ML20198R145

Rev 1 to Quad Cities,Unit 2 Cycle 15 Colr

ML20198R145
Person / Time
Site:	Quad Cities
Issue date:	11/04/1997
From:	COMMONWEALTH EDISON CO.
To:
Shared Package
ML19313D060	List: ML20198R145 ML20198R191 ML20198R233 SVP-97-248, Forwards Original & Rev 1 to Qcs Units 1 & 2 Colr.Original Rept Was Not Submitted Due to Administrative Oversight. Rev 2 to EMF-96-177 & Proprietary Rev 1 to EMF-96-180(P), Encl.Rev 1 to EMF-96-180(P) Withheld
References
NUDOCS 9711130167
Download: ML20198R145 (80)

Text

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ATTACHMENT A Quad Cities Nuclear Power Station Units 1 and 2 Q2C15 Core Operating Limits Report, Revision i NRC Docket Numbers 50 254 and 50-264 h

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Cora Opercting Limits RIport October 1997 i ISSUANCs OF CHANGES

SUMMARY

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IAffected '

Affected Section Pages Summary of Changec Date :

All : All Original Issue (Cycle 11). 10/89 All All Original Issue (Cycle 12), 11/91 Reissue. 2/92 All . All Originalissue (Cycle 13). 2/93 2 2 Added Section 2.3 on SLO. 3/93 References, iii,4 & 5 Revised References and Section 4.2 and added 4/94 4.2 & 5 Section 5.0 on Analytical Methods.

All All Originalissue (Cycle 14). 6/95 All All Latest Date Revised added to each COLR 5/96 page, Added SpecialInstructions and boxed in TSUP References for TSUP Implementation.

Added Control Rod Withdrawal Block Equation for Single Loop Operation.

All All OriginalIssue (Cycle 15). 5/97

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All All 1. " Cycle 15" Changed to " Cycle 15 Resivion 1" 10/97

2. Added references 10,11 and 12 to support Cycle 15 Revision 1 Core Loading to-Reference Section on Page iv

3. Page Dates Changed Quad Cities Unit 2 Cycle 15 Revision 1 4- ~ . _ . . _. _

0 Core Operating Lirnits R: port October 1997

.. TABLE OF CONTENTS ,

S P EC I AL I N ST R U CTl O N S . . . . . .. . . . . . .. . . . . . . . . . . . . . . . . . . .. . . . . .. . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . .. . . . . . .. . . . . iii

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REFERENCES............................................................................................................lv L I ST O F TA B L E S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . V LI ST O F F I G U R E S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

1.0 CONTROL ROD WITHDRAWAL BLOCK INSTRUMENTATION....................1 1.1. TECHNICAL SPECIFICATION

REFERENCE:

....................................... 1

1.2. DESCRIPTION

(Two Loop Operation) ............. .......................... ......... 1

1.3 DESCRIPTION

(Single Loop Operation) ........ .............. ..................... 1 2.0 AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)......... 2 ,

2.1 TECHNICAL SPECIFICATION

REFERENCE:

. ...................................... 2 2.2 DESCRIPTlON:...................................................................................2 2.3 SINGLE LOOP OPERATION MULTIPLIER: .......................................... 2 3.0 LINEAFt HEAT GENERATION RATE (LHGR)..... ........... .......................... .. 3 3.1 1 ECHNICAL SPECIFICATION

REFERENCE:

...................................... 3 3.2 D E S C R I P TI O N : . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 4.0 MINIMUM CRITICAL POWER RATIO (MOPR) .............................................. 4 4.1 TECHNICAL SPECIFICATION

REFERENCE:

....................................... 4 4.2 D E S C R I PT I O N : . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 1

5.0 A N A LYTl C A L M ET H O D S . . . . . . . . . ... . . . .. . . . . . . . . . . . . . . . ... . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . .. . 5 6.0 ATTA C H M E NT S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 ATTACHMENT 1 (RELOAD ANALYSIS REPORT) ......................................... A1 ATTACHMENT 2 (PLANT TRANSIENT ANALYSIS REPORT) ....................... A2 11 Quad Citias Unit 2 Cycle 15 Revision 1 e i

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= Core Op2 rating Limits Rtport

- October 1997

- SI'ECIAL INSTRUCTIONS l

- 1. This Core Operating Limits Report (COLR) contair.s the applicable reactor core limits . l and operational information mandated by Technical Specifications Section 6.9.A.6.

- -When the COLR is referenced by appl; cable Technical Specificailons or procedures fer  ;

. Technical Specification compliance, a controlled copy of this report shall be used as the 1 official source of the applicable limit or requirement.

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Quad Cities Unit 2 Cycle 15 Revision 1

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Core Op:rcting Limits Rrport October 1997-l

REFERENCES:

1 l 1. - Commonwealth Edison Company and lowa Illinois Gas and Electric Company Docket  !

No. 50 265, Quad Cities Station, Unit 2 Facility Operating License, i License No. DPR-30. l

2. Letter from D. M. Crutchfield to All Power Reactor Licensees and Applicants, Generic Letter 8816; Concerning the Removal of Cycle Specific Parameter Limits from i

Technical Specifications.

3. Quad Cities Nuclear Power Station, Units 1 and 2, SAFER /GESTR - LOCA Loss of-Coolant Accident Analy:is, NEDC 31345P, Revision 2, July 1989 (as amended).

4. Lattice Dependent MAPLHGR Report for Quad Cities Nuclear Power Station, Unit 2, Reload 13 Cycle 14,24A5161-AA Revision 0, December,1994.

5. SPC Document, EMF-96-185(P) Rev.1," Quad Cities LOCA ECCS Analysis MAPLHGR Limits for ATRIUM 98 Fuel", March,1997,

6. SPC Document EMF-96~ DP) Rev.1 "Ouad Cities Extended Operating Domain >

(EOD) and Equipment Om f Jervice (EOCS) Safety Analysis for ATRIUM-98 Fuel",

September,1996.

7. SPC Document, EMF 96177 Rev. 2, NDIT970069, " Quad Cities Unit 2 Cycle 15 Reload Analysis", April,1997 (COLR Attachment).

8. SPC Document EMF-96180 (P) Rev.1 NDIT970068 " Quad Cities Unit 2 Cycle 15 Plant Transient Analysis", April,1997 (COLR Attachment)

9. SPC Letter JHR:97:203," Analysis to Support Operation with Reactor Pressure Less than the Design Pressure at Quad Cities", May 27,1997.

10. SPC Letter DEG 97:156," Quad Cities Unit 2 Cycle 15 Asses.sment of Licensing Analyses After Replacement of Failed Fuel Assembly (Revision)", September 30,1997.

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11. NDIT 970159 Revision 0," Quad Cities 2 Cycia 15A Design Basis Loading Plan (DBLP)",

September 29,1997.

12. BNDO:97 062, *OC2 CIS Licensing Analysis Application to C15A (Assuming YJ8327 is the Leaker and LYU336 is Reinserted)", September 30,1997.

iv Quad Cities Unit 2 Cycle 15 Revision 1 W - - . .- -- -- -

Core Operv.ing Limits R: port October 1997 b

LIST OF TABLES Table , Title Page 21 Maximum Average Planar Linear Heat Generation 21 Rate (MAPLHGR) vs. Average Planar Exposure for GE98-P8DWB310-9GZ-80M 145 T.

22 Maximum Average Planar Linear Heat Generation 2-2.

Rate (MAPLHGR) vs. Average Planar Exposure for GE9B-P8DWB299-11GZ 80M-145-T.

23- Maximum Average Planar Linear Heat Generation 23 Rate (MAPLHGR) vs. Average Planar Exposure for GE9B-P8DWB286-7G3.0-80M-145 T.

24 Maximum Ave. age Planar Linear Heat Generation 24 Rate (MAPLHGR) vs. Average Planar Exposure for GE98 P8DWB286-9GZ-80M-145-T.

.25 Maximum Average Planar Linear Heat Generation 2-5 Rate (MAPLHCR) vs. Average Planar Exposure for GE9B-P8DWB310-7G3.0-80M-145-T.

2-6 Maximum Average Planar Linear Heat Generation 26 Rate (MAPLHGR) vs. Average Planar Exposure for GE98-P8DWB308-10GZ1-80M 145-T.

27 Maximum Average Planar Linear Heat Generation 27 Rate (MAPLHGR) vs. Average Planar Exposure for GE10-P8HXB316-8GZ 100M 145 T, 28 Maximum Average Planar Linear Heat Generation 2-8 Rate (MAPLHGR) vs. Average Planar Exposure for GF10-P8HXB312-7GZ-100M-145-T.

29 Maximum Average Planar Linear Heat Generation 2-9 Rate (MAPLHGR) vs. Average Planar Exposure for SPCA9-372B-11GZH-ADV.

2 10 Maximum Average Planar Linear Heat Generation 2-9 Rate (MAPLHGR) vs. Average Planar Exposure for SPCA9 358B-11GZL-ADV.

3 41 Automatic Flow Control MCPRr Results, Base Case 4-2 and EOOS/EOD.

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Quad Cities Unit 2 Cycle 15 Revision 1 L ,

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Cora Operating Limits Rtport .

October 1997 '

< UST OF FIGURES i i

Figure : Title = Page

.3-1 Protection Against Power Transient LHGR Limit- 31 for ATRIUM-98 Offset 41 MCPR,, Manual Flow Control ~ 4-1 n

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4-vi 1 l-Quad Cities Unit 2 Cycle 15 Revision 1

( __, , , ., w .- , m- ~.-,-g- o w .-4 a m-n., iA7 946a = we ,- -wwy -y,-mm,--.qw-W.,.

y.iw.p.--y g g g. zyy e-wr ? " qP e'T E D w V

Core Oper: ting Limits R: port October 1997-1.0 CONTROL ROD WITHDRAWAL BLOCK INSTRUMENTATION 1.1. TECHNICAL SPECIFICATION

REFERENCE:

Technical Specification: Table 3.2.E-1 [CCLR 1.2), 3.6.A.1.c [COLR 1.3)

1.2. DESCRIPTION

The Rod Withdrawal Block Monitor Upscale Instrumentation Trip Setpoint for Two Recirculation Loop Operation is determined from the following relationship:

s (0.65)Wd + 43% "

1.3. DESCRIPTION

The Rod Withdrawal Block Monitor Upscale Instrumentation Trip Setpoint for Single Recirculation Loop Operation (SLO) is determined from the following relationship, s (0.65)Wd + 39% "

" Clamped, with an allowable value not to exceed the allowable value for recirculation loop drive flow (Wd) of 100%.

Wd is the percent of drive flow required to produce a rated core flow of 98 million Ib/hr. Trip level setting is in percent of rated thermal power (2511 MWth).

Page1 Quad Cities tlnit 2 Cycle 15 Revision 1

Cors Opercting Limits R: port October 1997 2.0 AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) 2.1 ' TECHNICAL SPECIFICATION

REFERENCE:

l Technical Specification 3.11.Al

2.2 DESCRIPTION

- MAPLHGR versus Average Planar Exposure for GE9B-P8DWB310-9GZ 80M-145-T is determined from Table S-1.

MAPLHGR versus Average Planar Exposure for GE9B-P8DWB299-11GZ 80M-145 T is determined from Table 2 2.

MAPLHGR versus Average Planar Exposure for GE98-P8DWB286-7G3.0-80M-145-T is determined from Table 2-3.

MAPLHGR versus Average Planar Exposure for GE9B-P8DWB286-9GZ 80M-145-T is determined from Table 2-4.

MAPLHGR versus Average Planar Exposure for GE98-P8DWB310 7G3.0-80M-145-T is determined from Table 2 5.

MAPLHGR versus Average Planar Exposure for GE98-P8DWB308-10GZ1-80M-145-T is determined from Table 2-6.

MAPLHGR versus Average Planar Exposure for GE10-P8HXB316-8GZ-100M-145-T is determined from Table 2 7.

MAPLHGR versus Average Planar Exposure for GE10-P8HXB312-7GZ-100M-145-T is determined from Table 2-8.

MAPLHGR versus Average Planar Exposure for SPCA9-3728-11GZH-ADV is determined from Table 2-9.

MAPLHGR versus Average Planar Exposure for SPCA9-3588-11GZL-ADV is determined from Table 2-10.

- 2.3 SINGLE LOOP OPERATION MULTIPLIER:

The tabulated values are multiplied by 0.85 for GE fuel and 0.90 for SPC fuel

- whenever Quad Cities enters Single Loop Operation.

Page 2 Quad Cities Unit 2 Cycle 15 Revision 1

Core Op; rating Limits R: port  !

October 1997 -

TABLE 2-1 MAPLHGR vs.~ AVERAGE PLANAR EXPOSURE FOR BUNDLE TYPE : GE98-P3DWB310 9GZ-80M-145 T LATTICE 731 : P8DWLO71-NOG-80M T LATTICE 847 : P8DWL:rM-7G3.0-80M T LATTICE 848 : P8D%. d) 7G3.0-80M T j LATTICE 849 : P8DWL354274.0/7G3.0-80M-T I l

LATTICE 846 : P8DWLO71-9GE1-80M-T AVERAGE MAPLHGR LIMITS (KW/FT)

PLANAR EXPOSURE 731 847 848 849 846 (GWd/ST) -

0.00 11.64 12.25 11.78 11.31 11.64 0.20 11.57 12.32 11.85 11.39 11.57 1.00 11.38 12.46 11.99 11.55 11.38 2.00 11.36 12.62 12.19 11.78 11.36 3.00 11.41 12.79 12.36 12.02 11.41 4.00 11.49 12.96 12.52 12.22 11.49 5.00 11.56 13.14 12.69 12.44 11.56 6.00 11.63 13.23 12.81 12.60 11.63 7.00 11.69 13.30 12.93 12.77 11.69 8.00 11,74 13.38 13.04 12.93 11.74 9.00 11.78 13.43 13.14 13.08 11.78 10.00 11.81 13.46 13.21 13.20 11.81 12.50 11.54 13.41 13.21 13.21 11.54 15.00 11.16 13.02 12.94 12.93 11.16 20.00 10.37 12.28 12.30 12.29 10.37 25.00 9.58 11.57 11.61 11.60 9.58 35.00 8.01 10.27 10.29 10.27 8.01 43.70 4.71 - - -

4.71 45.00 -

9.02 8.98 8.92 -

51.10 - - -

5.90 -

51.20 - 5,96 5.90 - -

Page 2-1 Quad Cities Unit 2 Cycle 15 Revision 1

Core Opercting Limits R: port October 1997 TABLE 2-2 MAPLHGR vs AVERAGE PLANAR EXPOSURE FOR BUNDLE TYPE : GE98-P8DWB29911GZ 80V '45-T LATTICE 731 : P8DWLO71-NOG-80M-T LATTICE 850 : P8DWL322 9G3.0-80M-T LATTICE 851 : P8DWL337 9G3.0-80M-T LATTICE 852 : P8DWL337 2G4.0/9G3.0-80M-T LATTICE 6G3 : P8DWLO71-11GE 80M-T AVERAGE MAPLHGR LIMITS (KW/FT)

PLANAR EXPOSURE 731 850 851 852 853 (GWd/ST) 0.00 11.64 . 11.49 11.31 10.87 11.64 0.20 11.57 11.52 11.38 10.96 11.57

- : .00 11.38 11.62 11.56 11.14 11.38 2.00 11.36 11.87 11.84 11.44 11.36 3.00 11.41 12.19 12.05 11.70 11.41 4.00 11.49 12.57 12.21 11.88 11.49 5.00 11.56 12.94 12.37 12.07 1146 6.00 11.63 13.04 12.54 12.27 1 1, 13 7.00 11.69 13.14 12.73 12.50 11.69 8.00 11.74 13.21 12.90 12.72 11.74 9.00 11.78 13.26 13.03 12.92 11.78 10.00 11.81 13.28 13.11 13.07 11.81 12.50 11.54 13.21 13.07 13.06 11.54 15.00 11.16 -12.82 12.81 12.81 11.16 20.00 10.37 12.07 12.30 12.30 10.37 25.00 9.58 11.36 11.66 11.64 9.58 35.00 8.01 10.07 10.29 10.27 8.01 43.70 4.71 - - -

4.71 45.00 -

8.61 8.85 8.82 -

50.60 -

5.89 - - -

51.00 - - -

5.84 -

51.20 - -

5.81 - -

Page 2-2 Quad Cities Unit 2 Cycle 15 Revision 1

Cora Operating Limits Rrport October 1997 TABLE 2-3 MAPLHGR vs. AVERAGE PLANAR EXPOSURE FOR BUNDLE TYPE : GE98-P8DWB286-7G3.0-80M 145-T LATTICE 731 : P8DWLO71 NOG-80M T-LATTICE 1059 : P8DWL306 7G3.0-80M-T LATTICE 1060 : P8DWL324 7G3.0-80M-T LATTICE 1061 : P8DWLO717GE1-80M T AVERAGE MAPLHGR LIMITS (KW/FT)

PLANAR EXPOSURE 731 1059 1060 1061 (GWd/ST) 0.00 11.64 12.38 11.95 11.64 0.20 11.57 12.45 12.03 11.57 1.00 11.38 12.54 12.11 11.38 2.00 11.36 12.65 12.30 11.36 3.00 11.41 12.75 12.57 11.41 4.00 11.49 12.85 12.88 11.49 5.00 11.56 12.94 12.97 11.56 6.00 11.63 12.95 13.01 11.63 7.00 11.69 12.97 13.09 11.69 8.00 11.74 13.01 13.17 11.74 9.00 11.78 13.05 13.21 11.78 10.00 11.81 13.09 13.23 11.81 12.50 11.54 13.04 13.16 11.54 15.00 11.16 12.71 12.79 11.16 20.00 10.37 12.06 12.06 10.37 25.00 9.58 11.43 11.37 9.58 35.00 8.01 10.04 10.12 8.01 43.66 4.71 - - 4.71 45.00 -

8.60 8.68 -

50.66 -

5.84 - -

50.75 - -

5.91 -

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Page 2 3 Quad Cities Unit 2 Cycle 15 Revision 1

Cora Operating Limits R port October 1997 TABLE 2-4 MAPLHGR vs. AVERAGE PLANAR EXPOSURE FOR BUNDLE TYPE GE98 P8DWB286-9GZ-80M 145-T LATTICE 731 : P8DWLO71 NOG-80M T LATTICE 1059 : P8DWL306 7G3.0-80M-T LATTICE 1060 : P8DWL324 7G3.0-80M-T LATTICE 1000 : P8DWL324 2G4.0/7G3.0-80M-T LATTICE 1001 : P8DWLO719GE 80M-T AVERAGE MAPLHGR LIMITS (KW/FT)

PLANAR EXPOSURE 731 1059 1060 1000 1001 (GWd/ST) 0.00 11.64 12.38 11.95 11.42 11.64 0.20 11.57 12.45 12.03 11.51 11.57 1.00 11.38 12.54 12.11 11.59 11.38 2.00 11.36 12.65 12.30 11.82 11.36 3.00 11.41 12.75 12.57 12.13 11,41 ,

4.00 11.49 12.85 12.88 12.51 11.49 5.00 11.56 12.94 12.97 12.69 11.56 6.00 11.63 12.95 13.01 12.80 11.63 7,00 11.69 12.97 13.09 12.92 11.69 8.00 11.74 13.01 13.17 13.04 11.74 9.00 11.78 13.05 13.21 13,14 11.78 10.00 11.81 13.09 13.23 13.20 11.81 12.50 11.54 13.04 13.16 13.14 11.54 15.00 11.16 12.71 12.79 12.77 11.16 20.00 10.37 12.06 12.06 12.04 10.37 25,00 9.58 11,43 11.37 11.36 9.58 35.00 8.01 10.04 10.12 10.11 8.01-43.66 4.71 - - -

4.71-45.00 -

8.60 8.68 8.65 -

50.66 -

5.84 - 5.92 -

50.75 - - 5.91 -

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l Quad Cities Unit 2 Cycle 15 Revision 1 l

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Cors Opercting Limits R: port ,

October 1997 TABLE 2 5 .

MAPLHGR vs. AVERAGE PLANAR EXPOSURE FOR BUNDLE TYPE : GE98-P8DWB310-7G3.0-80M 145-T LATTICE 731 : P8DWLO71 NOG 80M-T LATTICE 1644 : P8DWL334 7G3.0-80M T LATTICE 1645 P8DWL350-7G3.0-80M-T l LATTICE 1004 : P8DWLO717GE-80M-T AVERAGE MAPI.HGR LIMITS (KW/FT)

PLANAR EXPOSURE 731 1644 1645 1004 (GWd/ST) 0.00 11.64 12.25 11.78 11.64 0.20 11.57 12.32 11.85 11.57 1.00 11.38 12.46 11.99 11.38 2.00 11.36 12.62 12.19 11.36 3.00 11.41 12.79 12.36 11.41

!2.52 11.49 4.00 11.49 12.96 5.00 11.56 13.14 12.69 11.56 6.00 11.63 13.23 12.81 11.63 7.00 11.69 13.30 12.93 11.69 8.00 11.74 13.38 13.04 *1.74 9.00 11.78 13.43 13.14 11.78 10.00 11.81 13.46 13.21 11.81 12.50 11.54 13.41 13.22 11.54 15.00 11.16 13.03 12.95 11.16 20.00 10.37 12.29 12.31 10.37 25.00 9.58 11.58 11.62 9.58 35.00 8.01 10.28 10.31 8.01 43.66 4.71 - -

4.71 45.00 -

9.04 9.01 -

51.41 - -

5.88 -

51.43 -

5.94 - -

Page 2-5 Quad Cities Unit 2 Cycle 15 Revision 1

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Cora Operating Liraits Report October 1997 TABLE 2-6 MAPLHGR vs. AVERAGE PLANAR EXPOSURE FOR BUNDLE ' TYPE : GE98-P8DWB50810GZ1-80M-145-T LATTICE 731 : P8DWLO71-NOG 80M T LATTICE 1642 : P8DWL332-8G4.0/2G3.0 80M-T 2

LATTICE 1669 : P8DWL348-8G4.0/2G3.0-80M-T LATTICE 1188 : P8DWLO71 10GE-80M T AVERAGE MAPLHGR LIMITS (KW/FT)

PLANAR EXPOSURE 731 1642 1669 1188 (GWd/ST) 0.00 11.64 11.63 11.24 11.64 0.20 11.57 11.69 11.31 11.57 1.00 11.38 11.83 11.44 11.38 2.00 11.36 12.04 11.62 11.36 3.00 11.41 12.25 11.82 11.41 4.00 11.49 12.48 12.02 11.49 5.00 11.56 12.58 12.24 11.56 ,

6.00 11.63 12.69 12.44 11.63 7.00 11.69 12.86 12.62 11.69 8.00 11.74 13.04 12.75 11.74 9.00 11.78 13.19 12.90 11.78 10.00 11.81 13.31 13.05 11.81 12.50 11.54 13.32 13.14 11.54 15.00 11.16 12.96 12.92 11.16 20.00 10.37 12.22 12.25 10.37 25.00 9.58 11.53 11.57 9.58 35.00 8.01 10.24 10.26 8.01 43.66 4.71 - -

4.71 45.00 -

8.93 8.93 -

51.09 -

5.96 - -

51.15 - -

5.90 -

Page 2-6 Ouad Cities Unit 2 Cycle 15 Revision 1 x .- . _ _ -

l Coro Opercting Lirnits Report October 1997 TABLE 2 7.

I MAPLHGR vs. AVERAGE PLANAR EXPOSURE FOR BUNDLE TYPE : GE10-P8HXB316-8GZ-100M 145-T LATTICE 7400 : P8HXLO71-NOG-100M LATTICE 7467 : P8HXL341-6G4.0/2G3.0-100M LATTICE 7469 : P8HXL357 6G4.0/2G3.0100M LATTICE 7468 : P8HXL341-8G3.0-100M LATTICE 7404 : P8HXLO718GE-100M AVERAGE MAPLHGR LIMITS (KW/FT)

PLANAR EXPOSURE 7400 7467 7469 7468 7404 4 (GWd/ST) 0.00 11.85 12.00 11.11 12.08 11.85 0.20 11,78 12.06 11.14 12.15 11.78 1.00 11.59 12.18 11.24 12.30 11.59 2.00 11.57 12.36 11.44 12.53 11.57 3.00 11.61 12.50 11.70 12.68 11.61 4.00 11.68 12.60 11.99 12.82 i1.68 5.00 11.75 12.71 12.26 12.96 11,75 6.00 11.81 12.84 12.37 13.12 11.81 7.00 11.86 13.01 12.51 13.29 11.86 8.00 11/91 13.20 12.68 13.44 11.91.

9.00 11.94 13.39 12.86 13.53 11.94 10.00 11.97 13.52 13.01 13.55 11.97 12.50 11.75 13.44 13.09 13.44 11.75 15.00 11.38 13.06 12.84 13.07 11.38 20.00 10.59 12.32 12.21 12.33 10.59 25.00 9.81 11.60 11.54 11.61 9.81 35.00 8.26 10.21 10.25 10.22 8.26 44.89 4.93 - - -

4.93 45.00 -

8.72 8.82 8.72 -

50.76 -

- 5.87 -

50.78 -

5.86 - - -

50.08 - - 5.90 - -

Faye 2-7 Quad Cities Unit 2 Cycle 15 Revision 1 l

Coro Operating Limits R: port October 1997 l l

TABLE 2-8 MAPLHGR vs. AVERAGE PLANAR EXPOSURE FOR BUNDLE TYPE : GE10-P8HXB312 7GZ 100M-145-T LATTICE 7400 : P8HXLO71 NOG-100M LATTICE 7405 : P8HXL336-3G4.0/4G3.0-100M LATTICE 7406 : P8HXL3541G4.0/6G3.0-100M LATTICE 7407 : P8HXL336 7G3.0-100M LATTICE 7408 : P8HXLO717GE 100M AVERAGE V5 P'.HM LIMITS (KW/FT)

PLANAR EXPOSURE 7400 7405 7406 7407 7408 (GWd/ST) 0.00 11.85 12.01 11.27 12.04 11.85 0.20 11.78 12.08 11.31 12.11 11.78 1.00 11.59 12.23 11.42 12.27 11.59 2.00 11.57 12.43 11.65 12,49 11 37 3.00 11.61 12.65 11.93 12.72 11.61 4.00 11.63 12.88 12.24 12.96 11.68 5.00 11.75 13.09 12.58 13.15 11,75 6.00 11.81 13.22 12.94 13.30 11.81 7.00 11.86 13.32 13.15 13.41 11 S6 8.00 11.91 13.40 13.32 13.46 11.91' 9.00 11.94 13.45 13.43 13.47 11.94 10.00 11.97 13.47 13.50 13.45 11.97 12.50 11.75 i3.35 13.45 13.35 11.75 15.00 11.38 12.97 13.10 12.97 11.38 20.00 10.59 12.23 12.41 12.24 10.59 25.00 9.81 11.51 11.74 11.52 9.81 35.00 8.26 10.14 10.41 10.15 8.26 44.89 4.93 - - -

4.93 45,00 - 8.61 9.01 8.60 -

50.55 - - - 5.85 -

50.59 - 5.85 - - -

51.56 - -

5.86 - -

Page 2-8 Quad Cities Unit 2 Cycle 15 Revision 1

Core Op;r ting Limits R: port October 1997 TABLE 2 9 MAPLHGR vs. AVERAGE PLANAR EXPOSURE l FOR BUNDLE TYPE : SPCA9-372811GZH-ADV l

Average Planar ATRIUM-9B Exposu.'e MAPLHGR (kW/ft)

(GWd/MTU) 0.0 13.5 20.0 13.5 l 60.0 8.7 i

4 TABLE 2-10 MAPLHGR vs. AVERAGE FLANAR EXPOSURE FOR BUNDLE TYPE : SPCAE4588-11GZbnDV Average Planar ATRIUM-9B Exposure MAPLHGR (kW/ft)

(GWd/MTU) 0.0 13.5 20.0 13.5 60.0 8.7 T

Page 2-9 Quad Cities Unit 2 Cycle 15 Revision 1

a Core Operating Limits Report October 1997

- 3.0 . LINEAR HEAT GENERATION RATE (LHGR) 3.1 -TECHNICAL SPECIFICATION

REFERENCE:

Technical Specification 3.11.Dl --

3.2 . DESCRIPTION:

A. The LHGR limit is 14.4 Kw/ft for all GE fuel types:

1. GE9B-P8DWB310-9GZ-80M 145 T

2. GE9B-P8DWB299-11GZ-80M-145 T

3. GE9B-P8DWB286-7G3.0-80M-145-T

4. GE9B-P8DWB286 9GZ-80M 145-T

5. GE98-P8DWB310 7G3.0-80M-145-T

6. GE98-P8DWB308-10GZ1-80M 145-T

7. GE10-P8HXB316-8GZ-100M-145-T

8. GE10-P8HXB312 7GZ-100M 145 T B. The LHGR limits are provided in the table below for all of the SPC fuel types:

1. SPCA9-3728-11GZH-ADV

2. SPCA9-3588-11GZL-ADV Average Planar ATRIUM-98 LHGR Exposure (kW/ft)

_(GWd/MTU) 0.0 14.4 15.0 14.4 61.1 8.32 The Protection Against Power Transient LHGR Limit for ATRIUM-9B Offset fuel is provided in Figure 3-1.

Page 3 Quad Cities Unit 2 Cycle 15 Revision 1

Core Op:reting Umits Rrpo.1 Octob:r 1997 n

3 (0.19.4) (15.5.4)

W-

. I 4

3 ,

%= . '.

_q. =

S=

l l

6- ,

i 4- '

1 .

0 0 5 N N ID IB N M Ad U IO $ A0 N 70 MerOgg PlcrMIr @caurs,6 .

QUAD CITIES UNIT 2 CYCLE 15 REVISION 1 PROTECTION AGAINST POWER TRANSIENT LHGR LIMIT FOR ATRIUM-98 OFFSET FIG'JRE 3-1 1

Page 3-1 Quad Cities Unit 2 Cycle 15 Revision 1

Core OperIting Limits RIport October 1997 "1

4.0 MINIMUM CRITICAL POWER RATIO (MCPR) .

4.1 - TECHNICAL SPECIFICATION

REFERENCE:

l Technical Specification 3.11.Cl

4.2 DESCRIPTION

During steady state operation at rated core flow, the Operating Limit MCPR (OLMCPR) shall be greater than or equal to:

1.58 for GE9B fuel .

1.53 for GE10 fuel 1.48 for ATRIUM-9B fuel for:

5% insertion . tave s 0.375 seconds '

20% insertion tave s 0.90 seconds 50% insertion tave s 2.00 seconds 90% insertion tave s 3.50 seconds where tave = Technical Specification scram insertion times for all surveillance data from Technical Specification 3/4.3.E which has been generated in the current cycle. Where t... is less than or equal to the Technical Specification 5, cram Speed.

For core flows less than rated, reduced flow MCPR, curves for Manual Flow Control are provided in Figure 41. MCPR, values for Automatic Flow Control are provided in Table 41.

The OLMCPR limits stated above are valid for all planned Operational modes, including increased Core Flow (ICF) and Final Feedwater Temperature Reduction (FFTR). The value corresponds to the cycle specific determination of the t ounding event. For Unit 2 Cycle 15 Revision 1 this event is the Feedwater Controller Failure Event with the above Operational modes incorporated. This value was determined to be 1.48 for the ATRIUM-9B bundles,1.53 for the GE10 bundles, and 1.58 for the GE9B bundles, for the Technical Specification Scram Speeds. This limit is increased by 0.04 for any and all combinations of Quad Cities Equipment Out-Of Service / Extended Operating Domain analysis (Reference 6) excluding Single Loop Operation (SLO). For SLO, the OLMCPR is increased by 0.01. The above OLMCPR values include a 0.01 adder to permit operation with reactor steam dome pressure up to 15 psi below rated pressure from 90% to 100% CTP. Below 90% CTP, the 15 psi band requirement is not required (Reference 9).- Unit Two has been approved for operating up to 15%

(Reference 6) above equilibrium coastdown power le el with allowance for multiple control rods inserted. At End of Full Power Capability (EFPC) the generic EOOS/EOD MCPR operating limit MCPR penalty of 0.04 must be added to the operating limit MCPR in order to exceed equilibrium xenon coastdown power.

Page 4 Quad Cities Unit 2 Cycle 15 Revision 1

Cora Op:rt. ting Limits R: port October 1997 .

, E9 MCPR(f) Umit -

2.4 -

. E10 MCPR(f) Umit ATREAMB Offast .

ICPR(f) Umit -

+ e nasuits  ;

]u x GE10 Reedts .

a -

.- a mu-a offset .

2.0 - nasults .

a  ; - -

m 1.s -

.9 - -

h. -

a e -

1 1.s u .

a .

. +

l a:'

ta - a t . .

1: -

l i.o ". , . . . . . , , , . , , , , , , , , , , , , , , , , . .-

0 2D 40 W g 100 12D Totd Core Flow (= Rated) .

QUAD CITIES UNIT 2 CYCLE 15 REVISION 1 l MCPR; CURVE FOR MANUAL FLOW CONTROL FIGURE 4-1 i

Page 4-1 Quad Cities Unit 2 Cycle 15 Revision 1

Core Operating Limits Report _- l October 1997 i

.l Pa9e 4 Tat'le 4-1 Automatic Flow' Control MCPRr Results Ray Casa. OLMCPR EOD/EOOS OLMCPR Core Flow

(% rated) faER GfiQ A1TRIUM-98 fzEt GE10 ATTRIUM-93

-108 1.58 1.53 1.48 1.62 1.57- 1.52 30 2.95 2.85' 2.80 3.02 2.92 2.88 0 3.85 3.72 3.82 3.95 3.82 3.75 Quad Cities Unit 2 Cycle 15 Revision 1

Core Operating Limits Rtport .

October 1997 I

- 5.0 Analytical Methods -l I

The analytical tr,ethods used to determine the core operating limits shall be those  !

prevlevsly reviewed and approved by the NRC in the latest approved revision or supplement of the topical reports describing the methodology. For Quad Cities Unit 2, the topical reports are:

(1) GE Document, NEDE 24011 P-A ' General Electric Standard Application for Reactor Fuel," (latest approved revision).

(2) Commonwealth Edison Topical Report NF9R-0091, ' Benchmark of BWR Nuclear Design Methods ' (latest approvec revision).

(3) Commonwealth Edison Topical Report NFSR 0091 Supplement 1," Benchmark of CASMO/MICROBURN BWR Nuclear Design Methods", (As Supplemented)

(4) SPC document," Application of the ANFB Critical Power Correlation to Coresident GE Fuel for Quad Cities Unit 2 Cycle 15", EMF-96-051(P), May, 1996.

t 4

Page5 Quad Cities Unit 2 Cycle 15 Revision 1 y .a y 4 y -

m --w

Cora Oper!. ting Limits R: port October 1997 6.0 Attachments Attachment 1. Quad Cities Unit 2 Cycle 15 Reload Analysis Attachment 2. Quad Cities Unit 2 Cycle 15 Plant Transient Analysis 4

1 l

1 i

Page 6 Quad Citics Unit 2 Cycle 15 Revision 1

NFS:BSS:97-119 Date: October 7,1997 To: R. Fairbank -

T. Ciesta

1. Johnson

Subject:

Revised 02C15revi 10CFR50.59 Safety Evaluation

Reference:

NFS:BSS:97-113," Revised O2C15 COLR to Address Revision 1 Changes to -

Cycle 15",' (Attachments Q2C15revi COLR and associated 50.59), E.A. -McVey, October 2,1997.

The purpose of this memorandum is to transmit the revised 02C15revi 10CFR50.59 Safety Evaluation due to incorporation of station comments, This attached Safety Evaluation shs.ll replace the previous 10CFR50.59 Safety Evaluation issued with the Reference memo. Please -

ensure that this new O2C15revi 10CFR50.59 Safety Eval.sation is used for the on-site and off-site evaluation.

If you have any questions, please contact Nancy Buck at x3099

~ b%f Edward. A. McVey BWR Support Services Supervisor Nuclear Fuel Services EAM/NJB Attachment CC:

John K. Wheeler - NFS:BND Martin Santic - Quad Cities (w/o attachmunt)

Robert Canalas/ James Piekut - Quad Cities Mark E. Wagner - DG Licensing (w/o attachment)

Robert W. Tsai/J.M. Freeman - NFS.BSA (wlo attachment)

J. E. Thompson - NFS:BSS (w/o attachment)

R. J. Chin - NFS (w/o attachment)

A. D. Pallotta - NFS:BND (w/o attachment)

John A Silady - NFS (w/o attachment)

File: O2C15revi Licensing BSS CF (w/o attachment)

NFS-CF Document ID: tpCMI - o ca i :3 c,$

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

NSHP.A 04 Reveron 0 10 CFR30.39 Safty Evaluanon Process EXHIBIT H Page 1 of 2 VALIDATION OF PREVIOUSLY PERFORME )

SAFETY EVALUATIONS AND SCREENINGU

1. Proposed Actinty Number (e g. Prmdure, Change,005, etc.). 07ei.eeeo t cet4

2. Station / Unit: Q..teicea/t Applicable Plant Modes: 4 il fystem(s) Involved: r.,i Equiptr.:nt EID(s): -A

3. Proposed Activity

Description:

Re .fu. .a of ,o kkf C..I wea/ 7

4. a. For doniments prepared in accordance with this promdure or an earlier mision, list applicable reference or location number: ooiT oem

b. For documents not prepared using this procedure, list Originating Organization: M F.c Reference or Imauon W :

5. Pruious Actaity Type and Numberts):

De nL. t a r t ir cotR J

List niode restncuons, special condiuons, lineups, system interactions, additional outef senice equipment, etc. for the referenced Safety Evaluation or screemng:

0., . e .

6. Desenbe anf justify any differences betmen the current proposed actisity and the pmious activity resiewed in the applicable referenced doeurnent: Re,.L u ...+ 4 f.S t (s t m.J .y

7. Is a change to the UFSAR needed?

O Yes - UFS AR/ Changes have been iidtiated sia O 'Iracking Contro! No.:

OR O All applicable changes are attached.

Proceed to next step.

6 No- Proceed to next step.

I l

NSHP A 04 Rmsson 0 10 CFR30 39 Safrp' Evaluarson Process EXHIB!T H Page 2 of 2 VALIDATION OF PREVIOUSLY PERFORMED SAFETY EVALUATIONS AND SCREENINGS

8. The preparer ensures that the referenced or .tta:hed 10CF'R50.59 Screemng or Safety Evaluauon meets these enterw a The prnious Screemog or Safety Evaluation had been performed in accordance with NOD TS 11.

or a procedure that is at least as ngorous.

b. The current proposed actnity is within the scope of the prnsous Screening or Safety Evaluation.

c. The current methity does not extend beyond the plant mode bounds assumed in the prnious Screening or Safety Evaluation.

d. There are no Outef Senices, equipment lineups, Temporary Alterations or Modifications in place ,

that would invalidate the prmous Screemns or Safety Evaluation. This inchutes systems or components that interact with or affect the equipment being n sluated.

c. There Are no other plant changes taking place at the rame timo that could invalida e the conditions assumed in the prnious Screcrung or Safety Evaluation.

f. The conclunons are reasonable, well supported and documented.

Preparer: \; 8:M /I de ,

Date: t/2rls ?

Pnnt / 5 gnature Dept. / Location: 3 e s[ %J c:+:e c Phone: 2tn i

The rniewer agrees that the conclusions are reasonable, well supported and documented. ,

Rniewer: %A F 9dmdet [ #

Date; o/w/f/

r l Print i Signature l

Dept. / Location: Mb/ ne4 & l Tr e Phone: 2R2 o I

in ni C

, ~ILJ R5 Beur~ 5 kmEld% (!)lk(GVd e =M~ S'fr "YV

l O \ tXi hAo l

l 1

NSWP A 04Reecon 0 10 CFR30.39 Saktv Evalanarson Process l EXHIBIT H Page 2 of 2 VALIDATION OF PREVIOUSLY PERFORMED SAFETY EVALUATIONS AND SCREENINGS

8. The preparer ensures that the referenced or attached 10CFR$0.59 Screemng or Safety Evaluauon meets these entena:

a. The prmors Screemns or Safety Evaluation had been perfonnM in accordance uith NOD TS 11.

or a procedure that is at least as ngorous.

b. The current proposed activity is within the scope of the preious Scrwmng or Safety Evaluation.

c. The current actnity does not extend beyond the plant mode bounds assumed in the prmous Screening or Safety E*eluation.

d. There are no Out+f Senices, equipment lineups. Temporary Alterations or Modificatiens in place that would invalidate the prmous Screemag or Safety Evaluation.1his includes syste.m or components that interact with or affect the equapment being evaluated.

c. There are no other plant changes taking place at the same time that could invalidate the condit;ons assumed in the prnious Screemag or Safety Evaluation.

L The conclusions are reasonable, well supported and documented.

Preparer: \; #1ek/

t

/ b.s- Date: t/2rh1 Pnnt (gnature Dept. / Location: 5P AJ Cr+:e c Phone: 7 s't3 l

The reslewer agrees that the conclusions are reasonable, well supported and documented. f Resiewer: %A F. 9Amula / h./ Dr.te: <0/w/f7

1. /

Print i Signature Dept./ Location: SEb / na A C'I Tr e Phone. Q 8,2 0

/

'lld 0L 0Nw $l%ElM 3e  % "l' W1' L'116-(RWf Q) t)f. 3h,\ o

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

h NSIl'P.A-04 Reviskm 0 10 CFR${FSV Sqfety Evaluathm Procen EXHIBIT G 10CFR50.59 SAFETY EVALUATION Sarely livaluation Tracking No.: ._,_$J 'udIP - - -

List the documents implementing the proposed change. (Mod #, Temporary Alt # Procedure #, DCR #. NWR #,

etc.):

NFS:BND:97-088, Comed to Mr. D. E, Garber, Siemens Power Corp., " Quad Cities 2 Cycle 15 -Validity of Licensing Analyses after Core Alterations to Replace Failed Fuel Assembly", August 27,1997, NFS:BND:97-094, Comed to D. E. Garber, S;emens Power Corp., " Quad Cities 2 Cycle 15 - Core Follow and Target Rod Patterns Data to Support Evaluation of Licensing Analyses", September 12,1997.

NDIT 970175 Revision 0. Letter, D. E. Garber to Dr. R. J. Chin (DEG:97:156),-

dated October 1,1997 Subject " Quad Cities Unit 2 Cycle 15 Assessment of Licensing Analyses After Replacement of Failed Fuel Assembls;(Revision)".

NDIT 970159 Revisio'i 0, " Quad Cities Unit 2 Cycle 15A Design Basis Loading Plan (DBLP)", September 29,1997.

Stationtnit: Quad Cities / 2 Applicable Modes: Al.L Other Reles ont Plant Conditions (Permanentiemporary etc.)

Cycle 15 Hetision i Systen (s) nfreeted: Q2Cl5 Core liquipment Name(s): Q2C15revi Core

!!PN: Not Applicable N OTI'.

A Consider the need for this methit) on the other Unit or Train ir not already addressed.

l a. Describe the proposed activity.

The proposed change is to discharge a failed GE10 fuel assembly from the Quad Cities Unit 2 core.

make three assembly shuffles within the core and reinsert a GE9B fuel assembly into the core. All methodologies ar'd fuel types used are the same as the previous reload analysis for Cycle 15 This 10CFR50.59 evaluation addresses these changes to the original Cycle 15 core due to the leaking fuel assembly. A detailed description and anaiysis for the original Cycle 15, including the 10CFR50 69 evaluation, is desenbed in the Q2C15 Reload Evaluation (Reference 11). If fuel sipping determines the failed bundle (s) to be other than the suspect GE10 assembly, a revision to this evaluation will be provided, with relevant support analyses from Nuclear Design and Siemens j

Power Corporation.

l

l l

l 1

' NSHN 04 Revision 0 l t 10 CFR$0 39 Sakty Evaluation l'rocen i '

l This change will happen approximately 1750 MWD /MT into Cycle 15. in order to keep all documentation for Quad Cities Unit 2 Cycle 15 before the change separate from docv'N uca j  ;

after the change (which willinclude a revised core map and target rod pattems) it i' r. #M '

- expectation that Comed will be calhng the per failed fuel assembly replacement m QuM W - ,

i Unit 2 Cycle 15 Revision 1* (O2C15tevi) whw1 referring to the revised Cycle i' r m w,al / t communications. i The O2C15tevi core will consist of 216 ATRIUM-9B Offset fresh fuel assembhes, W e oted

GE10 (Offset) assemblies, and 365 Irradiated GE9B assembhes. The failed fuei assembly that is

expected to be removed from the core is of the GE10 design. The replacement fuel assemoly is of the GE90 design. The failed assembly will be removed from the core and placed in the spent fuel pool. Then three fuel assemblies will be moved to fill in the failed assembly's core location by moving bundles inward from adjacent core locations. Finally, a discharged GE9B assembly will be taken from the Spent Fuel Pool and placed in an edge core location to make up for the failed i

assembly that was removed and the shuffles inward. This prevloesly discharged assembly has been inspected by camera in the spent fuel pool to ensure that it is physically fit for reinsertion nd  !

was not damaged during its short duration in the spent fuel pool (Reference 15) Although the ,

previously discharged assembly is higher in exposure than the assembly it is replacing, it has teen i i reviewed to ensure that it will not reach its end of hfe exposure prior to EOC15revi Deference 9).

The methodology or basis for the thermallimits for 02C15revi will not change from 02C15. The hmiting event is still FWCF (Reference 9). Similar target control rod patterns have been developed

, for 02C15revi that demonstrate there will be adequate margin to thermal kmits and assembly exposure hmits will not be violated after the core alterations have been made (References 9 and 10). Therefore, the fuelin 02C15tevi will be protected from boihng transition,1% plastic strain, and

10CFR50A6 hmits, as was the previous reload.

A LOCA analysis has been performed for Quad Caties Unit 2 (Reference 6). LOCA analyses are not dependent on cycle design but fuel design. The LOCA analyses determines the fuel dependent i MAPLHGR limits. Since the 02C15 fuel design has not changed tha LOCA analyses will remain -

vahd for 02C15revi as will all MAPLHGR limits.

The change from the 02C15 core to the 02C15revi core is insignificant when considering design  ;

and analyses of the core.

The O2C15 COLR was revised to include the new references for 02C15revi and to rename the COLR from 02C15 to 02C15rev1. These changes were administrctive in nature and no thermal hmits or actual content in the COLR was changed. Therefore, this 10CFR50.59, which evaluates the revisions to the O2015 core, is also the basis for the revised COLR.

I b. Describe the reason for the proposed activity.

Approximately two months into Cycle 15 at Quad Cities Unit 2 offgas levels increased due to a leaking fuel asst mbly. Subsequent power suppression testing located the leaking assembly as a GE10 bundle (YJ8327) in core location 21 14. Fuel sipping will be performed to con 6rm the leaking fuel assembly, in order to complete Cycle 15, the leaking assembly must be removed from the core since further operation could lead to increased degradation of the failure leading to power reductions to maintain offgas levels. This redesign of the O2C15 core is to allow full power operation for the remainder of O2C15.

l' age 2 of $1

NSIt'I' A 04 Reviston 0 10 ClR30 39 Sqfety &aluation l' roms  :

[

EXHIBIT G f Page 2 of 10 i 10CFR50.59 SAFETY EVALUATION  ;

2a. 1.ist the SAR sections review ed. Including approved pending (Ji SAR changes, w hich describe the affected systems, struc tures, or components (SSCs) or activities. Also list the SAR accident ana'ysis sections that discuss the alTected SSCs or their operation.1,ist any other controlling documents such as SERs, previous modifications or Safety Evaluations, etc.

Chapter i Section 12.2.3

• Reactor System

• Chapter 4 Section 4.1

• Summary Description

• Section 4.2

• Fuel System Design

• Section 4.3

• Nuclear Design' Section 4.4

• Thermal and Hydraulic Design

• Chapter 6 Section 6.3

• Emergency Core Coohng System' ,

Chapter 7 Section 7.3

• Engineered Safety Feature Systems Instrumentation and Control' Chapter 9 Section 9.1

• Fuel Storage and Handling

• Section 9.3 5

• Standby Uquid Control System * ,

Chapter 10 Section 10.4.4

• Turbine Bypass System

• Chapter 15 Section 15.0

• Accident and Transient Analysis Section 15.1

• Increase in Heat Removal by the Reactor Coolant System

• Section 15 2 " Decrease in Heat Removal by the Reactor Coolant System

• Section 15.3

• Decrease in Reactor Coolant System Flow Rate

• Section 15.4 ' Reactivity and Power Distnbution Anomakes' Section 15.5

• Increase in Reactor Coolant Inventory

• Section 15.6

• Decrease in Reactor Coolant Inventory

• Section 15.7

• Radioactive Release from a Subsystem or Component

• Section 15 8

• Anticipated Transients Without Scram

• Pending UFSAR Changes - Reference 11 Pending Technical Specification Update - Reference 12 2b. State the SAR sections, paragraphs, or drawings that are impacted by this change.

The current SAR words and drawings, including pending updates (Reference 11), will not be impacted by this change since the UFSAR does not specify cycle specific details of number / type of fuel bundles. Analyses for 02C15 were venfred to remain applicable or were re-performed for 02C15revi (Reference 9 and 10 ).

2c. Describe the change to the SAR words and drawings resulting from the ;,roposed change described in item l a.

. ' None, 2d. Is this a:

@ Change to the facil;ty as described in the SAR or O Change to a procedure as described in the SAR P ge 3 of 51 1

. . . _ .m-. ._.. _ - .~ .. ._.7, , . _ _ y____.y_._,,,m,_ .-_ _ , ,,, ,. .__ ._,.y,,, ,, , , . - . ,,,,,, , _ , _ . , .

NSit1*-A-04 Reviston 0 10 ClR50 39 Safety Duluation l'nwess or O Test or esperiment M described in the SAR NOTI:

l'or design chu sges,1:shibit 1. Design issues Worksheets should lic resiewed and applicalile iterns should lie used to answer the remaining questions.

Page 4 of 51

NSWP A-04 Revision 0 10 CFR30.39 Safety &aluation Pnwess EXHIBIT G 10CFR50.59 SAFETY EVALUATION 1 Does the proposed activity involve an analog to digital conversion?

O YES- Review EPRI

• Guideline on Licensing Digital Upgrades" (t:PRI TR 102348) and continue to next Step.

@ NO. Continue to next Step.

4. Describe the functions of the afTected systems, structures or components.

The function of the Q2C15revi core is to provide enough energy to support Quad Cities 2 Cycle 15 operation while maintaining margh to the prescribed safety criteria.

5. Describe how the proposed activity will affect plant operation w hen the changed SSCs function as intended (i c., focus on system operation / interactions in the absence of equipment failures). Consider all applicable operating modes. Include a discussion of any changed interactions with other SSCs. For a test or experiment, discuss the impact on the safe operation of the plant of any new technique or new system configuration.

General Effects of Reload Fuel This is virtually the same reload as was used for 02C15 analysis. 02C15revi is a continuation of Cycle 15. The actual reactivity of the Q2C15revi reload is shghtly less than 02C15 due to a once burned GE10 bundle being removed from the core and replaced by a discharged thrice burnt GE9B i assembly. Core operating parameters (such as licensing energy, EOFP exposure, core average erposure, axial offset, void coefficient, void fraction and maximum radial power) have not significantly changed from O2015. Additionally, control rod patterns have been developed for Q2015tev1 to ensure the operating limits will be met through out the cycle. Therefore, removing the leaking fuel assembly and replacing it with a discharged fuel assembly will not adversely affect plant operations.

The replacement fuel bundle is a GE9B type assembly. There are no adverse effects to adding a GE9B assembly to the core as this type of fuelis already present in the O2015 core. The replacement fuel is also projected to be well below the peak pellet exposure kmit for GE9B fud at the end of O2C15rovi (Reference 9). Since MAPLHGR and LHGR hmits are fuel type dependent and not cycle dependent and have already been provided for GE90 fuel and the MCPR safety limit actually improves with the O2C15revi core loading, there will be no change in those thermal limits when adding this GE9B assembly to the core.

There will be no changed interactions with other plant systems. The replacement fuel assembly for 02C15tevi is of the same type as other assemblies currently in the O2C15 core. The channel on the replacement fuel assembly is also of the same type as other channels currently in the Q2C15 core so there will be no increased potential of control blade interference. The fuel shuffle is the only physical change to the plant for this modification so from a plant system perspective, the Q2C15revi core will be the same as the O2C15 core. No other safety related equipment is affected.

Page 5 of 31

-.+_-.---.-_ _ _ . - _ - -

NSif'P.A 04 Revision 0 10 CFR50.39 Sqkty Evaluation Process Effect of Chanae on Safety Analysis Methodoloav There will be no change in methodology for safety analyses from Q2C15 to 02C15revi. For 02C15, Comed and SPC determined the limiting Anticipated Operational Occurrence in the Reload Documents (the Reload Analysis and Plant Transient Analysis Document, References 7 and 8) and used the limiting event (s) to determine the cycle's thermallimits The AOOs analyzed for the O2C15 reload by SPC include generator load rejection with no bypass (LRNB), feedwater controller failure (FWCF), and recirculation pump slow flow run-up. The AOOs analyzed by Comed Nuclear Design for the 02C15 reload included RWE, Misplaced /Misoriented Fuel Assembly and LFWH.

SPC performed the 02C15 MCPR safety limit calculation with the input of up to 2 TIP machines out of service (or the equivalent number of TIP channels), up to 50% of the LPRMs out of service, and an LPRM calibration frequency of 2000 Effective Full Power Hours. The EOOS/EOD ana'/ sis and the cycle specific transient analysis were also performed with the assumption of up to 2 TIP machines out of service (or the equivalent number of TIP channels) and up to 50% of the LPRMs out of service in the standard, ICF, and El LLA regions on the power to flow map. Reference i provides the information supporting Quad Cities 2 Cycle 15 operation with 2 TIP machines out of service (or the equivalent number of TIP channels).

The EOOS/EOD scenarios that were analyzed for 02C15 were evaluated with a 10CFR50.59 safety analysis provided in Reference 14. This package found the SPC EOOS/EOD scenarios and the EOOS/EOD methodologies acceptable.

All these Safety Analyses were completed for 02C15 and have been venfied to be bounding for Q2C15revi via a comparison of key core parameters (Reference 10).

6. Describe how the proposed activity will affect equipment failures. In particular, describe any new failure modes and their ir, pact during applicable operating modes and applicable accident conditions.

No new modes of failure will be generated by replacing the leaker fuel assembly with a discharged

)

fuel assembly. The replacement fuel assembly for 02C15rev1 has been determined to meet the mechanical and thermal-hydraulic criteria established. Thermallimits have been established for each fuel type as required by the Quad Cities Technical Specifications. The replacement assembly has the same exterior channel as other assemblies cb7ently in the O2C15 core so control blade mo*/ement will not be affected. No other equipment is affected by the fuel replacement / shuffle.  ;

Therefore, this fuel replacement / shuffle will have no impact on the consequences, the probability of, nor the creation of any new equipment failure modes.

Page 6 of 51

NSIVP.A 04 Redston 0 10 CFR3039 S$ty Eyaluathm Pnwess EXHlBIT G 10CFR50.59 SAFETY EVALUATION

7. Identify each accident or anticipated transient, including LOCA and transient analysis, (i.e., large!small break LOCA, loss of load, turbine missiles, fire, flooding) described in the S AR w here any of the follow ing is true:

• The proposed activity alters the initial conditions used in the SAR analysis

• The changed SSC is explicitly or implicitly assumed to function during or after the accident Operation or failure of the changed SSC could lead to the accident ACCI Dl'_NT S AR SECTIOJ Decrease in Reactor Coolant inventory 156 Break in Reactor Coolant Pressure Boundary Instrument Line Outside Containment 15 6.2 Load Rejection (Generator Trip) Without Bypass 15.2.2.1 Increase in Feedwater Flow 15.1.2 Control Rod Drop Accident 15.4.10 Rod Withdrawal Error (at Power) 15.4.2 ASME Overpressurization Event 5.2.2.2 Decrease in Feedwater Heating 15.1.1 Mislocated Fuel Assembly Accident 15.4.7 Slow Flow Excursion 15 4.5*

Design Basis Fuel Handling Accidents inside Containment and Spent Fuel Storage Buildings 15.7.2 Anticipated Transient Without Scram 15 8

• The slow flow core excursion is being added to UFSAR Section 15.4.5 as part of Reference 11.

8. List each Technical Specification (Safety Limit, Limiting Safety System Setting or Limiting Condition for Operation) w here the requirement, associated action items, associated surveillances, or bases may be affected. To determine the factors alTecting the specification, it is necessary to review the SAR, including approved pending UTSAR changes, where the Bases Section of the Technical Specifications does not explicitly state the basis.

JFCllNICAL SPFCIFICATION SECTION 2.1.8 'Thennal Power, High Pressure And High Flow" 2.1.C

• Reactor Coolant System Pressute' 2.1.D ' Reactor Vessel Water Level' 3.3.A

• Shutdown Margin (SDM)*

3.3.B

• Reactivity Anomalies

• 3.3 L

• Rod Worth Minimizer" 3.4.A " Standby Liquid Control System (SLCS)*

3 6.A ' Recirculation Loops' 3.11.A *Averago Planar Linear Heat Generation Rate

• 3.11.B ' Transient Linear Heat Generation Rate

• 3.11.C ' Minimum Critical Power Rat #

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NSNN-04 Rechion 0 i 10 CFR30 $9 Sqfery Evaluation l'rocess  ;

3.11.D *Lir.eer Heat Generation Rate' i

9. Will the proposed activity require a new Technical Specification, Technical Specification revision, or other Operating License amendment? i NOTI:

If a Technical Speciflemtion resision or other Operating License amendment is required, contact Regulatory Assurance Wida completing Step 16, Indicate that a Technical Specification resision is required.

O. YES- A new Technical Specification will be required, or a limiting condition more conservative than an existing Tech Spec needs to be established, and the Tech Spec should be changed -

to bring it into conformance. Notify Regulatory Assurance that a License Amendment will ,

I be needed and proceed to the next step.

O YES. The proposed activity would be in CONrLICT with the existing Tech Spec or Operating License conditions, or require NEW Tech Specs or License conditions. Notify Regulatory Assurance that a License Amendment is needed and proceed to Step 16. NRC approval is required before implementation.

[

@ No. Proceed to next Step. -

Reference 13 documents methods for calculating inputs to the MCPRSL analysis used for O2C15 in Section 6 9.A.6.b of the Quad Cities Technical Specifications, The Quad Cities Technical Specifications also contain footnotes in both section 2.1.8 ano 6.9.A.6.b regarding the MCPR safety limit stating that the Quad Cities Unit 2 limits are applicable for 02C15 only. The Q2C15revi core loading has been shown to have a raclial power that will result in an improved MCPRSL over the original Q2C15. 02C15rev1 has also been reviewed against key core parameters that dictate that 02C15revi is indistinguishable from 02C15 from a licensing analyses standpoint (References 9 and 10) and therefore, a new licensing analyses will not need to be performed. Therefore, Reference 13 and the MCPRSL footnotes remain valid and no Technical Specification revision is needed.

Page 8 of 51

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NSil'I' A-04 Revision 0 10 CFR$0 39 Safety Evaluation l'nass EXHIBIT G 10CFR50.59 SAFETY EVALUATION

10. lo detennine if the activity may increase the probability or the consequences of an accident previously evaluated in the SAR, use one copy of this step to ansaer the following questions for each accident listed in Step 7 that is affected in a different manner. Provide the rationale for all answers.

Affected accidents: Decrease in Reactor Coolant inventory SAR Section: 15.6 10a. hiny the probability of the accident be increased? O YES @ NO The loss-of-coolant accident resulting from a pipe break inside the primary containment is classified as a hmiting fault (i e., an event that is not expected to occur but is postulated). The introduction of a discharged fuel assembly into Quad Cities Unit 2 has no effect on the piping integrity. Thus, the probabihty of a LOCA is not increased and the accident would retain the hmiting fault classification.

10b. hiny the consequences of the accident (off site dose) O yns a no be increased?

Siemens has performed a LOCA Break Spectrum Analysis (Reference 6) to determine the peak ,

clad temperature, hydrogen generation, local cladding oxidation, coro-wide metal-water reaction, '

long term cookng, coolable geometry criteria and MAPLHGR for the limiting LOCA (Reference 4)

With appropriate MAPLHGR limits, these parameters remain with in the acceptance enteria of 10CFR50 46. This analysis and the resulting MAPLHGR hmits for Q2Cib are applicable to Q2C15revi per Reference 10 which documents that the LOCA analysis is independent of cycle design.

Since Q2C15revi contains the same fuel types as Q2C15, the source terms are not affected. Since the source term associated with Q2C15revi is not significantly different than the source term associated with Q2C15, there will be no increase in off site cose or on the consequences of an accidefit.

Therefore, there is no increase in the conseq!Jences of a LOCA event.

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NSWl' A-04 Rerhkm 0 10 CFR30 SVSqfery &a!uarkm Process l

EXHlBIT G t 10CFR50.59 SAFETY EVALUATION In. To detennine i the artis ity may increaue the probability or the consequences of an accident previously evaluated in the SAR. use one copy of this step to answer the following questions for each accident listed in l Step 7 that is affected in a different manner. Provide the rationale for all answcrs.  ;

4 Affected accidents: Break in Reactor Coolant Pressure Boundary instrument Line Outside Containment SAR Section: 16,6.2 10a. May the probability of the accident be increased? O YES @ NO The introduction cf a discharged fuel assembly into Quad Cities Unit 2 will not affect the probabikty of a break in reactor coolant pressure boundary instrument line outside containinent. The inctrument Line Break Outside Containment is classified as a kmiting fault (i e., an event that is not expected to occur but is postuisted). The introduction of a discharged GE9B fuel assembly into the Quad Cities Unit 2 core has no effect on instrument kne integrity. Thus, the probability of an Instrument hne Break Outside Containment is not increased and the accident would retain the limiting fault classification.

10b. May the consequences of the accident (olT site dose) O YEs S NO be increased?

The off-site dose consequences of the instrument Line Break Outside Containment are unaffected by the revision to the 02C15 core design. All radiological consequences for the instrument hne break result from the primary coolant activity since there are no fuel failures as a result of the instrument kne break. A break in an instrument line is equivalent to a steam hne break of the same size. A break in an instrument hne connected to the reactor vessel below the liquid levelis bounded by a recirculation kne break of the same size. The Technical Specification limitations on the specific activity of the primary coolant ensure that the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> thyroid end whole body dose resu' ting from a main steam kne failure outside the containment during steady state operation will not exceed small fractions of the dose guidelines of 10CFR100. The values for the hmits on specific activity represent intenm hmits based upon a parametric evaluation by the NRC of typical site locations. These values are conservative in that specific site parameters, such as site boundary location and meteorological conditions were not considered in this evaluation. Therefore, consequences of the instrument l ine Break are not increased, I

Page 10 of 51

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NSWP A 04 Revaton 0 l 10 CFR$0 39 Sakty Evaluation Procen l

EXHlBlT G  !

10CFR50.59 SAFETY EVALUATION l

10. To detemiine if the actisity may increase the probability or ti.e consequences of an accident previously evaluated in the SAR, use one copy of this nep to answ er the following questions for each accide<it listed in Step 7 that is aflected in a ditTerent manner. Provide the ratior ale for all answ ers.

AITected accidents: Load Rejection (Generator Trip) Wishout Bypass SAR Section: 15.2.2.1 10a. May the probability of the accident be increased? O YEs S NO The introduction of a discharged fuel assembly into Quad Cities Unit 2 will not affect the probabikty of the LRNB event. This change does not affect the turbine's abikty to sense electricalload the turbine control valve or the signal to initiate a turbine control valve fast closure. Reliability of the turbine bypass system is unaffected. Thertsfore, the probabihty of the LRNB event !s not increased.

i 10b. May the consequences of ti.e accident (olT site dose) O Yi:s 8 NO be increased?

Bounding thermal hmits are estal;lished to protect tne fuel from approaching the MCPR safety hmit and the LHGR Protection Against Power Transient Limit, TS a LRNB event is one of the AOOs that is analyzed for determination of the bounding MCPR hmits. SPC has also provided bounding MCPR hmits for GE fuel types These thermal limits are applied to maintain fuelintegnty and protect the fuel from the consequences of potential accidents, including the LRNfs The thermal hmits apphed to 02C15tevi are the same as used for O2C15. The O2C15revi core des gn has been reviewed to r ensure that all 02C15 thermal hmit analyses are bounding for 02C15tew1 (Reference 9 and 10).  !

Because the core pa:ameters do not significantly change for O2C15tev1 ccsmpared to 02C15, tho l LRND analysis from 02C15 is bounding for 02C15revi (Reference 10). Additionally, since the fuel bundle is the only equipment being physically changed, no equipment or scram logic required to mitigate the consequences of this event is modi'ied by this change.

The off . site dose is not increased from the LRNB actient due to 02C15rev1 core design. Fuel failure is not expected to occur since violation of the MCPRSL ,1% plastic strain, and fuel centerkne melt are prevented, 4

Since equivalent hmits are applied and mitigating equipment is not affected, the consequences of the LRNB event are not increased.

Page 11 of $1 l

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i NSil'l* A 04 Revision 0  !

10 CTR$0 39 Safety baluation Process }

f EXHIBIT G !

10CFR50.59 SAFETY EVALUATION i

10. To detennine if the activity may increase the probability or the consequences of an accident previously evaluated in the SAR, use one copy of this step to answer the following questions for each accident listed in Step 7 that is affected in a different manner. Provide the rationale for all answers. ,

Affected accidents:Incrosse in Feedwater Flow l SAR Section: 16.1.2 l

10a. May the probability of the accident be increased? O YES @ NO i The introduction of a discharged fuel assembly into Quad Cities Unit 2 will not affect the probabikty of the increase in feedwater flow which is the Feedwater Controller Failure (FWCF) event. This change does not affect the feedwater control system. Therefore the probabihty of the FWCF event is not increased due to this change in core design.

10b. May the consequences of the accident (off site dose) O yEs @ No be increased?

Bounding thermal hmits are estabkshed to protect the fuel from approaching the MCPR safety !

and the LHGR Protection Against Power Transient Limit. The FWCF event is one of the AOOs that is analyzed for determination of the bounding MCPR. SPC has provided these thermallimits for ATRIUM 98 fuel (MCPR). SPC has also provided MCPR limits for each GE9B and GE10 fuel type.

These thermallimits are applied to maintain fuelintegrity and protect the fuel from the consequences of potential accidents, including the FWCF. The thermal limits applied to Q2C15revi  ;

are the same as used for Q2C15. The Q2C15revi core design has been reviewed to ensure that all Q2C15 thermal hmit analyses are bounding for 02C15tevi (Reference 9 and 10). Because the core parameters do not significantly change for Q2C15*evi compared to 02C15, the FWCF analysis from Q2C15 is bounding for Q2C15rev1. Additionally, since the fuel bundle is the only equipment being physically changed, no equipment or scram logic required to mitigate the consequences of this event is modified by this change.

The off site dose is not increased from the FWCF accident due to 02C15revi core design. Fuel failure is not expected to occur since violation of the MCPRSL,1% plastic strain, and fuel centerkne melt are prevented.

Since equivalent limits are applied and mitigating equtpoent is not affected, the consequences of the FWCF event are not increased.

5 Page 12 of 51 e- --, +. , w - - - -

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NSIf P 4 04 Revision 0 10 CFR$0 39.%) cry &aluation hocess EXHIBIT G 10CFR50.59 SAFETY EVALUATION t

10. To detennine if the activity may increase the proitability or the consequences of an accident previously evaluated in the SAR, use one copy of this step to answer the following questions for each accident listed in l Step 7 that is affected in a difTerent manner. Provide the rationale for all answers.

Affected accidents: Control Rod Drop Accident SAR Section: 15.4.10 10a. May the probability of the accident be increased? O YEs B NO The introduction of a discharged fuel assembly into Quad Cities Unit 2 will not affect the probabihty of the event. Thit change will not increase or decrease the probability that a control rod will fall out of the core after becoming disconnected from its drive after the drive has been removed to the fully withdrawn position. This change in fuelloading does not in anyway affect the control rod drive system or control rod latching mechanism Therefore, the probabihty of the Control Rod Drop Accident will not be increased.

10b. May the consequences of the accident (oft site dose) O YES @ No be increased?

Since the source term associated with the 02C15revi core is not significantly different than the source term of the O2C15 core, there will be no significant effect on the radiological consequences of a control rod drop accident.

The requirement of the CRDA is that it results in an enthalpy below 280 cal /gm. The enteria of 280 callgm is not dependent on fuel type and therefore remains applicable for both GE9 and 10 and SPC Atrium-98 fuel Comed has performed a control rod drop accident ar.alysis for Q2C15 and it was calculated that 788 rods exceeded 170 callgrarn and that no rods would exceed 280 callgm.

Reference 9 documents that the 02C15revi core CRDA is bounded by the O2C15 CRDA since the Q2C15revi core has been projected to be depleted with rod patterns relatively the same as those from the Q2C15 analysis. Because the O2C15 results are applicable to 02C15revi there is no increase in the consequences of a CRDA.

Page 13 of 5'

NSIYP-A-04 Revision 0 10 CFR30 39 Sqfety Evaluation Process EXHIBIT G 10CFR50.59 SAFETY EVALUATION

10. To deter.nine if the activity may increase the probability or the consequences of an accident previously evaluated in the SAR, use one copy of this step to answ er the following questions for each accident listed in Step 7 that is affected in a different manner. Provide the rationale for all answers.

AITected accidents: Rod Withdrawal Error (at power)

SAR Section: 16.4.2 10a. May the probability of the accident be increased? O YES @ NO The introdur: tion of a discharged fuel assembly into Quad Cities Unit 2 will not affect the probability of the event. This change will not iricrease or decrease the probabihty that the incorrect control rod would be selected for withdrawal. The RBM is not affected by this change; therefore, the syrtem that is designed to eventually inhibit withdrawal of the control rod and terminate the transient is assumed to function the same as before the change. Therefore, the probabikty of +.he RWE accident will not be increased.

10b. May the wnsequences of the accident (off site dose) O YES @ NO he increased?

The result of the O2C15 analysis of this event is a A CPR that will protect the fuel (all fuel types in the core) from exceeding the MCPR safety limit, which protects the fuel from departure from nucleate boiling should the event occur. Since the RWE analysis is dependent on the base rod pattern, and the case rod pattern is conservatively unreakstic compared to the actual rod pattern, the fact that the core loading pattern will be slightly different near the core periphery for 02C15tevi will have httle effect on the RWE results (Reference 9). Therefore, the Q2C15 ACPR results for the RWE are still valid and hence there are no increased consequences of the RWE event.

Page 14 of 51

NSif f A 04 Revision U 10 CFR30 39 Safety Evaluation Pnwess EXHlBIT G 10CFR50.59 SAFETY EVALUATION

10. To determine if the activity may increase the probability or the consequences of an accident previously evaluated in the SAR use one copy of this step to answer the following questions for each accident listed in l Step 7 that is afTected in a dilferent manner. Provide the rationale for all answers. 1 Affected accidents: ASME OVERPRESSURIZATION EVENT SAR Section: 6.2.2.2 lua. May the probability of the accident be increased? O Yl:S @ NO The introduction of a discharged fuel assembly into Quad Cities Unit 2 will not affect the probability of the ASME event. The change will not affect the Main Steam Isohtion Valves (MSIVs), the Safety Rehef Valves (SRVs), the Turbine Control Valve (TCV), the Turbine Stop Valve (TSV), or the scram logic. Therefore, the probability of the ASME overpressurization accident is not increased 10b. May the consequences of the accident (off site dose) O vns @ NO te increased?

The ASME event will not be affected due to this change. The results of the ASME Overpressuritation analysis for O2015 were within the 110% of design limit and therefore did not increase the consequences of an ASME Overpressurization event. Reference 10 contains a comparison of key transient analysis parameters from the 02C15 core de>ign and the 02C15rev1 core design. This assessment shows that 3 key parameters from O2C15revi for the ASME event, EOFP Exposure. Axial Offset, and Void Coefficient, are similar to the O2C15 values. Because these parameter values are similar to the same parameter values for 02C15, the ASME overpressurization analysis performed for 02C15 is bounding and apphcable to O2C15revi i

(Reference 9). Since the ASME Overpressurization limits are still met, there is no increase in the consequences of the event.

t Page 15 of 51

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i NSilT-A 04 Revision 0 10 CFR30 39 Safctr &aluation Pneess EXHIBIT G -

10CFR50.59 SAFETY EVALUATION 10- . To determine if the activity may increase the probability or the consequences of an accident previcusly eval.nated in the S AR. use one copy of this step to answ er the following questions for each ac:iden' listed in Step 7 that is affected in a different manner. Provide the rationale for all answ crs.

Affected accidents: Decrease in Feedwater Heat!ng SAR Section: 15.1.1 10a. May the probability of the accident be increased? O Yes a No l

The introduction of a discharged fuel aseembly into Quad Cities Unit 2 does not affect the feedwater heaters or the steam extraction lines to the heaters. The prco&bility of a loss of feedwater heaters occurring is not affected by the change in fuel. Theieforo, the probability of the Loss of Feedwater  !

Heaters event does not increase.

10b. May the consequences of the accident (off site dose) O YES @ NO be increased?

The Loss of Feedwater Heater event is analyzed each cycie for its effects on MCPR. The delta CPR from this event is used to determine the OLMCPR. The OLMCPR ensures the MCPR safety hmrt is not violated during anticipated operational occurrences. Fuel failure is not expected to occur since violation of the MCPPSt,. is prevented. Although the power distribution and exposure "

distribution are affected by the change in core loading for 02C15revi, there will not be a significant effect on the delta CPR for this event. The AMCPR result of the LFWH will not be sensitive to the relatively minor change in the core loading pattern for 02C15tevi (Refarence 9). There is significtnt margin between the LFWH event and the limiting transient event for 02C15 (Reference 9). Therefore, any affects the change in exposure and power distribution may have on the LFWH event will be bounded by the limiting transient event and the consequences of the Loss of Feedwater Heater event are not increas*d by the change.

The off - site dose is not increased from the Loss of Feedwater Heaters accident due to the change.

Fuel failure is not expected to occur since violation of the MCPASL is prevented.

Page 16 er$1

NSQ A-04 Rettston 0 10 CFR$0.59 S&ty Evaluation I wess EXHIBIT G 10CFR50.59 SAFETY EVALUATION ,

10. To determine if the aethity may increase the probability or the consequences of aa accident previously evaluated in the SAR, use one copy of this step to answ er the following questions for each accident listed in Step 7 that is affected in a difTerent manner, Provide the rationale for all answers.

Affected accidents: Mislocated Fuel Assembly Accident SAR Section: 15.4.7 10a. May the probability of the accident be increased? O YES @ NO The introduction of a discharged fuel assembly into Quad Cities Unit 2 does not affect the probabikty Of a Misplaced Fuel Assembly and Subsequent Operation. Since the replacement fuel assembly is of the same tyoc am other fuel assemblies currently in the Q2C15 core, there will be no change in fuel handiing techniques or equipment. A core venfication, including assembly height, orientation and serial number cht - k, will be performed following the core loading changes to ensure that the 02C15revi core is to,,ded in accordance with the analyzed loading pattern. Therefore, the probabihty of the Misplaced Fuel Assembly and Subsequent Operation event does not increase.

10b. May the consequences of the accident (off-site dose) O YES @ NO be increased?

T he Misplaced Fuel Assembly and Subsequent Operation event was analyzed for Q2C15 for its

, effects on MCPR (Reference 3). The delta CPR from this event is used to determine the OLMCPR.

The OLMCPR ensures the MCPR safety limit is not violated during an anticipated operational occurrence. Fuel failute is not expected to occur shce violation of the MCPRSL is prevented. As stated in Reference 9, the Q2015 results indicate that the delta CPR from this event is significantly less than the delta CPR from the limiting prescurization event. Based on these facts, engineering judgement indicates that the licensing basis for the fuel loading error continues to be met for Q2C15revi. Therefore, the consequences of the Misplaced Fuel Assembly and Subsequent d

Operation event are not increased by the addition of a (, ' charged GE9B fuel assembly into the Quad Citiu Unit 2 core.

The off site dose is not increased from the Misplaced Fue, Assembly and Subsequent Operation accident due to the change since the MCPRSL continues in be protected.

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, i NSIiN 04 Revision 0 '

10 CFR30 39 Safety Evaluation l'nwess i

EXHIBIT G 10CFR50.59 SAFETY EVALUATION

10. To determine if the activity may increase the probability or the consequences of an accident previously evaluated in the SAR, use noe copy of this step to answer the following questions for each accident listed in Step 7 that is af fected in a difTerent manner. Provide the rationale for all answ ers.

Affected accidents: Slow Core Flow Excursion SAR Sectlon: 15.4.5 (This is not an accident evaluated in the original UFSAR; however, for SPC methodology, flow dependent hmits are based on the scenario of a slow flow excursien, antiis therefore, addressed here as ifit were an accident )

Text has been added by Reference 11 to Section 15.4.5

• Recirculation Loop Flow Controller Failure with Increasing Flow" to desenbe SPC's approach to this event.

10a. May the probability of the accident be increased? O YES @ NO The insertion of a discharged GE98 fuel assembly into Quad Cities Unit 2 does not affect the recirculation system or any other plant systems or components that could initiate a slow core flow excursion. Therefore the probabikty of the accident is not increased by this change, 10b. May the consequences of the accident (off site dose) O YES @ NO be increased?

SPC's methodology is such that they provide flow dependent limits (MCPR, values for each fuel type) to protect the fuel from violating the operating kmit and safety limit depending on the control mode dunng the slow flow runup event. Therefore, the consequences of the slow core flow excursion event are protected by the thermal kmits that are applied. Fuel failure is not expected to occur since violation of the MCPRSL is prevented by the flow dependent MCPR operating limit.

The flow dependent MCPR limit for 02C15 encompasses 02C15revi (Reference 9 and 10).

Therefore, the consequences of the slow core flow excursion will not increase or be affected based on maintaining operation within the flow dependent MCPR operating limits.

page 18 of 51

O NSif N-04 Revision 0 10 CFR30 39 Safety Evaluation Prtwess EXHlBIT G 10CFR50.59 SAFETY EVALUATION

10. lo deterrnine if the activity may lucrease the probability or the consequences of an accident previously evaluated in the SAR. use one copy of this step to answer the follow ing questions for each accident listed in Step 7 that is affected in a different manner. Provide the rationale for all answers.

Affected accidents: Design Hasis Fuelllandling Accidents inside Containinent and Spent Fuel Storage fluildings SAR Section: 15.7.2 10a. May the probability of the accident be increased? O YES @ NO The insertion of a discharged GE98 fuel assembly into the Quad Cities Unit 2 core does not affect the probability of a Fuel Handkng Accident. GE9B fuelis currently being used at Quad Cities Unit 2 and has not expenenced difficulties interacting with the fuel handling equipment. This change to the Q2C15 core design does not affect any fuel handling equipment or procedures. Therefore, there is no increase in the probabihty of a fuel handling accident.

10h. May the consequences of the accident (ofi site dose) O YES @ NO tv increased?

The fuel handhng accident consequences are analyzed based on certain fuel assembly mechanical parameters and if that fuel assembly were dropped, what would be the resulting off-site dose Since the discharged GEDB bundle that is being inserted into the core is of the same type and similar exposure as other assemblies currently in the core, the consequences of a fuel handling accident (off site dose) would not increase from the current analyses for the GE9 fuel.

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NSit'l* A-04 Revision 0 10 CFR30 39 Safety Evaluation Process P

EXHIBIT G 10CFR50.59 SAFETY EVALUATION

10. To detennine if the activity may increase the probability or the consequences of an accident previously evaluated in the SAR, use one copy of this step to ansu er the following questiom for sach accident listed in Step 7 that is affected in a difTerent manner. Provide the rationale for all answ ers.

Affected accidents: Anticipated Transient Without Scram (ATWS)

SAR Section: 16.8 10a. May the probability of the accident be increased? O vES @ NO The insertion of a discharged GE9B fuel assembly into the Quad Cities Unit 2 core does not affect the control rods, the control rod drives, or ARI System. The replacement fuelis of the same type and is of similar exposure as fuel currently in the core for Cycle 15; therefore, there is no increased probability of control rod interference due to the change in fuel design. Therefore, the probabikty that an afiticipated transient without a SCRAM would occur is not increased.

10b. May the consequences of the accident (off site dose) O yes @ so be increased?

The primary core parameter t%t determines the effects of an ATWS event is the void coefficient.

The 02C15revi void coefficient did not change from the 02C15 analysis (References 10).

Therefore, the consequences of an ATWS event are not increased due to the introduction of the discharged GE9 fuel assembly into the Quad Cities Unit 2 core.

If any amwer to Sten 10 is Yrs.then an Unrniewed saleiv ouestion esists, Proceed to sten 16.

If all answers are NO. proceed to netI sten, Page 20 of 5I

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NSIiMO4 Rension 0 10 CFR30 39 Safety Evaluation Process EXHIBIT G 10CFR50.59 SAFETY EVALUATION

11. To determine if the activity increases the probability or consequences of a malfunction of equipment important to safety for either the analyicd accidents or other SAR requirements, answ er the questions below.

Provide the rationale for all answers.

Ila. May the probability of a malfunction of equipment O YES @ NO important to safety increase?

The insertion of a discharged GE9B fuel assembly into the Quad Cities Unit 2 core for O2015revi will not affect the probability of a rMlfunction of equipment important to safety. This change only alters the fuelin the core. This change does not physically alter any other plant components, systems, or structures nor after the performance of any equipment important to safety. No plant support equipment is affected by these changes. The same type of fuel assemblies with similar exposures that are currently employed in the 02C15 core are being used in the 02C15tevi design.

Therefore the probability of a malfunction of equipment important to safety does not increase.

l ib, May the consequences of a malfunction of equipment O YEs E NO impodant to safety increase?

The insertion of a discharged GE98 fuel assembly into the Quad Cities Unit 2 core for 02C15revi does net affect the performance of any equipment important to safety. No plant support equipment designed to mitigate consequences of accidents or malfunctions are affected by these changes.

The replacement fuel bundle is of the same type and similar exposure of fuel assemblies already in the 02C15 core. The consequences would be the same as with the O2C15 fuel since the 02C15revi MCPR safety limit, LHGR limit and APLHGR limits remain protected and bounded by the 02C15 analysis (Reference 9 &10). Hence the consequences of a malfunction of equipment important to safety will not increase due to the change.

If the rnswer to Sten l[ b YES. then an Unresiewed Safety Ouestion eslits, Proceed to Sten 16.

If the answer is NO. proteed to nest sten, Page 21 of 51

-y y -,- -. 9-

i NSitT A-04 Revisi<>n 0 10 CFR30 39 Safety &aluation 1%'ess EXHIBIT G 10CFR50.59 SAFETY EVALUATION NOTE Certain changes can impact the ensironment. They can include, but are not limited to:

Increase in noise les el or heat discharged, change in station discharge quality *

(nonradiologicalligeld or gaseous effluents), changes to pollution control equipment, methitles affecting gegetation or animallife, escavating land that w as not disturbed during site preparation and plant construction. Consult Regulatory Assurance or Environmental Sersices to determine whether an ensironmental tvaluation, or a permit or permit change is required.

12. Itased on your answers to Steps 2, $ and 6, may the change adversely impact systems or functions so as to create the possibility of an accident or malfunction of a type different from those evaluated in the SAI ?

O YES @ NO Describe the rationale for your answer:

General Effects of Replasement Fuel Assembiv The replacement fus: assembly has been determined to be mechanically and thermal hydraulically compatible with the current core assembhes. The design is the same as other GE9B assemblies currently in the O2C15 core and the exposure is similar to other assemblies currently in the O2C15 core. At identicellicensing end-of-cycle core average exposures, the O2C15tevi cycle design has a similar void reactivity and axial power shape as the O2C15 heensing design. Only the makimum radial power shows any significant difference with a slight increase between 02C15 and O2C15tevi, and the increase actually improves the MCPR safety limit since the core is less flat with respect to radial peaking The replacement fuel sssembly has also been selected due to its relatively low discharged exposure and its recent discharge status. The bundle has been examined with a camera in the spent fuel pool to ensure that its physical integrity has been maintained (Reference 15). Also, the O2C15revi core has been evaluated to ensure that the assembly will not violate EOL exposure prior to EOC15revi (Reference 9).

The O2C15revi reload is not expected to affect any systems or plant operation such that it creates an accident or a malfunction of equipment different than that evaluated in the UFSAR. Fuel failure potential remains.

Effect on Control Blades The exterior envelope of the replacement fuelis dimensionally the same as the previous GE9B fuel in Quad Cities Unit 2 core and the exposure of the replacement bundle is similar to other exposures currently in the O2C15 core. Therefore no new accident or malfunction of the control blades is expected to occur due to this change.

Page 22 of 51

NSIil'-A 04 Revision 0 10 CFR30 39 Safety baluation Process if the answer to Stro 12 is Ylliktp an Unresiewed Safety Ouestion esists. Proceed to Sten 16.

If the answeris NOAiroceed to nest step, Page 23 of 51 l

l

NSil'I' A-04 Revision 0 10 CFRSC.39 Sakty Evaluation l'rocess EXHIBIT G -

Page 8 of 10 10CFR50.59 SAFETY EVALUATION

13. Deteimine if parameters used to establish the lechnical Specification limits are changed. Use one copy of this page to answer Steps 13 and 14 for each Technical Specification listed in Step 8.

AIIccted lechnical Specifi, tion (s): 2.1.B " THERMAL POWER, High Pressure and High Flow" Cll0CK APPROPRI ATE CONDlilON C 1 he cha**e does not n'Tect any parameters upon w hich Ttchment Specifications are based; therefore, there is no rede en in the margin of safety proceed to Step 15.

O All changes to the parameters or conditions used to estabbsh the Technical Specification requirements are in a conservatige direction. Therefore, the actual acceptance limit need not be identified to detennine that no reduction in margin of safety esists proceed to Step 14.

@ 1 he Technical Specification or SAR prog ides a margin of safety or acceptance limit for the applicabk parameter or condition 1 ist the limit (s)' margin (s) and applicable reference for the margin of safety below .

proceed to Step 14.

O 1 he applicable parameter or condition change is in a potentially non-consery A e direction and neither the lechnical Specification. the SAR, or the SI:R progides a margin of safety or an acceptance limit. Request Regulatory Assurance assistance to identify the acceptance hmiumargin for the h; , ' of Safety determination by consulting the SAR. SER's or other appropriate references. List the agreed limit (s)'n.wgin(s) below.

Proceed to Step 14.

Acceptance 1.imit(srhiargin(s) of Safety SAR thwuments & Sections MCPR > 1.10 two loop UFSAR 4.4.1.2.1.2 1.11 single loop 14 Di'll:RhtlNIL based upon the response to Step 13 w hether or not the margin of safety is reduced (i c.: the new salues are outside of the acceptance limits) Include the rationale for the detennination and a discussion of compensating factors used to reach the conclusion.

O Yl:S - Atarrin or Sareiy is reauced.

@ NO hlargin of Safety is NOf reduced.

Proceed to nest Step.

Discussion.

The n ergin of safety for this Technical Specification is the MCPR safety limit. The MCPR safe ' timit did not change from the previous value of 1.10 (two-loop) and 1.11 (singit sp).

Reft ences 2 and 13 document the methodology used for modeling a mixed core ant.

calculating the O2C15 MCPRSL. ' References 9 and 10 d .cument that the 02C15 MCPRSL is applicable for O2015revi.

If the remonse teltsp 14 is "YES". the Marcin of Safety is reduced ano an Unreviewed Safety Ouestion esists.

I'rocred to Sten 16.

Page 24 of 51

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

1 NSliT A 04 Revision 0 10 CFR$0 39 Safety Dr.luation Process F

NOTE in sonne esses , the proposed metivity ticing evaluated may t>e a candidate for adding words to the Uf'SAR. Consideration should tie gh en to adding a discussion of key regulatory ,

issues, regulatory documents (Generic Letters, Regulatory Guides. NHC Hulletins, etc.), j station commitmeuts and new equipment. (See Regulatory Guide 1.70 for level of detall) '

l i

4 9

t 4

4 k

9 l

1 1

].

Page 25 of 51

l NSIl'P.A-04 Revision 0 10 CFP30 39 Safety Evaluction Pn. cess EXHIBIT G Page 8 of 10 10CFR50.59 SAFETY EVALUATION e

13. Determine if p rameters used to establish the Technical Specificatio t limits are changed. (Jsc one copy of this page to answer Steps 13 and 14 for each lechnical Specification listed in Step 8.

Affectcj Technical Speci6 cation (st 2.1.C REETOR COOLANT SYSTEM PRESSURE CllITK APPHOPRI A10 CONDITIO_N:

C 'I he change does not aflect any parameters upon which Technical Specifications are based, therefore, there is no reduction in the margin of safety . proceed to Step 15.

O All changes to the parameters or conditions used to establish the Technical Specification requirements are in a consen atis e direction. Therefore, the actual acceptance limit need not be identified to determine that no reduction in margin of safety exists - proceed to Step 14.

@ The Technical Specification or SAR prosides a margin of safety or acceptance limit for the applicable parameter or condition.1.ist the limit (symargin(s) and applicable reference for the margin of safety below -

proceed to Step 14.

O 1he arrlicable parameter or condition chanse is in a notentiall> non-consen atis e direction and neither the Techmcal Specification, the SAR, or the SER provides a margin of safety or an acceptance limit. Request Regulatory Assurance assistance to identify the acceptance hmit' margin for the Margin of Safety determination by consulting the SAR SER's or other appropriate references. I,ist the agreed limit (symargin(s) below.

Proceed to Step 14.

Acceptance 1.imit(srMargin(s) of Safety SAR Documents & Sections The reactor coolant system pressure pressure shall not exceed 1345 psig as measured in the reactor vessel steam dome. UFSAR 5.2.2 14, DET12MINE, based upon the response to Step 13, w hether or not the margin of safety is reduced (i e.: the new 5alues j ars outside of the acceptance limits). Include the rationale for the determination and a discussion of compensating factors I used to reach the conclusion.

O YES - Margin or Sarety iS reducco.

@ NO Margm of Safety is NOT reduced.

Proceed to next Step.

Discussion:

Reference 10 shows that the ASME overpressurization analysis for Q2C15 is boundir.g for Q2C15revi since the key ASME parameters (EOFP Exposure, Axial Offset, and Void Codficient) for Q2C15 are similar to Q2C15rev1. Because the ASME pressurization limits are also applicable for Q2C15revi, there is no reduction in the margin of safety for reactor pressure.

Page 26 of 51

NSit*P-A-04 Revision 0 10 CFR50.39 Safety Evaluation Proc < ss if thf f.t!Ponse to Stro 14 is "YES". the Marain >f Sa'ety is reduced and an 11nreviewed Safety Ouestion esists.

Proceed to Step 16.

NOTE

• In so:ac cases, the proposed activity being evaluated may be a candidate for adding words

. to the UFSAli Consideration should be given to adding a discussion of key regulatory issues, regulatory documents (Generic Letters, Hegulatory Guides, NRC llulletins, etc.),

station commitments and new equipment. (See ReguMory Guide 1.70 for level of detall)

Page 27 of 51

NSWP-A-04 Revision 0 10 CFR$0.39 Safety Evaluation Pwcess EXHIBIT G 10CFR50.59 SAFETY EVALUATION

13. Determine if parameters used to establish the Technical SpeJ0 cation limits are changed. Use one copy of this page to answer Steps 13 and 14 for each Technical Speci0 cation listed in Step 8.

Affected Technical Speci0 cation (s): 2.1.D REACTOR VESSEL WATER LEVEL CilECK APPROPPI ATE COND! TION:

0 'I bc change does not affect any parameters upon w hich Technical Specifications are based; therefore, there is no reduction in the margin of safety proceed to Step 15.

O All changes to the parameters or conditions used to establish the Technical Speci0 cation requirements are in a consen ative direction. Therefore, the actual acceptance limit need not be identified to determine that no reduction in margin of safsy exists procccd to Step 14.

@ The Technical Specincr. tion o: SAR pros ides ti margin of safety or acceptance limit for the applicable parameter or condition.1.ist the limit (s)' margin (s) and applicable reference for the margin of safety below -

proceed to Step 14.

The applicable parameter or condition change is ir. a potentially non-conservatis e direction and neither the Technical Specification. the SAR, or the SER prosides a margin of safety or an acceptance limit Request Regulatory Assurance assistance to identify tha acceptance limivmargin for the Margm of Safety determination by consulting the SAR, SER's or other appropriate references.1.ist the agreed limit (sVmargin(s) belcw.

Proceed to Step 14.

Acceptance 1.imit(srMargin(s) of Safety SAR Documents & Sections The reactor vessel water level shall be greater than twelve inches above the top of the active irradiated fuel. UFSAR 6.3 UFSAR 7.2

14. DETERMINE, based upon the response to Step 13, w hether or not the margin of safety is reduced (tc.: the new salues are outside of the acceptance limits). Include the rationale for the determination and a discussion of compensating factors used to reach the conclusion.

O YES - Margin or Sareiy IS reduced.

l @ NO - Margin of Safety is NOT reduced Proceed to next Step.

Discussion:

The replacement GE98 assembly is of the same type as other assemblies currently in the

core and therefore, has the same physical characteristics, Therefore, introducing this assembly into the core will not change the current top of active fuel. Plant instrumentation setpoints are not affected. Therefore, the margin of safety is not reduced.

l If the response to Sten 14 is "YES", the Marcin of Safety is reduced and an Unreviewed Safety Ouestion exists.

l Proceed to Sten 16.

Page 28 of 51 I

l

NSil'P-A-N Revision 0 10 CFR$0.39 Safety Evaluation Process -

NOTE In some cases, the piaposed methity being evaluated may be a candidate for adding words ,

to the UFSAR. Consideration should be gisen to adding a discussion oflicy regulatory inues, regulatory documents (Generic Letters, Regulatory Guides, NRC Bulletins, etc.),

station commitments and new equipment. (See Regulatory Guide 1.70 for level of detail)

Page 29 of 51 4 - -.,n.., , ,,

NSilN 04 Revisiont 10 CFR$0 $9 Safety Evaheation l'rocess EXHIBIT G 10CFR50.59 SAFETY EVALUATION

13. Determine if parameters uwd to establish the Technical Specification limits are changed. Use one copy of this page to answer Steps 13 and 14 for each Technical Specification listed in Step R.

A!Tected Technical Specification (s): 3/4.3.A Shutdown Margin (SDM)

CliFCK APPROPRI ATE CONDITION:

C ne change does not affect any parameters upon w hich Technical Specifications are based, therefore, there is no reduction in the margin of safety proceed to Step 15.

O All changes to the parameters or conditions used to estahtish the Technical Specification requirements are in a consen stive direction. Therefore, the actual acceptance limit need not be identified to determine that no reduction in n argin of safety exists . proceed to Step 14.

@ The Technical Specificcion or S AR pros ides a margin of safety or acceptance limit for the applicable parameter or condstion.1.ist the limit (symargin(s) and applicable reference for the margii of safety below -

proceed to Step 14.

O 'ihe applicable parameter or condition change is in a potentially non conservatise direction and neither the Technical Specification, the SAR, or the SI;R provides a margin of safety or an acceptance limit. Request Regulatory Assurance assistance to identify the acceptance hmit/ margin for the Margin of Safety determination by consulting the SAR, SER's or other apptopriate references. List the agreed limit (symargin(s) below.

Proceed to Step 14.

Acceptance 1.imit(s)'hlargin(s) of Safety SAR Documents & Sections The shutdown margin shall be greater than or equal to 0.38% Ak/k with highest worth Control rod analytically determined or greater than or equal to 0.28% Ak/k by test. UFSAR 4.3.2.1.3 UFSAR 15.1

14. DETERhllNE, based upon the response to Step 13, whether or not the margin of safety is ret uced (i.e.: the new $alues are outside of the acceptance limitst include the rationale for the determination and a discusssa of compensating factors used to reach the conclusion.

O YES htargin or Sarety iS reauced.

@ NO - hlargin of Safety is NOT reduced.

Proceed to next Step.

Discussion:

The change to the fuel loading pattern for 02C15rev1 will be expected and verified to meet the appropriate shutdown margin of either analytical method where the requirement is 0.38%Ak, or by test where the SDM requirement is 0.28% Ak (Reference 16). Reference 9 documents that although there are relatively minor changes associated with the O2C15rev1 core, the SDM trend with cycle exposure will be the same as Q2C15. Therefore, the SDM calculations for Q2C15 are applicable to 02C15rev1.

Pege 30 of 51

. . ~ . . . _ _

t

.i NSWP-A-04 Revision 0 i

10 CFR$0 $9 Safety Evaluation Process if the response to Sten 14 is "\T.S".the Martin of Safety is reduced and an Unreviewed Safety Question exists.

Proceed to Sten 16.

NOTE in some cases, the proposed metisity being evaluated may be a candidate for adding words to the UFSAR Consideration should be gis en to adding a discussion of key regulatory issues, regulatory documents (Generie t,etters, Regulatory Guides, NRC Hulletins, etc.),

station commitments and new equipment. (See Regulatory Guide 1.70 for lesel of detail) i l

Page 31 of 51  :

i

NSitT-A-04 Revision 0 10 CFR30.39 Safety &aluation l'rocess EXHIBIT G 10CFR50.59 SAFETY EVALUATION

13. Determine if parameters uvd to establish the Technical Syci0 cation limits are changed. Use one copy of this page to 4

answer Steps 13 and 14 for each lerhnical Specification listed in Step 8.

Aficcted Technical Speci0 cation (s)- 3/4.3.B REACTIVITY ANOMALIES CilECK APPROPRI ATF CONDITIGNt C t he chanFe does not affect any parameters upon which Technical Speci0 cations are based; therefore, there is no reduction i the margm of safety proceed to Step 15.

O All changes to the parameters or conditions used to establish the Technical Speci0 cation requirements are in a consersatise direction. Therefore, the actual acceptance limi: need not be identined to determine that no reduction in margin of safety exists proceed to Step 14.

@ The Technical Speci0 cation or SAR provides a margin of safety or Prceptance limit for the applicable parameter or condition.1.ist the hmit(symargin(s) and applicable reference for the margin of safety below -

proceed to Step 14.

The applicahle parameter or condition change is in a potentially non consenatise direction and neither the Technical Speci0 cation, the SAR, or the SER provides a margin of safety or an acceptance limit. Request Regulatory Assurance assistance to identify the acceptance limit' margin for the Margin of Safety determination by consulting the SAR, SER's or other appropriate references.1.ist the agreed limit (symargin(s) below.

Proceed to Step 14.

Acceptance ! imit(sVMargin(s) of Safety SAR Documents & Sections The reactivity equivalence of the difference between the actual critical control rod configuration and the predicted critical control rod configuration shall not exceed 1% delta k/k TS 133/4.3.11

14. DI:1 ERMINE, based uptm the response to Step 13, w hether or not the margin of safety is reduced (i c.: the new salues are outside of the acceptance limits). Include the rationale for the determination and a discussion of compenssg factors used to reach the conclusion.

O Yl'S Margin or Sarciy iS reduced.

@ NO - Margm of Safety is NOT reduced.

Proceed to next Step.

Discussion:

The Q2C15revi design is required to ineet the 1% reactivity anomaly criteria when the unit operates, which protects the unit from operation when a large reactivity difference is calculated. Because this criterion is unchanged, there is no reduction in the margin of safety for the reactivity anomaly.

If the response to Sten 14 is "YES". the Marcin of Safety is reduced and un Unreviewed Safety Ouestion esists, Proceed to Sten 16.

Page 32 of 51

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

WA NSlYP-A-04 Revhlon 0 10 CFR30.39 Safety Evaluation Process t

. NOTE in some cases, the proposed activity being evaluoted may be a candidate for adding words to the UFSAR. Consideration should he given to adding a discussion of key regulatory issnes, regulatory documents (Generic Letters, Regulator- Guides, WRC Bulletins, etc.),

station commitments and new equipment. (See Regulatory Guide 1.70 for level of detail)

Page 33 of 51

NSil7'-A-04 Redsion 0 10 CFR30.39 Sakty Evaluation Process _

EXHIBIT G 10CFR50.59 SAFETY EVALUATION ,

13. Determine if parameters used to establish the Technical Specification limits are changed. Uw one copy of this page to answer Steps 13 and 14 for each Technical Speci0 cation listed in Step 8.

AITeeted Technical Specincation(s): 3/4.3.L Rod Worth Minimizer EllECK APPROPRI ATE CONDITION:

[] The change does not aff ect any parameters upon which Technical Speci0 cations are based, therefore, there is no reduction in the margin of safety proceed to Step 15.

O All changes to the parameters or conditions used to establish the Technical Speci0 cation requirements are in a consers utis e direction. Therefore, the actual acceptance limit need not be identined to determine that no reduction in margin of safety exists proceed to Step I4.

@ The Technical Specincation or SAR provides a margin of safety or acceptance limit for the applicable pararneter or condition.1.ist the limit (symargin(s) and applicable reference for the margin of safety below -

proceed to Step 14.

The applicable parameter or condition change is in a potentially non-conservatise direction and neither the Technical Speci0 cation, the SAR, or the SER prosides a margin of safety or an acceptance limit. Request Regulatory Assurance assistance to identify the acceptance limit / margin for the hlargin of Safety determination by consulting the SAR, SI:R's or other appropriate references. I.ist the agreed limit (symargin(s) below.

Proceed to Step 14.

Acceptance Limit (s)'hlargin(s) of Safety SAR Documents & Sections RWM shall be operable when power is UFSAR 4.3.2.1.4 less than or equal to 10% of rated. UFSAR 4.4.3.1.5 UFSAR 4f. 4.2 UFSAR 7.6.1 UFSAR 7.7.1.2.2 UFSAR 14.2.12.1.28 UFSAR 15.4.10

14. DI:T ERh11NE, based upon the response to Step 13, whether or not the margin of safety is reduced (i.e.: the new values are outside of the acceptance limitst include the rationale for the determination and a discussion of compensating factors used to reach the conclusion.

YES hlargin of Safety IS reduced.

@ NO Atargin of Safety is NOT reduced Proceed to next Step.

Discussion:

There is no change in the margin of safety for the requirement of RWM operability. The Bases of this Technical Specification discusses the 280 cal /gm entnalpy limit on the fuel.

There is no reduction in the margin of safety for the requirement that the RDA result in an enthalpy below 280 callgm. The enteria of 280 cal /gm is not dependent on fuel type and Page 34 of 51

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

.i NSill'-A 04 Revision 0 10 CFR30395afetypvaluation i'rocess therefore rerf.ains applicaMe for both GE9 and 10 and SPC Atrium-9B fuel. Comed has performed a control rod drop accident analysis for Q2C15 and it was calculated that 788 rods exceeJed 170 cal / gram and that no rods would exceed 280 caVgm. Reference 9 documents that the Q2C15revi core CRDA is bounced by the O2C15 CRDA since the Q2C15revi core has been projected to be depleted with rod patterns relatively the same as .

. those from the 02C15 analysis. Because the 02C15 results are applicable to 02C15revi the margin of safety has not been reduced for this Technical Specification.

jfLhe response to Sten 14 is YES" tiie Martin of Safety is reduced and an Unres icwed Safety Ouestion psists.

Proceed to Sten 16, NOTE '

in seine cases, the proposed activity being esaluated may be a candidate for adding words to the UFSAlt. Consideration should be given to adding a discussion of key regulatory luues, regulatory documents (Generic 1.etters, Regulatory Guides, NRC Hulletins, etc.),

station commitinents and new equipment. (See Regulatory Guide 1.70 for level of detail) i Page 35 of $1 w* - w

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NSIl'P-A-04 Revision 0 10 CFR50 39 Safety &aluation Process EXHIBIT G 10CFR50.59 SAFETY EVALUATION

13. Determine if parameters used to establish the l echnical Specincatbn limits are changed the one copy of this page to answer Steps 13 and 14 for each Technical Specification listed in Step 8.

AITected lechnical Speci6 cation (s): 3I4.4.A Standby Liquid Control System (SLCS)

CllECK APPROPRI ATE CONDITION:

O 1 hc change does not alTect any parameters upon w hich Technical Specincations are based, therefore, there is no reduction in the margin of safety proceed to Step 15.

All chanFes to the parameters or conditions used to establish the Technical Specification requirements are in a consersatise direction. Therefore, the actual acceptance limit need not be ident Ged to determine that no reduction in margin of safety exists proceed to Step 14.

@ The Technical Specincation or SAR prosides a margin of safety or acceptance limit for the applicable parameter or condition List the limit (s)' margin (s) and applicable reference for the margin of safety below -

proceed to Step 14.

O The applicable parameter or condition change is in a poteatially non-conscrs atis e direction and neither the fechnical Speci0 cation, the SAR, or the SER prosides a margin of safety or asi acceptance limit. Request Regulatory Assurance assistance to identify the acceptance limit / margin for the Margin of Safety determination by consulting the SAR, SER's or other appropriate references. I,ist the agreed limit (s)' margin (s) below.

Proceed to Step 14.

Acceptance 1.imit(s)' Margin (s) of Safety SAR Documents & Sections Bring the reactor from full power to a 3% delta k/k or more suberitical condition with 600 ppm concentration of sodium pentaborate . UFSAR 1.2.1.1 UFSAR 1.2.2.3 UFSAR 3.1.5.2 UFSAR 3.9.5.2 UFSAR 4.2.1.3 UFSAR 4.3.2.1.3 UFSAR 4.6.2 UFSAR 6.0.1.10.2 UFSAR 6.1.1.2 UFSAR 7.7.3.5 UFSAR 7.8 UFSAR 9.3.5 UFSAR 15.8.1.3

14. DETERMlNE:, based upon the response to Step 13, whether or not the margin of safety is reduced (i c.; the new salues are outside of the acceptance limits). Include the rationale for the determination and a discussion of compensating factors used to reach the conclusion O YI:s - Margin of Safety is reduced.

Page 36 of 51

NSWP A-04 Revision 0 10 CFR30.39 Safety Evaluation Process

@ NO Margin of Safety is NOI reduced.

Proceed to next Step.

Discussion:

The requirement of at least 3% Ak subenticality with 600 ppm concentration of sodium pentaborate during an ATWS remains but with the minor changes to the Q2C15 core loading pattern for Q2015revi, the exposure / reactivity will have a different distribution than that expected in the Q2015 licensing calculations. However, since the SLCS is a core wide analysis, and tha Q2C15revi core is less reactive than that assumed in the licensing calculat: ens for Q2015 (References 9 and 10), the Q2015 SLCS analysis bounds the O2C15rev1. Thcrefore, the margin to safety is not decreased for SLCS shutdown capability.

jithe response to Sten 14 is "YES" the Marein of S:sfety is reduced and an Unresiewed Safety Ouestion exists, Proceed to Sten 16, NOTE in some cases, the proposed artisity being evaluated may be a candidate for adding words to the UFSAR. Consideration should be given to adding a discussion of key regulatory issues, regulatory c:ocu*nents gGenetic Letters, Regulatos) Guides, NRC Ilulletins, etc.),

station cominitments and new equipment. (Set Regulatory Guide 1.70 for leselof detail)

Page 37 of 5I

l NSH'P-A-04 Revision 0  ;

10 CFRSO 39 Safety Evaluation Pwess I I

EXHIBIT G 10CFR50.59 SAFETY EVALUATION

13. Detenmne if parameters used to establish the Technical Speci0 cation limits are changed Use one copy of this page to answer Steps 13 and 14 for each Technical Specification listed in Step 8.

Aff ected Technical Specification (s): 3.6.A Recirculation Loops CHITK APPROPRI ATE CONDITION:

O Ihe change does not afTect any parameters upon which Technical Specifications are based; therefore, there is no reduction in the margin of safety proceed to Step 15.

All changes to the parameters or conditions used to establish the Technical Specification requirements are in a consers atis e dirculon. Therefore, the actual acceptance limit. need not be identified to determine that no reduction in margin of safety exists - proceed 'o Step 14.

@ The Technical Specification or SAR provides a margin of safety or acceptance limit for the applicable parameter or condition. I,ist the limit (symargin(s) and applicable reference for the margin of safety below .

proceed to Step 14.

O The applicable parameter or conditio:n change is in a potentially non-conservative direction and neither the Technical Specification, the SAR, or the SLR prm ides a margin of safety or an acceptance limit. Request Regulatory Assurance assistance to identify the acceptance limivmarFi n for the Margin of Safety dstermination by consulting the SAR, Si,R's or other appropriate references, l.ist the agreed limit (symargin(s) below.

Proceed to Step 14.

Acceptance 1 imit(s)' Margin (s) of Safety SAR Documents & Sections Two reactor coolant system recirculation loops shall be in operation or increase the MCPRSL by 0.01, increase the MCPROL by 0.01, reduce the APLHGR to single loop limits as specified in the COLR. and the APRM flow biased SCRAM signal and the rod telock monitor trip setpoints shall be reduced UFSAR 42 UFSAR 4.4.4.2.5 14- DETERMINE, based upon the response to Step 13, w hether or not the margin of safety is reduced (i e.: the new values are outside of the acceptance limits) include the rationale for the determination and a discussion of compensating factors used to reach the conclusion.

O yES Margin of Safety is reducea.

@ NO - Margin of Safe.y is NOT icduced.

Proceed to next Step.

Discussian:

The margin of safety is not reduced for this Technical Specification. 02C15 was analyzed for single loop operation and the 0.01 adder to the MCPR Safety Limit was found to be applicable to both O2C15 and O2C15revi (Reference 10) The requirements of this Technical Specification are not changed due to the addition of a discharged GE9B assembly into the O2C15 core. A MAPLHGR multiplier for single reenculation loop operation was established for 02C15. The O2C15revi Page 38 of 51

NSWP A 04 Revision 0 10 CFR30 39 Safety Evaluation Process MAPLHGR multipist remains the same as the Q2C15 MAPLHGR multipher since the Q2015revi .

fuel lattice typs i:ive not changed from O2C15 based on the replacement fuel being the same type

- as fuel currently in the Q2 core (Reference 10). The MAPLHGR multipi er and the effect on the Operating Limit MCPR for SLO is reflected in the COLR, Therefore, there is no decease in the .

margin of safety.

If the regnonse to Sten 14 is "YES". the Mornin of Safety is reduced and an Unreviewed Safety Oucstion esists.

Proceed to Sten 16 NOTE

- In some cases, the proposed activity being evaluated may be a candidate for adding wt,rds to the UFSAR, Consideration should be gisen to adding a discussior, of key regulatory issues, regulatory documents (Generic Letters, Regulatory Guides, NRC llulletins, etc.),

station commitments and new equipment. (See Regulatory Guide 1.70 for level of detail) 1 Page 39 of 51 Lw..__

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f NSIIP-A-04 Revision 0 10 CFR$0.59 Safety Evaluation Process EXHIBIT G 10CFR50.59 SAFETY EVALUATION 13 Determine if parameters used to establish the Technical Speci6 cation liri.its are changed Use one copy of this page to answer Steps 13 and 14 for each Technical Specincation listed in Step 8.

Alfccted lechnical Speci0 cation (s) 3/4.11.A AVERAGE PLANAR LINEAR HEAT GENERATION RATE CitrCK APPROPRI ATE CONDillONt C 1 he change does not aff ect any parameters upon u hich Iechnical Speci0 cations are based, therefore, there is no reduction in the margin of safet) . proceed to Step 15.

All changes to the parameters or conditions used to establish the Technical Specincation requirements are in a consenatis e direction Therefore, the actual acceptance limit need not be identined to determine that no reduction in margin of safety exists . proceed to Step 14.

@ 1 he 'l echnical Specineation or SAR prog ides a margin of safety or acceptance limit for the applicable parameter or condition List the limit (simargin(.) and applicable reference for the margin of safety below .

proceed to Step 14.

'I he applicable parameter or condition change is in a potentially non-consen atis e direction and neither the Technical Specineation, the SAR, or the SER prmides a margin of safety or an acceptance limit _ Request Regulatory Assurance assistance to identify the acceptance limit' margin for the htargin of Safety determination by consulting the SAR, SF.R's or other appropriate references, l.ist the agreed limit (sFmargin(s) below.

Proceed to Step 14.

Acceptance Limit (srhiargin(s) of Safety SAR Documents & Sections MAPLHGR limits based on Average Planar Exposure and A single loop multiplier specified in The COLR shall not be exceeded UFSAR 4.4 2.2 UFSAR G.3.3.2.2.2

14. DI:TERhtlNI:, based upon the response to Step 13, whether or not the margin of safety is reduced (i c.: the new salues are outside of the acceptance limits) include the rationale for the determination and a discussion of compensating factors used to reach the conclusion.

O Yi:S - htarsin of Sarci3 IS reduced

@ NO - hlargin of Safety is NO F reduced.

Proceed to next Step.

Discussion:

The actual Technical Specifications that require APLHGR be maintained within specificahon and actions be taken when it is not within the specification have not changed.

The APLHGR hmits ensure that the fuel rod mechanicalintegnty is maintained du ing normal and transient operation. Operating the fuel within its APLHGR kmits ensures that Page 40 of $1 l

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

i NSH'P-A-04 Rerbian 0 10 CFR30.39 Safety Evaluation Process 100FR50.46 limits are maintained during a loss of coolant accident.- SPC has provided MAPLHGR limits for the ATRIUM-9B fuelin Q2C15. The GE98 and GE10 fuel have MAPLHGR values GE determined from the cycle in which that batch of GE9B and GE10 fuel was loaded. - These MAPLHGRs have not changed due to the revised core design since MAPLHGR are fuel (lattice) dependent limits based on exposure and not core design dependent (Reference 10). Therefore, there is no redurfdon in the margin of safety, jf the response to St.n 14 is "YES". the Marnin of Safety is reduced and an Unreviewed Safety Ouestion exists.

Proceed to Sten 16

~

NOTE i

In some cases, the proposed activity being evaluated may be a candidate for adding words to the UFSA'e Consideration should be given to adding a discussion of key regulatory issues, regula9ry documents (Generic Letters, Regulatory Guides, NRC llulletins, etc.),

station commitments and new equipment. (See Regulatory Guide 1.70 for level ordetail)

Page 41 of 51 I

I NSWP A-04 Revision 0 10 CFR$0.39 Safm Evaluation l'rocess EXHIBIT G 10CFR50.59 SAFETY EVALUATION

13. Determine if parameters used to establish the Technical Specification limits are changed. Use one copy of this page to answer Steps 13 and 14 for each Technical Specification listed in Step 8.

Affected Technical Specification (s): 3/4.11.B Transient Linear Heat Generation Rate ClllTis APPROPRI ATE CONDITION:

0 1 he change does not affect any parameters upon which Technical Specifications are based; therefore, there is no reduction in the margin of safety proceed to Step 15.

O All changes to the parameters or conditions used to establish the Technical Specification requirements are in a consenatise direction. Therefore, the actual acceptance limit need not be identified to determine that no reduction in margin of shiety exists proceed to Step 14.

@ The Technical Sprainewon or SAR piovides a margin of safety or accepta ice limit for the applicable parameter or condition. List the limit (syma gin (s) and applicable reference for the margin of safety below -

proceed to Step 14.

O The upplicable parameter or condition change is in a potentially non-consen atis e direction and neither the Technical Specification, the SAR, or the SER provides a margin of safety or an acceptance limit, Request Regulatory Assurance assistance to identify the acceptance limit' margin for the Margin of Safety determination b) c onsulting :he SAR, SER's or other appropriate references. List the agreed limit (symargin(s) below.

Proceed to Step 14.

Acceptance 1.imit(sVMargin(s) of Safety SAR Documents & Sections The transient LHGR limit shall be maintained Such that the fuel'fesign limiting ratio For centerline melt is less than or equal to 1.0. UFSAR 4.4.1.2.1.3

14. DETLRMINE, based upon the response to Step 13, whether or not the margin of safety is reduced (i e.: the new salues are outside of the acceptance hmits). Include the rationale foi the de'ermination and a discussion of compensating factors used to reach the conclusion.

O YES Margin or Safety iS reduced.

@ NO Margin of Safety is NOT reduced Proceed to next Step.

Discussion:

The Fuel Design Limiting Ratio for Centerkne Melt (FDLRC) provides adequate protection to ensure that the Transient Linear Heat Generation Rate is nc,t violated for any power distribution. The enteria in the Technical Specifications that requires that the transient LHGR shall be maintained such that the fuel design limiting ratio for centerline melt (FDLRC) is less than or equal to 1.0 has not been changed or affected. The corrective actions required if FDLRC becomes greater than 1.0 also have not changed. Mrintaining FDLRC less than or equal to 1.0 ensures that greater than or equal to 1% plastic stain will not occur end the fuel will not experience centerline melt dunng anticipated operational occurrences beginning at any power level and terminating at 120% of rated thermal power.

Page 42 of 51

NSWP-A-04 Revision 0 10 CFR$0.59 Safety Evaluatioeuments & Sections MCPR shall be greater than or equal to the MCPROL in the COLR. UFSAR 4.2.1.1 UFSAR 4.4.1 UFSAR 15

14. Di'.TERht!Nii, based upon the response to Step 13, w hether or not the margin of safety is reduced (i.e.: the new L. lues are outside of the acceptance hmits). Include the rationale for the determination and a discussion of compensating factors used to reach the conclusion.

O Yl'.S . hiargin of Safety IS reduced.

@ NO hlqin of Safety is NOT reduced.

Proceed to next Step.

Discussion:

The actual Technical Specifications that requ:re MCPR be maintained within specification and actions must be taken when it is not within the specification have not changed.

The MCPR safety limit for Quad Cities Ur it 2 of 1.10 fer two loop operation and 1.11 for single 1000 operation did not change from 02C15 operation to 02C15revi operation. The MCPR safety limit is set so that boiling transition is avoided in 99 9% of the fuel rocs if the kmit is not viola %d. Therefore, provided the MCPR safety limit is not violated, the margin cf safety remains the same. In addition to tne MCPR safety hmit, SPC provided MCPR Page 44 of 51

NSWP.4-04 Raision 0 10 CFRUJ9 Safety Evaluarlon Process operating limits, including flow dependent MCPR limits for both GE and SPC fuoi to protect - )

the MCPR safety limit, as well as MCPR operating limits for equipment out of service and extended operating domains t'or Q2C15, These MCPR lim:ts remain applicable to Q2015:evi since the O2015 an#ys!s is botading for Q2C15rovi per Referenceo 9 and 10.

The piant will operate with a e1CPR opertting liriit that is greater than the safety 1.mit, and .

provides margia to the MCPR safety limit. Therafare, the margin of safet/ is not reduced for the O2C15tevi MCPR safety limit.

i If the resc9ane to Stepj4 is , i:S". the Marnin of Safety is reduct d and aO'nreviewed Sa(sty Ouestion exists.

Proceed to Stere la.

< NOTE in some cases, the proposed activity being evaluated may be s candidate for adding words to the UI'SAR Consideration should be given to adding a discussion of key regulatory iss;ies, regulatory documenis (Generic Letters, Segulatory Gdides, NRC Bulletins, etc.),

sta.clon commitments and new equipment. I'see Regulatory Guide 1.70 for les el of detail)

.t Page 45 of 51 m - - - ~ p

NSIIN-04 Revision 0 10 Cr R30.39 Safetr Evaluation Process EXHIBIT G 10CFR50.59 SAFETY EVALUATION

13. Ihtermine if parsmeters used to establish the Technical Specification limits are changed. Use one copy of this page to answer Steps i3 and 14 for each Technical Specification listed in Step 8.

Affected fechnical Specification (s): 3/4.11,D LINEAR HEAT GENERATION RATE ClllTN Af PROPRI ATE CONDITIONt O 'I he change does not atTect any parameters upon w hich Technical Specifications are based, therefore, there is no reduction in the margin of safety . proceed to Step 15.

O All chz.nges to the parameters or conditions used to establish the Technical Specification requirements are in a conservatise direction. Therefore, the actual acceptance limit need not be identified to determine that no reduction in margin of safety exists proceed to Step 14.

@ The Technical Specification or SAR provides a margin of safety or acceptance limit for the applicable parameter or condition. List the limit (s1' margin (s) and applicable reference for the margin of safety below -

proceed to Step 14.

O 'I he applicable parameter or condition change is in a potentially non-conservatis e direction and neither the Technical Sp;cification, the SAR, or the Sl;R provides a margin of safety or an acceptance limit. Request Regulatory Assurece assistance to identify the acceptance hmit' margin for the Margin of Safety determination by consult ing the dAlt, SER's or other appropriate references. List the agreed limit (s)' margin (s) below.

Proceed to Step 14.

Acceptance Limit (s)/ Margin (s) of Safety SAR Documents & Sections LHGR shall not exceed the Limits in the COLR. UFSAR 4.2.1.1 UFSAR 4.4.1.2.1.3 UFSAR 15.6 14 DETERMINIL based up(m the response to Step 13, whether or not the margin of safety is reduced (i.e.: the new values are outside of the acceptance hmits). Include the rationale for the determination and a discussion of compensating factors used to reach the conclusion.

O YES Margin of Sareiy iS redueca.

@ NO Margin of Safety is NOT reduced.

Proceed to next Step.

Discussion:

The fuelis required to remain belaw a set LHGR rate dunng operation. The LHGR limit ensures that fuel mechanical design enteria limits will not be exceeded for the ft'el; therefore, the fuel cladding will act be overstressed. The LHGR limits are fuel dependent.

Therefore, the GE fuelin Q2C15revi has the same LHGR limit as the GE fuelin Q2C15.

The SPC ATRIUM-90 in Q2C15rev1 fuel has tha same LHGR hmit as the O2C15 ATRIUM-98 fuel. Therefore, there is no change in the LHGR limits and no reduction in the margin of safety for the LHGR.

Page 46 of $1

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

l NSWP-A-04 Revision 0 10 CFR$0.59 Safety Evaluation rocess If the response to Sten 14 is "YES". the Martin of Safety is reduced and an Unreviewed Safety Ouestion esists.

Proceed to Stco 16.

NOTE in some cases, the proposed activity being evaluated may be a candidate for adding words to the UFSAR. Consideration should be given to adding a discussion of key regulatory

. Issues, regulatory documents (Generic Letters, Regulatory Guides, NRC Bulletins, etc.),

station commitments and new equipment. (See Regulatory Guide 1,70 for level of detail) t F

i -

Page 47 of 51 L

NSifP-A-04 C := ton 0 10 CFR30 59 Safety Evaluation Process EXHIBIT G 10CFR50.59 SAFETY EVALUATION

15. Is a change to the UFSAR needed?

O YES- UFSAR/ Changes have been initiated via 0 Tracking Control No.:

QB 0 All applicable changes are attached.

Proceed to next step.

NO- Proceed to next step.

16. Check one of the following:

@ No Unreviewed Safety Question will result (Steps 10,11,12, and 14) Ak? no Technical Specification revision (Step 9) is required. 'Ihe proposed activity may be implemented under the provision of 10CFR50.59 in accordance with applicable procedures.

An Unreviewed Safety Question was identified in either Steps 10,11,12, or 14. The proposed activity cannot be implemented under the provisions of 10CFR5 . 59 and SilALL NOT be implemented u ithout NRC approval.

O A new Technical Specification, Technical Specification revision (Step 9), or other License Ainendment is reqiiired. Notify Station Regulatory Assurance that a License Amendment is required. Mark below as applicable.

The change provides more conservative operational restrictions than those in the cuaent Technical Specifications. In order to assure safe operation the following shall be performed:

• A Reportability Review e A License Amendment Request. Regulatory Assurance concurs that administrative controls shall be implemented until receipt of the approved License Amendment.

Regulatory Assurance Concurrence:

Print / Signature / Date The proposed activity is not a Facility change and shall not be implemented under 10CFR50.59.

O The proposed activity is a Facility change. Mark below as applicable.

A revision to an existing Technical Specification is required The change SilALL NOT be installed until receipt of License Amendment O The installation work for this change will not conflict with any existing Technical Specifications. On Site Review may authorize installation, but not operation, prior to receipt 44NRC approval of the License Amendment.

Page 48 of 51

T' NSilT-A 04 Revision 0 10 CFR$0 $9 Safety Evaluation Process EXHIBIT G 10CFR50.59 SAFETY EVAL.UATION NOTE l Partial modifications and/or separate 10CFR50.59 reviews for portions of the work may be used to facilitate installatiori.

I

17. Assign a Safety Evaluation tracking number and write it on page 1 of this form. Si;; natures below inay be obtained prior to assigning a tracking number.

18. Ihe preparer has determined that the documentation is ade uate to support the abos e conclusion.

bj p J (5c< bbomL A. ei hingh Preparer:

are: O/kM Print i SignatT/J' s CG 1(o Dept. /1.ocation: ~"b NESf M _ Phone: x 378 X 3032_

19. The reviewer has determined that the documentation is adequate to support the abose conclusion and agrees with the conclusion, Reviewer: W"* MA ~

Date: ~

N# ~ 2- 8 2 Print ' / [ignature Own6/I orr

• Dept. /1.ocation: Phone:

Page 49 of 51 L

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NSWP-A 04 RsvisIon 0 10 CFR30.39 Safety Evaluation Process

REFERENCES:

1, SPC document, *lmpact of Failed / Bypassed LPRMs and TIPS and Extended LPRM Calibration Interval  :

on Radial Bundle Power Uncertainty", EMF-1903(P); Revision 2, October 1996.

2. SPC' document, " Application of the ANFB Critical Power Correlation to Coresident GE Fuel for Quad Cities Unit 2 Cycle 15", EMF-96 051(P), May,1996.

3. Comed Document, BNDO.96-049," Quad Cities 2 Cycle 15 Fuel Assembly Mislocation Calculat;ons".

4. SPC Document, " Quad Cities LOCA-ECCS Analysis MAPLHGR Limits for ATRillM-98 Fuel', EMF-96 185(P) Revision 1, NDIT 970015, Siemens Power Corporation, Nuclear Division, March 1996.

5. SPC Letter

• Disposition of ATWS Event at Outd Cities", JHR:96:205, Siemens Power Corporation,

( RicVrand WA, May 31,1996.

6. SPC Document,'LOCA Break Spectrum Analysis for Quad Cities Units 1 and 2', EMF-96-184(P), NDIT 970014; Siemens Power Corporation, December 1996.

7, Quad Cities Unit 2 Cycle 15 Reload Analysis, EMF 95 /7, Revision 2, Siemens Power Corporation -

Nuclear Div'sion, i April 1997

8. Quad Cities Unit 2 Cycle 15 Plant Transient Analysis, EMF-96-180 (P), Revision 1, Siemens Power Corporation - Nuclear Division, April 1997,

9. Comed Calculation Number BNDO:97-062, *Q2C15 Licensing Analyses Application to CiSA (Assuming -

YJ8327 is the Leabr and LYU336 is Reinserted", September 29,1997.

10. NDIT 970175 Revision 0, Letter, D. E. Garber to Dr. R. J. Chin (DEG:97:156), dated October 1,1997.

Subject " Quad Cities Unit 2 Cycle 15 Assessment of Licensing Analyses After Replacement of Failed Fuel Assembly (Revision)".

ii. Letter, E. A. McVey to J. Hutchinson (NFS:BSS:97-056), dated May 14,1997, Subject " Revised Q2C* 5 NFS Reload Evaluation and UFSAR Changes" and Letter E.A. McVey to J. Hutchinson (NFS:BSS:97-028), dated April 28,1997, Subject 'O7-Site and Off-Sito Reviews of the Quad Cities Unit 2 Cycle 15 (Q2C15) Reload Licensing Package under 10CFR50.59".

12J Technical Specification Amendment, "Outd Cities Nuclear Power Station Units 1 and 2, Dresden Nuclear Power Station Units 2 and 3, LaSalle County Nuclear Power Station Units 1 and 2, Application for Amend.nent Request to Facility Operating Licenses, DPR-29 and DPR 30, DPR-19 and DPR 25, and NPF-11 and NPF-18, respectively, Technical Specification Changes for Transition to Siemens -

Power Corporation ATRIUM 9B Fuel, Docket Nos. 50-254 and 50-265,50-237 and 50-249, and 50-373

, and 50-373, respectively", dated August,29,1997.

13. Comed letter," Comed Response to NRC Staff Request for Additional Information (RAI) Regarding the Application of Siemens Power Corporation ANFB Critical Power Correlation to uc-asident General Electric Fuel for LaSalle Unit 2 Cycle 8 and Quad Cites Unit 2 Cycle 15, NRC Docket No.'s 50-373/374 and 50-254/265*, J.B. Hosmer to U.S. NRC, July 2,1996, transmitting t.1e topical report; Application of the'ANFB Critical Power Correlation to Coresident GE Fuel for Quad Cites Unit 2 Cycle 15, EMF . 051(P), Siemens Power Corporation - Nuclear Division, May 1900, and n ! lated information.

Page 50 of 51 i

I NSliP-A-04 Revision 0 10 CFR50.59 S$ty Evaluatior; Process

14. NDIT 96-134 Revisb11,'OC EOD and EOOS Safety Analysis for ATRIUM-9B Fuel, EMF-96-037(P),

Rev 1", September 16,1996.

15. Memo D. A. Pearson to T. Rieck, "Quac Cities Unit 2 Cycle 15 Inspections of Assembhes to be Reinserted to Replace Failed Fuel Assemblies", September 25,1997.

16. Corned Memo, NFS BSS:97111, T.A. Rieck to L.W. Pearce,'Ouad Cities Unit 2 Cycle 15 Rev 1 Startup Testing Recommendations" October 1,1997.

PageSiof51