PLA-5509, Supplement to Proposed Amendments 243 & 207 Adoption of NRC Approved Generic Changes to Improved Technical Specifications

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Supplement to Proposed Amendments 243 & 207 Adoption of NRC Approved Generic Changes to Improved Technical Specifications
ML022480229
Person / Time
Site: Susquehanna  Talen Energy icon.png
Issue date: 08/23/2002
From: Shriver B
Susquehanna
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
PLA-5509
Download: ML022480229 (18)
v • d • e


Text

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%sit Bryce L. Shriver PPL Susqiehanna, LLC S"33-Senior Vice President and 769 Salem Boulevard Chief Nuclear Officer Berwick, PA 18603 Tel 570 542 3120 Fax 570 542.1504 blshnver@pplweb corn P12 AUG 2 3 2002 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Mail Station OP 1-17 Washington, DC 20555 SUSQUEHANNA STEAM ELECTRIC STATION SUPPLEMENT TO PROPOSED AMENDMENT NO. 243 TO LICENSE NFP-14 AND PROPOSED AMENDMENT NO. 207 TO LICENSE NFP-22: ADOPTION OF NRC APPROVED GENERIC CHANGES TO IMPROVED TECHNICAL SPECIFICATIONS Docket Nos. 50-387 PLA-5509 and 50-388

Reference:

1) PLA-5372, R G. Byram (PPL) to USNRC, "ProposedAmendmentNo. 243 to License NFP-14 andProposedAmendmentNo. 207 to License NFP-22: Adoption of NRC Approved Generic Changes to Improved Technical Specifications,"dated October 16, 2001
2) PLA-5467, B. L. Shriver (PPL)to USNRC, "ProposedAmendment No. 211 to Unit 2 Licensee NPF-22: MCPR Safety Limits andReference Changes," datedJuly 17, 2002 The purpose of this letter is to provide supplemental information necessary for the NRC staff to complete its review of the license amendment proposed in Reference 1. The information provided was discussed via telecon between PPL and NRC on July 31, 2002.

Attachment I to this letter contains the responses to each of the NRC questions discussed during the telecon. The responses to NRC Questions 2 and 3 require PPL to revise some of the camera-ready pages provided with Reference 1. These revisions also affect the markups in Reference 2; however, no changes will be made to the Reference 2 submitted Technical Specification pages at this time. A separate submittal will be provided for those changes to the previously submitted markups and camera-ready TS pages.

Attachment 2 to this letter contains the Technical Specification page markups from Reference 1, including new camera-ready pages depicting those changes.

_ýDO I

Region I PLA-5509 Should you have any questions or require additional information, please contact Mr. Duane L. Filchner at (610) 774-7819.

Sincerely,

(. Shriver Attachments NRC Region I Mr. S. Hansell, NRC Sr. Resident Inspector Mr. T. G. Colbum, NRC Sr. Project Manager Mr. R. Janati, DEP/BRP Mr. E. M. Thomas, NRC Project Manager

BEFORE THE BEFORE THE UNITED STATES NUCLEAR REGULATORY COMMISSION In the Matter of PPL Susquehanna, LLC: Docket No. 50-388 SUPPLEMENT TO PROPOSED AMENDMENT NO. 207 TO LICENSE NFP-22: ADOPTION OF NRC APPROVED GENERIC CHANGES TO IMPROVED TECHNICAL SPECIFICATIONS UNIT NO. 2 Licensee, PPL Susquehanna, LLC, hereby files a revision to its Facility Operating License No. NPF-22 dated March 23, 1984.

This amendment involves a revision to the Unit 2 Technical Specifications.

Y 3/4- -

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PPL Susquehanna, LLC Notarial Seal

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-~. Colgitanp Carver, Notary Public

  • -,*. / l . Salem3-Wp., Luzerne County My Cdo'nrmuslon Expires Aug. 30, 2003 SMember, Pennsylvanla Association o? Notaries v tSr. Vice President and Che"NucleaOfie Sworn to and subscribed before me this,?3 day ofe:c4 , 2002.

Notary Public Notarv Public

BEFORE THE BEFORE THE UNITED STATES NUCLEAR REGULATORY COMMISSION In the Matter of PPL Susquehanna, LLC: Docket No. 50-387 SUPPLEMENT TO PROPOSED AMENDMENT NO. 243 TO LICENSE NFP-14: ADOPTION OF NRC APPROVED GENERIC CHANGES TO IMPROVED TECHNICAL SPECIFICATIONS UNIT NO. 1 Licensee, PPL Susquehanna, LLC, hereby files a revision to its Facility Operating License No. NPF-14 dated July 17, 1982.

This amendment involves a revision to the Unit 1 Technical Specifications.

PPL Susquehanna, LLC

- Notarial Seal S.. Colleen MCarver, Notary Public

,....>-Saler~ Twp.,Luzerne County

,MyCommlssi*riExpires Aug. 30, 2003 Member, Pecnsylvanla Association of Notaries BI L. ShýiAvei` tý1256-'

Si. Vice President and Chief Nuclear Officer 1"Latý-

Sworn to and subscribed before me this *5,day of , 2002.

$04 0 ;)-

Notary Public

Attachment I to PLA-5509 Response to Request for Additional Information

Attachment 1 to PLA-5509 Page 1 of 4 Attachment 1 - Response to Request for Additional Information NRC Question 1 As stated in Generic Letter 88-16, "Removal of Cycle-Specific Parameter Limits from Technical Specifications," the analytical methods used to determine the cycle-specific core operating limits shall be those previously reviewed and approved by NRC. Five core operating limits are specified in TS 5.6.5.a and seventeen approved analytical methods are identified in TS 5.6.5.b to support those cycle-specific core operating limits specified in TS 5.6.5.a. Please identify which cycle-specific core operating limit specified in TS 5.6.5.a is supported by those seventeen approved analytical methods identified in TS 5.6.5.b.

PPL Response The following table provides a cross-reference listing of the core operating limit specified in TS 5.6.5.a and the applicable reference from the Unit 2 TS 5.6.5.b. This table is provided to clarify the relationship between references and the applicable core operating limits. The table was not required for the Reference 1 submittal.

TS 5.6.5.a TS 5.6.5.b

1. The Average Planar Linear 2. XN-NF-80-19 (P)(A), "Exxon Nuclear Heat Generation Rate for Methodology for Boiling Water Reactors," Exxon Specification 3.2.1; Nuclear Company, Inc.
6. NEDC-32071P, "SAFER/GESTR-LOCA Loss of Coolant Accident Analysis," GE Nuclear Energy.
7. NE-092-001A, "Licensing Topical Report for Power Uprate With Increased Core Flow,"

Pennsylvania Power & Light Company.

9. NEDE-240 11-P-A- 10, "General Electric Standard Application For Reactor Fuel."
11. ANF-91-048 (P)(A), "Advanced Nuclear Fuels Corporation Methodology for Boiling Water Reactors EXEM BWR Evaluation Model."
15. Caldon, Inc., "TOPICAL REPORT: Improving Thermal Power Accuracy and Plant Safety While Increasing Operating Power Level Using the LEFM41 System," Engineering Report - 80P.
16. Caldon, Inc., "Supplement to Topical Report ER-80P: Basis for a Power Uprate with the LEFM'J'm or LEFM CheckPluslm System,"

Engineering Report ER- 160P.

Attachment 1 to PLA-5509 Page 2 of 4 TS 5.6.5.a TS 5.6.5.b

2. The Minimum Critical Power 1. PL-NF-90-001-A, "Application of Reactor Ratio for Specification 3.2.2; Analysis Methods for BWR Design and Analysis."
2. XN-NF-80-19 (P)(A), "Exxon Nuclear Methodology for Boiling Water Reactors," Exxon Nuclear Company, Inc.
4. ANF-524 (P)(A), "Advanced Nuclear Fuels Corporation Critical Power Methodology for Boiling Water Reactors."
5. ANF-l 125 (P)(A), "ANFB Critical Power Correlation."
7. NE-092-001A, "Licensing Topical Report for Power Uprate With Increased Core Flow,"

Pennsylvania Power & Light Company.

9. NEDE-2401 1-P-A-10, "General Electric Standard Application For Reactor Fuel."
13. XN-NF-79-71 (P)(A), "Exxon Nuclear Plant Transient Methodology for Boiling Water Reactors."
14. EMF-1997 (P)(A) "ANFB-10 Critical Power Correlation."
15. Caldon, Inc., "TOPICAL REPORT: Improving Thermal Power Accuracy and Plant Safety While Increasing Operating Power Level Using the LEFM4ThM System," Engineering Report - 80P.
16. Caldon, Inc., "Supplement to Topical Report ER-80P: Basis for a Power Uprate with the LEFM4P or LEFM CheckPlusTM System,"

Engineering Report ER-160P.

3. The Linear Heat Generation 3. XN-NF-85-67 (P)(A), "Generic Mechanical Rate for Specification 3.2.3; Design for Exxon Nuclear Jet Pump BWR Reload Fuel," Exxon Nuclear Company, Inc.
7. NE-092-001A, "Licensing Topical Report for Power Uprate With Increased Core Flow,"

Pennsylvania Power & Light Company.

8. PL-NF-94-005-P-A, "Technical Basis for SPC 9x9-2 Extended Fuel Exposure at Susquehanna SES."
9. NEDE-2401 1-P-A-10, "General Electric Standard Application For Reactor Fuel."
10. ANF-89-98 (P)(A), "Generic Mechanical Design Criteria for BWR Fuel Designs," Advanced Nuclear Fuels Corporation.
17. EMF-85-74 (P)(A), "RODEX 2A (BWR) Fuel Rod Thermal-Mechanical Evaluation Model."

1.

Attachment 1 to PLA-5509 Page 3 of 4 TS 5.6.5.a TS 5.6.5.b

4. The Average Power Range 3. XN-NF-85-67 (P)(A), "Generic Mechanical Monitor (APRM) Gain and Design for Exxon Nuclear Jet Pump BWR Reload Setpoints for Specification Fuel," Exxon Nuclear Company, Inc.

3.2.4; and 7. NE-092-001A, "Licensing Topical Report for Power Uprate With Increased Core Flow,"

Pennsylvania Power & Light Company.

8. PL-NF-94-005-P-A, "Technical Basis for SPC 9x9-2 Extended Fuel Exposure at Susquehanna SES."
9. NEDE-240 11-P-A-10, "General Electric Standard Application For Reactor Fuel."
10. ANF-89-98 (P)(A), "Generic Mechanical Design Criteria for BWR Fuel Designs," Advanced Nuclear Fuels Corporation.
17. EMF-85-74 (P)(A), "RODEX 2A (BWR) Fuel Rod Thermal-Mechanical Evaluation Model."
5. The Shutdown Margin for 1. PL-NF-90-001-A, "Application of Reactor Specification 3.1.1. Analysis Methods for BWR Design and Analysis."

It should be noted that the Unit 2 TS 5.6.5.b Reference 12 is a repeat of TS 5.6.5.b Reference 2. On July 17, 2002 PPL submitted PLA-5467 (Ref. 2) which includes removal of Reference 12. Therefore, Reference 12 is not included in the above table.

Also note that the Caldon Reports, (Reference 15 and 16), were approved by the NRC Staff in Safety Evaluations dated March 8, 1999 and November 24, 2000 respectively as documented in Section 3.4.2 of the Safety Evaluation for Unit 1 Amendment 194 and Unit 2 Amendment 169 dated July 6, 2001. These amendments approved the increase in licensed power level from 3,441 megawatts thermal to 3,489 megawatts thermal.

The Unit 1 list is similar to the Unit 2 list, except that Unit 2 Reference 9 (NEDE-2401 1-P-A-10) is not included in Unit 1 TS 5.6.5.b. Unit 2 Reference 9 was required to support four GE lead use assemblies that were only inserted in Unit 2.

Therefore, Unit 2 Reference 9 is not applicable to Unit 1. This results in Unit 1 containing 16 references with References 9 through 16 equivalent to the Unit 2 References 10 through 17.

Attachment 1 to PLA-5509 Page 4 of 4 NRC Question 2 Please identify the current fuel types loaded in Susquehanna Units 1 and 2 and provide justification that GE topical reports - NEDC-32071P and NEDE-2401 l-P-A-10 are needed to support the cycle-specific core operating limits specified in TS 5.6.5.a.

PPL Response Currently, Susquehanna Units 1 and Unit 2 contain full cores of the FRA-ANP ATRIUM-10 fuel design. GE topical report NEDC-32071P was required to support the FRA-ANP 9x9-2 fuel design. GE topical report NEDE-2401 1-P-A-10 was required to support GE lead use assemblies. At this time neither Susquehanna unit contains FRA-ANP 9x9-2 or GE lead use assemblies. Therefore, these references can be removed. The references were not removed as part of the Reference 1 submittal because at the time there was a possibility that the GE lead use assemblies or FRA-ANP 9x9-2 fuel assemblies may be reinserted in the Susquehanna cores.

Based on a Telecon between NRC (T. Colbum and T. Huang) and PPL (D. Filchner, A. Dyszel, and J. Smith) on July 31, 2002, it was concluded that PLA-5372 (Ref. 1) will be revised to eliminate both reports from the Technical Specifications as shown on the markup in Attachment 2. A revision to the markups and camera-ready TS pages in PLA-5467 (Ref. 2) is also required as a result of the changes to Reference 1, however, that revision will be submitted to the NRC via separate letter.

NRC Question 3 Provide justification that EMF-85-74 (P), "RODEX 2A (BWR) Fuel Rod Thermal-Mechanical Evaluation Model" is needed to support current Susquehanna Units 1 and 2 operation.

PPL Response As previously noted on page 3 of Attachment 1 to Reference 1, EMF-85-74 provides the latest approved burnup limits for Framatone-ANP fuel. This reference supports the fuel assembly bumup limits for the fuel types currently loaded in Susquehanna. This is an NRC approved topical and an (A) will be added to the listed reference as shown on the markup in Attachment 2.

Attachment 2 to PLA-5509 PLA-5372 Markups and Camera-Ready Pages

fWO, e of fLA- 67-37, A 4-4c. / e, 5.6 Reporting Requirements 5.6 Reporting Requirements (continued)

(102% of 3441 MWt) remains the initial power level for the bounding licensing analysis.--.......

Future revisions of approved analytical methods listed in this Technical Specification that are currently referenced to 102%

of rated thermal power (3510 MWt) shall include reference that the licensed RIP is actually 3489 MWt. The revisions shall document that the licensing analysis performed at 3510 MWt bounds operation at the RIP of 3489 MWt so long as the LEFM/'

.sy-stemis-used as the feedwater flow measurement input into the core thermal power calculation.

The approved analytical methods are described in the following

- documents, the approved version(s) of which are specified in the COLR.

1. PL-NF-90-001-A, "Application of Reactor Analysis Methods for BWR Design and Analysis."
2. XN-NF-80-19(P)(A), "Exxon Nuclear Methodology for Boiling Water Reactors," Exxon Nuclear Company, Inc.
3. XN-NF-85-67(P)(A), "Generic Mechanical Design for Exxon Nuclear Jet Pump BWR Reload Fuel, "Exxon Nuclear Company, Inc.
4. ANF*524(P)-(A), "Advaficed Nuclear Fuels Corporation Critical Power Methodology for Boiling Water Reactors."
5. ANF-1125(P)(A), "ANFB Critical Power Correlation."

NE-Og2-OO1A, "Licensing Topical Report for Power Uprate

- With Increased Core Flow," Pennsylvania Power & Light Company.

I __ ( PL-NF-94-005-P-A, "Technical Basis for SPC 9x9-2 Extended-Fue-Exposure at Susquehanna SES."

. ANF-89-98(P)(A), "Generic Mechanical Design Criteria for BWR Fuel Designs," Advanced Nuclear Fuels Corppration.

(continued)

SUSQUEHANNA - UNIT 1 TS / 5.0-22 Amendment

5. &i Reo f Y, - G37- AeAa-Awe 4(- Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 COL R (continued)

ANF-91-048(P)(A), "Advanced Nuclear Fuels Corporation a 1>0 Methodology for Boiling Water Reactors EXEM BWR Evaluation Model."

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

rI4 . . _N=NF-=7 4-Jl*P_-4A),-!xxon-Nucl ear--Plant Transient

'-Methodology for Boiling Water Reactors." I EMF-1997(P)(A), "ANFB-IO Critical Power Correlation."

Caldon, Inc., "TOPICAL REPORT: Improving Thermal Power Accuracy and Plant Safety While Increasing Operating Power Level Using the LEFM/"' System," Engineering Report

- 80P.

1 Caldon, Basis forInc., "Supplement a Power Uprate withto Topical Report the LEFM/M ER-80P:

or LEFM CheckPlusm System, "Engineering Report ER-160P.

0 EMF-85-74(PY', "RODEX 2A (BWR) Fuel Rod Thermal Mechanical Evaluation Model."

-c. - The core operating limits shall be determined such that all

_____ applicable limits (e.g., fuel ther-al- mechanical limits, core th-eYira-hy-drauiic ii-t-sEmergency Core o-oling -

Systems. (ECCS) limits,- nuclear limits such as SDM, transient analysis limits, -and accident analysis limits) of the safety analysis are met.

d. The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

(continued)

SUSQUEHANNA - UNIT I TS / 5.0-23 Amendment

Reporting Requirements 5.6 5.6 Reporting Reqdirements (continued) 5.6.5 COLR (continued)

(102% of 3441 MWt), remains the initial power level for the bounding licensing analysis.

Future revisions of approved analytical methods listed in this Technical Specification that are currently referenced to 102% of rated thermal power (3510 MWt) shall include reference that the licensed RTP is actually 3489 MWt. The revisions shall document that the licensing analysis performed at 3510 MWt bounds operation at the RTP of 3489 MWt so long as the LEFMVTM system is used as the feedwater flow measurement input into the core thermal power calculation.

The approved analytical methods are described in the following documents, the approved version(s) of which are specified in the COLR.

1. PL-NF-90-001 -A, "Application of Reactor Analysis Methods for BWR Design and Analysis."
2. XN-NF-80-19(P)(A), "Exxon Nuclear Methodology for Boiling Water Reactors," Exxon Nuclear Company, Inc.
3. XN-NF-85-67(P)(A), "Generic Mechanical Design for Exxon Nuclear Jet Pump BWR Reload Fuel, "Exxon Nuclear Company, Inc.
4. ANF-524(P)(A), "Advanced Nuclear Fuels Corporation Critical Power Methodology for Boiling Water Reactors"
5. ANF-1 125(P)(A), "ANFB Critical Power Correlation."
6. NE-092-O01A, "Licensing Topical Report for Power Uprate With Increased Core Flow," Pennsylvania Power & Light Company.
7. PL-NF-94-005-P-A, "Technical Basis for SPC 9x9-2 Extended Fuel Exposure at Susquehanna SES.",
8. ANF-89-98(P)(A), "Generic Mechanical Design Criteria for BWR Fuel Designs," Advanced Nuclear Fuels Corporation.

(continued)

SUSQUEHANNA- UNIT 1 TS / 5.0-22 Amendment

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 COLR (continued)

9. ANF-91-048(P)(A), "Advanced Nuclear Fuels Corporation Methodology for Boiling Water Reactors EXEM BWR Evaluation Model.
10. XN-NF-80-19(P)(A), "Exxon Nuclear Methodology for Boiling Water Reactors.
11. XN-NF-79-71 (P)(A), "Exxon Nuclear Plant Transient Methodology for Boiling Water Reactors."
12. EMF-1 997(P)(A), "ANFB-1 0 Critical Power Correlation."
13. Caldon, Inc., 'TOPICAL REPORT: Improving Thermal Power Accuracy and Plant Safety While Increasing Operating Power Level Using the LEFMVTM System," Engineering Report - 80P.
14. Caldon, Inc., "Supplement to Topical Report ER-80P: Basis for a Power Uprate with the LEFMVTM or LEFM CheckPlusTM System,", Engineering Report ER-160P.
15. EMF-85-74 (P)(A), "RODEX 2A (BWR) Fuel Rod Thermal-Mechanical Evaluation Model."
c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.
d. The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

(continued)

SUSQUEHANNA - UNIT 1 TS / 5.0-23 Amendment

  • k4 0 LA-a 5 T372 A+1Lvmeo1 +4 Reporting Requirements 5.6 5.6 Reporting Requirements (continued) core thermal power level may not exceed the originally approved RTP of 3441 MWt, but the value of 3510 MWt (102%

of 3441 MWt) remains the initial power level for the bounding licensing analysis.

Future revisions of approved analytical methods listed in this Technical Specification that are currently referenced to 102% of rated thermal power (3510 MWt) shall include reference that the licensed RIP is actually 3489 MWt. The r-ev]ons-shaldocument-that-theJ-lcensing analysis performed at 3510 MWt bounds operation at the RIP of 3489 MWt so long as the LEFM/N system is used as the feedwater flow measurement input into the core thermal power calculation.

The approved analytical iethods are described-in the following documents, the approved version(s) of which are specified in the COLR.

1. PL-NF-90-001-A, "Application of Reactor Analysis Methods for BWR Design and Analysis."
2. XN-NF-80-19(P)(A), "Exxon Nuclear Methodology for Boiling Water Reactors," Exxon Nuclear Company, Inc.
3. XNAfE_-85-67(P}(A),-"Generic Mechanical Design for Exxon Nuclear Jet Pump BWR Reload Fuel," Exxon Nuclear Company, Inc.
4. ANF-524(P)(A), "Advanced Nuclear Fuels Corporation Critical Power Methodology for Boiling Water Reactors."
5. ANF-1125(P)(A), "ANFB Critical Power Correlation."
6. C-071 "1 E E 0C S '- *NE-092-001A, "Licensing Topical Report for Power Uprate With Increased Core Flow,"-Pennsylvania Power &

Light Company.

PL-NF-94-005-P-A, "Technical Basis for SPC 9x9-2 Extended Fuel Exposure at Susquehanna SES."

(continued)

SUSQUEHANNA - UNIT 2 TS / 5.0-22 Amendment

6LA-SY12-, "A+me-** Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 COLR (continued)

ANF-89-98(P)(A), "Generic Mechanical Design Criteria for BWR Fuel Designs," Advanced Nuclear Fuels Corporation.

rfN ANF-91-048(P)(A), "Advanced Nuclear Fuels Corporation Methodology for Boiling Water Reactors EXEM BWR

-- - E-val uati on-Model." ......

A--, XN-NF-80-19(P)(A), "Exxon Boiling Water Reactors."

Nuclear Methodology for

.hJ ... XN-NF_-79-T7](P.)(A), ".Exxon Nuclear Plant Transient Methodology for Boiling Water Reactors."

EMF-1997 (P)(A), "ANFB-1O Critical Power Correlation." I I

Caldon, Inc., "TOPICAL REPORT: Improving Thermal Power Accuracy and Plant Safety While Increasing Operating Power Level Using the LEFM/TM System,"

Engineering Report - 80P.

Caldon, Inc., "Supplement to Topical Report ER-80P:

- -Basis-for a- Power Uprate-with-t-he-L-EFM/IM-or-LEFM CheckPlusW--System,-!'Engineering Report ER-160P."

  • -EMF-85-7W-,-RODEX 2A (BWR) Fuel Rod Thermal

.. . MechanicalEvaluation-Model." -.

c. -- The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.
d. The COLR, including any midcycle revisions or supplements, shall-be- *oVid*ed up-on issuance for ea*h*l*bIb-af-&-yd-le-to the NRC.

(continued)

SUSQUEHANNA - UNIT 2 TS / 5.0-23 Amendment

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 COLR (continued)

6. N E-092-001A, "Licensing Topical Report for Power Uprate With Increased Core Flow," Pennsylvania Power & Light Company.
7. PL-NF-94-005-P-A, "Technical Basis for SPC 9x9-2 Extended Fuel I Exposure at Susquehanna SES".
8. ANF-89-98(P)(A) "Generic Mechanical Design Criteria for BWR Fuel Designs," Advanced Nuclear Fuels Corporation.
9. ANF-91-048(P)(A), "Advanced Nuclear Fuels Corporation Methodology for Boiling Water Reactors EXEM BWR Evaluation Model."
10. XN-NF-80-19(P)(A), "Exxon Nuclear Methodology for Boiling Water Reactors.

(continued)

SUSQUEHANNA - UNIT 2 TS / 5.0-23 Amendment

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 COLR (continued)

11. XN-NF-79-71 (P)(A) "Exxon Nuclear Plant Transient Methodology for I Boiling Water Reactors."
12. EMF-1 997 (P)(A) "ANFB-10 Critical Power Correlation."
13. Caldon, Inc., "TOPICAL REPORT: Improving Thermal Power Accuracy I and Plant Safety While Increasing Operating Power Level Using the LEFMV/TM System," Engineering Report - 80P.
14. Caldon, Inc., "Supplement to Topical Report ER-80P: Basis for a Power Uprate with the LEFMVTM or LEFM CheckPlusTm System, "Engineering Report ER-1 60P.
15. EMF-85-74 (P)(A), "RODEX 2A (BWR) Fuel Rod Thermal-Mechanical Evaluation Model."
c. The core operating limits shall be determined such that all applicable limits (e.g.,

fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d. The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

(continued)

SUSQUEHANNA-UNIT2 TS / 5.0-24 Amendment

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