ML20236T833

COLR for Dresden Station Unit 3,Cycle 15

ML20236T833
Person / Time
Site:	Dresden
Issue date:	06/30/1998
From:	COMMONWEALTH EDISON CO.
To:
Shared Package
ML17191A805	List: ML20236T833 ML20236T839
References
NUDOCS 9807290059
Download: ML20236T833 (27)

Text

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ATTACilMENT A l

CORE OPERATING LIMITS REPORT I

FOR DRESDEN STATION UNIT 3, CYCLE 15 (Revised) i i

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9807290059 990724 PDR ADOCK 05000249 P

PDR Y

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SECTION A Core Operating Limits Report l

DRESDEN STATION UNIT 3 CYCLE 15 l

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1 June 1998

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ISSUANCE OF CHANGES

SUMMARY

Affected Section Affected Pages Summary of Changes Date All 1-1 through 5-6 Incorporated Reference to TSUP Section 06/97 Number / Deleted References to Custom TS.

References iii identified Analyses of Record for D3C15.

06/97 2.2 and 2-1, 2-2 and 2-3 Included MAPLHGR limits for D3C15 9x9-2 06/97 Figure 2.2-1 and and ATRIUM-98 reload fuel.

Table 2.2-1 3.2 and 3-1 and 3-2 included SLHGR limits for D3C15 06/97 Figure 3.2-1 ATRIUM-9B reload fuel.

4.2 and 4-1 and 4-2 Included TLHGR limits for D3C15 06/97 Figure 4.2-1 ATRIUM-9B reload fuel.

l 5.2 and 5-1 and 5-2 Simplified from Figure 5.2-1 A, sinca 06/97 Table 5.2-1 OLMCPRs are not scram time dependent.

5.2 and Table 5-2, 5-4 and 5-5 Revised to reflect new Operating Limit 06/97 5.2-1, Figure MCPRs for 9x9-2 and ATRIUM-9B reload I

5.21 and 5.2-2 fuel, Deleted previous Figure 5.2-1, because Operating Limits MCPR's were performed using only the Technical Specification scram j

times and, thus, are not scram time dependent Table 5.2-2 5-3 Added a table of the OLMCPR adders for 06/97 turbine bypass valve opening time degradation l

Figure 2.2-1, 2 2, 3-2, and 4-2 The table of information in Figure 2.2-1 was 06/97 Figure 3.2-1, split into two tables, Figures 3.2-1 and 4.2-1 and Figure 4.2-1 were changed to say N/A if an limit did not l

exist at that exposure.

l References iv The SPC letter documenting the 0.01 adder 9/97 for the reduced dome pressure operation was i

added Table 2.3-1 2-3 Corrected action step to be consistent wnh 9/97 TSUP by changing 3.6.A Action d to 3.6.A.1d.

Done at Dresden.

Dresden Unit 3 Cycle 15 i

June 1998 t

l Table 5.2-1 5-2 increased the Operating Limit Minimum 9/97 Critical Power Ratio by 0.02 due to operation at up to 15 psi below the analyzed pressure (0.01) and a conservative (0.01) for future potential additional MCPR penalties (i.e. for operation at a higher steam flow rate).

Table 5.2-1 5-2 Added statement that no NSS limits are 9/97 presented for this cycle. Only TSSS limits are presented in the COLR lssuance of i, ii The table of changes increased by one page, 9/97 Changes therefore affecting the page numbering of the

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Summary following pages, and the Table of Contents References iv The D3C15 Reload Licensing documents 6/98 (Reference numbers 4 and 5) were updated for their most recent revisions due to the incorporation of an increased steam flow rate in the analyses.

Table 5.2-1 5-2 Added a footnote explaining that the 9/97 6/98 additional 0.01 MCPR penalty is not necessary to support the operation in increased steam flow.

Section B Attachments 2 & 3 Attached the latest revision of the D3C15 6/98 SPC Reload Analysis and excerpts from the SPC Transient Analysis i

l Dresden Unit 3 Cycle 15 ii June 1998

SECTION A TABLE OF CONTENTS Page REFERENCES.

.iv LIST OF FIGURES..

.v LIST OF TABLES.

.vi 1.0 ROD BLOCK MONITOR (RBM).

1-1 1.1 TECHNICAL SPECIFICATION REFERENCE.

1-1

1.2 DESCRIPTION

1-1 2.0 AVERAGE PLANAR LINEAR HEAT GENERATION RATE

.2-1 2.1 TECHNICAL SPECIFICATION REFERENCE.

.2-1

.2-1

2.2 DESCRIPTION

2.3 MAPLHGR LIM!T EQU!PMENT OUT OF SERVICE MULTIPLIERS.

.2-1 3.0 STEADY STATE LINEAR HEAT GENERATION RATE.

.3-1 3.1 TECHNICAL SPECIFICATION REFERENCE.

.3-1

3.2 DESCRIPTION

.3-1 4.0 TRANSIENT LINEAR HEAT GENERATION RATE.

.4-1 1

4.1 TECHNICAL SPEC.FlCATION REFERENCE.

. 4-1

4.2 DESCRIPTION

.4-1 S.O MINIMUM CRITICAL POWER RATIO.

.5-1 5.1 TECHNICAL SPECIFICATION REFERENCE.

.5-1

5.2 DESCRIPTION

.5-1 6.0 ' METHODOLOGIES.

.6-1 Dresden Unit 3 Cycle 15 iii June 1998 i

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REFERENCES 1.

Commonwealth Edison Company Docket No. 50-249. Dresden Nuclear Power Station, Unit 3, Facility Operating License DPR-25.

2.

Letter, D.M. Crutchfield to All Power Reactor Licensees and Applicants. Generic Letter 88-16, Concerning the Removal of Cycle-Specific Parameter Limits from Technical Specifications.

3.

EMF-97-031(P), Dresden LOCA-ECCS Analysis MAPLHGR Limits For ATRIUM-9B And 9x9-2 Fuel, Siemens Power Corporation, May 1997, NFS NDIT # 970081.

4.

EMF-96-141, Revision 1, Dresden Unit 3 Cycle 15 Reload Analysis, Siemens Power Corporation, June 1998 NFS NDIT# NFS-97-0085, Sequence 1.

l S.

EMF-97-047, Dresden Unit 3 Cycle 15 Plant Transient Analysis with !ncre:ised Steam Flow, Siemens Power Corporation, June 1998, NFS NDIT # NFS-97-0084, Sequence 1.

6.

Dresden Unit 3 Cycle 15 Neutronic Licensing Report, NFS NDIT # 97007.8.

7.

EMF-92-149(P) And EMF-92-149(P) Supplement 1, Revision 1, Dresdea Units 2 And 3 Generic Coastdown Analysis With ATRIUM-98, Siemens Power Corporation, September 1996, NFS NDIT # 960137 t

8.

SPC letter, Dresden Unit 3 Cycle 15 MAPLHGR Limits Versus Asserably Average Exposure, DEG:97:048. D.E. Garber to R.J. Chin, June 5,1997 NFf NDIT # 970111.

9.

SPC letter, Dresden Reduced Dome Pressure Analyses, DEG:97:102, D.E. Garber to R.J. Chin, August 6,1997, NFS NDIT #970117 Rev.1.

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l Dresden Unit 3 Cycle 15 iv June 1998 L______________________

LIST OF FIGURES Figure Title / Description Page f

2.2-1 MAPLHGR Limit vs. Bundle Average Exposure 2-2 l

3.2-1 Steady State LHGR (SLHGR) Limit vs. Average Planar Exposure 3-2 l

4.2-1 Transient LHGR (TLHGR) Limit vs. Average Planar Exposure 4-2 5.2-1 Operating Limit MCPR for Manual Flow Control 5-4 5.2-2 Operating Limit MCPR for Automatic Flow Control for ATRIUM-9B and 9x9-2 5-5 Fuel l

l June 1998 Dresden Unit 3 Cycle 15 v

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LIST OF TABLES Table Title / Description Page 1.2-1 Control Rod Withdrawal Block Instrumentation Setpoints 1-2 2.3-1 Equipment Out Of Service MAPI.HGR Limit Multipliers 2-3 5.2-1 Operating Limit MCPR for 9x9-2 and ATRIUM-9B Reload Fuel 5-2 l

5.2-2 Turbine Bypass Valve Degradation OLMCPR Adders 5-3 l

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Dresden Unit 3 Cycle 15 vi June 1998

1.0 ROD BLOCK MONITOR (RBM) 1.1 Technical Specification Reference Technical Specification 3.3.M. - Rod Block Monitor (RBM) 1.2 Description The Rod Block Monitor Upscale Instrumentation Setpoints are determined from the relationships shown in Table 1.2-1.

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Dresden Unit 3 Cycle 15 1-1 L_ _

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TABLE 1.2-1 CONTROL ROD WITHDRAWAL BLOCK INSTRUMENTATION SETPOINTS l

TRIP FUNCTION:

TRIP LEVEL SETTING.

Rod Block Monitor Upscale I

l (Flow Bias)

Dual Loop Operation Less than or equal to (0.65 W plus 55)*

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Single Loop Operation Less than or equal to l

(0.65 W plus 51)*

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• W - percent of drive flow required to produce a rated core flow of 98 Mlb/hr.

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i Dresden Unit 3 Cycle 15 1-2 June 1998 u-__________.--____----

l 2.0 AVERAGE PLANAR LINEAR HEAT GENERATION RATE 2.1 Technical Specification Reference Technical Specification 3.11.A - AVERAGE PLANAR LINEAR HEAT GENERATION RATE 2.2 Description The Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) Limit versus Bundle Average Exposure for each fuel type is determined from Figure 2.2-1.

2.3 MAPLHGR Limit Eauipment Out of Service Multipliers The appropriate multiplicative factors, during power operation with equipment out of service, to apply to the base MAPLHGR limits specified in Section 2.2 are shown in Table 2.3-1.

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Dresden Unit 3 Cycle 15 2-1 June 1998 1

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l FIGURE 2.2-1 MAPLHGR LIMIT VS. BUNDLE AVERAGE EXPOSURE 14 l

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212 x

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- 9x9 2 s

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+ AT RIU M-9 B h

y RELOAD 7

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5 10 15 20 25 30 35 40 Bundle Average Exposure (GWd/MTU)

Bundle Average Exposure MAPLHGR Limit (GWD/MTU) 9x9-2 j

(kW/ft) l 0

12.5 8.8 12.5 39,3 7.7 i

Bundle Average Exposure MAPLHGR Limit (GWD/MTU)

ATRIUM-98 Reload Fuel (kW/ft) 0 13.5 11.1 13.5 37.5 8.7 l

Dresden Unit'3 Cycle 15 2-2 June 1998

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i TABLE 2.31 EQUIPMENT OUT OF SERVICE MAPLHGR LIMIT MULTIPLIERS Technical Title of Technical Scenario Multiplicative Factor, Specification Specification 9x9-2 and ATRIUM-98 3.11.A &

Average Planar Linear Single Loop 0.90 3.6.A.1.d Heat Generation Rate Operation (SLO) and Recirculation Loops i

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l Dresden Unit 3 Cycle 15 2-3 June 1998

3.0 STEADY STATE LINEAR HEAT GENERATION RATE l

t 3.1 Technical Specification Reference Technical Specification 3.11.D - STEADY STATE LINEAR HEAT GENERATION RATE 3.2 Description I

The Steady State LHGR (SLHGR) limit versus Average Planar Exposure for each fuel type is determined from Figure 3.2-1.

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I June 1998 Dresden Unit 3 Cycle 15 3-1

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l FIGURE 3.2-1 STEADY STATE LHGR (SLHGR) LIMIT VS. AVERAGE PLANAR EXPOSURE l

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9x9 2

+ ATRIUM 9B 8

Reload Fuel N

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10 20 30 40 50 60 Average Planar Exposure (GWD/MTU)

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i' Average Planar Exposure SLHGR Limit SLHGR Limit (GWD/MTU) 9x9-2 ATRIUM-98 Reload Fuel (kW/ft)

(kW/ft) 0-14.5 14.4 5.0 14.5 14.4 15.0 12.6 14.4 25.2 10.8 13.0 i.

48.0 7.2 10.0 55.0 N/A 9,1 l

' Dresden Unit 3 Cycle 15 3-2 June 1998

4.0 TRANSIENT LINEAR HEAT GENERATION RATE l

4.1 Technical Specification Reference Technical Specification 3.11.B - TRANSIENT LINEAR HEAT GENERATION RATE 1

4.2 Description The Transient LHGR (TLHGR) limit versus Average Planar Exposure for each fuel type is determined from Figure 4.2-1.

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Dresden Unit 3 Cycle 15 4-1 June 1998

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FIGURE 4.21 TRANSIENT LHGR (TLHGR) LIMIT VS AVERAGE PLANAR EXPOSURE l

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19 C T

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~NA\\

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14 N

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(3 13

9x9 2

=J 12

+ ATRIUM 9B 99 Reload Fuel i

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9 L

0 10 20 30 40 50 60 l

Average Planar Exposure (GWD/MTU)

I Average Planar Exposure TLHGR Limit TLHGR Limit (GWD/MTU) 9x9-2 ATRIUM-9B (kW/ft)

Reload Fuel (kW/ft) 0.0 19.2 19.4 15.0 17.8 19.4 25.4 16.9 17.5 43.2 10.8 14.3 48.0 10.0 13.5 55.0 N/A 12.3 l

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~ Dresden Unit 3 Cycle 15 4-2 June 1998

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5.0 MINIMUM CRITICAL POWER RATIO I

5.1 Technical Specification Reference Technical Specification 3.11.C - MINIMUM CRITICAL POWER RATIO 5.2 Description a.

The Operating Limit MCPRs for D3C15 are listed in Table 5.2-1. The

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OLMCPRs calculated for D3015 are based on Technical Specification Scram insertion Speeds (3.3 E).

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For operation with a degraded turbine bypass valve opening time, the t

OLMCPR adder in Table 5.2-2 must be added to the Operating Limit l

MCPR determined from Table 5.2-1. Linear interpolation between the data points is permissible.

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During Manual Flow Control, the Operating Limit MCPR for each fuel type at reduced core flow conditions can be determined from whichever is l

greater:

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i. Figure 5.2-1 using the curve and the appropriate flow rate.

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ii. The Operating Limit MCPR determined from Table 5.2-1, and supplemented by Table 5.2-2 when appropriate.

d.

During Automatic Flow Control, the Operating Limit MCPR for each fuel -

type at reduced flow rates can be determined from Figure 5.2-2 using the appropriate flow rate and the Operating Limit MCPR, which is obtained from Table 5.2-1, and supplemented by Table 5.2-2 when appropriate.

Linear interpolation between the curves on Figure 5.2-2 is permissible.

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Dresden Unit 3 Cycle 15 5-1 June 1998 l

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

TABLE 5.2-1 OPERATING LIMIT MCPR f

FOR 9x9-2 AND ATRIUM-9B RELOAD FUEL l

Operating Scenario Operating Limit MCPR'

• l 1.48 Normal Operation f~

Normal Operation with Feedwater Heaters Out of Service 1.48 1.49 Single Loop Operation Coastdown:

1.52 Coastdown and SLO Operation' 1.53

' Note that the Operating Limit MCPR is not a function of the average CRD scram insertion time for the current operating cycle other than assuming the Technical Specification average CRD scram insertion time limits (3.3.E) are met. For simplification of implementing the limits for D3C15, only limits corresponding to the Technical Specification Scram Speeds have been specified. The MCPR Operating Limits presented are based on Technical Specification Scram

' Speeds and bound the Nominal Scram Speed Operating Limits.

The 0.04 MCPR penalty during Coastdown includes the effects of Feedwater Heater (s) Out of 2

Service and Single Loop Operation.

For coastdown and SLO, the 0.01 adder to the MCPR Safety Limit is still necessary.

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'The MCPR Operating Limits contain a 0.01 adder to the reload licensing results due to -

operation at slightly less than design pressure. Originally (the September 1997 version of the D3C15 COLR), the MCPR Operating Limit had an additional 0.01 conservatism for future potential operation with increased steam flow (for a total 0.02 adder to the reload licensing results). However, when the operation at increased steam flow was analyzed, the extra 0.01 in

- the MCPR Operating Limit was not necessary and the MCPR Operating Limits were not changed. Therefore, the above MCPR Operating Limits contain a 0.01 adder for operation at slightly less than design pressure and a 0.01 generic conservatism, that can be used for future operational flexibility.

Dresden Unit 3 Cycle 15 5-2 June 1998

I TABLE 5.2-2 TURBINE BYPASS VALVE DEGRADATION OLMCPR ADDERS l

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Equivalent Bypass Valve OLMCPR Adder for Delay Time (msec)*

ATRIUM-9B and 9x9-2 Fuel 50 0.00 150 0.02 250 0.03 350 0.04 450 0.04 550 0.05 700 0.05 900 0.05 l

No Bypass 0.05

• Delay is relative to the time of TSV full closure.

Linear interpolation can be used for purposes of selecting a conservative OLMCPR adders for equivalent delay times not specifically listed in the table.

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Turbine Bypass Valve Degradation OLMCPR Adders 0.05 0.04 8

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0.03 l

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0.02 0

0.01 i

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0 100 200 300 400 500 600 700 800 900 1000 Equivalent Bypass Vahe Delay Time (msec)

Dresden Unit 3 Cycle 15 5-3 June 1998 E___________________________________

l l-FIGURE 5.2-1 i

OPERATING LIMIT MCPR FOR MANUAL FLOW CONTROL 2.4 2.3 2.2 2

E

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g 2

s 2,,,

\\

- 1.8 m

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17 en 3

1.6 1.5 i

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O 1.4

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1.3 1.2 w

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10 20 30 40 50 60 70 80 90 100 l

Total Core Flow (% rated,98 Mlb/HR)

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110% Maximum Flow l

I Total Core Flow Operating Limit MCPR for

(% Rated) 9x9-2 and ATRIUM-98 Reload Fuel 100 1.18 l~

30 1.86 i

O 2.35 l

Dresden Unit 3 Cycle 15 5-4 June 1998 1

f FIGURE 5.2-2 OPERATING LIMIT MCPR FOR AUTOMATIC FLOW CONTROL FOR ATRIUM 9B and 9x9-2 FUEL' 3

2.9 2.8 CC 2.7 h 2[6 2

YA s

A 2.4

+-

2.3 2.2 m2 NNN l

55 h\\

2 E

3.,

-+- 1.46 NAN l

$ 1.8

-e-1.5 NNN I

O 1.7

--+- 1.5 5 -

16 L

1.5 w r l

1.4 O

10 20 30 40 50 60 70 80 90 100 Total Core Flow (% Rated,98 MLB/HR) 110% Maximum Flow Total Core Flow Operating Limit MCPR

(% Rated) 9x9-2 and ATRIUM-98 Reload Fuel i

1.46 1.50 1.55 100 1.46 1.50 1.55 l

30 2.29 2.35 2.45 00 2.79 2.87 2.97 i

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' Although analyzed for core flows from 0% to 100%. Technical Specification 3.3.N prohibits AFC operation below 65% core flow.

Dresden Unit 3 Cycle 15 5-5 June 1998

S.0 METHODOLOGIES The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC in the latest approved revision or supplement of the topical reports describing the methodology. For Dresden Unit 3 the NRC approved topical reports are:

1)

ANF-1125(P)(A) and Supplements 1 and 2. " Critical Power Correlation - ANFB",

April 1990.

2)

ANF-524(P)(A), " Advanced Nuclear Fuels Corporation Critical Power Methodology for Boiling Water Reactors: Methodology for Analysis of Assembly Channel Bowing Effects /NRC Correspondence", XN-NF-524(P)(A) Revision 2, Supplement 1 Revision 2, Supplement 2, November 1990.

3)

XN-NF-79-71(P)(A), " Exxon Nuclear Plant Transient Methodology for Boiling Water Reactors," Revision 2 Supplements 1,2, and 3, March 1986.

4)

XN-NF-80-19(P)(A), " Exxon Nuclear Methodology for Boiling Water Reactors -

Neutronic Methods for Design and Analysis", Volume 1 and Supplements 1 and 2, March 1983.

5)

XN-NF-80-19(P)(A), " Advanced Nuclear Fuels-Methodology for Boiling Water Reactors " Volume 1 Supplement 3, Supplement 3 Appendix F, and Supplement 4, November 1990.

6)

XN-NF-80-19(P)(A), " Exxon Nuclear Methodology for Boiling Water Reactors:

EXEM BWR ECCS Evaluation Model." Volumes 2,2A,2B,2C, September 1982.

7)

XN-NF-80-19(P)(A), " Exxon Nuclear Methodology for Boiling Water Reactors THERMEX: Thermal Limits Methodology Summary Description", Volume 3 Revision 2, January 1987.

8)

XN-NF-80-19(P)(A), " Exxon Nuclear Methodology for Boiling Water Reactors:

Application of the ENC Methodology to BWR Reloads", Volume 4, Revision 1, June 1986.

9)

XN-NF-85-67(P)(A), " Generic Mechanical Design for Exxon Nuclear Jet Pump Boiling Water Reactors Reload Fuel," Revision 1, September 1986.

10)

ANF-913(P)(A), "COTRANSA2: A Computer Program for Boiling Water Reactor Transient Analyses," Volume 1 Revision 1 and Volume 1 Supplements 2,3, and 4, August 1990.

11)

XN-NF-82-06(P)(A), " Qualification of Exxon Nuclear Fuel for Extended Burnup Supplement 1 Extended Burnup Qualification of ENC 9x9 BWR Fuel," May 1988.

' Dresden Unit 3 Cycle 15 6-1 June 1998

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12)

ANF-89-014(P)(A)," Advanced Nuclear Fuels Corporation Generic Mechanical Design for Advanced Nuclear Fuels Corporation 9x9-IX and 9x9-9X BWR Reload Fuel" October 1991.

13)

ANF-89-98(P)(A)," Generic Mechanical Design Criteria for BWR Fuel Designs,"

Revision 1 and Revision 1 Supplement 1, May 1995.

l 14)

ANF-91-048(P)(A), " Advanced Nuclear Fuels Corporation Methodology for l

Boiling Water Reactors EXEM BWR Evaluation Model," January 1993.

15)

Commonwealth Edison Company Topical Report NFSR-0091, " Benchmark of CASMO/MICROBURN BWR Nuclear Design Methods," and associated Supplements on Neutronic Licensing Analyses (Supplement 1) and LaSalle l

County Unit 2 Benchmarking (Supplement 2).

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Dresden Unit 3 Cycle 15 6-2 June 1998

ATTACilM ENT B RELOAD LICENSING ANALYSIS l

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Section B Dresden Unit 3 Cycle 15 Reload Transient Analysis Results June 1998 r

1 Dresden Unit 3 Cycle 15 June 1998

Section B Table of Contents Attachment Document 1

Neutronics 1.icensing Report 2

Reload Analysis Report 3

Excerpts from Plant Transient Analysis Report I

Dresden Unit 3 Cycle 15 June 1998

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l Dresden Unit 3 Cycle 15 Neutronics Licensing Report l

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June 1998 Dresden Unit 3 Cycle 15 L______________________________________-.--

NUCLEAR FUEL SERVICES DEPARTMENT

' NUCLEAR DESIGN INFORMATION TRANSMITTAL

@ SAFETY RELATED Onginsung Organtuuon NDIT No.

970028 O NON-SAFE'IY RE!ATED

@ Nuclear Fuel Services Rev. No.

O O REGULATORY RELATED 0 Other(spectfy)

Page 1 of 19 Station Dresden Unit 3 Cycle 15 Genene To. Russell D. Freeman Subject Dresden Unit 3 Cvele 15 Neutronie Ucensine Report (NLR)

[

$['f[$ @

David A Pheelev 1

Preparer Pfe;6ter's Signature d [

Date Theodore P Shannon C

f/l 3[1]

Reviewer

[ Reviewer's Sipiature Date Adelmo S Pallotta

/

I f) 7 NFS Supervisor NFS Supervisor's Signature Date Status of informauon:

@ Verined O unvenfied O Engineenng Judgernent Method and Schedule of Vennestion for Unvenfied NDfTs; Desenpuon of Informanon: Results and basis of neutronic licensing caelutsuons for Dresden Unit 3 Cycle 15 l

Purpose of information: Provide Station and NFS BSS group with neuuonic licensing results t

1 i

Darce ofinformauon Calculauen Notes NFS.BNDD.97 015 and NF3.BNDD.97 03S i

Supplemental Distnbution:

J.W. Keffer (Dresden). D. A. Worthington. Dresden Central File. D3C15 Letternook DG Cenud F.le

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[0 "Y O C A// '9 l}f k

Dresden 3 Cycle 15 NDIT c 970028 Neutronic Licensing Reoort Revision 0 Licensing Basis This document, in conjunction with References 1 and 2, provides the licensing basis for Dresden Unit 3 Reload 14, Cycle 15. The calculations that support this report are given in References 3 through 10.

Table of Contents 1.

Nuclear Design Analysis 3

1.1 Fuel Bundle Nuclear Design Analysis 3

1.2 Core Nuclear Design Analysis 4

1.2.1 Core Configuration and Licensing Exposure Limits 4

1.2.2 Core Reactivity Characteristics 5

11.

Control Rod Withdrawal Error 5

l 111.

Fuel Loading Error 6

IV. -

Control Rod Drop Accident 7

V.

Loss of Feedwater Heating 7

VI.

Maximum Exposure Limit Compliance 8

Vll.

Spent Fuel Pool and New Fuel Vault Criticality Compilance 8

Vll.1 New Fuel Vault Criticality Compliance 8

Vll.2 Spent Fud Pool Criticality Compliance 9

Vill.

References 9

Attachment A Neutronic Licensing Procedure References 19

($

[4 N

Page 2 of 19 7b 3/W

viesuen 2 uyue i a

. e w.. -o,vv m

Neutronic Licensing Report Revision 0 l.

_ Nuclear Desian Analysis 1.1 Fuel Bundle Nuclear Desian Analysis Assembly Average Enrichment (w/o U-235)

SPC ATRIUM-9B 3.26 9Cd3.5/11Gd5.5/9Cd4.5 (DRC 8L) 3.26 SPC ATRIUM-9B 3.26 9Cd3.5/11Cd5.5 (DRC-8H) 3.26 SPC ATRIUM 98 3.39 6Cd3.5/6Cd4.5/6Cd5.5 (URC 8A) 3.39 Radial Enrichment and Burnable Poison Distribution SPC ATRIUM-9B 3.43 9Cd3.5 (DRC-8L and 8H )

Figure 1 SPC A' TRIUM-9B 3.4411Cd5.5 (DRC-8L and 8H )

Figure 2 SPC ATRIUM-98 3.78 9Cd4.5 (DRC 8L)

Figure 3 SPC ATRIUM 9B 3.7811Cd5.5 (DRC-8H )

Figure 4 SPC ATRIUM 98 3.62 6Cd3.5 (DRC-8A )

Figure 5 SPC ATRIUM 9B 3.62 6Cd4.5 (DRC 8A )

Figure 6 i

SPC ATRIUM 9B 3.88 6Cd5.5 (DRC-8A )

Figure 7 1

Axial Enrichment and Burnable Poison Distribution SPC ATRIUM 9B 3.26 9Cd3.5/11Cd5.5/9Cd4.5 Figure 8 SPC ATRIUM 9B 3.26 9Cd3.5/11Cd5.5 Figure 8 SP.C ATRIUM 9B 3.39 6Cd3.5/6Cd4.5/6Cd5.5 Figure 8 I

Page 3 of 19 gW M

Neutronic Licensing Report Revision 0 i

1.2 Core Nuclear Desion Analysis 1.2.1 Core Configuration and Licensina Exoosure Limits l

Cycle Number Bundle Tvoe Loaded in Core SPC 9x9 2 3.13 8Cd2.0/8Cd3.0 12 84 l

SPC 9x9-2 3.13 8Cd2.0/8Cd4.0 12 12 SPC 9x9-2 2.95 8Cd3.0 13 48 SPC 9x9-2 2.95 9Cd3.5 13 116 SPC 9x9-2 2.95 7Cd3.0 13 52 SPC 9x9 2B 3.13 7Cd3.5 14 72 l

SPC Sx9 2B 3.13 8Cd3.5/9Cd4.5 14 108 I

SPC ATRIUM 98 3.26 9Cd3.5/11Gd5.5/9Cd4.515 104 l

SPC ATRIUM 9B 3.26 9Cd3.5/11Gd5.5 15 72 l

SPC ATRIUM 9B 3.39 6Cd3.5/6Cd4.5/6Cd5.5 15 56 ore Cycle l.

Average incremental Exoosure Exnosure l

Exposure at EOC N-1 Nominal EOC N-1 (MWD /MTU) 26199.6 8900.0 Short EOC N 1 (MWD /MTU) 25699.6 8400.0 i

Exposure at EOC N 1 l

Shutdown Reactivity Calculations, l

(MWD /MTU) 25394.9 8095.2 Cycle 15 neutronics analyses are valid for EOC N-1 l

exposures greater than 8400.0 MWD /MT, The exposure window that validates the pressml. ation transients can be found in Reference 2.

l l

l I

i DWG/9/W Page 4 of 19

/ bn5/i2/9~,

Neuwonic Licensing Report Revision b i

'l.2.2 Core Reactivity Characteristics s

I j.

All values reported below are with zero xenon and are for i

l 68'F moderator temperature. The MICROBURN B cold BOC

{

K-effective bias is 1.0070 (see Reference 17).

l l

BOC Cold K Effective, All Rods Out 1.10471 BOC Cold K-Effective All Rods in 0.96027 l

BOC Cold K-Effective, Strongest Rod Out 0.99691 BOC Shutdown Margin, % AK 1.00 Minimum Shutdown Margin, % AK 1.00 l

Reactivity Defect (R-value) Total, % AK 0.04 Baron Slumping, % AK 0.04 SDM Decrease from BOC, % 4K 0.00 Standby Liquid Control System Shutdown Margin, Cold Condition,600 ppm (% AK) 4.767 11.

Control Rod Withdrawal Error l

Analysis was performed at 100% power,100% flow, unblocked conditions only.

Distance Withdrawn (ft)

ACPR 12 0.32 The design complies with the SPC 1% plastic strain criteria via conformance to I

the transient LHGR limits.

l f-i

/N %h Page 5 of 19 n

V 5/stl9i Q_____________--_----_----

aw es,

-, s s Neutronic Licensing Report Revision 0 Ill.

Fuel Loadina Error f

The fuel loading error, including fuel dislocation and disorientation, is 1.

classified as an accident. By demonstrating that the fuel loading error meets i

the more stringent Anticipated Operational Occurrence (AOO) requirements, the offsite dose requirement is assured to be met. Because the events listed below result in a ACPR value that is less than that of the limiting transient, the AOO requirements and hence the off-site dose requirements are met for the fuel loading error.

l The values reported below bound all fuel types found in the core.

Event ACPR Mislocated Bundle 0.28 i

Disoriented Bundle 0.10 For the fuel loading error, the design, complies with the SPC 1% plastic strain criteria via conformance to the transient LHCR limits.

l l

i I

l l

l i

i Page 6 of 19 gl 7PE4-.1/2

vie w eisa w ue is

.cw..

2, v v o Neutronic Licensing Report Revision 0 L

IV. Control Rod Dron Accident i.

l This Analysis was performed using a rod sequence that bounds the Dresden-supplied rod sequence (to be used in D3C15) as described by References 22 and 23 (rod arrays 1-4) and Reference 18 (rod arrays 5-8). Note that the 0.32Ek adder mentioned below is included in this analysis to account for possible rod L

l mispositioning errors.

d I

I Dropped Control Rod Worth without 0.32 Ek adder. Ek 0.793 i

i Dropped Control Rod Worth l

with 0.32 Ek adder, Ek 1.113 l

Doppler Coefficient,1/k Ak/dT

-10.44E-06 (*F) 1 Effective Delayed Neutron Fraction used 0.0055 Four Bundle Local Peaking Factor l.28 l

Maximum Deposited Fuel Rod Enthalpy with 0.32 Ek adder, (Cal /gm) 168 l

Number of Rods Creater than 170 Cal /gm with 0.32Ek adder 0

l V.

Loss of Feedwater Heatina The loss of feedwater heating event is analyzed at 100% of rated power,87% and 100% of rated flow and an assumed inlet temperature decrease of 200 F.

l Event ACPR Loss of Feedwater Heating 0.22 i

The design complies with the SPC 1% plastic strain criteria via conformance to' the transient LHCR limits.

l pySNn l

Page 7 of 19 -

74 g

Dresden 3 Cycle 15 NDIT # 970028 Neutronic Licensing Report Revision 0 VI. Maximum Exposure Limit Compliance Note that these exposures are based on the nominal Cycle 14 exposure, 8900 MWD /MT, and an End Of Cycle 15 core exposure of 26,386 MWD /MT (this is the l

licensing basis core exposure at EOC15 per Reference 2). See References 19 21 for fuel assembly exposure limits.

Projected Peak Assembly Exposure (MWD /MTU) 39,330 (Assembly A3UOl3 @ 35 9x9-2 Fuel) l SPC 9x9 2 Assembly Exposure Limit (MWD /MTU) 40,000 Projected Peak Pellet Exposure (MWD /MTU)

(Assembly A3U006 @ 25-36 9x9-2 Fuel) 53,811 SPC 9x9-2 Pellet Exposure Limit (MWD /MTU) 55,000 l

The Data found above is for 9x9-2 fuel. The ATRIUM 9B fuel will not be near the Exposure limits set for it at the end of Cycle 15. The projected peak exposures for ATRIUM 9B fuel are listed below.

Projected ATRIUM-9B Peak Assembly Exposure (MWD /MT) 18,004 (Assembly A3X198 @ 23-36)

SPC ATRIUM-9B Assembly Exposure Limit (MWD /MT) 48,000 Projected ATRIUM-9B Peak Pellet Exposure (MWD /MT) 27,810 (Assembly A3XO62 @ 51-28-5)

SPC ATRIUM 9B Pellet Exposure Limit (MWD /MT) 60,000 Vll. Spent Fuel Pool and New Fuel Vault Criticality Compliance For the D3C15 reload, there are three new SPC ATRIUM-9B assembly types consisting of 7 unique lattices, as identified in Section 1.1. As described in the Reference 13 and 14 transmittals, all three fresh bundle types comply with the spent fuel pool and new fuel vault criticality limits.

Vll.1 New Fuel Vault Criticality Compliance All the new assemblies comply with the fresh fuel vault criticality limits of enrichment less than 5.00 wt% U-235 (lattice average) and gadolinia content greater than 6 rods at 2.0 wt% Gd,0. Reference 11 details the analysis showing 3

that the above enrithment/Gd limits insure compliance with the Reference 15 UFSAR section.

l Page 8 of 19

[

L

Neutronic Licensing Report Revision 0 4

l Vll.2 Spent Fuel Pool Criticality Comoliance All the new assemblies comply with the spent fuel pool criticality limits of enrichment less than 4.30 wt% U-235 (lattice average) and gadolinia content j

greater than 6 rods at 2.0 wt% Gd 0. Reference 12 details the analysis showing i

2 3 that the above enrichment /Gd limits insure compliance with the Reference 16 Technical Specifications section.

I Vill. References 1.

EMF-96-139,"Dresden Unit 3 Cycle 15 Plant Transient Analysis," May 1997.

2.

EMF-96-141,"Dresden Unit 3 Cycle 15 Reload Analysis," May 1997.

3.

Calcnote NFS:BNDD:97-002,."Dresden Unit 3 Cycle 15 Neutronic Licensing Basepoint and SDM Calculations," Revision 0,1/29/97.

4.

Calcnote NFS:BNDD:97-003,"Dresden Unit 3 Cycle 15 CRDA Analysis,"

l Revision 0,3/6/97.

5.

Calcnote NFS:BNDD:96-042,"Dresden 3 Cycle 15 Fuel Assembly Dislocation Calculations," Revision 0,3/4/97.

6.

Calcnote NFS:BNDD:96-044,"Dresden 3 Cycle 15 Loss of Feedwater Heating Analysis," Revision _0.2/1/97.

f

\\

7.

Calcnote NFS:BNDD:97-006,"Dresden Unit 3 Cycle 1S Rod Withdrawal Error Analysis," Revision 0, 2/28/97.

8.

Calcnote NFS;BNDD:97-011,"Dresden Unit 3 Cycle 15 Standby Liquid Control Calculations," Revision 0,2/24/97.

9.

Caltnote NFS:BNDD:97-010,"Dresden Unit 3 Cycle 15 Disoriented Bundle Analysis," Revision 0, 3/7/97.

10. Calcnote NFS:BNDD:97-016,"Dresden Unit 3 Cycle 15 Analysis of 030697 Loading Pattern," Revision 0, 3/13/97.

11. EMF-96148(P)," Criticality Safety Analysis for ATRIUM-9B Fuel, Dresden and Quad Cities New Fuel Storage Vault," Revision 1, September 1996.

12. EMF 94-098(P)," Criticality Safety Analysis for ATRIUM 9B Fuel, Dresden Units 2 and 3 Spent Fuel Storage Pool," Revision 1, January 1996.

13. Letter NFS:BSS:96-005,"D3Cl 5 New Fuel Storage Vault and Spent Fuel Storage Pool Criticality Analysis (Current Technical Specification),"

C.H. Nguyen to J.W. Keffer, January 16,1996.

I Page 9 of 19 9

j

'Neutronic Licensing Report Revision 0 l

14. Letter NFS:B55:96 098,"D3C15 New Fuel Storage Vault and Spent Fuel Storage Pool Criticality Analysis (Current Technical Specification)- Extra 56 Fuel Bundles," C.H. Nguyen to R. Kundalkar, May 28,1996.

l

15. "Dresden Updated Final Safety Analysis Report," Section 9.1.1.3.

16. "Dresden Units 2 and 3 Technical Specifications," Section 5.6.A.

17. Calcnote NFS:BND:95-038," Selection of D3C15 Target Eigenvalues,"

Revision 0,4/12/96.

L

18. Letter "Dresden Control Rod Withdrawal Sequences," from Carlos de la Hoz to Ronald J. Chin,9/16/94.

19. XN NF 85 67(P)(A), Revision 1," Generic Mechanical Design for Exxon Nuclear jet Pump BWR Reload Fuel," September 1986.

20. XN NF 82 06(P)(A), Supplement 1, Revision 2," Qualification of Exxon Nuclear Fuel for Extended Burnup, Supplement i Extended Burnup Qualification of ENC 9x9 BWR Fuel," May 1988.

21. EMF-96-040(P), Rcvision 1, " Fuel Design Report for Dresden Unit 3, Cycle 15 ATRIUM-9B Fuel Assemblies," August 1996.

22. Letter " Control Rod Arrays and Bank Positions for D3C15," from R.D.

Freeman to Dr. R. Chin, July 16,1996, Doc. l.D.# 5028404.

23. Letter "New BPWS 'X' Rod Pull Order," from R.D. Freeman to Dr. R. Chin, August 10,1995.

Page 10 of 19 5/d

& r/J/47

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Fiaure 1 SPC ATRIUM-9B 3.43 9Gd3.5 (DRC-8L and 8H)

Enrichment Distribution 3lt' Page il of 19

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Rods (32) at 4.06 w/o U-235 Fiaure 2 SPC ATRIUM-98 3.4411Gd5.5 (DRC-8L and 8H)

Enrichment Distribution Page 12 of 19 7F.j:'/.

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Rods ( 2) at 2.70 w/o U-235 Fiaure 3 SPC ATRIUM-9B 3.78 9Gd4.5 (DRC-8L)

Enrichment Distribution

$lt' Page 13 of 19

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Enrichment Distribution

@&l Page 15 of l9 l

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Rods ( 2) 2.30 w/o U-235 3

Rods ( 2) 2.75 w/o U-235 4

Rods (21) 3.53 w/o U-235 G

Rods (6) 3.00 w/o U-235 + 4.5 w/o Gd 5

Rods (28) 4.30 w/o U-235 7

Rods (12) 3.00 w/o U-235 Figuie 6 SPC ATRIUM-9B 3.62 6Gd4.5 (DRC-8A)

Enrichment Distribution i

j Page 16 of 19

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Neutronic Licensing Report Revision 0

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Rods (21) 3.78 w/o U-235 G

Rods (6) 333 w/o U-235 + 5.5 w/o Gd 5

Rods (26) 435 w/o U-235 7

Rods ( 2) 4.30 w/o U 235 8

Rods ( 4) 3.00 w/o U-235 9

Rods ( 2) 333 w/o U-235 Figure 7 SPC ATRIUM-98 3.88 SGd5.5 (DRC-8A)

Enrichment Distribution

$0 Page 17 of 19 n

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Dresaen 3 Cycle 15 Neutronic Licensing Report Revision 0 i

Figure 8. DR315 ATRIUM-9B Bundle Axial Designs Natural U 11" Natural U 11" Natural U i

S.43 w/o 3.43 w/o 3.62 w/o i

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Page 18 of 19 b,1/

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urescen nyue -> a Neutron!c Licensing Report Revision 0

' ATTACHMENT A - NEUTRONIC LICENSING PROCEDURE REFERENCES 1.

NFS-ND-900," Nuclear Design Procedures," Revision 4,8/20/96, Appendix A,

" Performing Nuclear Design Routine Controlled Analysis using SPC Methods," Section 5.9," Design Shutdown Margin Calculations," Revision 1, 9/2/92.

~2.

NFS-ND-900," Nuclear Desig'n Procedures," Revision 4,8/20/96, Appendix A,

" Performing Nuclear Design Routine Controlled Analysis using SPC Methods," Section 8.2," Bundle Disorientation Calculations," Revision 0, 8/9/93.-

~

3.

FS-ND 900," Nuclear D'esign Procedures," Revision 4,8/20/96, Appendix A,

" Performing Nuclear Design Rputine Controlled Analysis using SPC Methods," Section 8.3," Fuel Assembly Dislocation Calculations," Revision 0, 8/10/93.

4.

NFS-ND-900," Nuclear Design Procedures," Revision 4,8/20/96, Appendix A,

" Performing Nuclear Design Routine Controlled Analysis using SPC Methods," Section 8.4, " Rod Withdrawal Error Calculations," Revision 0, 9/21/93.

5.

NFS-ND 900," Nuclear Design Procedures," Revision 4,8/20/96, Appendix A,

" Performing Nuclear Design Routine Controlled Analysis using SPC Methods," Section 8.5 " Control Rod Drop Accident Analysis," Revision 0, 7/22/94.

6.

NFS-ND-900," Nuclear Design Procedures," Revision 4,8/20/96, Appendix A,

" Performing Nuclear Design Routine Controlled Analysis using SPC Methods," Section 8.7," Standby Liquid Control System (SBLQ Worth Calculaticris, Revision 0,4/6/93.

7.

NFS ND-900," Nuclear Design Procedures," Revision 4,8/20/96, Appendix A,

" Performing Nuclear Design Routine Controlled Analysis using SPC Methods," Section 8.8," Loss of Feedwater Heating Transient Analysis,"

Revision 2, 5/20/94.

l 8.

NFS-ND-900," Nuclear Design Procedures," Revision 4,8/20/96, Appendix A.

" Performing Nuclear Design Routine Controlled Analysis using SPC

- Methods," Section 8,6," Reload Licensing Report," Revision 0,11/2/93.

f.

t/

Page 19 of 19 yz /

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