ML17180A828

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Summary of 940505 Meeting W/Util Re Util Plan,Approach & Schedule for BWR Safety Analysis Vendor Independence Program.List of Attendees & Viewgraphs Encl
ML17180A828
Person / Time
Site: Dresden, Byron, Braidwood, Quad Cities, Zion, LaSalle  
Issue date: 07/12/1994
From: Dick G
Office of Nuclear Reactor Regulation
To:
Office of Nuclear Reactor Regulation
References
NUDOCS 9407150116
Download: ML17180A828 (76)
v • d • e


Text

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THEY HAVE BEEN CHARGED TO YOU. *

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-NOTICE-

ENCLOSURE 2.

Commonwea_lth Edison Company BWR_Safety Analysis Vendor lndeperid.ence Program NRC/CECo Meeting May 5, 1994 Terry Rieck John Freeman Bob Tsai Hossein Youssefnia Nuclear Fuel Services Commonwealth Edison Company

BWR SAFETY ANALYSIS VENDOR INDEPENDENCE PROG.RAM NRC/CECo Mee.ting.- 5/5/94 AGENDA

1. Introduction.

Review Agenda, Meeting Objectives

2. Update on Submittal Schedule Transient, Core Thermal Limit & Reload Application Topicals.
3. Methodology Overview Approach, Computer Codes Used, Planned Topical Reports &

Review of 9/23/93 Meeting

4. Training and Quality Assurance Internal & External Trainings, Quality Assurance Program *
5. Transient Analysis Topical Update Preliminary Results on Core Thermal Hydraulics and System transients
6. Summary CECo Summary
7. NRC Discussions NRC Feedback on CECo BWR SA VIP BNRCB.DOC

BW~ SAFETY ANALYSIS VENDOR -INDEPENDENCE PROGRAM

- NRC/CEca* Meeting - 5/5/94

- (Introduction)

MEETING OBJECTIVES To update NRC on CECo's Plan, approach

& schedule for the CECo BWR Safety Analysis Vendor Independence Program (VIP).:

To obtain NRC feedback on CECo plan for BWR SA VIP.

BNRCC.doc

/

I BWR SAFETY ANALYSIS VENDOR INDEPENDENCE PROGRAM NRC/CECo Meeting - 5/5/94 UPDATE OF SUBMITTAL SCHEDULE

  • Methods, Models & Quad Cities, Dresden & Peach Bottom Benchmarking
  • LaSalle Results Supplement
  • NRC/CECo Meeting
  • NRC Audit, If Necessary
  • Core Thermal Limit & Reload Application Topical
  • Topical Submittal
  • NRC/CECo Meeting
  • NRC Audit, If Necessary
  • Start of D3C 16 Analyses
  • D3Cl 6 Startup BNRCG.DOC

. 12/94 2Q/95 2Q/95 3Q/95 12/95 06/96 3Q/96 4Q/96 06/97 06/97 06/98

I BWR SAFETY.ANALYSIS VENDOR INDEPENDENCE PROGRAM

. NRC/CECo Meeting - 5/5/94

.*.. METHODO-LOGY OVERVIEW.

  • In-House Neutronics Analysis: Vendor (SPC) Nuclear Design Methods
  • CASMO/MICROBURN
  • NRC Approved CECo Use in 03/93
  • Safety Analysis: CECo/EPRI Methods BNRCDA.doc
  • PETRA/RETRAN/FIBWR2
  • Other Utility Topicals Approved by NRC TVA,YAEC,GPUN,PECo,PP&L (GSU, WPSS)

I BWR SAFETY ANALYSIS VENDOR INDEPENDENCE PRO~RAM

. NRC/CECo Meeting - 5/5/94 (Methodology Overview)

METHODOLOGY OVERVIEW FUEL DATA PLANT DATA MICROBURN 1 r 1 r

, r

, r.

ESCO RE PETRA FIBWR2S.S.

TRANSIENT INPUIS CHANGE DECK CORE AVERAGE Hgap HOT CHANNEL Hgap

, r, r

, r, r RETRAN

  • ONE DIMENSIONAL HYDRAULICS
  • ONE DIMENSIONAL KINETICS PLANT MODEL
  • CONSIDERS SYSTEM AS A WHOLE

, r CORE BOUNDARY CONDITIONS

-_. CORE MODEL FIBWR2

  • MUL Tl-CHANNEL HYDRAULICS
  • HOT CHANNEL HYDRAULICS ACPR UNADJUSTED FUEL VENDOR STATISTICAL ANALYSIS &

SAFETY LIMIT _.

TREATMENT FOR UNCERTAINTIES u

OPERA TING LIMIT MCPR BSANRCS2.DOC

I BWR SAFETY ANALYSIS VENDOR INDEPENDENCE PROGRAM NRC/CECo Meeting - 5/5/94 9/23/93 NRC/CECo MEETING REVIEW

1. NRC/CECo Interaction on Submittal Plan & Schedule.
2. Treatment of Uncertainties
3. Computer Code Applicability (conditions in SERs)
4. Thermal Limit Methodology Benchmarking
5. Mixed Core Effects in the Application Topicals BNRCDC.DOC

BWR SAFETY ANALYSIS VENDOR INDEPENDENCE PROGRAM NRC/CECo Meeting -.5/5/94

. TRAINING & QUALITY ASSURANCE

  • CE'.Co Internal Training
  • Computer Code Applications
  • Reactor Systems
  • Nuclear Power Plant Operation &

Simulator

  • External Training
  • Design Participation Training Programs (fuel vendor assignments)
  • Fuel Vendor Reload Safety Analysis Courses
  • Computer Code Vendor Workshops
  • Industry Workshops
  • Quality Assurance Program BNRCF.DOC
  • Computer Code Certification
  • Perform, Review & Approval of Controlled Analysis per QP's &

Department Procedures

BWR SAFETY ANALYSIS VENDOR INDEPENDENCE PROGRAM

. NRC/CECo* Meeting - 5/5/94 (Transient Analysis Topical Update)

  • Cote Thermal Hydraulics Mo'del Developm~nt & Benchmarking BNRCEl.DOC
  • FIBWR2 *summary
  • CEC*o FIBWR2 Applications

. ***Steady State Analysis Qualification *

  • CPR Correlation Implementation &

Benchmarking

  • Core Thermal Hydraulic Summary

History BWR SAFETY ANALYSIS VENDOR INDEPENDENCE PROGRAM NRC/CECo Meeting - 5/5/94_

FIBWR2 Summary

  • Development Began in 1 988 Sponsored by the FIBWR2 Owner's Group
  • Steady" State: Extended Capabilities of FIBWR Code.

Funded by Yankee Atomic Electric Company First Review by NRC early 1 980's (Part of YAEC's Submittal)

  • Transient: BWR Core Model Under Non-LOCA Conditions Some Important Features
  • Predicts Flow Distribution for a Given Power Distribution Uses Total Core Flow or Pressure Drop Condition
  • Axially Varying Flow Geometry
  • Fuel Rod Model
  • Linkage to RETRAN Code Allows Time Varying Axial Power Shape BNRCElA.DOC

I.

BWR SAFETY ANALYSl.S VENDOR INDEPENDENCE PROGRAM NRC/CECo Meeting - 5/5/94 CECO FIBWR2 Applications Steady State Core Thermal Hydraulics

  • Calculates Inputs *to Initialize RETRAN System Model Core Equivalent Nodal Pressure Drop Total Bypass Flow Channel Dependent Flow Transient Core Thermal Hydraulics
  • Single Hot Chann.el Model

~CPR Calculation Multi-Pin Heat Flux and LHGR Edits

  • Multi-Channel CPR Calculations Mixed Core Loading BNRCEl B.DOC

BWR SAFETY ANALYSIS VENDOR INDEPENDENCE PROGRAM

..

  • NRC/CECo Meeting - 5/5/94 Steady State Analysis Qualification
  • .Comparison to FIBWR Quad Cities Cycle l Core Model Dresden Cycle l Core Model
  • Comparison to GE's Steady State Thermal Hydraulic Code (ISCOR)

Quad Cities Cycle 1 Core Model Dresden Cycle 1 Core Model

  • Coding of GE Correlation (GEXL) in FIBWR2 MCPR Prediction vs. GE's Method (Quad Cities Cycle 1)

BNRCEl C.DOC

BWR SAFETY ANALYSIS VENDOR INDEPE;NDENCE PROGRAM NRC/CECo Meeting* - 5/5/94 *

  • FIBRW2 VS. FIBWR RESULTS

!Example 1

!Quad Cities Cycle 1: 100% pow~r/100% flow (Central Orificing)

  • FIBWR2 FIB WR DIFFERENCE Core dp (psi) 21.5602 21.6978
  • 0.1376 Core Plate dp (psi) 16.9917 17.1455

-0.1538 Bypass Fraction 0.1119 0.1105 0.0014 Average Void 0.3839 0.3848

-0.0009 HEATED CHANNEL Friction dp (psi) 2.7746 2.7492 0.0254 Elevation dp (psi) 2.494 2.4854 0.0086 Local dp (psi) 3.8868 3.9009

-0.0141 Acceleration dp (psi) 0.8187 0.8135 0.0052 Total dp (psi) 9.9741 9.949 0.0251 Non-Boiling Length (in.)

9.1428 8.9942 0.1486 Quad Cities Cycle 1 - Steady State (100%/100%)

0.7 0.6 0.5 c c

  • -a 0.4 a:l u..

'tl 0.3 c > 0.2 0.1 BNRCEl D.DOC FIBWR

- - - - -

  • FIBWR2 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 Node

BWR SAFETY ANALYSIS VENDOR iNDEPENDENCE PROGRAM NRC/CECo Meetin_g *- S/S/94 FIBRW2 VS. FIBWR RESULTS

~IE_x_am_p_le_2 _____ ____.IQuad Citie~ Cycle 1.: 75~/opower/75% flow (Central Orificing)

Core dp (psi)

Core Plate dp (psi)

Bypass Fraction Average Void HEATED CHANNEL Friction dp (psi)

Elevation dp (psi)

Local dp (psi)

Acceleration dp (psi)

Total dp (psi)

Non-Boiling Length (in.)

FIBWR2 13.7614 9.1718 0.1009 0.3797 1.7574 2.5146 2.2298 0.4717 6.9735 12.4664 FIB WR DIFFERENCE 13.7855

. -0.0241 9.2154

-0.0436 0.0933

. 0.0076 0.3816.

-0.0019 1.7465 0.0109 2.5055 0.0091 2.2515

-0.0217 0.47 0.0017 6.9735 0

12.5414

-0.075 Quad Cities Cycle 1 -Steady state (75%/75%)

0.7 0.6 z

0.5 0

b 0.4 cl LL 0.3 0

0 0.2 0.1 0

BNRCEl E.DOC FIBW R FIBW R 2 1 2 3

4 5

6 7

8 9 1 0 11 12 13 14 15 16 17 18 19 20 21 22 23 24 NODE

BWR SAFETY ANALYSIS VENDOR INDEPENDENCE PROGRAM

. NRC/CECo Meeting - 5/5/94 FIBRW2 VS. FIBWR RESULTS

,_E_x_a_m~le_3 _____

~* Quad Citi~s Cycle 1: 50%power/100% flow (Central _Orificing).

Core dp (psi)

Core Plate dp (psi)

Bypass Fraction.

Average Void.

HEATED CHANNEL Friction dp (psi)

Elevation dp (psi)

Local dp (psi)

Acceleration dp (psi)

Total dp (psi)

Non-Boiling Length (in.)

FIBWR2*

20.0475 15.4736 0.095 0.2621 2.1246 2.9614 2.9655 0.4431 8.4946 14.8592 FIB WR DIFFERENCE 20.1262

-0.0787 15.5669

.:.0.0933 0.095 0

0.2645.

-0.0024 2.1087 0.0159 2.9433 0.0181 2.9708

-0.0053 0.4386 0.0045 8.4614 0.0332 14.8177 0.0415 Quad Cities Cycle 1 - Steady state (50% /100%)

0.5 0.45 0.4 z

0.35 0

b 0.3 Cf 0.25 u...

0 0.2 0

0.15 0.1 0.05 0

BNRCEl F.DOC r

~

i f-I I

~

I I f-I I r I r I

~

i 1 2 3

FIBWR FIBW R 2 4

5 6

7 8 9 101112131415161718192021222324 NODE

J BW.R SAFETY ANALYSIS VENDOR _INDEPENDENCE PROGRAM NRC/CECo Meeting - 5/5/94.

FIBRW2 VS. FIBWR RESULTS

~IE_x_am~pl_e_l ____ ___.IDresden Cycle 1: 100% power/120% flow (Central Orificing)

Core dp (psi)

Core Plate dp (psi)

Bypass Fraction Average Void HEATED CHANNEL Friction dp (psi)

Elevation dp (psi)

Local dp (psi)

Acceleration dp (psi)

Total dp (psi)

Non.:.Boiling Length (in.)

!Example 2 Core dp (psi)

Core Plate dp (psi)

Bypass Fraction Average Void HEATED CHANNEL Friction dp (psi)

Elevation dp (psi)

Local dp (psi)

Acceleration dp (psi)

Total dp (psi)

Non-Boiling Length (in.)

BNRCEl G.DOC FIBWR2 FIB WR DIFFERENCE 28.3621 28.5676

-0.2055 23.802

-24.0342

-0.2322

. O.l 152 0.1153

-0.0001 0.3535 0.3542

-0.0007 3.4479 3.4163 0.0316 2.6062

  • 2.5958 0.0104 4.9909 4.998

-0.0071 0.9928 0.9856 0.0072 12.0378 11.9957 0.0421 8.7511 8.5863 0.1648

  • !Dresden Cycle 1: 100%power/75% flow (Central Orificing)

FIBWR2 FIB WR DIFFERENCE 14.3506 14.3077 0.0429 9.7676 9.7456 0.022 0.1014 0.101 0.0004 0.4325 0.4356

-0.0031 2.074 2.0408 0.0332 2.3188 2.2992 0.0196 2.5061 2.5564

-0.0503 0.6156 0.6089 0.0067 7.5145 7.5053 0.0092 10.0651 9.8424 0.2227

BWR SAFETY ANALYSIS VENDOR INDEPENDENCE PROGRAM NRC/CECo Meeting - 5/5/94 Exam.pie: FIBWR2 - /SCOR Com.parison for Quad-Cities Cycle 1 (Full Power & Flow Conditions) i Variable FIBWR2 ISCOR Difference Total Bypass Flow 10.78 11.21

-0.43 Plenum to Plenum Press.

21.53 21.64

-0.11 Total Exit Steam Flow 9.76 9.63 0.13 Core Avg Void Fraction 0.65 0.64 0.01 Pressure Drop Components: Central Orificing Variable FIBWR2*

ISCOR Difference Friction 3.18 3.32

-0.14 Elevation 3.07 2.70 0.37 Local losses 15.09 15.07 0.02 Acceleration 0.18 0.55

-0.37 Press.ure Drop Components: Peripheral Orificing Variable FIBWR2

!SCOR Difference Friction 1.36 1.33 0.03 Elevation 2.80 2.47 0.33 Local losses 17.50 17.60

-0.10 Acceleration 0.13 0.25

-0.12 Quad Cities Cycle 1-Steady State (100%/100%)

0.7....---------------------------

0.6 I

I C 0.5

  • ~ 0.4*

I I ~ 0.3

> 0.2 I

I 0.1

- - - - -

  • FIBWR2

ISCOR 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 Node BNRCEl H.DOC uriits Mlb/hr psi Mlb/hr Units psi psi psi psi Units psi psi psi psi

BWR SAFETY ANALYSIS VENDOR INDEPENDENCE PROGRAM

.* NRC/CECo Meeting - 5/5/94 Example: _Steady. State MCPR *calculation FIBWR2 vs. *1scoR (Quad Cities Cycle.1 Core)

Power/Flow 100/100.

50/100 50/50 100/120 100/50 75/50 25/75 100/100 50/100 50/50 100/120 100/50 75/50 25/75 100/100 50/100 50/50 100/120 100/50 75/50 25/75 BNRCEl I.DOC Average Channel (Central Orificing)

FIBWR2 ISCOR 1.92 1.92 3.77 3.77 3.13 3.12 1.97 1.97 1.60 1.60 2.11 2.11 6.99 6.99 Average Channel (Peripheral Orificing)

FIBWR2 ISCOR 2.24 2.24 4.43 4.41 3.46 3.43 2.36 2.37 1.82 1.81 2.37 2.36 7.82 7.75 Hot Channel (Central Orificing}

FIBWR2 ISCOR 1.32 1.31 2.64 2.62 2.15 2.13 1.37 1.36 1.06 1.05 1.42 1.41 4.90 4.88 Difference FIBWR2-ISCOR.

o.oo 0.00 0.01 0.00 0.00 0.00 0.00 Difference FIBWR2-ISCOR 0.00 0.02 0.03

-0.01 0.01 0.01 0.07 Difference FIBWR2-ISCOR 0.01 0.02 0.02 0.01 0.01 0.01 0.02

BWR SAFETY ANALYSIS VENDOR INDEPENDENCE PROGRAM NRC/CECo Meeting - 5/.5/94. -.

  • CPR Correlation Implementation & Be*nchma_rking.
  • Implementation of the SPC's Correlation (ANF-8)

Application within the Correlation Limits

  • Comparison of Calculated MCPR vs. Vendors' Method Steady State Transient
  • Benchmarking against Appropriate Test Data
  • Qualification of the Vendors' Correlations in a Mixed Core Loading BNRCElJ.DOC

BWR SAFETY ANALYSIS VENDOR INDEPENDENCE PROGRAM NRC/CECo Meet!ng - 5/5/94 Core Thermal Hydraulic -Summary

  • *CECo FIBWR2 Qualified for RETRAN Initialization Comparison to FIBWR Comparison to GE's Steady State Thermal Hydraulic Method
  • GE Correlation (GEXL) Coding Tested in FIBWR2 Comparison to GE's Steady State MCPR Predictions
  • CPR Correlation Implementation and Benchmarking Comparison to Vendors' Methods (Steady State &

Transient)

Benchmarking against Appropriate Test Data Qualification of a mixed Core Loading BNRCEl K.DOC

BWR 'SAFETY ANALYSIS VENDOR INDEPENDENCE PROGRAM

. NRC/CECo Meeting - 5/5/94 (T.rans.ient'Analysis Topical Update)

Transient Analysis Model Development &

Benchmarking *

  • Model Development Status
  • Benchmarking Results

-Quad-Cities Startup Test

-Peach Bottom Turbine Trip

  • Contractor Review of CECo's One Dimensional Neutronics Collapsing Methods

BNRC0-1 BWR SAFETY ANALYSIS VENDOR.INDEPENDENCE PROGRAM NRC/CECo Meeting - 5/5/94.

MODEL DEVELOPMENT s*TATUS QUAD-CITIES RETRAN MODEL:

  • Base Model with Cycle l C9re Completed
  • Cycle l Start Up Test Plant Data Collected
  • Cycle l Start *Up Test Benchmark Completed (sample results will be presented today)

DRESDEN RETRAN MODEL :

  • Base Model with Cycle l Core Completed
  • Cycle l Start Up Test Plant Data Collected
  • Cycle l Start Up Test Benchmark in Progress LASALLE RETRAN MODEL :
  • Base Model in Progress
  • Cycle l Start Up Test Plant Data Collection Nearly Complete MODEL ADJUSTMENTS:
  • Final modeling techniques from completed Peach Bottom Benchmark will be incorporated into the Quad-Cities, Dresden and LaSalle RETRAN models

I BWR SAFETY ANALYSIS VENDOR INDEPENDENCE PROGRAM NRC/CECo Meeting - 5/5/94 Quad-Cities.Prelim.inary Results List of Startup Test Initial Conditions.

Start Up Test Acceptance Criterion for used to Benchmark Quad-Ci.ties RETRAN Model Benchmark RETRAN Models Rating Initial

'Units PRSC RWLSC BPVC Parameter

(+)

(0)

(*)

Condition Steam Dome

<lOpsi

<20psi

>20psi Reactor

(%)

22.5 91.5 68.0 Pressure Power Downcomer Level

<Sin

<lOin

> lOin Core Flow (MLb/hr) 36.5 98.0 55.0 Steam Flow Rate

<5%

<10%"

>10%

Reactor (PSIG) 958.0 998.0 972.0 Feedwater Flow

<5%

<10%

>10%

Pressure Rate Feed Water (MLb/hr) 2.21 8.8 6.40 Recirculation Loop

<5%

<10%

>10%

Flow Flow Rate Reactor (ln.N.R.l 29.0 32.5 30.0 Core Flow Rate

<5%

<10%

>10%

Water Level Reactor Power

<3%

<6%

>6%

Steam Flow (MLb/hr) 2.64 8.8 0.00 Turbine Control,

<0.5%

<l.0%

>l.0%

Duration for (sec) 20 70 70 Valve Position Main Steam Flow

<10%

<20%

>20%

Benchmark Bypass Valve Position Instrument Accuracv Data Plant Accept.

Measured Full Scale Criteria Parameter Span Accuracy Reactor Power 0-125%

<3%

of (APRM/LPRMl Rated Core Flow 0-80

<5%

of Mlb/hr per Rated loop Dome 0*1200

+/-2%FS

<10 PSI Pressure PSIG Reactor Water 0-60 INWC

+/-2%FS

<5 Level, NR Inches Reactor Water

-42 to 358,

+/-2%FS

<5 Level, WR -340 to 60 Inches INWC Feed Water 0-6

<5%

of Flow Mlbm/hr Rated Main Steam 0-3Mlb/h

+/-8.5%FS

<5%

of Flow Rated Main Steam

+/-lOmV N/A Bypass Valve 0-5V i.e.+/-0.2%

Position Turbine

+/-lOmV N/A Control Valve 0-5V i.e.+/-0.2%

Position

  1. Indicates GE proprietary data omitted

~NKLU-L

BWR SAFETY ANALYSIS VENDOR INDEPENDENCE PROGRAM

. NRC/CECo.Meeting - S/S/94 Quad-Cities Prelimin*ary Results Pressure.Regulator Setpoint Change (Case 1)

Pressure Regula1or (*) 10 psi Step

!MIO llyp* sVlive ~D

~g

\\*,

\\.

&i.tUpC..*r M

RETRAH02

\\..

'\\

~!i f

  • ~*

M6 I

M2 llYPass' lii!veCla es MO ro u

u tt a

Time (5ec)

Reactor Dome Pressure for PRSC Pressure Regulator H 1 O psi Step 2

~

~ 2.T +-l'~-->---H--=-+-1----1-->---+--+---l--J c:

E

~

10 12 14 18 18

~

Ti*~(Hc)

Main Steam Flow for PRSC Reactor Water Change (Case 2)

Level.

Ji LO e

~.

~ 7.5

~

7 i

.% 1.5 r

.J Level Se1poin1 (+) 6 Inch Step

+

1 1--SU:rt Up Data!

RETRAN>2 50 Time Csac)

Feed Water Flow for RWLSC Level Setpoint (+) 6 Inch Step I-Start Up Daul RETRAN'.J2

.I~ -

JY.V: -.

.... ir

~.,-

21 21 50 Time (MC)

Reactor Water Level for RWLSC Setpoint

~. -.

71 u..

A

\\t-..

~--..

71 BNRCl -5

BWR SAFETY ANALYSIS VENDOR INDEPENDENCE PROGRAM NRC/CECo Meeting - 5/5/94...

Quad-Cities Preliminary Results 30 29.5 29 e

c 28.5 E

~

28

.... > 27.5 e

~

27 u.,

c i! 26.5 f--

26 25.5 25 975 974 973 972

~ 971 a

~

970 969 z

968 967 966 965 Bypass Valve Change (Case 3)

Bypass Va.Ive Open_ing at 70% Power I --start Up Data I I-RETRAN02 I Full Closed '

~

Bypass Val a Starts Op! 'Ing

~

Bypass Valv Starts Clos ng I

~

Ono\\ alvo Full Op*'

f I

~

,,_11:w.. -

~~

10 20 30 40 50 60 70 Time (sec)

Turbine Control Valve Position for BPVC Bypass Va.Ive Opening at 70% Power

~

)....

...... ~

\\

~

~,,

... ~

~

... "-*. ****Ir*... -........

"'==-,.

,-start Up Dalal RETRAN02 10 20 30 40 50 60 70 Time (sec)

I BWR SAFETY ANALYSIS VENDOR INDEPENDENCE PROGRAM NRC/CECo Meeting - 5/5/94 MO.DEL DEVELOPMEN-T STATUS PEACH BOTTOM RETRAN MODEL :

  • Base Model at Nominal Conditions Completed.
  • Benchmark in Progress

-Cycle 2 Startup Turbine Trip Test # 1

-Cycle 2 Startup Turbine Trip Test # 2*

-Cycle 2 Startup Turbine Trip Test# 3

-NRC Licensing Problem, Turbine Trip W/out Bypass (sample results will be presented today)

  • Final Adjustments to models pending
  • Sample Results Follow

45 4

35

.!:! ii E

0 3

z a::

a..

~5

.J..

[!..

ci:

~

0 1 5 u

05 0

0 70 0 600 500 c

iii

!!:. 40.0 ii:

~ " 30.0

~

a.

20.0 100 00 0 000 BWR SAFETY ANALYSIS VENDOR INDEPENDENCE PROGRAM

. NRC/CECo Meeting - 5/5/94 Peach Bottom Preliminary Results*

Peach Bottom Turbine Trip #2., c'ore Avg LPRM i--EPRI NP-564 Measured data1 1- - RETRANData 0.2 0.4 0.6 0.8 1.2 14 1.6 18 2

TIITlll(sec)

Peach Bottom Turbine Trip #2., Pressure Rise Reactor Dome

--PB TT2Measured1 RETRAN Data 0.500 1.000 1500 2.000 2.500 3.000 3.500 4.000 4.500 5.000 Time(secl

5 iii ii:

~

ii:

5

  • iii

"' ii:., :;

~

a.

70 0 60 0 50.0 400 BWR SAFETY ANALYSIS VENDOR l_NDEl'ENDENCE PROGRAM NRC/CECo Meeting - 5/5/94.

Peach Bottom Preliminary Results Peach Bottom Turbine Trip #2, Pressure Rise Core Exit I

300 ----..........-ttt-tt-i-...+--+--'---------'--PB TI2 Measured '-----,----1

  • RETRAN Data
o 0 100+----+--------~----------i----'-----------------t I

0.0......................... ______________________________________________________ ___.

0.000 0.500 120.0 1100 100.0 90.0 80.0 i

70.0 60.0 50.0 40 0 I

I I

30.0 20.0.

I I

I 10 0.

I I

0.0

-10 0

-20.0

-30.0 i

[

-40.0

-50.0 I

-&l.O 0.000 0.500 1.000 1.500 2.000 2.500 Till"lll (sec) 3.000 3.500 4.000 4.500 Peach Bottom Turbine Trip #2, Pressure Rise Turbine Inlet I

' I I

I

~ ~ v.

~

I I

6 l

I I I I

I I

I I

1.000 1.500 I

I i

i i

I i

I I

I I

rd\\"

I I

l I ~ MV ~ V..\\Jvv\\ \\rJV-.. "lV'!~

I I V1

~,

I I

i I

2.000 i

i I I

2.500 Tlme(sec)

I I

.. -,.v -

I

--PB TI2 Measured i

--RETRAN Data I '

I i

I I

I I

I I

i I

I I

I I

I I

I 3.000 3.500 4.000 4.500 5.000 5.000

BWR SAFETY ANALYSIS VENDOR INDEPENDENCE PROGRAM

. NRC/CECo Meeting - 5/5/94 ONE DIMENS.IONAL METHODOLOGY REVIEW.

  • CECo has initiated an Independent review of its 3-D to 1-D cross section collapsing methods (MICROBURN-->PETRA-->RETRAN)

BNRCl -4

-Contract with Computer Simulation Analysis (Developers of RETRAN)

-Utility Code MICPET Documents reviewed

-Utility Code WIDE Documents reviewed

-Methodology Calculation Note Review in Progress

-Sample Peach Bottom model Results Review in Progress

-Final comments from CSA are pending completion of CECo Peach Bottom Studies

BWR SAFETY ANALYSIS VENDOR INDEPENDENCE PROGRAM NRC/CECo Me~ting - 5/5/94

SUMMARY

  • Update on Submittal Schedule
  • Training & Quality Assurance Program
  • Update on Methods & Progress
  • NRC Feedback on CECo Plan & Schedule on Submittals BNRCH.doc

CECo Nuclear Fuel Services Department Overview NRR Presentation May 5, 1994

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CECO NUCLEAR FUEL SERVICE*s

  • VENDORINDEPENDENCE PROGRAM
  • ENGINEERING AND OPERATIONAL SUPPORT
  • NUCLEAR FUEL AND REACTOR TERRY RIECK:

KEN KOVAR KEVIN RAMSDEN ENGINEERING SUPPORT JACK COLTER 2:15PM BREAK

2:30PM

  • BWR SAFETY ANALYSIS PR.OGRAM UPDATE *
  • INTRODUCTION
  • SUBMITTAL SCHEDULE
  • METHODS OVERVIEW
  • TRAINING & QA
  • TOPICAL PROGRESS

SUMMARY

TERRY RIECK BOB TSAI HOSSEIN YOUSSEFNIA JOHN FREEMAN TERRY RIECK

NUCLEAR FUEL SERVICES OVERVIEW A. Vision B.

Key Expectations C. History D. Organization o:briefbk:overvwcv: I

NFS VISION

  • Professional Partner on NOD Team
  • Proactive Response
  • Excellence
  • Impeccable Nuclear Engineering
  • Stimulating Work Environment Centered Around Our People

KEY EXPECTATION~

VENDOR INDEPENDENT RELOADS Development and use of analytical methods for reactor neutronic, thermal hydraulic, and transient analyses in order to safely and efficiently design, license, and operate reload cores.

IN-HOUSE ENGINEERING AND OPERATIONAL SUPPORT Application of in-house analytical tools and e~pertise to support plant design changes, equipment problems, and other engineering and operational needs of the nuclear stations.

RESPONSIBLE DESIGN AUTHORITY FOR FUEL AND CORE COMPONENTS Implementing safe, economic, and reliable fuel and core component designs which meet changing station needs, improve product performance, and reduce product and fuel cycle costs by working with the Company's vendors and site engineers.

DIRECTION AND MONITORING OF ON-SITE FUEL ACTIVITIES Directing, monitoring, and assessing on-site fuel related activities, including reactivity management, fuel reliability, and core component performance.

o:briefbk:sec I a&b:keyexpct: I

1966 1967 1968 1970 1974 1975 1977

    • \\

1978 1980 1983 1988 1990

(

1989 KEY EVENTS IN DEVELOPMENT of NUCLEAR FUEL SERVICES at CECO

-Corporate Nuclear Fuel Committee formed

-Four Production Dept. people sent to Purdue Fu.el Management course

-Task Force on Nuclear Fuel Management Planning officially formed

-United Nuclear (UNC) computer programs ~btained

-Production Nuclear Reactor Analysis (PNRA) formed

-Approximately 1 0 people in PNRA

-PNRA name changed to Nuclear Fuel Servies (NFS)

  • Major Dresden-3 7x7 fuel failure event (departure from preconditioning rules)

-Major Quad Cities-2 7x7 fuel failure (departure from preconditioning rules)

-BWR Qualified Nuclear Engineer (QNE) Program Initiated by NFS

-Parallel of vendor fuel management on large BWR's using UNC codes

-Full scope Dresden-1 fuel management by NFS using UNC codes

-Completed development and implemented a computer based nuclear material accountability system called the Nuclear Fuel Data Bank

-Carroll County contract provides rights to use W neutronics methods for PWR's

-Exxon contract provides rights to use Exxon neutronics methods for BWR's

-Design Participation Program at W (3 engineers and supervisors for 1 year)

-Approximately 25 people in NFS

-Began development of safety analysis methods using EPRI codes

-Intensive effort to develop and implement POWERPLEX advanced core monitoring system for Dresden transition to Exxon fuel supply

-Intensive effort to implement W codes on CECO IBM and benchmark/validate for NRC Topical Report

-NRC approval obtained for NFS to use W neutronics methods

-INPO push for utility Reactivity Management Program

-PWR Qualified Nuclear Engineering Training Program begins

-Approximately 70 people in NFS

-NRC approval of Zion DNB limit for PWR In-house safety analysis methods o:briefbk:kevevnts: I

1992

-NRC approval obtained for NFS to use GE ne.utroriics methods for BWR's 1993

-NRC approval obtained for NFS to. use SPC neutronics methods for BWR's

  • 1994

-NFS *completes 40th in-house reload design

-NRC approval obtained for NFS to use EPRI transient analysis methods (RETRAN/VIPRE) for Zion o:brietblt:lteyevnts:2

Nfclear Engineering Technology Ser~es I

Electrical/

l&C Engineering Supervisor G. WaQner I

Mechanical Structural Engineering Supervisor M. Reed o:brlefbk:charts:organ4.p1111:1 I

PAA&

Reliability Engineering Supervisor F. Lentine Senior Vice President M. Wallace I

Vice President J. Hosmer I

Manager D. Shamblin I

II I

Engineering Performance

& lmprvmnt Manager E.Zebus l

Regulatory Assurance Supervisor S. Stimac Nuclear Store Supervisor W. Sheldon I

Nuclear Construction Supervisor L. f?etrie

  • Nuclear Fuel Services Department I

P. c. leBlond Chief Nuclear Engineer Direction of BWR/PWR Common Issues Reactivity Management Oversije Technical Direction of Reactor Engineers I

R.J. Chin Nuclear Design Supervisor Nuclear Fuel Management Neutronlc Reload Licensing Calculation Nuclear Core Operational Data t 6 Staff Members o:brfefbk:charta:nfafunc.DJS:1 I

KB.Ramsden Reactor Systems Engineer Transient Analysis Consultant Containment Performance I

H. S. Kim Computer Methods Development Supervisor NFS Sotwara System Developmert & Support Station Core Monitoring Sys. lmplemenfzdion &

Maintenance NFS Local Area Nel"Mlrl<N/orkstation Management t 1 Staff Members I

D. R. O' Boyle Core Materials Engineer Core Materials Corrosion Fuel and Core Component Performance In-Core Fuel I Component Tests K. N. Kovar Safety Analysis Supervisor T. A. Rieck Manager Reactor Safety Analysis NOD Safety Technical Support Fuel Vendor Safety T"chnical Review 16 Staff Members I

R. W. Tsai BWR Safety Analysis DtlWllopment Supeivlsor I

E.A.Armstrong Fuel Reliability Engineer Fuel Reliability J. A. Sllady Reload Licensing Engineer L Gutierrez Office

.__supe*rvi-so*r -

Oerlcal Support Failed Fuel Action Plan Reload Design I Safety Licensing Issues Coordination Departmental Training and Trip Coordination Fuel Fabrication Reviews Fuel Bid Evaluation Fuel Contract Input New Employe Orientation Fuel Tests &

Inspections Review New Fuel Designs I

E. H. Young PWRSupport Services Supervisor Station NE Support &

Assistance Reload Licensing Package Development Outage & Startup Support Core Monitoring &

Operation Software Support Core Component Acquisition PWR ONE Training 11 Staff Members Custodian of.

Confidential and/or Proprietary Data I

J.M. Ocher BWRSupport Services. Supervisor Station NE Support &

Assistance Reload Licensing Package Development Outage & Startup

  • Support Core Monitoring &

Operation Software Support Core Component Acquisition.

  • BWR ONE Training 9 Staff Members

PE.RSONNEL EDUCATION....

MS uclear Fuel Services o:briefbk:hgS:pnseddeg:1 b.y degree

  • BS 40.6%

Ph.D 15.9%

other 4.3%

30 28 26 24 22 20 en 18 a:

16 us 14 12 10 8

6

  • 4 2

NUCLEAR EXPERIENCE

  • 1n years o~__.___..___._--4-__.__.__._~---+-~~__.___._ __

0 5 10 15 20 25 30 35 40 45 50 55 60 65 PEOPLE o:brie1bk:hg3:nuce>ep.pra: 1

Vendor Independence Program (VIP)

NRR Presentation May 5, 1994 Ken Kovar, Safety Analysis Supervisor I

I

  • 1

What is it?

Development of CECo reload neutronic and safety analysis capabilities Production of the analyses of record in these areas Vendor Independence Program (VIP)

Slide 1

Why do this?

To save$$$

Fuel savings Avoided analysis cost To provide in-depth technical support to the stations Vendor Independence Program (VIP)

Slide 2

How do we do it?

General plan - Four phases:

PWR Neutronic Analysis PWR Safety Analysis BWR Neutronic Analysis BWR Safety Analysis Common App~oach:

Vendor or EPRI methods and software On-the-job training at vendors (Design Participation Training)

Software installed at CECo Topical reports sub1nitted for NRC approval Vendor Independence Program (VIP)

Slide 3

  • /,,*;,

PWR Neutronic Analysis Westinghouse methods and software Started in '78 Eight engineers trained at Westinghouse Total project scope - thirty-five person years NRC approval in '83 NRC approval of methods upgrade in '91 Thirty-two (32) reload designs Technical support (e.g. Byron bents-pin reanalysis)

Savings: $15.5 million Vendor Independence Program (VIP)

Slide 4

PWR Safety Analysis Vendor Independence Program (VIP)

Slide 5

BWR Neutronic Analysis GE and SPC methods and software Started in early '80's Eleven engineers trained at GE and SPC Total project scope - fifty person years NRC approval for GE applications in '92 NRC approval for SPC applications in '93 Eleven (11) reload designs Technical support (e.g. Quad Cities high flux trip evaluation)

Savings: $19.3 million Vendor Independence Program {VIP)

Slide 6

BWR Safety Analysis Details on niethods and schedule to be presented later today.

t SPC methods and EPRI software Started in '90 One enginee~ trained at GE Total project scope - sixty person years (estimated)

Topicals to be submitted starting later this year**

Technical Support (e.g. Post-LOCA suppression pool temperature monitoring)

Savings: $4.1 million Vendor Independence Program (VIP)

Siide.7

What's in the future?

CECo rejoining EPRI (nuclear)

Left EPRI in '92 due to financial situation Remained as members of code maintenance groups Access to research and deyelopn1ent Joint development with fuel vendors Transition from GE to SPC fuel at LaSalle and Quad Cities PWR multi-dimensional kinetics (Rod Eject)

Unified neutronic 1nethods Continuing ties with Nuclear Engineering Schools CECo/DOE matching grants Parallel projects Sun11ner intern program Vendor Independence Program (VIP)

Slide 8

Engineering and Operational Support NRR Presentation 5 May, 1994 Kevin B. Ramsden, Reactor Systems Engineer

Introduction NFS Performs a Wide Variety of Support Functions Some are highly visible Some are transparent This discussion ~ill highlight the more important functions and services NFS performs Engineering and Operational Support*

Slide 1

Tools System Transient Analysis RETRAN02 RELAP5 TRAC Vendor tools (ODYN/LOFTRAN/TWINKLE)

Core T/H VI PRE COBRAIIICMIT FIBWR2 CONTAINMENT CONTEMPT4M5 GO'fl IIC Engineering and Operational Support Slide 2

Tools, contd.

SPECIAL COMMIXIB HEATING?

RELAP4M6 MATHCAD ANALYSIS PLATFORMS IBM Mainframes HP 735 Workstations Vax Pritne PCs Engineering and Operational Support

.I Slide 3

NFS Customers

Nuclear Stations Nuclear Groups Site Engineering Systems Engineers Operations Regulator):' Assurance Training Nuclear Licensing Offsite Review AEs and Consultants Vendors Engineering and Operational Support Slide 4

,;_.~'

NFS Design Ownership Reload N eutronic Analyses LOCA Analyses Vendor performs/NFS controls Containment Analyses Vend or or NFS may perform Reload Transient Analyses Engineering and Operational Support Slide 5

. ~.,. '...

NFS Design Ownership, contd.

EQ Compartment Analyses BIB Steam Tunnel Superheat Review/performance of new analyses Special Analyses PWR ECCS Flow Balance AFW Performance BIB SG Tube Rupture BIB UHS analysis BWR ECCS room cooling requirements Engineering and Operational Support el Slide 6

Safety Evaluation/Support Perform Input Reviews (OPL3/4)

Review MODS/Changes with Sites Close Relationship with Vendors/Consultants Review of Vendor/AE analysis Ge~erate/support preparation of 50.59 for Sites Maintain understanding of margins LOCA rackups MCPR/DNBR penalties Engineering and Operational Support Slide 7

. ***l Operability/Plant Problems Assist System Engineers and SEC Personnel Develop JCO/BCOs when appropriate Perform Operability Evaluations Support Safety Significance Determination LERS Enforcement Conferences

!'Informal Support" AEOD Licensing Engineering and Operational Support I_

. Slide R

On-Site Investigations NFS Pr~vides Dir~ct Support Analytical Supp.ort Provided QC HPCI Exhaust Failure Event LaSalle RCIC Exhaust Failure Event Undervoltage Issues Battery Loading Profiles MOY concerns Dresden SORV Dresden FW Transient

~ngineering and Operational Support Slide 9

Generic Issue Support Station Blackout Coping Studies Dresden Quad Cities Containment Analysis for D/QC/8&8 Review of Zion submittal BWR Stability Owners Group Activities Analysis Subcommittee MOV Issues Review of Generic BCOs Review of Valve Priority Assignments ECCS Strainers Engineering and Operational S~pport Siide.10

I Operations Support EOPs Simulator Support Startup support( Physics/criticality etc.)

Tech Spec Interpr~tation/Improvement Direct Assistance of Site NEs GSEP Engineering and Operati~~al Support o!

  • Slide. I I

Observations Need to shift more work from ~'Reactive" to

Preventive~'

Earlier engineering involvement may prevent/reduce "fire drills'.'

"Excuse Engineering" is challenging, but it doesn't improve the plant._

Communications

. Welcome more early interaction with NRR to prevent misunderstandings, particularly on older plant design basis questions.

Engineering and Operational Support I

Slide 12

I'.,

  • Summary NFS Provides a Variety of Seryices Much of our work is '!behind the scenes" NFS growth has been deli~erately controlled Engineering and Operational Support I

I Slide 13

Nuclear Fuel and Reactor Engineering Support NRR Presentation May 5, 1994 By: Jack M. Colter

BWR-& PWR Support Services

Purpose:

Support Services provides technical support to the Site_

Reactor Engineers and others *for the nuclear fuel and core components.

Key Responsibilities:

  • Reactivity Management
  • Station Nuclear Engineer Training and Qualification
  • Reload Licensing
  • Core Monitoring Codes
  • Fuel Reliability
  • Core Components
  • Nuclear Material Control and Accountability
  • Spent Fuel and Criticality Analysis
  • Operations Support
  • Safety Evaluations for Plant Transients and Accidents Personnel ~ 17 Full-time Engineers 3 Engineering Assistants Nuclear Fuel and Reactor Engineering Support Slide 1

Reactivity Management Policy:

The Nuclear Fuel must be operated and handled in such a way that unplanned criticalities and fuel failure. from oper_atiori beyond design or operational* limits can never be permitted. e All planned reactivity changes shall be conducted in a controlled manner, the effects of reactivity changes are known and monitored and any anomalous indication is met with conservative action.

Activities:

  • Develop Policy
  • Review Station Procedures
  • Review License Training Lesson Plans
  • Event Investigation
  • Develop Guidelines/Procedures
  • Review Industry Events
  • Recommend SNE Staffing Levels
  • BWR & PWR Owner's Groups Committees (GE & W)
  • ANS Standards Committees Nuclear Fuel and Reactor Engineering Support Slide 2

Reactor Engineering Support Training and Qualification of site NE's Ensure the station NE's can perform their routine duties and can respond appropriately to any credible core problem.

e BWR & PWR NE Qualification Program Vendor Nuclear Engineering Course On-the-job training CECo SNE Course Qualification Oral* Board

  • Tech Specs/Bases
  • Situations
  • Job Responsibilities Normal (Xenon)

Abnormal On-Going Communication Lead Nuclear Engineer Meetings Weekly Conference Calls Site Visits Nuclear Fuel and Reactor Engineering Support Slide 3

Reload Licensing Goal: Provide coordination of reload licensing among stations/vendors/NFS groups.

Develop Schedule Submit Anticipatory Tech Spec Changes Develop Safety Evaluation Report Prepare Reload Package Develop or Review COLR Answer On-Site and Off-Site Review Questions GE/SPC Transition Nuclear Fuel and Reactor Engineering Support Slide 4

Fuel Reliability Goal: No defective fuel assemblies in any of Edison's Units and*

operate the fuel unrestricted to its end of life.

.~

Focus on technical aspects of design, manufacture and operation of nuclear fuel.

Manufacturing Plant Reviews New Fuel Designs

~ Fuel Design Changes Parts List Reviews Site Inspections (FME Control)

Reactor Water Chemistry Lost Parts Analysis Failed Fuel Action Plan Failed Fuel Inspections Vendor Technical Review Meetings Nuclear fuel and Reactor Engineering Sup po~

Slide 5

~

I

Core Monitoring Goal: Provide the site with the most accurate and useful information to support operation of the units.

Pioneer in core monitoring code development

~ early BO's POWERPLEX (with Exxon Nuclear)

Currently BEACON (Westinghouse)

Initial Testing at NFS Site Testing/Parallel Run Software Problem Reports Revisions to Core Monitoring Codes Cycle Data Updates Nuclear Fuel and Reactor Engineering Support Slide 6

---------~

Core Components Goal: Provide the sites with the best core components at the most reasonable prices. Encourage Innovation - Non-Original Equipment Control Blades, & RCCAs.

e BWR's Control Blades, LPRM's, Channels, and Channel Fasteners PWR's RCCA's Bid Specification I.

Bid Technical Evaluations/Recommendations Technical Issues End Of Life Tracking Vendor Meetings Nuclear Fuel and Reactor Engineering Support Slide 7

~~

Nuclear Material Control & Accountability Technical Content of CECo's Nuclear Material Control Procedures Technical Consultant to Stations Changes/Process Improvements Nuclear Fuel Data Bank Nuclear Fuel and Reactor Engineering Support Slid~-8

Conclusion Duties -

Diverse, Point of Contact for Stations &

Vendors Staff Mostly Former Station NE's Goal Support Nuclear Fuel & Reactor Engineering Staffs Nuclear Fuel and Reactor Engineering Support Slide 9 I"

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