of BWR Licensing

Enclosures:

1. List of Attendees for Meeting
2. The Licensee's Proposed Meeting Agenda
3. The Licensee's Slide presentation cc w/ enclosures:

See next page Distribution Docket file NRC PDR L PDR Memo file PD#2 G. Gears D. Muller E. Jordan 0GC-Bethesda ACRS 10 B. Grimes U. Cheh H. Richings D. Latze C. Graves L. Phillips V. DeMasi N. McCoy

.hd

  1. 2 DNU)FbA2 DB SNorr Ris GGears-
i. 8/11/86 8/[l/86 l

L- J

Mr. E. G. Bauer, Jr. Peach Botton Atomic Power Station, Philadelphia Electric Company Units 2 and 3 cc: Mr. R. A. Heiss, Coordinator Mr. Eugene J. Bradley Pennsylvania State Clearinghouse Assistant General Counsel Governor's Office of State Planning Philadelphia Electric Company and Development 2301 Market Street P.O. Box 1323 Philadelphia, Pennsylvania 19101 Harrisburg, Pennsylvania 17120 Troy B. Conner, Jr., Esq. Mr. Thomas M. Gerusky, Director 1747 Pennsylvania Avenue, N.lf.

Washington, D.C. 20006 Bureau of Radiation Protection Pennsylvania Department of Thomas A. Demino, Esq.

Environmental Resources Assistant Attorney General P.O. Box 2063 Harrisburg, Pennsylvania 17120 Department of Natural Resources -

Annapolis, Maryland 21401 Mr. Albert R. Steel, Chairman Board of Supervisors Mr. R. Fleischmann, II, Manager Peach Bottom Township Peach Bottom Atomic Power Station R. D. #1 RD #1 Delta, Pennsylvania 17314 Delta, Pennsylvania 17314 Mr. G. M. Leitch, Superintendent Nuclear Generation Division S7-1 Philadelphia Electric Company 2301 Market Street Philadelphia, Pennsylvania 19101 Mr. Anthony J. Pietrofitta, General Manager Power Production Engineering Atlantic Electric P. O. Box 1500 1199 Black Horse Pike Pleasantville, New Jersey 08232 Resident Inspector U.S. Nuclear Regulatory Commission Peach Bottom Atomic Power Station P.O. Box 399 Delta, Pennsylvania 17314 Regional Administrator, Region I U.S. Nuclear Regulatory Commission 631 Park Avenue King of Prussia, Pennsylvania 19406 .

W - , - - - - . - - - , - - - ., - ~ . - - . - - - . . - - , - - , .

n,

~ ATTACHMENT 1 List of Meeting Attendees July 8, 1986 Names Organization:

A. Olson Philadelphia Electric L.' Rubino Philadelphia Electric W. Alden Philadelphia Electric W. Lee Philadelphia Electric C. Cowan Philadelphia Electric L. Hemler Philadelphia Electric S.-Hesse Philadelphia Electric P. Wen NRC/ REGION 1 U. Cheh NRC/ DBL /RSB H. Richings NRC/ DBL /RSB D. Katze NRC/ DBL /RSB C. Graves- NRC/ DBL /RSB L. Phillips NRC/ DBL /RSB G. Gears NRC/ DBL /PD#2 V. DeMasi NUCOMP/PECO CONSULTANT M.-McCoy NRC/ DBL /RSB

g ,p-. . x - - s - <

h

$~

ATTACHMENT 2 PHILADELPHIA PROPOSED MEETING AGENDA 1

a i

I I

4 4

4 l

TRANSMITTAL OF PECO RELOAD LICENSING INFORMATION The Fuel Management Section of the Philadelphia Electric Company (PECo) is in the process of developing reload licensing capability. PECo plans to develop six topical reports in 1986/1987 and to perform initial 'in-house' reload analyses in support of a Peach Bottom 3 Cycle 8 startup in September of 1988.

Attached for your information are brief descriptions of six topical repor ts along with estimated submittal dates for NRC review. The titles of the topicals along with submittal dates are as follows:

Available for ,

Topical Report Submittal to NRC Steady State Thermal Hydraulics 2nd Quarter 1986 Thermal Margins 2nd Quarter 1986 Physics 4th Quarter 1986 Transient Analysis 4th Quarter 1986 Fuel Performance 4th Quarter 1986 Reload Safety Evaluation 1st Quarter 1987 We ask that these topical report descriptions be reviewed for the July 8th NRC/PECO meeting so t! sat an overall acceptable d

schedule of our reload licensing plan can be established.

Attachment 1 is an agenda of topics to be discussed at the requested meeting.

Attachment 2 is a brief description of the Philadelphia Electric Company Quality Assurance Plan for reload licensing.

Figure 1 shows a layout of computer codes use*6 in the PECo licensing process. The use of these codes is described in the attached topical repor t summaries.

f O

~

PHYSICS TOPICAL REPORT PECO STEADY STATE REACTOR PHYSICS METHODS FOR RELOAD DESIGN AND LICENSING APPLICATIONS ,

Philadelphia Electric Company is currently developing a Physics Method Report to be submitted to the NRC. PECo steady state reactor physics methods can be classified according to application as:

. 1. Reactor Operations Support.

2. Re~ load Core Design
3. Reload Safety Evaluation The Fuel Management Section/ Electric Production Department has provided reactor operations and fuel management support services for Peach Bottom Units 2 and 3 since 1977. These services include development of control rod patterns, predictions of criticality and cycle length, as well as monitoring and verification of Technical Specifications parameters such as plant thermal margins and shutdown margins. We are applying this operations support experience and data for purposes of physics methods qualification, to be used as the foundation of our physics topical report. From this experience base, we are also developing procedures to design reload cores and to perform the associated reload safety evaluations.

The following is a list of codes used to generate the physics topical report. With the exception of RWRA.SY and PESIGMA, these codes are extensively used in the nuclear i

industry.

Page 2 of 15

MICBURN - Davalops cross caction input to CASMO for gedolinia fuel rods.

CASMO - Single assembly fuel assembly lattice physics model.

Develops few group cross sections for SIMULATE (3-D nodal code) and PDQ (multi-assembly lattice code) .

NORGE - Processes CASMO output for input into SIMULATE.

COPHIN - Processes CASMO pin cell cross sections for input into

. PDQ-7.

PDQ-7 - Multi-assembly lattice physics code. Develops local pin peaking factor methodology to b'e used in SIMULATE.

SIMULATE - 3D steady state reactor physics model. Performs nuclear calculation with thermal and hydraulic feedback iterations to predict core criticality, power and flow distributions, thermal margins, and preconditioning.

FIGUR - 3D stes:1y state hydraulics code, embedded in SIMULATE.

Calculates 2D in-channel flow and core pressure drop distributions, as well'as core bypass flow.

RWEASY - SIMULATE output post processor used to perform rod withdrawal error analysis.

PESIGMA - Compares plant measured and SIMULATC predicted transverse incore probe (TIP) data to quantify the statistical differences between the model predictions and plant measurements.

Page 3 of 15

_ . - . - _ . _ . __m., ,, . _ _ . _ _ _ _ . _ , _ , . _ _ _ - , . _ _ - _ _ _ . _ , . . _ ,. _

The qualification of physics models for reload design and safety evaluation is a stepwise progression from our operations support experience and our use of vendor, consultant, and EPRI computer code analysis systems. Part 1 of the physics Methods report describes the benchmarking of the models for prediction of steady state thermal margins, power distributions, and hot and cold criticality. Part 2 will include statistical analysis to determine uncertainties associated with model predictions. Part 3 will describe physics methods employed in the calculation of Reload Safety Evaluation (RSE) parameters and input data to the transient analysis models. Appendices of t$e physics topical will describe specific metho~ds employed for the analysis of the Standby Liquid Control System (SLCS) Shutdown Capability, Rod Withdrawal Error (RWE), Mislocated Dundle Loading Error (M3LE),

and Rotated Bundle Loading Error (RBLE).

]

The target date for submittal to the NRC is December, 1986.

d i

Page 4 of 15

FR(YSRCS 9I6PRM MJDlWi>R$ D Rod Withdrawal Error (RWE) Analysis The RWE event is an abnormal operational transient that examines the consequences of a sudden insertion of positive reactivity to a localized area of the core. The reactor and core are assumed to be operating at rated power / flow conditions in accordance with normal operating limits. In addition, the core is conservatively assumed to be in a xenon free state before the withdrawal of the error control rod. The transient scenario is an operator-initiated event .whereby the operator inadvertently selects (procedural error) the highest

+

worth control rod and proceeds to withdraw 'it at maximum withdrawal speed from the fully ins,erte'd to the fully withdrawn position. The e thermal limiting core locations are conservatively modeled close to the highest worth rod.

The RWE analysis is performed with the 3-D model reactor simulator code SIMULATE. SIMULATE calculates MCPR and LHGR values at

each state point of the analysis and also calculates LPRM readings used to predict the Rod Block Monitor (RBM) system response.

The RWEASY code was developed to simplify the data reduction process associated with the RWE analysis. RWEASY utilizes LPRM data to calculate RBM 'A' and RBM 'B' responses as a function of error rod notch position and failed local power range string (s) . A thermal limit summary table (MCPR, Delta-MCPR, and MLHGR values) is also l

generated.

Demonstration trial runs will be performed and. the results will be compared to the licensing basis for previous cycles.

Page 5 of 15

~

PHYSICS TOPICAL REPORT APPENDIX B MISLOCATED BUNDLE LOADING ERROR (MBLE)

The mislocated bundle loading error (MBLE) is an abnormal event analyzed with the SIMULATE 3-D nodal code. This event analysis is essentially a comparative analysis focusing on a cycle depletion utilizing the design loading pattern and the same pattern with a single,mislocated bundle event. The delta-CPR is found by subtracting the CPR corr,esponding to the mislocated assembly from the design pattern CPR at the same location. A procedure will be applied to a bounding number of possible core locations where the MBLE limiting delta-CPR is determined.

Demonstration trial runs will be performed and the results will be compared to the licensing basis for previous cycles.

Page 6 of 15

~ ~

~

PHYSICS TOPICAL APPENDIX C Rotated Bundle Loading Error (RBLE) Analysis e

The RBLE is a localized abnormal operational transient which results from the loading of a single fuel assembly in an improper orientation. It is assumed that the loading error is not detected, and the plant operates on thermal limits with the design core configuration throughout the operating cycle, assuming'all other fuel

- assemb' lies are correctly loaded.

Because of the enrichment distribution with D-lattice fuel assemblies (Peach Botton Units 2 and 3), the 180 degree rotation which results in higher enrichment fuel rods adjacent to the wide water gaps is the most limiting case. Because of the uniform water gaps and 'the symmetry of the fuel rods in the fuel assembly, rotation of a C-lattice fuel assembly results in a small change in thermal margins and is not generally considered a limiting event.

The rotated fuel assembly is analyzed using CASMO (the two-dimensional fuel assembly burnup program). The three-dimensional l

l nodal reactor simulator code, SIMULATE, calculates CPR and LHGR values at desired exposure points throughout the cycle. The analysis will be evaluated at beginning of' cycle (BOC), at end of cycle (EOC) and at peak cycle reactivity exposure conditions.

Demonstration trial runs will be performed and the results will be compared to the licensing basis for previous cycles.

Page 7 of 15

PHYSICS TOPICAL - APPENDIX D STANDBY LIQUID CONTROL SYSTEM (SLCS) SHUTDOWN CAPABILITY ANALYSIS Three separate analyses will be described in the SLCS Topical Repo r t.

The first analysis consists of deriving the Standby Liquid Control System (SLCS) Shutdown Margin (SDM) Design Criteria for the Philadelphia Electric Company (PECo) SIMULATE Model.

This entails per forming a k-effective vs. temperature sensitivity study utilizing CASMO. CASMO will be run for all fuel types from 68 degrees Fahrenheit to approximat,ely '265 degrees Fahrenheit and the CASMO results compared to develop corresponding uncertainty factors (in %

delta-K) . This temperature uncertainty will be combined with the uncertainty in the NORGE computer code, the SIMULATE Physics Model and the Non-Haling Exposure Depletion. This combined uncertainty (%

delta-K) will be used as the Design Criteria for SLCS.

The second analysis utilizes the PECo SIMULATE computer code to verify that a subcritical condition exists with All Rods Out (ARO),

and 660 PPM of Boron at the most reactive point in the cycle and the Design Criteria mentioned previously. This analysis will be accomplished by executing SIMULATE at (approximately) every 1000 MWD /t exposure step in the reference cycles (PB2 Cycle 7, PB3 Cycle 7) to determine the minimum SDM.

Page 8 of 15

The third analysis utilizes the PECo SIMULATE-E computer code to determine the Boron concentration necessary to satisfy adequate SDM.

This is accomplished by holding K-effective constant (or SDM) and allowing SIMULATE-E to search for the minimum Boron concentration necessary to satisfy the SLCS SDM Design Criteria.

The SLCS-SDM analyses will be per formed for both Peach Bottom 2 and 3 reference cycles and the results will be compared to the current licensing basis.

t Page 9 of 15

STEADY STATE THERMAL HYDRAULICS TOPICAL REPORT A topical report has been completed describing the method for calculating reactor core steady state thermal hydraulics.

Philadelphia Electric Company used the FIBWR stand-alone computer code to perform this analysis. This stand-alone FIBWR computer code is logically identical to the steady state hydraulic models and sub programs used in the 3-D steady state physics code, S IMULATE. To qualify the model developed, actual plant data were

. compar'ed to corresponding FIBWR predictions at various operating statepoints. The parameters used as the basis of these comparisons were the core support plate pressure drop and the core bypass flow. The power shape input to FIBWR in this work i

was generated in a consistent manner using the SIMULATE 3-D core simulator code. A linkage code titled SIMFIB (Figure 1) was written to read the SIMULATE restart file and create a FIBWR overlay deck of the power shape.

Results from this study have been compiled and included in the steady state thermal hydraulics topical report to be i

submitted to the NRC.

( ..

Page 10 of 15

THERMAL MARGINS TOPICAL REPORT A topical repor t has been completed describing the method for calculating transient Critical Power Ratios (CPR's).

Philadelphia Electric Company uses the TCPPECO computer code for calculating CPR's. TCPPECO is a backend code which uses a restart file created by the transient analysis code RETRAN (run in a hot channel mode using boundary conditions from the RETRAN core-wide transient analyses) . To qualify the method, compar'isons were made to GE ATLAS Test Loop data. A RETRAN model of the ATLAS Test Loop was developed and run to generate restart files for TCPPECO CPR calculations. Comparisons were made between the observed time to boiling transition and the predicted time to boiling transition (CPR of 1.0). Sensitivity studies were performed by varying the modeling options of the RETRAN ATLAS model.

i l

. Page 11 of 15

TRANSIENT ANALYSIS TOPICAL REPORT Philadelphia Electric Company is currently in the process of developing methods for reload transient analysis of the, Peach Bottom Atomic Power Station. Qualification of these methods is to be documented in a topical report to be submitted to the NRC for review in 1986 and includes analyses of plant startup tests, the Peach Bottom Unit 2 Cycle 5 S/RV tests, the Peach Bottom Unit 2 Cycle 2 Turbine Trip tests, and the NRC test problem. The

'- topica'l report will also include a . description of the Peach Bottom RETRA'N model.

l The Peach Bottom RETRAN model is in its second revision with model improvements incorporated as experience with the RETRAN code and its applications has been gained. All RETRAN model calculations have been documented in a model calculation document (MCD), checked and independently reviewed. RETRAN code verification and validation has been performed and documented by EPRI and an SER has been issued by the NRC.

The PECo RETRAN model has been used to provide plant operations support including plant start-up analysis, loss of of f-site power analysis, ATWS analysis, MSIV pressure setpoint reduction analysis, and extended load line limit analysis.

Page 12 of 15

~ L1 1_ . _ . _ _ . _ _ __ .11 _ . _ __ _ _ _ _ _. . _ _ _ . _ _ _ _ ___ f . __ _ _ _ _ _ _ _ _ _ . _ _ __

RELOAD SAFETY EVALUATION (RSE)

The Reload Safety Evaluation (RSE) topical report will define and qualify the overall methodology to be used ig the RSE process. It will describe computer codes and technical procedures used to standardize and control the methods used in j the generation of RSE parameters related to both technical I specifications and reactor kinetics.

i The RSE technical specification parameters are Minimum -

Critical Power Ratio (MCPR), Linear Heat Generation Rate (LHGR),

l Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) , and 1

Shutdown Margin (SDM). ,

The' technical specification parameters are evaluated in the course of reload design and licensing process by both the steady state physics and transient analysis

- models. The results of these analyses demonstrate either that the present operating limits are sufficient to ensure the safe operation of the reload design or that a technical specification l

change for the reload design will be required. Steady state initial conditions for the transient evaluation of these parameters will be provided using the 3-D core physics model SIMULATE. The transient evaluation of technical specification ,

parameters and the determination of associated operating limits will be performed with the RETRAN and TCPECO computer codes.

The reactor kinetics parameters include void, Doppler and control rod scram reactivities, as well as the delayed neutron fraction and neutron lifetime. The reactor kinetics parameters are reload specific input to the RETRAN Lafety analysis model.

They determine the relative change in the plant conditions and Page 13 of 15

.--5 +- a ei*=e------ wmy -- ,----e r -e-i--e,-n-ye-r--------w w-&-- ----wr-+--w-- -------v3-m-----,.- ,,,-,,,.e=--,r------ . we= = - . . - - ew-~

l the consequent severity associated with the various transients. ,

The reactor kinetics parameters will be evaluated with the j S'IMULATE and SIMTRAN computer models.

Physics and transient model uncertainties will be considered in the overall evaluation of each of the transient events.

The following is a list of computer codes to be used for the  ;

Reload Safety Evaluation: ,

SIMULATE - dalculates steady state 3-D power distributions and peaking, as well as technical specification parameters: MCPR, LHGR, MAPHLGR, SDM. Also calculates reactor kinetics parameters for use in RETRAN.

RETRAN - Safety analysis and transient code used to determine NSSS system response to various anticipated operational transients (AOT's). Uses averaging techniques to model the reactor core and thermally induced nuclear feedbacks in either one dimensional

, (1-D) or point (0-D) kinetics formulations.

4 TCPPECo - Evaluates thermal margins and transient MCPR using results from a RETRAN hot channel model.

SIMTRAN - Collapses and averages 3-D nuclear data from SIMULATE for later input to the RETRAN-lD and -OD (point) transient models.

The target date for submittal to the NRC is the 1st quarter of 1987.

l i Page 14 of 15

FUEL PERFORMANCE METHODS REPORT The Fuel Performance Methods Report will be based primarily on the fuel performance code FROSSTEY (NRC Docket No. 50-271) developed by Yankee Atomic Electric Company (YAEC) and approved by the NRC in June, 1984. The FROSSTEY (Fuel Rod Steady State Thermal Effects) code was acquired from YAEC and placed on the PECo computer system during January of this year. Validation and verification of the code, culminating in its placement in the Reload Analysis Licensing production library, was completed by the beginning of April.

The FROSSTEY code will be used to determine gap conductance and fuel temperatures for input to the RETRAN safety analysis model. We do not intend to use FROSSTEY for licensing of fuel thermal limits, mechanical design limits or LOCA analysis. The code may be used for in-house audits of fuel vendor calculations, but all fuel related thermal or mechanical safety limits are provided by the fuel vendor.

We are currently in the process of selecting fuel rods for use in the methods report. We hope,to have completed work on the methods report by the end.of the year with an anticipated l submittal date to the NRC in the 4th quarter of 1986.

f I

l Page 15 of 15

ATTACHMENT 1 AGENDA FOR MEETING WITH NRC ,

TOPICS PECO PERSONNEL Overview - PECo Reload Licensing Program Lou Rubino 3.-

Reload Licensi'g Methodology Walt Lee Physics Topical , Steve Hesse Steady State Thermal Hydraulics Topical Andy Olson '

1 Transient Analysis Topical Andy Olson Thermal Margin Topical Andy Olson Fuel Performance Topical Walt Lee Reload Safety Evaluation Topical Walt Lee QA Program Chuck Cowan s

i l General Discussion All l ...

- ------w w- -e _w, -e--, - - - - - - - - - ------.,-r -

, + - --.-w---.,e-, ,- . - - - - - - - - - , - , - - - - - - -- --- m -- +--- ---+

ATTACHMENT 2

]

An impor tant part of the development of all topical repor ts is the Quality Assurance Program used to ensure procedure control over the development and implementation of computer codes and models. Reload-specific physics methods activities are controlled by Reload Analysis Licensing (RAL) technical and administrative procedures. These procedures are reviewed and approved by a RAL Review Committee bEfore issuance and implementation.

With regard to plant model development, the RAL procedures control the development of Model Calculation Documents (MCD's) for each reload-specific model. Each MCD has a review process i governed by RAL administrative procedures.

With regard to computer codes development, the RAL procedures control verification, validation and documentation as i well as ongoing control and maintenance. Computer code models are secured on a ' read only' library for Fuel Management Section use.

O.

I

PECO LICENSING CODE SEQUENCE PRI11ARY ANALYSIS COMPUTER CODES : .'  : LICENSING FUNCTIONS CODE I LINKAGE ,

CODE WICSURN (CAD CROSS-SECTION) l 1. CROSS SECTION i GENERATION

{ i 2. PEAKING FACTORS O

ICOPHIN &

I S. DETECTOR PARAMETERS (LATTI P YSICS)

Y 8

PDO-7 NORGE l

l 1

1 I

8

. I f I f 8

STEADY STATE SHUFFLE ytAWACARDj SIWULATE Q IPAVEj l 1. RELOAD DESIGN PHY3 LINK

  • 3-0 SIWULATOR 1 2. CORE DESIGN
1r p 3. FUEL PE-SIGWA PERFORMANCE l
  • RWEASY (STATISTICS) 4. RELOAD SAFETY (ROD WITHDRAWAL
  • s 8 EVALUATION

+

l ERROR) l PARAMETERS l TOPS l l 5. STARTUP TEST 8

ANALYSIS i

l 6. WODEL STATISTICAL 3 , J I r l QUALIFICATION IWFIS SIWTRAN

+ (FUEL PERFORWANCE) i i

i i

I I I 1 r

~

FIBWR -

l (THERWL- HGEIGH - I HYDRAULICS) - 8 l 8 I

I i f U 8

, TRANSIENT ANALYSIS l

RETRAN 8 HCORECALCl 8 (TRANSIENT ANALYSIS) 8

% TUR8EHC l ,

g i 1. RELDAD SAFETY 1 . . . ANALYSIS

INITIAL l 8

2. TRANSIENT

, , l THERMAL MARGIN 8

'T I TCP P

= ,,yO,T

,t , p TuE, m .ECO.IN) ,

WGL/ JPW 4/24/86 fico86*1 . .. - -

ATTACMENT 3: THE LICENSEE *S SLIDE PRESENTATION

4 PECo OVERVEW 3

PECo Reoad Licensing rogram O

L F. RUBINO Engineer-in-Charge Fuel Management Section Philadelphia Electric Co.

s ATTACHMENT 1 4

AGENDA FOR MEETING WITH NRC i TOPICS PECO PERSONNEL Overview - PECo Reload Licensing Program Lou Rubinc l

Reload Licensing Methodology Walt Lee Physics Methods Steve Hesse Steady State Thermal Hydraulics Methods Andy Olson Transient Analysis Methods Andy Olson Thermal Margin Methods Andy Olson Fuel Performance Methods Walt Lee Reload Safety Evaluation Methods Walt Lee QA Program Chuck Cowan General Discussion All l

l

PURPOSE OF THIS MEETING e INTRODUCE TO NRC FECO'S FLANS /

SCHEDlJLES TO FERFORM RELOAD LICENSING ANALYSIS FOR PEACH BOTTOM UNITS e PROVIDE DESCRIPTION OF RELOAD LICENSING METHODOLOGY l e OBTAIN NRC COMMENTS ON METHODOLOGY e OBTAIN NRC CONCURRENCE FOR PERFORMING REVIEW OF THE PROPOSED METHODOLOGY l

0 OBTAIN NRC CONCURRENCE ON PROPOSED LICENSING REVIEW SCHEDlJLE l

W SCODE e ViA4AGEMEN- COMM~~viENT e ORGAN ZK iOh e PECO EXPERENCE

~

e 3ROPOSED STRATEGY e ME- -iOJS REPORTS e 3RO30SEJ _CENSbG SC-EJU_E l

\

MAN!AGEVEN-~ COtv v'- vEN-~

e Parr Specific Analysis e Provices good uncers ancing of t,e a ant e Timey tumaround o

- siRC cuestions

- Acministrative concerns e Better contro on opercring l

margins l

l r

e Cos e'fective in the onc_

i run cue o multioe uni"s i

l

I l

ELECTRIC PRODUCTION DEPARTMENT ,

i i

l FUEL MANAGEMENT SECTION

( s l L. F. RUBINO i

l 1

l FUEL SAFETY & TECHNOLOGY CORE DESIGN & OPERATIONS l

B g l

W. G. LEE H. J. DIAMOND ,

l l l l l I I I I METHODS & SAFETY NUCLEAR FUEL CORE CORE ANALYSIS MATERIALS LICENSING DESIGN OPERATIONS e

S.R. HESSE C.E. COWAN A.M. OLSON

7

(~ 1 l

l ilu i -

1 l1 : <aaa

4. 4 e

e i

i l

I g*,!

sisilli 4.J d.44 l

,,e w- , , , - - - - , - - - , . - , , - - , , - - , - , - - , , , - , . - - - - , - , , _ - - , , . , ~ , , . , - - , - -

l

. 1 PECO EXPERIENCE i

  • CORE OPERATIONS SUPPORT l '

- Extensive reactor operations support  ;

using CASMO/ SIMULATE-E

  • for Peach Bottom Units 2 & 5 since 1977  !

l

  • for Limerick Unit 1 since 1985 i

I

- Participation in EPRI ARMP package for past 10 years e TRANSIENT ANALYSIS SUPPORT

- Participation in RETRAN development

since 1978

- Plant specific analysis using RETRAN:

l

  • MSiV Set Pt. Change
  • PB2 Cycle 5 SRV Tests
  • Extended Load Une Limit Analysis l

PROPOSED STRATEGf FOR RELOAD LICENSING ANALYSIS

\<

F e Preposed Methodology will be submitted to l

' the NRC for review and approval as Methods Reports

?

e Analyses to be performed by PEC3 as part of re!oad licensing submittal:

  • Core Design

5 Operation Limit (CPR's) e Analyses to be cerformed by vendor (GE)

  • LOCA Analysis e Special Analyses

_ - Stability

- Rod Drop Accident

  • Cycle specific CRDA onclysis has been discontinued for BPWS plants based on the fact that in all cases peck enthcipy is much less than 280 caVgm. PB 2 & 3 plan to implement BPWS (NEDE-24011 L

Ammendment #9 item 9E) l

METHODS REPORTS

  • Reports currently being developed by PECo:

i PROPOSED SCHEDULE FOR

METHODS REPORT SUBMITTAL TO NRC Stecdy Sicle T-H August 1986 Tnermal Margin . August 1986 4

Fuel Performance 4th qtr 1986 Phyoles 1st qtr 1987 Transient Analysis 1st qtr 1987 Reload Safeb / Evoluotion _- 1st qtr 1987 e , - , - - . , - - -

~

PROPOSED SCHEDULE .

NRC Approval \

l PECo Submittal y \

k RELOADk i

SUBMITTAL l PB3C8 l

PB3C8 RELOAD DESIGN l

! l & LICENSING PECo '

I Methods Report 1 Submittal NRC Approvals fr y -

f BEGINS ENDS e.................................e.yc.............................

NRC REVIEW OF METHODS REPORTS l l

PB3C8 STARTUP

- as m v k l4 I

JJASONDJFMAMJJASONDJFM MJ[AS, 1986 1987 1988

RECo Reload _icensing viemocoogy WALTER G. LEE Senior Engineer Philadelphia Electric Co.

l

AREAS TO BE AJJRESSEJ l

e _f- CE P-YS CS e ST ADY S- A t. Ab ALYSS e S I - AJY S X E ~~O - RANSIN-AN A_YS S _bKAGE e ~~RANS N AN A_YSS e VENJOR slTERFACE

4 l

l I

t LATTICE PHYSICS 8 Gadolinia Micro Cross Sections MICBURN

  • Cross Section Generation CASMO/NORGE

- Hot Cross Sections

- Cold Cross Sections o Peaking Factors CASMO/PDO e Defector Parameters CASMO g

l

STEADY STATE ANALYSIS

  • Core Power Distributions SIMULATE-E
  • Exposure Distributions SIMULATE-E
  • Rod Worths SIMULATE-E G Reactivity Coefficients SIMULATE-E
  • Core Thermal Hydraulics FIBWR e Fuel Performance FROSSTEY e Core Design SIMULATE-E/PE-SHUF e RSE Parameters SIMULATE-E e Model Stailsticci Qualification SIMULATE-E/PE-SIGMA l

O STEADY STATE TO TRANSIENT ANALYSIS UNKAGE e 3-Dimensionci ==> 1-Dimensional Cross-Section Generofion SIMTRAN-E e 3-Dimensionci ==> Point Kinetics

- Reactivity data (SIMULATE)

- Kinetics data (SIMTRANE) l l

^- - - - - - , , . . _ _ , , , _

E TRANSIENT ANALYSIS AND DETERMINATION OF OPERATING UMITS (CPR'S) e Operationd Transients RETRAN-02 8 Transien! Therrnal Margins RETRAN-02/TCPPECO

t VENDOR TEFACE DECO Owes Vencor:(Mkl e Reload Schedules e Vateria Recuiremen s e Numoer/Tyoe Fue Assemolies e _.oading Da "e m o Exoec"ec Opera"iona Conciiions l

l l

l

i VEN JOR N ERFACE (corr.)

- Vendor Owes U-ihy:[ht orue Assemoy Jesign Data e _icensaae Uec,anical Jesign e T,erma ycrauic Parame"ers

- GEX_ Corre a ion

-R " actors i e3rocess Comou"er Cons"an"s l

l

l 3ECo Reacor 3hysics Vethods Report SIE.VEN R. HESSE Engineer Philadelphia Electric Co.

PHILADELPHIA ELECTRIC COMPANY (PECo)

REACTOR PHYSICS METHODS REPORT

Purpose:

Define and qualify the methodology for performing steady-state physics -

calculations in support of reload core design and licensing analysis Scope:

- PECo Experience

- Reload Design / Licensing Process (Physics)

- Description of the Proposed Methodology

- Qualification of the Proposed Methodology l

l l

.-. n . _ -

PECO REACTOR PHYSICS EXPERIENCE I

i 1

e Operations Support (1976- ) l l

  • Rod Pattems
  • Hot /Cald Criticals
  • Power Distribution
  • Anomoiy Curves
  • Thermal Margins
  • PCIOMR Restrictions
  • Operating Strategies
  • Cycle Management Report Verification
  • Process Computer Data Bank Verification e Fuel Management (1977- )
  • Review of Vendor Fuel Designs
  • Evaluate Refueling Strategies l
  • Energy Utilization Plans
  • Determhation of fuel Requirements
  • Fuel Economics Evaluation e Reload Design l/ icensing Verification (1981- )

l

~"

.? ~ ~ _ _ . - - _

PHYSICS RELOAD DESIGN /LICENS!NG PROCESS

- Batch Size / Enrichment Requinements

- Core Design ,

- Ha!ing Analysis

- Shutdown Margin Andysis 1

- Hot Excess Reactivity Analysis

- Rcdded Cycle Depletion

- Thermai Limits Evaiuotion

- Evduction of Recc4Mty Anomoly Events

  • Rod Withdrawl Error
  • Mislocoted Bundle Loading Error
  • Rotated Bundle Loading Error

- Generation o' Reactivity input for Sofety Analysis Calculations

- Compilation of Data for Reload Licensing Submittals

, PHYSICS METHODOLOGY PRIMARY COMPUTER PROGRAMS e MICBURN:

Gadclinia pin microscopic bum-up program.

MiCBURN generates effective cross section l data for fuel pins containing homogeneously distributed bumable absorbers.

r e CASMO:

2-D multigroup transport theory fuel assembly burn-up program. CASMO is used to generate cross section data for the 3-D l nodal power distribution program SIMULATE.

O PDQ-7:

Few group diffusion theory program. PDQ is used to develop local peaking factor methodology for the SIMULATE program.

1 l

l i

Ip PHYSICS METHODOLOGY PRIMARY COMPUTER PROGRAMS (cont.)

e SIMULATE:

Steady state 3-D core physics simulator.

SIMULATE ca!culates steady state power /

flow distnbutions as well as tech. spec.

parameters and reactivity coe'ficients.

e PESIGMA/ TOPS: Or-W Power distributien statisticci analysis I

programs. PESIGMA and TOPS are used to determine the uncertainty associated with SIMULATE's 3-D power distribution cciculation.

t

.~......-

QUALIFICATION OF PHYSICS METHODS a i i

1 3 e Hot Critical K-Effective Comparision 3

- Benchmark to 5 Peach Bottom Cycles

  • Start-Up
  • Steady-State

- Development of Trends, Bias Uncertainties e Cold Critied K-Effective Comparisions

- Benchmark to 5. Peach Bottom Cycles

  • Moderator Temperature Effects (68 F-220 F)
  • Period Correction

- Development of Trends, Bias, Uncertainties e Power Distribution Comparisions

- Benchmark to 5 Peach Boticm Cyc!ss

  • Full Power

- Development of Trends, Bias, Uncertainties 1

l

1 QUALIFICATION OF PHYSICS METHODS (cont.)

I

)

e Thermd Umits Comparision (To Process Computer Vdues) i

- Benchmark to 5 Peach Bottom Cycles

- Development of Trends, Sios, Uncertainties e Reactivity Parameters

- Void Reactivity

- Doppler Reactivity

- Scram Reactivity

- Delayed Neutron Parameters

- Consideration of Calculational Uncertainties

QUAUF! CATION OF PHYSICS METHODS (cont.)

i e Rod Withdrawal Error (Appendix B)

- Comparison to Ucensing Easis for y Peach Bottom 3 Cycle 7

)

  • Peak Reactivity -
  • Umifing Rod Pattem/Umiting Rod i;)
  • No XenorVFull Power

' - SensitMty on Power. Exposune s Mislocated Bund le Loading Error (Appendix C)

- Comparison to Ucensing Basis for Peach Bottom 3 Cycle 7

- Sensitivity on Bundle Type, Selection Criteria e Rotated Bundle Loading Error (Appendix 0)

- Comparison to Ucensing Basis for Peach Bottom 3 Cycle 7

  • Fresh Fuel

- Sensitivity on Bundle Type e Standby Uquid Control Analysis (Appendix E)

- Comparison to Ucensing Basis for Peach Bottom 3 Cycle 7

  • All Rods Out
  • 68 F

- Considemtion of Cciculational Uncertainties

i t

[ PECO PHYSICS METHODS REPORT 5

PROPOSED TABLE OF CONTENTS 1.0 Introduction 2.0 Description of PECo Calculational Methods and Computer Code Sequence 3.0 PECo Experience in Model Applications 3.1 Prediction of Hot & Cold Reactor Criticals 3.2 Verification of Operating Thermal Margins

PECO PHYSICS METHODS REPORT PROPOSED TABLE OF CONTENTS (cont.)

4.0 Steady State Physics Model Qualification 4.1 Nodal and Assembly Power Distribe'lon Comparisons 4.2 Sub-Assembly Local Pin Power Distributlan Comparisons 4.3 Doppler and Void ReactMty Bonchmarking 4,4 Control Rod Worth Benchmarking 4.5 Delayed Neutron Parameters and Effective Neutron Lifetime 4.6 isotopics

PECO PHYSICS METHODS REPORT PROF 0 SED TABLE OF CONTENTS (cont.)

5.0 Model Applications to Reload Design and Safety Evaluation Calculations 5.1 Maximum Average Planor Linear Heat Generation Rote (MAPLHGR) 5.2 Peak Pin Linear Heat Generation Rate (LHGR) 5.3 Min! mum Crit!cd Power Ratio (MCPR) 5.4 Shutdown Margin 5.5 Doppler Reactivity vs. Fue! Temperature .

5.6 Void Reactivity vs. Moderator Density 5.7 Scram Curve Reactivity vs. Time 5.8 Effective Delayed Neutron Fraction. BETA-EFF 5.9 Prompt Neutron Genemtion Time, L 5.10 Analysis of FSAR Reactivity Anomoly Events 5.10.1 Rod Withdrawl Error (RWE) 5.10.2 Mislocated Bundle Loading Ermr (MBLE) 5.10.3 Rotated 5.10.4 Standby LiquidBundle Control System Loading (SLCSError (RBLE))

Analysis

PECO PHYSICS METHODS REPORT PROPOSED. TABLE OF CONTENTS (cont.)

6.0 References APPENDIX A: A Discussion of Statistical Methods Employed for Model Qualification APPENDIX 8: A Discussion of the Rod Withdrawal Error Analysis APPENDIX C: A Discussion of the Mislocated Bundle Loading Error Analysis APPENDIX D: A Discussion of the Rotated Bundle Loading Error Analysis AFPENDlX E: A Discussion of Standby Liquid Control System Analysis I

- . - _ ..,m -- --...<,-..,-m. ,-y.e-_, _ - . _ - - , . . . _ _ _ . _ , . - _ . . . _ _ - - _ - . - - - - - . . - _ - - . - - - , - -

3 Co

~~

3WR S eacy-S ate

,erma ycrauic Anaysis ve"nocs Recor-ANDY M. OLSON Engineer Philadelphia Electric Co.

l l

l PHILADELPHIA ELECTRIC COMPANY (PECo)

BWR STIADY-STATE THERMAL HYDRAUUC ANALYSIS METHODS REPORT

Purpose:

Define and qualify a methodology for steady-state thermal hydraulic analysis for the determination of core de'ta P and core bypass flow required to pedorm core reload design and licensing analysis Scope:

- Description of the proposed methodology

- Qualification of the proposed methodoiogy l

DESCRiFTION STEADY STATE THERMAL HYDRAUUC ANALYSIS METHODS e FIBWR

- FiBWR is a computer code developed for steady state thermal-hydraulic enclysis of BWR's

- FIBWR wcs developed by Yankee Atomic Electric Co. for EPRI i - Tne NRC has issued an SER for application j of FIBWR by Vermont Yankee 1

e SIMEFlBR

- SIMEFIBR is a linkage code between SIMULATE-E ond FIBWR used to provide the 3-D power distribution

QLA_ r CC ON COVPAR SON ~~O DLANF JATA e

1. Peac, Bot om Uni s 2 & 3 Core Suoport Pate Jifferen-ia Pressure (osic)
2. Peac, 3ct om Uni s 2 & 3 Core Byoass ow (W om/hr)

h I

I

._ _b 1

4 1XCe en" CiCreernent wrh measured ca"a.

d i

e i ~

I i

BWR STEADY-STATE THERMAL HYDRAUUC ANALYSIS METHODS REPORT TABLE OF CONTENTS 1.0 Introduction 1.1 Purpose 1.2 Description of FIBWR 2.0 FIBWR Comparisons to Peach BoMom Data 2.1 Plant Specific FlBWR Model 2.2 Comparison of the FIBWR Model Predictions Against Measured Plant Data 3.0 Summary and Conclusions 4.0 References l

l 1

TCo BWR Sys ems -ransient Anaysis Vemocs Recor" ANDY M. OLSON l

Engineer Philadelphia Electric Co.

l l

i

I I

PHILADELPHIA ELECTRIC COMPANY (PECo)

BWR SYSTEMS TRANSIENT ANALYSIS METHODS REPORT

Purpose:

Define and qualify the PECo BWR transient analysis methods required to perform reload design and licensing analysis Scope:

- Description of the proposed methodology

- Qualification of the proposed methodology l

1 l

DESCRIPTION PECO TRANSIENT ANALYSIS METHODS e RETRAN-02

- RETRAN-02 is a computer code developed for thermd-hydraulic analysis of comp!sx fluid flow systems

- Constituitve mode!s to be used in performing transient ondysis ine:Udes:

  • One-dimensional space-time kinetics
  • Non-equilibrium model for steam-water interface
  • Algebraic slip model
  • Sub-cooled woid mode!
  • Implicit ilme step control

- An SER has been issued for RETRAN-02 by the HRC

- RETRAN-02 is a widely used and Industry accepted loci for performing ircnsient thermd-hydrau'ic andysis .

e Auxillary Codes

- CORECALC

- INITIAL

- TURBEHC

- These are interface codes thof facilliate '

the generation of cycle specific inputs

QUAFCATIOb Comparison o measurec oarr co a

- P.B. Unit 2 Cycle 2 Turbine Trip es~

3 B. Uni" 2 Cyc e 5 Sve y/Reief Va ve ~~es"s

- Plant Start-uo Data

2. NRC Standard Test Problem
3. Comparison to Vendor Calculations

- Anaysis o' imi"ing ransient in 3AR even" ccragories (in RSE ve"iocs Reoor)

L. E RAb-02 has been ex ensivey voica"ed anc veri'lec*oy PR anc u~ii~y users group' L a

y -

..m_______.. . . .. .. .

RESlJLTS Preliminary results have indicated-excellent agreement with the measured data. Further qualification and validation is under way p  :- : -. _

BWR SYSTEMS TRANSIENT ANALYSIS METHODS REPORT TABLE OF CONTENTS (lPREUMINARi')

1.0 :ntroduction 1.1 Purpose 1.2 Model Ouclification 1.3 Model Application 2.0 Model Descripticn 2.11ntroduction 2.2 Model Geometry 2.2.1 Steam Unes 2.2.2 Feedwcter Lines 2.2.3 Recetor Vessel 2.2..i Reactor Recirculation Loops 2.2.5 Core Region 2.3 Componeni Models 2.3.1 Safety / Relief Valves 2.3.2 Stecm Separctors 2.3.3 Recin:uletion Pumps 2.3.4 Jet Pumps 2.3.5 Core Hydrau!ics 2.4 Trip Logic 2.5 Contrcl Logic 2.5.1 Sensed Parameters and Miscellcneous Cdculations 2.5.2 Reactor Water Level Cdeu!ations 2.5.3 Feedwater Control System 2.5.4 Recirculation Control System i

2.5.5 Turbine Electro-Hydraulic Control System

r t

TABLE OF CONTENTS (cont.)

3.0 Qualification 3.1 Peach Bottom Start-lJp Tests 3.1.1 Feedwater System Transients 3.1.2 TJrbine Electro-Hydraulic Control Transients 3.1.3 Reactor Recirculation Transients 3.2 Peach Bottom Unit 2 Cycle 5 Safety / Relief Valve Test 3.3 Peach Bottom Unit 2 Cycle 2 Turbine Trip Tests 3.4 NRC Standard Test Problem l

i ,

i o I {

i DECO i BWR -ransient J l Critical Power Rcrio Analysis j i Vemods Recor"

\

f l .

ANDY M. OLSON Engineer Philadelphia Bectric Co.

e i

I l

1 1

FHILADELPHiA ELECTRIC COMPANY (PECo)

BWR TRANSIENT CRITICAL POWER PATIO ANALYSIS METHODS REPORT l

Purpose:

Define and qualify a methodology for the cdculation of transient critical power ratios for the determination of operating limits (CPR's)

Scope; 4

- Description of the proposed methodology

- Qualification of the proposed methoddogy 4

DESCRIPTION PECo Transient CPR Methodology -

  • A RETRAN-02 system level calculations is performed e Using the boundary conditions (cressure, normalized power) from the system level calculation and limiting values for bundle power, flow, and He-gap conductivity, a RETRAN-02 ' hot-channel' limiting bundle analysis is performed ,

e Using the boundary conditions (bundle flows and enthalpies) from the ' hot-channel' calculation, TCPPECO calculates the fransient CPR's using the GEXL correlation and the bundle R-factor c

s

l BOUNDARY CONDITIONS FROW RETRAN SYSTEM LEVEL WODEL RESTART FILE ,

.J l

CHANNEL FLOW "

SINGLE  : RETRAN CHANNEL HOT CHANNEL FIBWR  : WODEL WODEL ,

H u i

RETRAN ESTIMATED RE-EDIT FOR CHANNEL -

BOUNDARY POWER CONDITIONS n

hY TCPPECO .

o NO CPR=1.0 YES o

TRANSIENT A CPR WETHODOLOGY FOR CALCULATING TRANSIENT ACPR FIGURE 33 PAGE 80

s ,.

! l l

1 QUAUFICATION COMPARISON TO GE-ATLAS LOOP FACILITY DATA i'

1. 50 Steady-State Data Points ,

2.13 Constant Power, Flow Decay Tests .

3.12 Power and Flow Decay Tests

4. 2 Power and Flow Increase Tests l

RESULTS i

Excellent agreement with the above GE experimental data t

s l

l l

l l

M EWR TRANSIENT CRITICAL POWER RATIO ANALYSIS l

METHODS REPORT TABLE OF CONTENTS 1.0 Introduction -

1.1 Purpose 1.2 Brief Description 2.0 Description 2.1 RETRAN Model 2.2 TCPPECO Methodology 3.0 Qualification 3.1 Steady-State Comparisons 3.2 Transient Comparisons 3.3 Sensitivity Studies 3.4 Resu!ts 4.0 Method for Determining MCPR's 5.0 Application -

5.1 RETRAN Hot Channel Model for Pecch Bottom

5.2 Results 5.2 Range of Applicability 6.0 Summary and Conclusions 7.0 References I

i I

3ECo .

ruel Perbrmance Anaysis Me-hocs Reoort e WALTER G. LEE Senior Engineer Philadelphia Electric Co.

PECo Fuel Performance Analysis Methods Topical Report Purocse:

Define and qualify ihe methodology for determining fuel temperatures and '

Gap conductance required for performing reload licensing anclysis.

Scope:

  • Description of the proposed rnethodology e Qualification and Applicofion

[

.cf the proposed methodology

\

i

DESCRIPTION e FROSSTEY The FROSSit.Y package is specifica!b/

designed to provide fuel rod temperatune distributions, fuel-to-cladding gap conductance, etc., as a function of fuel rod operating history.

FROSSTEY has been obtained under license fmm Yankee Atomic Electric l

Company.

I I

t

1 i

i i

QUALIFICATION l

ED

! APPLICATION e NRC issued an SER on app!icaiion of FROSSTEY to Vermont Yankee based on extensive quellflection by Ycnkee Atomic Electric Company e Currently, PECo is preparing the methodology to utilize the FROSSTEY code in performing Peach Boliom raiood calculations

  • The qudification of FROSSTEY will be demonstreted by comparing FROSSTEY predictions to test data in the

~

following areas:

FlSSION GAS RELEASE FUEL TEMPERATURES Additionally, Sensitivity Studies will be performed on various power shapes to observe changes in delta CPR.

4 RESULTS The application of FROSSTEY wi!! be demonatrated by utilizing FROSSTEY to obicin core average l gcp conductance, hot channe! gcp conductances, and fuel tempemt!res for Peach Bottom Unit 3 Cyde 7.

in-house results are expected by Dec.,1986 l

. . . - . . _ . ~. .

TEST DATA

  • We have not yet se!ected those rods that ws will use for the Quclification section of the Methods Report. ,

e The test data from which we plan to select experimental data for the '

qualification of FROSSTEY was compiled by Combustion Engineering for EPRI e The rods for which we have fission gas data are from:

HALDEN TEST ASSEMBUES: ,

IFA-11.21,431, and 432 <

MONTICELLO HBEP PEACH BOTTOM 2 RISO RISO FISSION GAS PROJECT and STUDSVIK $150 4 The rods for which we have fuel temperature dato are from:

HALDEN TEST ASSEMBUES:

IFA-11,21,430,431,432, 513, and 527 l STUDSVIK S150 PBF GAP CONDUCTANCE TEST RODS l

i Table of contents (Preliminary)

1.0 INTRODUCTION

j 1.1 Purpose 1.2 Brief Description

2.0 DESCRIPTION

3.0 QUAUFICATION s 3.1 Test Data Comparison 3.2 Sensitivity Studies 4.0 APPLICATIONS 4.1 Gap Conductance i 4.2 Fuel Temperatures 5.0

SUMMARY

AND CONCLUSIONS

6.0 REFERENCES

l l

l

l f

i 3ECo i Reloac Sa=e y Evauation (RSE) vemocs Reaort 0

s f

WALTER G. LEE Senior Engineer Philadelphia Bectric Co.

/

i l

PHILADELPHIA ELECTRIC COMPANY (PECc)

RELOAD SAFETY EVALUATION (RSE)

METHODS REPORT

?.

f i

Purpose:

1. Define and qualify the proposed methodology for performing 13 reload licensing analysis
2. Establish the licensing bases for Peach Bottom units

~

Scope:

1. PECo RSE Coordination l

l 2. RSE Parameters l

3. Qualification and application of

( the RSE methodology l

l

, . . . - , _ _ - _ = - _ _ _ - - . - - - . - - . . . . . - . . . - -

y u

PECo RSE Coordination SAFETY $AFETY l  !

TASK ANALYSIS EVALUATION j  ;

TRANSIENT ANALYSIS (ANTICIPATED OPERATIONAL TRANSIENTS) j increase in Reactor Pressure Events PECo PECo j Decrease in' Core Coolant Temperature Events PECo PECo l Reactivity / Power Distribution Anornoly Events PECo PECo

! Decrease in Reactor Coolant inventory Events PECo PECo Decrease in Reactor Coolant Flow Rate Events PECo PECo increase in Reactor Coolant Flow Events PECo PECo i l ACCIDENT ANALYSIS l

Control Rod Drop Accident ** - -

l  ;

Loss of Coolant Accident GE PECo i

j SPECIAL ANALYSIS Shutdown Margin PECo PECo l Stability * - -

l l

l

  • Analysis will not be needed O I M4w4% "h
    • Not needed pending inatallati{on of BPWS [at Peach Bottom 1

r ]

. 1 RSE PARAMETERS 7_

  • {
  • TECHNICAL SPECIFICATICN PARAMETERS

- Minimum Critical Power Ratio (MCPR)

- Linear Heat Generation Rate (LHGR)

- Maximum Average Plonar LHGR (MAPLHGR) *

- Shutdown Margin (SDM) e TRANSIEh'T/klNETICS PARAMEILHS b

P'

- Doppler Cross Sections & Reactivity

- Void Cross Sections & Reactivity

- Control Rod (Scrom) Cross Sections

& Reactivity

- Delayed Neutron Fraction

- - Neutron Velocities

- Neutron Generation Time d

e W

. - .- . - - - - - - _ - - - - - . - - - ~ - . - . . - -

I

\

l I

l QUAUFICATION AND APPUCATION

, OF RSE METHODS

^

e INTEGRATED QUAUFICATION

- Plant Test Data '

  • Peach Bottom 2 Start Up Tests d
  • Peach Bottom 2 Cycle 5' SRV Uft Tests

- FSAR A.O.T. Analyses J

  • Peach Bottom 3 Cycle 7 Refenance Cycle
  • Parametric Studies ~& Uncertainty Analyses For Umiting Transients e SIMTRAN QUAUFICATION

( '

- Verify SIMULATE (3-D) to RETRAN (1-D and 0-D) Normalizations

  • Initial Power Distribution
  • Reactivity Driving Functions i

,- 3 3 ; _ 7 ---;-------- _;

PLANT REFERENCE CYCLE ANALYSES

- Increase in Reactor Pressure

  • Generator Load Rejection w/o Bypass

- Decrease in Core Coolant Temperature

- ReactMty/ Power Distribution Anomolles

/

  • Rotated Bundle Error
  • Mislocated Bundle Error

- Decrease ir, Reactor Coolant inventory

- Decrease in Reactor Coolant Flow Rate

  • Recinculation Pump Seizure

, - Increase in Reactor Coolant Flow Rate

  • Recirculation Flow Controller Failure e PARAMETRIC STUDIES PERFORMED TO:

-identify sensitive parameters which influence the limiting transients.

- Consider Uncertainties

  • Steady State Model

- ~

D W - -- - . . . m..=a..g

r u

PECO RSE METHODS REPORT TABLE OF CONTENTS (Freliminary) 1.0 Introduction 2.0 Description of PECo Methods and Computer Models Employed for Reload Safety Evaluation.

3.0 Description of PECo Methods for Calculation of Reload Safety Andysis Input and RSE Parameters.

3.1 Technical Specification Parameters 4

3.1.1 Minimum Critical Power Ratio (MCPR) 3.12 Maximum Average Planar Linear Heat Genenation Rate (MAPLHGR) 3.1.3 Peak Pin Linear Heat Generation Rate (LHGR) 3.1.4 Shutdown Margin (SDM) -

3.2 Transient and Kinetics Parameters 3.2.1 Scrum Reactivity l 3.2.2 Void Reactivity 3.2.3 Doppler Reactivity 3.2.4 Delayed Neutron Froction 3.2.5 Effective Neutron Lifetime 3.2.6 Neutron Velocities a

3 I

T

=i I

h l

I TABLE Oi CONTENTS (cont.)

4.0 Qudificction of PECo RSE Methodology 4.1 Steady-State Physics to Tmnsient Model Linkage: 3-D to 1-D Normalization Studies.

4.2 Test Qualification of Integrated Steady State /

Transient Analysis Models. D 4.2.1 Peach Bottom 2 End of Cycle 2 Turbine Trip Tests: Measurements vs. Model Predictions 4.2.2 Peach Bottom 2 Beginning of Cycle 5 Safety Relief Valve (SRV) Lift Test Andysis:

Measurements vs. Model Pmdictions 4.3 Analysis of Peach Bottom PSAR Translents 4 4.3.1 Parametric Studies and Uncertainty Andysis 4.3.2 Peach Bottom 3 Cycle 7 Analyses t - < ., e l *

=a m. m. . . = * = .eae~---e . - - . __w w ,g g , ,,, , m . . , , , ,

7 TABLE OF CONTENTS (cont.)

5.0 Reload Sofety Evoluotion Procedures for FSAR Anticipated Operational Transients.

5.1 !ncrease in Reactor Pressure Events 5.1.1 Generator Load Rejection Without Eypass 5.2 Decrease in Core Coolant Temperature Events 5.2.1 Loss of Feedwater Heating 5.2.2 Feedwater controller failure (Max. Demend) 5.3 ReactMty and Power Distribution Anomaly Events 5.3.1 Control Rod Withdrawl Error - Power Range Operation 5.3.2 Fuel Loading Eror - Rotated 5.3.3 Fuel Loading Error - Mlsiccated 5.4 Deereose in Reccier Coolant inventory Events -

5.4.1 Loss of Feedwoier Flow l 5.5 Decrease in Reactor Coolont Flow Rate Events 5.5.1 Recirculation Pump Seizure 5.6 increase in Recefor Coolant Flow Rate Events 5.6.1 Recircu'ution Flow Controller Failura 5.7 ASME Vesse! Overpressure Protection Events i 5.7.1 MSIVC With Position Switch Foilure l

(I.e., H!gh Flux Sensm)

I

- - ~ - - - - - . - - - _. _= - . .

y-1 l

TABLE OF CONTENTS (cont.)

6.0 Accident and Special Analyses l 6.1 Shutdown Man 2n i 6.2 LOCA (Vendor Will Perform S.A.)

6.3 Control Rod Drop 6.4 Stability 7.0 References

( .

l

r - _

o

i I

l PECo QUAUTY ASSURANCE Program implementation Reload Ucensing Program I

i x l CHARLES S. COWAN Engineer Ph!!cde!phie Electric Cc.

b

- ---__%_m 7 M w ygpe D 9-@ # 'W * * *'

(

PURPOSE:

To implement the Pecch Ecttom OA program for performing in-house reload Ik:ensing onelysis.

SCOPE:

e Relood Ana!ysis and Ucensing Committee (RALC) l

  • Computer Code Committee (CCC) l
r.
  • e _ e_ ** v_e fa p w e 9 ~~^nma=^-- -m e m -w + - = = * * - , e *

-^^'"^S.* .F .

i i

i i

RA OAD ANALYS!S AND LICENSING COMMii ILE I .

I i (RALC) l i

I i i

I Fesocrsioilities inc!ude:

t e Review and approva: of technicci and administrative procecures  !

developed for pedorming reloac licensing analysis e Reviea and approvd of Methods Reports (MR's) e Review and approvo! of Model Co!culation Documents (MCD's) i s

** ~ "~ TR*2_* M ___c_'__~_l__ 2 :

~ - - -

4 r

o

, i l l

! COMPUTER CODE COMMil It I

l (CCC)

- Responsible for review and approval i

of all the computer codes used for performing reload licensing calculations. This includes:

  • Code Modifications .

i  !

e Code improvements and enhancements Computer Codes approved by CCC for use in Reload Licensing Analysis are stored in a controlled software library in the PECo corporate computer.

%

Retrieved from "https://kanterella.com/w/index.php?title=ML20212K310&oldid=3098603"