ML20059K281

Summary of 931027 Meeting w/ABB-CE in Rockville,Md Re CE Sys 80+ Protection Against Containment Bypass During SGTR & Fuel Design.List of Meeting Attendees & ABB-CE Handouts Encl

ML20059K281
Person / Time
Site:	05200002
Issue date:	11/10/1993
From:	Mike Franovich Office of Nuclear Reactor Regulation
To:	Office of Nuclear Reactor Regulation
References
NUDOCS 9311150293
Download: ML20059K281 (39)

Text

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-Docket No.52-002 APPLICANT: ABB-Combustion Engineering, Inc. (ABB-CE)

PROJECT: CE System 80+

SUBJECT:

PUBLIC MEETING OF OCTOBER 27, 1993, 10 DISCUSS CE SYSTEM 80+

PROTECTION AGAINST CONTAINMENT BYPASS DURING A STEAM GENERATOR TUBE RUPTURE (SGTR) AND FUEL DESIGN On October 27, 1993, a public meeting was held at the U.S. Nuclear Regulatory Commission (NRC) offices in Rockville, Maryland, between representatives of NRC and ABB-CE. Enclosure 1 provides a list of attendees. Enclosure 2.is the material presented by ABB-CE.

The purpose of the meeting was to discuss the results of the ABB-CE final report on SGTR containment bypass (DSER Open Item 15.3.8-1). The principal concern for this issue involves the potential for a main steam safety valve (MSSV) sticking open during an SGTR resulting in an unisolable release' path from the reactor coolant system to the environment.

In addition to SGTR, the design change process for the System 80+ fuel design was discussed under the provisions of Tier 2. The staff's DSER stated that any change to the fuel design would constitute an unreviewed safety question.

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ABB-CE believes that there should be some flexibility in this area to permit design changes to the fuel under the 50.59 like-process, especially since there is a high probability that future developments in the fuels area will cause ABB-CE to modify their design.

SGTR Containment Bvoass By letter dated October 6,1993 (LD-93-145), ABB-CE provided the NRC.a final report entitled " Evaluation of the System 80+ Standard Design for Steam Generator Tube Rupture Events." In the report, ABB-CE claimed that the proba-bility that an SGTR event will result in containment bypass is reduced by.

current System 80+ design features. With the System 80+ design.featuri.s (including recent design changes), the conditional probability for containment bypass via opening of the MSSVs during an SGTR event becomes lower by two orders of magnitude versus the earlier System 80 design.

Based on the SGTR analysis, ABB-CE has made two modifications to the Sys-tem 80+ design. ABB-CE committed to add two nitrogen-16 (N-16) monitors (one for each steam generator), with a latching mechanism, (indication / alarm lock)

.to assist operators in their diagnostic functions. ABB-CE also modified the component cooling water system (CCWS) to permit the steam bypass control system (SBCS) to continue to operate after a safety injection actuation. signal (SIAS). This design change was necessary to ensure cooling water is main-tained for the air compressors following an SIAS so that actuation air remains available to the turbine bypass valves throughout the event. l 120013 9311150293 931110 PDR ADOCK 05200002-NRC HLE CBmR COPY J i

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g- U g November 10, 1993 Based on the staff review of the ABB-CE report on SGTR, the staff concluded that the issue of SGTR containment bypass is technically resolved. However, several significant areas and further clarification must be addressed by ABB-CE prior to closure of the issue.

The staff expressed concern that the report does not explain how operating System 80 plants and the previous System 80+ design's instrument air system (IAS) and associated cooling water for the air compressors react under an SIAS. A clear description of how the SBCS, CCWS, and IAS function under an SIAS would be needed in order to substantiate the applicant's conclusion that this design change to the CCWS is the most important design feature that reduces the likelihood of the bypass sequence.

The staff also noted that issues concerning technical specifications and operability requirements for the N-16 monitors must be-addressed by ABB-CE.

Availability of N-16 monitors would provide early warning and an event precur-sor indication of a potential SGTR due to early primary to secondary leakage detection assuming setpoints are adjusted for realistic operating conditions and not offsite dose limits for Part 100. The staff also stated that operat-ing experience has not provided solid indication that secondary system detectors are adequately maintained without technical specification require-ments. Therefore, the staff stated that technical specifications for the N-16 monitors would be in order.

Finally, the staff stated that the emergency procedure guidelines (EPGs) and the inspections, tests, analyses, and acceptance criteria (ITAAC) need to factor in the design changes and cooldown strategy. Design features such as the rapid depressurization system (RDS) and N-16 monitors will need to be incorporated into the EPGs. For ITAAC, ABB-CE stated they believe the N-16 monitors are addressed in ITAAC; however, the staff requested verification since the design change was made two months after the ITAAC submittal. The design change to the CCWS should also be reflected in the associated ITAAC.

In addition, the staff stated that the N-16 monitors should be considered for the minimum inventory of alarms for the main control room ITAAC, especially since the main steam area radiation monitors have made the minimum inventory list.

Fuels Area In the DSER, the staff designated the fuels area as Tier 2 prime. Although not specifically called " Tier 2 prime" in the DSER, Tier 2 material which the staff concludes may not be changed without prior NRC approval has been termed as " Tier 2 prime" for discussion purposes only. Tier 2 prime will be identi-fied in the staff's FSER as Tier 2 material that if changed, would involve an unreviewed safety question and, therefore, requires NRC review and approval prior to implementation. Any requested change to this material shall either be specifically described in the combined license (COL) application or submitted for license amendment after COL issuance.

ABB-CE presented their definition of what should be covered by Tier 2 prime for the fuel and core design that would allow for design changes without NRC review. These parameters for both fuel and core design cover a broad range of

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.t November 10, 1993 parameters (see pages 23 and 24 of Enclosure 2). Fuel and cladding design -

limits, fuel material and structural properties, and control element assem- '

blies were some of the parameters identified in the fuels area. Burnable ,

poisons, reactivity and fuel burnup, minimum departure from nucleate boiling  !

ratio (DNBR), and power distribution were some of the parameters identified for the core design area.

ABB-CE attempted to define the appropriate numerical envelopes for these parameters that would allow for future design changes under the 50.59-like process. ABB-CE also stated that if design changes do not exceed the speci- i fled envelopes and are covered by an applicable NRC approved topical report, ,

design changes should not require NRC review and approval. '

The staff did not agree with ABB-CE's approach in defining what parameters may be changed in Tier 2. The staff stated that the ABB-CE approach would require nearly a year to review and did not appear to be achievable. The staff stated that a more viable approach would be to define what parameters clearly should .

not involve an unreviewed safety question. ABB-CE agreed with the staff's position and will propose a table of approximately five core / fuel design parameters that envelope design changes that could be made unaer the 50.59- g like process. ABB-CE estimated that this table would include items such as burnable poisons (if in accordance with NRC approved topical), burnup, and  :

core load patterns.

Commitments The following commitments were made during the meeting: .,

(1) ABB-CE commit i to evaluate technical specifications for the N-16 monitors. i (2) For the SGTR event, ABB-CE committed to provide updated EPGs that have factored in the design changes such as the N-16 monitors and System 80+

design features which may be used as a contingency action such as'the  ;

rapid depressurization system (RDS).

(3) ABB-CE committed to clearly explain how the previous System 80+ design's ,

SBCS, IAS, and CCWS respond under an SIAS during the SGTR event. ABB-CE >

should also explain how operating System 80 plant's turbine-bypass valves and air supply system perform during an SGTR event.

(4) ABB-CE committed to confirm that the N-16 instrumentation is shown on the i figure (s) for the appropriate system (s) ITAAC. In addition, ABB-CE should update the minimum inventory list of alarms for the main control room ITAAC to reflect the N-16 monitors. ABB-CE should also modify the CCWS ITAAC if needed.  :

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November 10, 1993- '

f In the fuels area, ABB-CE committed to provide a table of parameters '

'(5) where fuel or core design changes would be permitted under the 50.59-like process.

ABB-CE stated.their intent is to fulfill these commitments by November 15, 1993.

r Oright Swr o.,.

Michael X. Franovich, Project Manager Standardization Project Directorate Associate Directorate for Advanced Reactors and License Renewal Office of Nuclear Reactor Regulation

Enclosures:

As stated cc w/ enclosures:

See next page DISTRIBUTION w/ enclosures:

'Docketffile' PDST R/F DCrutchfield TKenyon PDR MFranovich PShea i

DISTRIBUTION w/o enclosures:

RBorchardt JMoore, 15B18 TMurley/FMiraglia WRussell ,

RJones, 8E23 MRubin, 10E4 SSun, 8E23 GGrant, 17G21 -

GHsii, 8E23 RArchitzel flSaltos, 10E4 TBoyce  ;

AThadani, 8E2 RPalla,10E4 TCollins, 8E23 LPhillips, BE23 SMagruder RArchitzel LKopp, 8E23 KShembarger AEl-Bassioni, 10E4 KEccleston, 1004 WTravers ACRS (11) AChu, llE22 .

OFC: LA:PDST:ADAR PM:PpMADAR SC:PDST:ADAR NAME: PShea qu? MXFfno{vich:tz RArcpitzel DATE: 11/y;/93 T17i/93 11/4/93 1

0FFICIAL RECORD COPY: DOCUMENT NAME: MSUM1027.MXF t

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-ABB-Combustion Engineering, Inc. ' Docket No.52-002 cc: Mr. C. B. Brinkman,cActing Director Nuclear Systems Licensing ABB-Combustion Engineering, Inc.-

e: - - 1000 Prospect Hill' Road Windsor, Connecticut 06095-0500

Mr. C. B. Brinkman, Manager Washington Nuclear Operations ABB-Combustion Engineering,.Inc.

12300 Twinbrook Parkway, Suite 330 Rockville, Maryland 20852 Mr. Stan Ritterbusch Nuclear Systems Licensing ABB-Combustion Engineering, Inc.

1000 Prospect Hill Road Post Office Box 500 Windsor, Connecticut 06095-0500 Mr. Sterling Franks U.S. Department of Energy NE-42 Washington, D.C. 20585 Mr. Steve Goldberg Budget Examiner 725 17th Street, N.W.

Washington, D.C. 20503 Mr. Raymond Ng 1776 Eye Street, N.W.

Suite 300 Washington, D.C. 20006-Joseph R. Egan, Esquire Shaw, Pittman, Potts & Trowbridge 2300 N Street, N.W.

Washington, D.C. 20037-112B Mr. Regis A. Matzie, Vice President Nuclear Systems Development ~_

ABB-Combustion Engineering, Inc.

1000 Prospect Hill Road Post Office Box 500 Windsor,; Connecticut 06095-0500 >

Mr. Victor. G. Snell, Director Safety and' Licensing AECL Technologies 9210 Corporate Boulevard Suite 410 Rockville, Maryland 20850

, :* .' - AEB-CE SYSTEM 80+..  :

Meeting -  !

October 27,1993

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Rockville, Maryland

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'NAME ORGANIZATION -  : i c

A. Thadani . NRR/DSSA' [

M. Franovich NRR/PDST {

N. Saltos NRR/SPSB  :

R.- Jones NRR/SRXB ,!

T. Collins NRR/SRXB i M. Rubin NRR/SPSB i T. Boyce NRR/PDST A. El-Bassioni NRR/DSSA '  !

T. Wambach NRR/PDST

. Y. G. Hsii NRR/DSSA .

L. Kopp NRR/SRXB .

L. Phillips NRR/DSSA - 1 '

D. Finnicum ABB-CE C. Brinkman ABB-CE j '

M. T. Cross ABB-CE  ;

M. Kantrowitz ABB-CE -;

T. Rudek ABB-CE  !

J. Iongo, Jr. ABB-CE -

P. M. Lang DOE ,

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SYSTEM 80+

CONTAINMENT BYPASS ISSUE ABB-CE/NRC MEETING OCTOBER 27,1993 i-

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CONTAINAIENT BYPASS ISSUE 1

OUTLINE OF PRESENTATION: .

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1. PURPOSE ,

2. SUABIARY OF EVALUATIONS OF SGTR EVENTS

3. RESULTS OF PROBABILISTIC EVALUATIONS -!

4. SUABIARY. OF BENEFITS & LBIITATIONS.0F POTENTIAL I DESIGN CHANGES i j

5. RESULTS OF BEST ESTBIATE SGTR ANALYSES l

6. SUABIARY OF CASES ANALYZED FOR SGTR EVENTS P

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CONTAINhENT BYPASS ISSUE 'i y

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TO DESCRIBETHE STATUS OF COMMITAENTS FROM 9/23/93 NRC LETTER  !

THE COMMITAENTS WERE: j 1

(1) QUALIFY & QUANTIFY THE IMPROVED CAPABILITY OF - l SYSTEM 80+ TO COPE WITH AN SGTR (DISCUSSED. l IN SGTR REPORT) . j y

f (2) INVESTIGATE POTENTIAL HIGH CAPACITY SG LIQUID - 1 BLOWDOWN SYSTEM WHICH DISCHARGES TO THE IRWST. .i (DISCUSSED IN SGTR REPORT)  !

l (3) . PROVIDE COMPUTER PLOTS TO DESCRIBE SGTR TRANSENTS l (PROVIDED IN SGTR REPORT) {

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-SYSTEM 80+ N CONTAINMENT BYPASS ISSUE ,

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SUMMARY

OF EVALUATIONS OF SGTR EVENTS 5

e THE PROBABILITY THAT SGTR EVENTS WILL RESULT IN- 'i CONTAINMENT BYPASS IS SIGNIFICANTLY REDUCED -  ;

VIA THE USE OF SPECIFIC SYSTEM 80+ DESIGN FEATURES.  :

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(a) N-I6 RADIATION MONITORS (WITH LATCHED SIGNAL),  ;

ONE PER SG, TO. ALLOW QUICK DIAGNOSIS OF AN SGTR (b) CCW FLOW TO AIR COMPRESSORS FOLLOWING -

SIAS SO THAT ADEQUATE AIR SUPPLY IS AVAILABLE a FOR ACTUATION OF STEAM BYPASS VALVES.- a (c) AUTOMATIC TERMINATION OF MAIN FEEDWATER-FOLLOWING REACTOR TRIP AND WITH REDUCED PRIMARY COOLANT TEMPERATURES SO AS TO EXTEND  !

THE RUPTURED SG FILL TIME (d) RDS WHICH IS ACTUATED BY THE OPERATOR IF A -

POTENTIAL FOR MSSV CHALLENGE EXISTS (e) NUPLEX 80+ CONTROL ROOM TO ASSIST THE +

OPERATOR ~IN QUICKLY DIAGNOSING THE EVENT ~ 1 l

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CONTAINMENT BYPASS ISSUE  !

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SUMMARY

OF EVALUATIONS OF SGTR. EVENTS (CONT'D)  !

e SPECIFIC FEATURES (CONT'D):

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(O STEAM BYPASS SYSTEM, FOR WHICH ALL VALVES .!

DIRECT SECONDARY FLOW TO THE CONDENSER,  !

ELIMINATES TWO DIRECT PATHS FOR ATMOSPHERIC DISCHARGE j (g) IRWST WHICH HAS MORE THAN ADEQUATE INVENTORY-  !

(500,000 GAL MIN) OF BORATED WATER FOR SEVERAL rj HOURS OF SAFETY INJECTION j

'l (h) LARGER SECONDARY VOLUME OF SYSTEM 80+ SGs PROVIDES EXTRA CAPACITY FOR STORAGE OF TUBE j RUPTURE FLOW, THUS EXTENDING THE TIME OF.-  :

OPERATOR ACTION TIME PRIOR TO MSSV CHALLENGE

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SYSTEh! 80+

CONTAINh1ENT BYPASS ISSUE - 1

'i RESULTS OF PROBABILISTIC EVALUATION i I

e THE CONDITIONAL PROBABILITY OF A CONTAINMENT BYPASS VIA THE MSSVs GIVEN AN SGTR FOR SYSTEM 80+ IS ABOUT TWO ORDERS OF MAGNITUDE LOWER THAN FOR SYSTEM 80. 1 i

t e MOST IMPORTANT DESIGN FEATURE IS ENSURING THAT THE .i STEAM BYPASS SYSTEM CONTINUES TO OPERATE FOLLOWING l SIAS. .

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TABl.E 2 -1 Evaluation of Henefii of Adding Design Features to System 80 to Pieclude Lilting MSSVs on SGTR PLANT FEATURE DENEFIT Conditional Probability  %

of Stuck Open MSSV . Reduction System 80 n/a 6.96E-02 -

(Base)

Add Only N16 Detectors plus Faster end more accurate identification 6.30E-02 10 %

Advanced Cor trol Room of SGTR and statt.s of equipment.

Reduce operator error rates by 50%

Add Only Cooling Wate to Turbine Uypass Control System not 9.59E-03 86 %

Instrument Air failed on SIAS Compressors not lost on SIAS Add Only Safety Depresturization Alternato Methed to achieve RCS 5.63E-02 19 %

System pressur6 control in time to prevent lifting of MSSVs Add Only Larger Steam Generators Approximately 5 minutes longer before not quantified MSSVs lift. Slight increase in operator reliability Add Only All Turbine Dypass Eliminate potential containment bypass not quantified -

Valves discharge only to path containment System 80 + Includes all adited , -

2.28E-04 99.7 %

features above

TABLE 1.3-1

SUMMARY

OF THE BENEFITS & LIMITATIONS OF POTENTIAL DESIGN C11ANGES SECTION NO. & DESIGN CHANGE BENEFIT (S) LIMITATION (S) 3.2. Automatically bypass e Extends MSSV lift time e Redesign of the turbine MSIS on high SG 1evel. from 30 to 50 minutes for bypass system is 5 tubes ruptured. necessary.

e Steam lines will flood e Minimal hardware changes causing significant required. equipment damage.

• Conflicts with criterion 24 of Reg. Guide 1.153 which requires separation of Protection and Control Systems.

3.3 Automatically initiate e Very small extension of e Reduces RCS subcooling auxiliary pressurizer spray MSSV lift time (1 minute) e Increases pressurizer (APS). for 5 tubes ruptured. level early complicating e Minimal hardware changes diagnosis.

required.

3.4 Automatically open the e Very small extension of e Reduces RCS subcooling Reactor Coolant Gas Vent MSSV lift time (3 minutes) e Increases pressurizer System (RCGVS) for 5 tubes ruptured, level early complicating e Minimal hardware changes diagnosis.

required. e Conflicts with criterion 24 of Reg. Guide 1.153 which requires separation of Protection and Control Systems.

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TABLE 1.3-1 (Cont'd)

SUMMARY

OF THE BENEFITS & LIMITATIONS OF POTENTIAL DESIGN CHANGES BENEFIT (S) LIMITATION (S)

SECTION NO. & DESIGN CHANGE Automatically open e MSSV will not lift for 5 e New single failure safety 3.5 analysis concerns.

the SG liquid blowdown ruptured tubes beyond 10,000 seconds. e New containment penetra-system. tion piping & valves.

e Will require exemption from regulatory isolation requirements for the containment.

3.6 Automatically reduce e Small extension of MSSV 'e Pressure reduction post-trip SBCS pressure to lift time (3 minutes) for limited to remain above 5 tubes ruptured. the low SG pressure MSIS 900 psia (vs. 1100 psia) setpoint (850 psia).

• Complicates SBCS with potential impact on plant availability.

3.7 Automatic opening of e May extend MSSV lift time.

• Conflicts with required Turbine Eypass System and safety function for main Bypass of MSIS on low steam steam line breaks.

generator pressure. e High steam generator level MSIS still occurs.

e Conflicts with criterion 24 of Reg. Guide 1.153 which requires separation of Protection and Control Systems.

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SUMMARY

OF THE BENEFITS & LIMITATIONS OF POTENTIAL DESIGN CIIANGES SECTION NO. & DESIGN CILANGE BENEFIT (S) LIMITATION (S) 3.8 Automatically open the e Will prevent MSSV lift for e core subcooling lost Rapid Depressurization >10,000 seconds for a five e Large reverse flow of System (RDS) on the tube rupture. unborated water.

pressurizer. . Conflicts with criterion 24 of Reg. Guide 1.153 which requires separation of Protection and Control Systems.

• RDS is not currently designed for design basis events.

• Challenge to environ-mental qualification of containment equipment.

3.9 Increase main steam o Small extension of MSSV e Will require a redesign safety valve setpressure. lift (5 minutes) for a of steam generator &

five tube rupture. secondary systems.

e Higher RCS temperatures &

pressures for decreased heat removal events.

e The small break LOCA peak clad temperature will increase significantly.

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SUMMARY

OF THE BENEFITS & LIMITATIONS OF POTENTIAL DESIGN CHANGES T

SECTION NO. & DESIGN CHANGE BENEFIT (S) LIMITATION (S) e MSSV will not lift for 5 e New valves, piping sup ,

3.10 Automatic blowdown of ruptured tubes beyond ports, extensive first steam generator liquid to 10,000 seconds. of a kind engineering.

JRWST. e Challenge to' environ-mental qualification of j containment equipment.

o Overheating'of'the IRWST.

e Potential'for boron dilution.

3.11 Automatic initiation. e Minimal hardware changes .o Steam lines will flood of the atmospheric dump required. causing' equipment damage.

p valves, o Extension of MSSV lift e Vents tol atmosphere time, e Conflicts with criterion 24'of Reg; Guide 1.153' I-which require separation of protection and control systems.

3.12 Passive secondary e Extension of.MSSV' lift e' Steam lines will flood' cooling system. time. causing equipment-damage.

e Extensive redesignIo'f-containment and_ design of suppression tank.

o Extensive'first.of.a kind'.-

engineering.

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CONTAINMENT BYPASS ISSUE j 1

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RESULTS OF BEST ESTIMATE SGTR ANALYSES '

.i (1) ONE TUBE RUPTURE CASE ALLOWS > 4 HOURS BEFORE MSSVs ARE CHALLENGED  !

j (2) FIVE TUBE RUPTURE CASE ALLOWS > 30 MINUTES BEFORE  ;

MSSVs ARE CHALLENGED j

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(3) SOhE POTENTIALLY BENEFICIAL DESIGN CHANGES RISK l NEW PROBLEMS 1 a) REVERSE LEAKAGE INTO RCS -

b) SG DRYOUT j l

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d) HEATUP OF IRWST FLUID [

q (4) OTHER POTENTIALLY BENEFICIAL DESIGN CHANGES REQUIRE  !

RELAXATION OF NRC REQUIREhENTS (E.G., SEPARATION OF; i SAFETY AND NON-SAFETY GRADE EQUIPMENT) AND/OR ARE: '

COMPLEX FIRST-OF-A-KIND DESIGN OPTIONS.

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SUMMARY

OF CASES AMAI.YZED FOR MULTIPLE. STEAM GENERATOR TUBE RUPTURE Case # of SBCS' Auto' Auto' Auto' Auto SG' Auto' Auto' MSSV Approximate Tubes Dypass APS HCGV Dlowdown HDS ADV  !.1 I t MSSV Lift Time 4

MSIS Setpoint (Minutes)

( HSGI. )

(psia) _

Auto e no no no no ro no 1200 167+

11 1 900 psia Auto e no no no no no no 1200 33 12 5 900 t

psia 5 Auto e no no no no yes no 1200 167+

13 1100 psia 14 5 Auto e no no no yes, no no 1200 167+

1100 blowdown psia to 1RWST 15 5 Auto no no no no no yes 1200 tio MSSV lift 9 1100 ADV lifts at psia 28 mins.

16 5 Auto e no no no no no no 1400 35 1100 psia

1. SDCS . Steam Dypass Control System

2. Auto Bypass MSIS (11SGL) ... Automatic Dypass of the fligh .SG Level Initiation of Main Steam Line Isolation

3. Auto APS . Automatic Initiation of Auxiliary Pressurizer Spray

4. Auto RCGVS ... Automatic Initiation of the Reactor Coolant Gas Vent System

5. Auto SG Dlowdown . Automatic Initiation of the SG Liquid tslowdown System (to condenser)

6. Auto RDS .. Automatic Initiation of the Rapid Depressurization System (to IRWST) 7 Auto ADV Automatic Initiation of the ADV on fligh SG Pressure (setpoint of 1160 psia)

LICENSIBILITY THROUGH PART 52 -

1 4

NRC REVIEW OF PLANT _

DESIGN l CERTIFICATION g

DESIGN LOAD O = A = FUEL '

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NRC REVIEW EARLY SITE OF SITE

PERMIT C

LICENSING INSTABILITY CAUSED BY UNRESTRICTED RE-REVIEWS ANY NRC KEVIEW  :

FUEL OR CORE DESIGN CHANGE = OF FUEL OR .b 7 CORE DESIGN ) i!

U NRC REVIEW OF PLANT DESIGN I ,

DESIGN CERTIFICATION  : ,

O A LOAD '

= A FUEL L

EARLY SITE i NRC REVIEW OF SITE  : PERMIT

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. a LICENSING STABILITY RETAINED. ' f.'.

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WITH PRE-AGREED CHANGE PROCESS '

ANY PHE-AGREED FUEL OR CORE DESIGN CHANGE = PROCESS FOR CHANGES 4

NRC REVIEW OF PLANT DESIGN I CERTIFICATION  :

DESIGN T

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. L EARLY - SITE C

NRC REVIEW OF SITE = PERMIT 7

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PROPOSED PRE-AGREED CHANGE PROCESS 1 i FOR FIRST REFERENCE PLANT INITIAL CORE .

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" EFFECT OF

NO NRC DESIGN RE-REVIEW CHANGE E_______________.________________________________________________ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ ____________=___________

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CATEGORIZATION OF FUEL / CORE DESIGN CHANGES .

BEFORE:FIRST REFERENCELPLANT INITIAL. CORE .

. Topical Report Reviews Methodology Changes MechanicalRdaterial Design Changes

. Allowable Burnup Increases

2. NRC Re-reviews CESSAR-DC Tech Spec Changes Design Bases Changes Other Tier 2-Prime Changes

3. No Further NRC Review No change to CESSAR-DC Design Negative 50.59 Finding

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TIER 2-PRIME FOR FUEL AND CORE DESIGN TIER 2-PRIME FOR FUEL AND CORE DESIGN DEFINED BY ACCEPTANCE CRITERIA GIVEN IN TABLE 1 CHANGES WHICH DO NOT SATISFY TABLE 1 ACCEPTANCE CRITERIA BUT ARE COVERED- BY APPLICABLE NRC-APPROVED TOPICAL REPORTS WILL

. NOT REQUIRE NRC REVIEW CHANGES WHICH DO NOT SATISFY TABLE 1 ACCEPTANCE CRITERIA AND ARE NOT COVERED BY NRC-APPROVED TOPICAL REPORTS WILL REQUIRE NRC REVIEW 12-

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+ a Table 1  !

Acceptance Criteria for Fuel and Core Design  ;

Acceptance Criteria CESSAR-DC Section  !

_ Fuel Desian Fuel Assembly Structural Integrity 4.2.1.1.1 e

Fuel Cladding Design Limits 4.2.1.2.1 .

Fuel Rod Cladding Properties 4.2.1.2.2 .,

UO2 Fuel Pellet Properties 4.2.1.2.4 Burnable Poison Rod Design Bases 4.2.1.3 Control Element Assembly 4.2.1.4 Design Bases Fuel Assembly Design Evaluation 4.2.3.1  ;

Fuel Rod Design Evaluation 4.2.3.2 i

Burnable Poison Rod Design Evaluation 4.2.3.3 CEA Design Evaluation 4.2.3.4

-1 i

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. Core Desian Excess Reactivity and Fuel Burnup 4.3.1.1.1-Negative Reactivity Feedback 4.3.1.3 ,

t Reactivity Coefficients 4.3.1.4 Burnable Poison Requirements 4.3.1.5 Stability Criteria 4.3.1.6 Maximum Controlled Reactivity insertion 4.3.1.7 Rate ,

Power Distribution Control 4.3.1.8 i Excess CEA Worth 4.3.1.9.1  !

Chemical Shim Control 4.3.1.10 Maximum CEA Speed 4.3.1.11.1 Power Distribution Limits 4.3.2.2.1.1, 4.3.2.2.1.4.1 i

i K-effective During Refueling 4.3.2.6.1 Peak Reactor Vessel Wall Fluence 4.3.2.8.1 ]

Minimum DNBR 4.4.1.1.1 :r Flow Stability 4.4.1.2 .

Peak Fuel Temperature 4.4.1.3A f

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.s. 4.3.1.1 Excess Reactivity and Fuel Burnu_a 4.3.1.1.1 Fuel Burnup Design Bt. sis The average fuel burnup is chosen to ensure that the peak burnup is within the limits discussed in Paragraph 4.2.3.2.10. This design basis, along with the design basis in Paragraph 4.3.1.8, satisfies General Design Criterion 10.

4.3.1.1.2 Excess Reactivity and Fuel Burnup Design The excess reactivity provided for each cycle is based on the depletion characteristics of the fuel and burnable poison and on the desired burnup for each cycle. The desired -

burnup is based on an economic analysis of the fuel cost and the projected operating load cycle for System 80+.

k

. 4.3.1.9 Excess CEA Worth with Stuck Rod Criteria  !

4.3.1.9.1 Excess CEA Worth Design Basis  :

The amount of reactivity available from insertion of withdrawn CEAs assures that,  ;

under conditions of normal operation and anticipated operational occurrences, and I with appropriate margin for stuck rods, specified acceptable fuel design limits are not exceeded. This basis, along with Paragraph 4.3.1.10, satisfies General Design Criteria 26 and 27.

4.3.1.9.2 Excess CEA Worth Design The amount of reactivity available from insertion of withdrawn CEAs under all power operating conditions, even when the highest worth CEA f ails to insert, will provide for '

at least 2% excess CEA worth after cooldown to hot zero power, plus any additional shutdown reactivity requirements assumed in the safety analyses.

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4.3.1.11 Maximum CEA Speed I

1 4.3.1.11.1 Maximum CEA Speed Design Basis j l

The maximum CEA speed is consistent with the maximum controlled reactivity l insertion rate design basis discussed in Paragraph 4.3.1.7.  !

4.3.1.11.2 Maximum CEA Speed Design l Additional discussion of maximum CEA speed is presented in Section 4.2.

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E 4.3.2.2 Power Distribution i

4.3.2.2.1 General ,

4.3.2.2.1.1 Power Distribution Control Power distribution and coolant conditions are controlled so that the peak linear heat rate and the minimum departure from nuclear boiling ratio (DNBR) are maintained within operating limits supported by the safety analyses (Chapters 6 and 15)yith due regard for the correlations between measured quantities, the power distribiudon and uncertainties in the determination of the power distribution. '

4.3.2.2.1.2 Methods of Control Methods of controlling the power distribution include: the use of full- or part-strength l' CEAs to alter the axial power distribution; decreasing CEA insertion by boration, thereby improving the radial power distrib: tion; and correcting off-optimum

• conditions which cause margin degradations (e.g., CEA misoperation).

4.3.2.2.1.3 Core Operating Limit Supervisory System (COLSS) l i

4.3.2.2.1.3.1 COLSS Design Basis The Core Operating Limit Supervisory System (COLSS) indicates to the operator how ,

far the core is from the operating limits and provides an audible alarm _should an operating limit be exceeded. Such a condition signifies a reduction in the capability ,

of the plant to withstand an anticipated transient, but does not necessarily imply a violation of fuel design limits.

4.3.2.2.1.3.2 COLSS Design The COLSS, described in Section 7.7 and Reference 1, continually generates an assessment of the margin to linear heat rate and DNBR operating limits. The data ,

required for these assessments include rnessured in-core neutron flux data, CEA positions, and coolant inlet temperature, pressure and flow rate. In the event of an alarm indicating that an operating limit has been exceeded, power must be reduced -  !

unless the alarm can be cleared by improving either the power distribution or another i process parameter. The accuracy of the COLSS calculations are verified periodically. '

4.3.2.2.1.4 Reactor Protective System (RPS) i 4.3.2.2.1.4.1 RPS Design Basis if the margin to fuel design limits continues to decrease beyond the COLSS operating limits, the Reactor Protective System (RPS) assures that the SAFDLs are not exceeded j by initiating a trip.

t 28 j

J e 4.3.2.2.1.4.2 RPS Design

- The RPS Core Protection Calculators (CPCs, see Section 7.2) continually infer the core  ;

power distribution and DNBR by processing reactor coolant data, signals from ex-core -  !

neutron flux detectors, each containing three axially stacked elements, and input from l redundant reed switch assemblies to indicate CEA position. In the event the power distributions or other parameters are perturbed as a result of an anticipated operational occurrence that would violate fuel design limits, the high local power density or low DNBR trips in the RPS will initiate a reactor trip.

4 I

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. 4.3.2.6 Criticality of Reactor Durina Refuelina 4.3.2.6.1 Criticality Design Basis During Refueling i The soluble boron concentrations during refueling ensure that the k,n of the core l during refueling does not exceed 0.95.

4.3.2.6.2 Criticality Design During Refueling The soluble boron concentrations during refueling are shown in Table 4.3-1. These concentrations ensure that the k,n of the core during refueling does not exceed 0.95.

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. Vessel Irradiation 4.3.2.8.1 Vessel Irradiation Design Basis  !

The reactor design ensures that the peak vessel wall fluence for the design life of the  ;

2 plant is less than 1.0 x 10 n/cm for neutrons of energy greater than 1 MeV.

2 4.3.2.8.2 Vessel Irradiation Design  ;

t The design of the reactor internals and of the water annulus between the active core -

and vessel wallis such that for reactor operation at the full power rating and an 80%

capacity factor, the vessel fluence greater than 1 MeV at the. vessel wall is not-  ;

expected to exceed 6.2 x 10'8 n/cm2 over the 60-year design life of the vessel. The calculated exposure includes a 30% uncertainty factor.

The maximum fast neutron flux greater than 1 MeV incident on the vesselID is based on a time-averaged equilibrium cycle radial power distribution and an axial power distribution with a peak-to-average planar power ratio of 1.15. The models used in .j these calculations are discussed in Section 4.3.3.3. >

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7-4.4.1.1 Minimum Deoarture from Nucleate Boilina Ratio (DNBR) 4.4.1.1.1 Minimum DNBR Design Basis The minimum DNBR shall be such as to provide at least a 95% probability with 95%

confidence that departure from nucleate boiling (DNB) does not occur on a fuel rod having that minimum DNBR during steady-state operation and anticipated operational occurrences.

4.4.1.1.2 Minimum DNBR Design A minimum DNBR value of 1.24 using the CE-1 correlation coupled with the CETOP code provides at least 95% probability with 95% confidence that DNB does not occur ,

during stead-state operation and anticipated operational occurrences.

31