View source for PLA-6132, Susquehanna Steam Electric Station Proposed Amendments No. 259 to Unit 2 License NPF-22: MCPR Safety Limits and Reference Changes

{{Adams
| number = ML070670263
| issue date = 11/16/2006
| title = Susquehanna Steam Electric Station Proposed Amendments No. 259 to Unit 2 License NPF-22: MCPR Safety Limits and Reference Changes
| author name = McKinney B T
| author affiliation = PPL Susquehanna, LLC
| addressee name = 
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000388
| license number = NPF-022
| contact person = 
| case reference number = PLA-6132
| document type = Letter type:PLA, License-Application for Facility Operating License (Amend/Renewal) DKT 50, Technical Specification, Bases Change
| page count = 43
}}

=Text=
{{#Wiki_filter:Brltt 1. McKInney Sr. Vice President 
& Chief Nuclear Officer PPL Susquohanna, LLC 769 Salem Boulevard Berwick, PA 18603 Tel. 570.542.3149 Fax 570.542.1504 btmckinney 
@ pplweb.com ELM el~~,1 NOV 1 6 2006 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Mail Stop OPI-17 Washington, DC 20555 SUSQUEHANNA STEAM ELECTRIC STATION PROPOSED AMENDMENT NO. 259 TO UNIT 2 LICENSE NPF-22: MCPR SAFETY LIMITS AND REFERENCE CHANGES.PLA-6132 Docket No. 50-388 In accordance with the provisions of 10 CFR 50.90, PPL Susquehanna, LLC is submitting a request for an amendment to the Technical Specifications for Susquehanna Unit 2.The purpose of this letter is to propose changes to the Susquehanna Steam Electric Station Unit 2 Technical Specifications.
Included is a revision to Section 2.1.1.2 which reflects the Unit 2 Cycle 14 (U2C14) Minimum Critical Power Ratio (MCPR) Safety Limits for two-loop and single-loop operation.
Additionally, Section 5.6.5.b is revised to reflect the NRC-approved methodology used in the MCPR Safety Limit Analysis.The enclosure to this letter contains PPL's evaluation of this proposed change. Included are a description of the proposed change, technical analysis of the change, regulatory analysis of the change (No Significant Hazards Consideration and the Applicable Regulatory Requirements), and the environmental considerations associated with the change.Attachment 1 to this letter contains the applicable pages of the Susquehanna SES Unit 2 Technical Specifications, marked to show the proposed change.Attachment 2 is included which identifies that there are no new regulatory commitments associated with this change.Attachment 3 contains the applicable pages of the Susquehanna SES Unit 2 Technical Specifications Bases, marked to show the proposed changes (Provided for Information). Document Control Desk PLA-6132 Attachment 4 has been provided as a description of the U2C14 preliminary core composition to assist in your review.Attachment 5 provides the preliminary U2C14 Core Loading Pattern.Attachment 6 contains preliminary descriptions of the reload fuel bundles utilized in U2C14.Attachment 7 contains a listing of the AREVA NP, approved methodology, and the applicable Technical Specification LCO's.Attachment 8 provides a diagram of the NRC-approved MCPR Safety Limit Methodology.
The proposed changes have been approved by the Susquehanna SES Plant Operations Review Committee and reviewed by the Susquehanna Review Committee.
PPL plans to implement the proposed changes in the spring of 2007 to support the startup of U2C14 operation.
Therefore, we request NRC complete its review of this change by February 16, 2007 with the changes effective upon startup following the Unit 2 13ah Refueling and Inspection Outage.Any questions regarding this request should be directed to Mr. Duane L. Filchner at (610) 774-7819.I declare under penalty of perjury that the foregoing is true and correct.Executed on: it B.T. Document Control Desk PLA-6132 

==Enclosure:==

PPL Susquehanna Evaluation of the Proposed Changes Unit 2 Cycle 14 MCPR Safety Limit and COLR References Attachments:
Attachment 1 -Proposed Technical Specification Changes Unit 2 (Markups)Attachment 2 -List of Regulatory Commitments (Unit 2)Attachment 3 -Proposed Changes to TS Bases Pages Unit 2 (Markup)Attachment 4 -Preliminary Description of Unit 2 Cycle 14 Core Composition Attachment 5 -Preliminary Unit 2 Cycle 14 Core Loading Pattern Attachment 6 -Preliminary Descriptions of Unit 2 Cycle 14 Fresh Fuel Description Attachment 7 -Listing of AREVA NP, Inc. Approved Methodology and Applicable LCOs Attachment 8 -MCPR Safety Limit Methodology cc: NRC Region I Mr. A. J. Blarney, NRC Sr. Resident Inspector Mr. R. V. Guzman, NRC Project Manager Mr. R. Janati, DEP/BRP Enclosure to PLA-6132 PPL Susquehanna Evaluation of the Proposed Changes Unit 2 Cycle 14 MCPR Safety Limit and COLR References
: 1. DESCRIPTION
: 2. PROPOSED CHANGE 3. BACKGROUND
: 4. TECHNICAL ANALYSIS 5. REGULATORY SAFETY ANALYSIS 5.1 No Significant Hazards Consideration

===5.2 Applicable===

Regulatory Requirements/Criteria
: 6. ENVIRONMENTAL CONSIDERATIONS
: 7. REFERENCES Enclosure to PLA-6132 Page 1 of 14 PPL EVALUATION

==Subject:==
Unit 2 Cycle 14 MCPR Safety Limit and COLR 

==References:==

TS Section 2.1.1.2 and 5.6.5.b.

==1.0 DESCRIPTION==

This letter is a request to amend Operating License NPF-22 for PPL Susquehanna, LLC (PPL), Susquehanna Steam Electric Station Unit 2 (SSES).The proposed changes would revise the Susquehanna Unit 2 Technical Specifications (TS) Section 2.1.1.2 to reflect the Unit 2 Cycle 14 (U2C14) Minimum Critical Power Ratio (MCPR) Safety Limit for two-loop and single-loop operation.
The change to Section 2.1.1.2 is necessary because, as a result of U2C14 cycle specific calculations, the two-loop and single-loop operation MCPR Safety Limits increased relative to the existing Unit 2 TS values. The proposed changes also would revise Susquehanna Unit 2 TS Section 5.6.5.b. TS 5.6.5.b lists the NRC-approved analytical methods used to determine the core operating limits contained in the unit / cycle specific Core Operating Limits Report (COLR). The proposed change to TS 5.6.5.b removes the AREVA ANFB-10 critical power correlation methodology report. The NRC-approved AREVA SPCB critical power correlation replaces the ANFB-10 critical power correlation for Unit 2 Cycle 14.It should be noted that the calculations performed to support the proposed changes to the two-loop and single-loop MCPR Safety Limits include the effects of ARTS / MELLLA (submitted to NRC via PLA-593 1, Reference 7.4) and extended power uprate (EPU)(submitted to NRC via PLA-6076, Reference 7.5). The MCPR Safety Limit calculations account for the increased operating domain (e.g., lower reactor core flow) allowed by ARTS / MELLLA and the increased power level for EPU. Therefore, two-loop and single-loop MCPR Safety Limits presented herein are applicable to current conditions, ARTS / MELLLA, and EPU.The proposed changes are described in detail in Section 4.0.The requested approval date (February 16, 2007) will allow time for the Core Operating Limits Report to be prepared and reviewed by the Plant Operation Review Committee (PORC) prior to the Spring 2007 Unit.2 refueling outage.
Enclosure to PLA-6132 Page 2 of 14 2.0 PROPOSED CHANGE Specifically, the proposed changes would revise the following:
2.1 TS 2.1.1.2 The Minimum Critical Power Ratio (MCPR) Safety Limits (two-loop operation and single-loop operation) are revised from 1.09 (two-loop operation) and 1.10 (single-loop operation) to 1.11 (two-loop operation) and 1.14 (single-loop operation) to reflect results of the cycle specific MCPR Safety Limit analysis for Unit 2 Cycle 14.2.2 TS 5.6.5.b Core Operating Limits Report (COLR) references are revised to remove the ANFB-10 critical power correlation methodology report (EMF-1997(P)(A)).
For Unit 2 Cycle 14, the AREVA SPCB critical power correlation (EMF-2209(P)(A))
was used to calculate the MCPR Safety Limit and will be used to determine the MCPR operating limits. The core monitoring system will use the SPCB critical power correlation to ensure compliance with Unit 2 TS 3.2.2, "Minimum Critical Power Ratio (MCPR)." The SPCB critical power correlation is AREVA's most recent critical power correlation for the ATRIUMTM-10 fuel design and can be used to accurately predict assembly critical power. The SPCB critical power correlation has been reviewed and approved by the NRC and was previously added to the Unit 2 TS 5.6.5.b via PLA-5793 dated September 8, 2004 (Reference 7.2). The remaining references in TS 5.6.5.b were renumbered.
In summary, the proposed changes would revise the Susquehanna Unit 2 Technical Specifications (TS) Section 2.1.1.2. TS Section 2.1.1.2 is revised to reflect the Unit 2 Cycle 14 (U2C14) MCPR Safety Limit for two-loop and single-loop operation.
TS Section 5.6.5.b is revised to remove the ANFB-10 critical power correlation.
The TS Bases changes corresponding to the proposed TS change are included for information.
Enclosure to PLA-6132 Page 3 of 14 3.0 BACKGROUND 3.1 MCPR SAFETY LIMIT CHANGE Excessive thermal overheating of the fuel rod cladding can result in cladding damage and the release of fission products.
In order to protect the cladding against thermal overheating due to boiling transition, Safety Limits (Section 2.1.1.2 of the Susquehanna SES Unit 2 Technical Specifications) were established.
The change to Section 2.1.1.2 reflects the change from the previous Unit 2 MCPR Safety Limits to the U2C 14 MCPR Safety Limits.NUREG-0800, Standard Review Plan Section 4.4, specifies an acceptable, conservative approach to define this Safety Limit. Specifically, a Minimum Critical Power Ratio (MCPR) value is specified such that at least 99.9% of the fuel rods are expected to avoid boiling transition during normal operation or Anticipated Operational Occurrences (AOOs). Boiling transition is predicted using a correlation based on test data (i.e., a Critical Power Correlation).
The Safety Limit MCPR calculation accounts for various uncertainties such as feedwater flow, feedwater temperature, pressure, power distribution uncertainties (including the effects of fuel channel bow), and uncertainty in the Critical Power Correlation.
AREVA calculated both two-loop and single-loop Safety Limit MCPR values for Unit 2 Cycle 14 using NRC-approved analytical methods with the SPCB critical power correlation for ATRIUMrm-10 fuel. The proposed Safety Limit MCPR values (1.11 for two-loop operation and 1.14 for single-loop operation) assure that at least 99.9% of the fuel rods are expected to avoid boiling transition during normal operation or anticipated operational occurrences.
The MCPR Safety Limit analysis is the first in a series of analyses that assure the new core loading for U2C14 is operated in a safe manner. Prior to the startup of U2C14, other licensing analyses are performed (using NRC-approved methodology referenced in Technical Specification Section 5.6.5.b) to determine changes in the critical power ratio as a result of anticipated operational occurrences.
These results are combined with the MCPR Safety Limit values to generate the MCPR operating limits in the U2C14 COLR.The COLR operating limits thus assure that the MCPR Safety Limit will not be exceeded during normal operation or anticipated operational occurrences, thus providing the required protection for the fuel rod cladding.
Postulated accidents are also analyzed prior to the startup of U2C14 and the results shown to be within the NRC-approved criteria.
Enclosure to PLA-6132 Page 4 of 14 3.2. CHANGES TO COLR REFERENCES Core operating limits are established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and are documented in the Core Operating Limits Report (COLR). Technical Specification Section 5.6.5.b contains the NRC-approved methodology used to determine the core operating limits.The AREVA ANFB-10 critical power correlation was used in the previous Unit 2 analyses to determine the MCPR Safety Limit and the MCPR operating limits, and to ensure compliance with Unit 2 TS 3.2.2. For Unit 2 Cycle 14, the AREVA SPCB critical power correlation (EMF-2209(P)(A))
was used to calculate the MCPR Safety Limit and will be used to determine the MCPR operating limits and ensure compliance with Unit 2 TS 3.2.2. The SPCB critical power correlation is AREVA's most recent critical power correlation for the ATRIUMTM-10 fuel design, and can be used to accurately predict assembly critical power.The remaining references in Section 5.6.5.b following the ANFB-10 methodology report were renumbered.
Attachment 7 provides the relationship between the AREVA references in Section 5.6.5.b and the applicable Technical Specification Limiting Condition for Operation.
Enclosure to PLA-6132 Page 5 of 14 4.0 TECHNICAL ANALYSIS 4.1 MCPR SAFETY LIMIT CHANGE This Technical Specification change increases the MCPR Safety Limits from the current Unit 2 limits of 1.09 for two-loop and 1.10 for single-loop to the proposed limits of 1.11 for two-loop and 1.14 for single-loop.
The MCPR Safety Limit changesoccur due to cycle-to-cycle variation, changing the critical power correlation from ANFB-10 to SPCB, and consideration of the predicted percentage of rods in boiling transition.
A more detailed description of the reasons for the increase in the MCPR Safety Limit is provided below in sections "Cycle-to-Cycle Variation," "SPCB Critical Power Correlation" and"Results Summary." In addition, due to recent control cell interference issues due to channel bow on Unit 2, the channel bow assumptions for the Unit 2 Cycle 14 MCPR Safety Limit are discussed below ("Channel Bow" section).Cycle-to-Cycle Variation The preliminary Unit 2 Cycle 14 core consists of a full core of AREVA's ATRIUM T M-10 fuel design. The preliminary core composition is provided as Attachment 4 and the corresponding core loading pattern is provided as Attachment 
: 5. The fresh fuel for Unit 2 Cycle 14 is split into three different assembly types as described in Attachment 
: 6. The descriptions of the previous Unit 2 Cycle 13 core loading and exposed fuel assemblies used in both Unit 2 Cycle 13 and Unit 2 Cycle 14 can be found in the Susquehanna FSAR Section 4.3.As described previously in PPL correspondence with the NRC (e.g., PLA-5702 dated December 22, 2003, Reference 7.1) changes in both the two-loop and single-loop MCPR Safety Limits due solely to cycle-to-cycle variation are estimated to range from -0.01 to +0.01.SPCB Critical Power Correlation The current Unit 2 MCPR Safety Limit is based on AREVA's NRC-approved ANFB-10 critical power correlation, EMF-1997(P)(A), which is referenced in TS 5.6.5.b.Beginning with Unit 2 Cycle 14, PPL plans on using AREVA's NRC-approved SPCB critical power correlation, EMF-2209(P)(A), for MCPR Safety Limit determination, reload licensing analyses, and MCPR monitoring.
SPCB is currently included in TS 5.6.5.b per Section 4.2.Previous correspondence with the NRC (i.e., PLA-5990, dated December 1, 2005, Reference 7.3) demonstrated that the change to the SPCB critical power correlation may increase the single-loop MCPR Safety Limit by +0.01. The sensitivity provided in PLA-5990 is the only sensitivity performed by AREVA for a transition from ANFB-10 to Enclosure to PLA-6132 Page 6 of 14 SPCB since PPL was the only utility to utilize the ANFB-10 critical power correlation for reloads of ATRIUM'r M-10 fuel. Therefore, an increase in the single-loop MCPR Safety Limit of +0.01 to +0.02 is not unexpected due to a change from the ANFB-10 to SPCB critical power correlation.
Channel Bow The impact of channel bow on the MCPR Safety Limit is included due to recent fuel channel / control rod interference observed during previous Unit 2 operating cycles.The channel bow assumptions used for Unit 2 Cycle 14 are consistent with the previous Unit 2 Cycle 13 assumptions.
Therefore, the change in the two-loop and single-loop MCPR Safety Limit is not the result of a change in channel bow assumptions.
NRC Bulletin 90-02 was issued to ensure that the effects of channel box bow on the critical power ratio (CPR) calculations are properly taken into account. In response to NRC Bulletin 90-02, AREVA issued Supplement 1 to their CPR Methodology, ANF-524(P)(A).
The methodology described in ANF-524 has been reviewed and approved by the NRC and incorporated in Section 5.6.5.b. The ANF-524(P)(A) methodology incorporates the effects of channel bow on CPR through the MCPR Safety Limit (SL) calculation.
Based on fuel channel / control rod interference observed during previous Unit 2 cycles which may indicate that fuel channel bow is larger than the AREVA nominal database, PPL requested AREVA to increase by a factor of two the amount of channel bow assumed in the Unit 2 Cycle 14 MCPR Safety Limit calculation.
PPL requested that AREVA use a mean channel bow for the highly exposed Unit 2 Cycle 14 fuel assemblies of 122.6 mils (from an initial value of 61.3 mils). The assumption of two times nominal bow for the Unit 2 Cycle 14 MCPR Safety Limit is consistent with the channel bow assumption used for Unit 2 Cycle 13 MCPR Safety Limit. As part of the continuous validation of safety analyses assumptions, PPL will confirm that the actual Unit 2 Cycle 14 mean channel bow is less than or equal to the mean channel bow assumed.The confirmation will rely on performance data from previously measured fuel channels that were operated in a manner consistent with projected Unit 2 Cycle 14 operation and a potential channel measurement and re-channeling campaign during the refueling and inspection outage preceding Unit 2 Cycle 14 operation.
PPL will continue to monitor fuel channel performance in conformance with PPL's fuel channel monitoring program.
Enclosure to PLA-6132 Page 7 of 14 Results Summary A summary of the MCPR SL calculations performed by AREVA are provided in the following tables: Percentage of Pins in Boiling Transition for Two-Loop Operation Proposed MCPR Safet Limit % of Pins In Boiling Transition 1.09 0.1713 1.10 0.0971 1.11 < 0.0971 Percentage of Pins in Boiling Transition for Single-Loop Operation rProp s lMPRSafety, Limit. ]a gj. %~of, Pins I J!01 ig TapnsItio 1.12 0.1799 1.13 0.0996 1.14 < 0.0996 The above tables demonstrate that MCPR Safety Limits of 1.10 (two-loop) and 1.13 (single-loop) result in < 0.1% of pins in boiling transition.
However, MCPR Safety Limits of 1.11 (two-loop) and 1.14 (single-loop) are proposed to reduce the predicted percentage of pins in boiling transition.
Additional Discussion for MCPR SL Change The proposed change to the MCPR Safety Limits does not directly or indirectly affect any plant system, equipment, component, or change the processes used to operate the plant.As discussed above, the reload analyses performed prior to U2C14 startup will meet all applicable acceptance criteria.
Therefore, the proposed changes do not affect the failure modes of any systems or components.
Thus, the proposed change does not create the possibility of a previously unevaluated operator error or a new single failure. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
Since the proposed change does not alter any plant system, equipment, or component, the proposed change will not jeopardize or degrade the function or operation of any plant system or component governed by Technical Specifications.
The proposed MCPR Safety Limits do not involve a significant reduction in the margin of safety as currently defined Enclosure to PLA-6132 Page 8 of 14 in the Bases of the applicable Technical Specification sections because the MCPR Safety Limits calculated for U2C14 preserve the required margin of safety.Operator performance and procedures are unaffected by these proposed changes since the changes are essentially transparent to the operators and plant procedures and do not change the way in which the plant is operated.
The MCPR Operating Limits to be incorporated in the Core Operating Limits Report (determined from the MCPR Safety Limits and U2C 14 transient analysis results) may be different from the previous Unit 2 limits. Following use of the methodology to analyze the Unit 2 Cycle 14 core design and future Unit 2 reloads, the reload cycle specific results are incorporated into the FSAR via inclusion of the COLR in the Technical Requirements Manual (TRM).4.2 CHANGES TO COLR REFERENCES The AREVA ANFB-10 critical power correlation was used in the previous Unit 2 analyses to determine the MCPR Safety Limit and the MCPR operating limits and to ensure compliance with Unit 2 TS 3.2.2. For Unit 2 Cycle 14, the AREVA SPCB critical power correlation (EMF-2209(P)(A))
was used to calculate the MCPR Safety Limit and will be used to determine the MCPR operating limits and ensure compliance with Unit 2 TS 3.2.2. The SPCB critical power correlation is AREVA's most recent critical power correlation for the ATRIUMTM-10 fuel design and can be used to accurately predict assembly critical power.The references in Section 5.6.5.b are renumbered following the removal of EMF-1997(P)(A).
Attachment 7 provides the relationship between the references generated by AREVA in Section 5.6.5.b and the applicable Technical Specification Limiting Condition for Operation.
Attachment 8 provides the calculational flowpath and methodology reports used in the MCPR Safety Limit. The attachment shows that the entire MCPR Safety Limit calculation is based on AREVA analytical methods.

==4.3 CONCLUSION==

The changes to Section 5.6.5.b references reflect the NRC-approved methodology which will be used to generate Core Operating Limits for Unit 2 Cycle 14.The proposed change to the MCPR Safety Limit does not affect any plant system, equipment, or component.
Therefore, the proposed change will not jeopardize or degrade the function or operation of any plant system or component governed by Technical Specifications.
The proposed MCPR Safety Limit change does not involve a significant reduction in the margin of safety as currently defined in the Bases of the applicable Technical Specification sections because the .MCPR Safety Limits calculated for U2C14 preserve the required margin of safety.
Enclosure to PLA-6132 Page 9 of 14 Licensing analyses will be performed (using methodology referenced in Technical Specification Section 5.6.5.b) to determine changes in the critical power ratio as a result of anticipated operational occurrences.
These results are added to the MCPR Safety Limit values proposed herein to generate the MCPR operating limits in the U2C 14 COLR. Thus, the MCPR operating limits assure that the MCPR Safety Limits will not be exceeded during normal operation or anticipated operational occurrences, providing the required protection for the fuel rod cladding.
The proposed change to the MCPR Safety Limits will have a negligible impact on the results of postulated accident analyses.Therefore, the proposed action does not involve an increase in the probability or an increase in the consequences of an accident previously evaluated in the SAR. Thus, the proposed changes are in compliance with applicable regulations.
The health and safety of the public are not adversely impacted by operation of SSES as proposed.
Enclosure to PLA-6132 Page 10 of 14 5.0 REGULATORY SAFETY ANALYSIS 5.1 No Significant Hazards Consideration The proposed changes would revise the following:
TS 2.1.1.2 The two-loop and single-loop Minimum Critical Power Ratio (MCPR) Safety Limits are revised to reflect results of the cycle-specific MCPR Safety Limit analysis for Unit 2 Cycle 14. The two-loop MCPR Safety Limit increases from 1.09 to 1.11. The single-loop MCPR Safety Limit increases from 1.10 to 1.14.TS 5.6.5.b Core Operating Limits Report (COLR) references are revised to remove the ANFB-10 critical power correlation methodology report, EMF-1997(P)(A).
The remaining references following EMF-1997(P)(A) were renumbered.
PPL Susquehanna, LLC (PPL) has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below: 1. Does the proposed change involve a significant increase in the probability of occurrence or consequences of an accident previously evaluated?
Response:
No.The proposed change to the two-loop and single-loop MCPR Safety Limits do not directly or indirectly affect any plant system, equipment, component, or change the processes used to operate the plant. Further, the proposed U2C14 MCPR Safety Limits were generated using NRC-approved methodology and meet the applicable acceptance criteria.
Thus, this proposed amendment does not involve a significant increase in the probability of occurrence or consequences of an accident previously evaluated.
Prior to the startup of U2C 14, licensing analyses are performed (using NRC-approved methodology referenced in Technical Specification Section 5.6.5.b) to determine changes in the critical power ratio as a result of anticipated operational occurrences.
These results are added to the MCPR Safety Limit values to generate the MCPR operating limits in the U2C14 COLR. These limits could be different from those specified for the previous Unit 2 COLR. The COLR operating limits thus assure that the MCPR Safety Limit will not be exceeded during normal operation or anticipated Enclosure to PLA-6132 Page 11 of 14 operational occurrences.
Postulated accidents are also analyzed prior to the startup of U2C14 and the results shown to be within the NRC-approved criteria.The changes to the references in Section 5.6.5.b were made to properly reflect the NRC-approved methodology used to generate the U2C14 core operating limits. The use of this approved methodology does not increase the probability of occurrence or consequences of an accident previously evaluated.
Therefore, this proposed amendment does not involve a significant increase in the probability of occurrence or consequences of an accident previously evaluated.
: 2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?
Response:
No.The changes to the two-loop and single-loop MCPR Safety Limits do not directly or indirectly affect any plant system, equipment, or component and therefore does not affect the failure modes of any of these items. Thus, the proposed change does not create the possibility of a previously unevaluated operator error or a new single failure.The changes to the references in Section 5.6.5.b were made to properly reflect the NRC-approved methodology used to generate the U2C14 core operating limits. The use of this approved methodology does not create the possibility of a new or different kind of accident.Therefore, this proposed amendment does not create the possibility of a new or different kind of accident from any previously evaluated.
: 3. Does the proposed amendment involve a significant reduction in a margin of safety?Response:
No.Since the proposed changes do not alter any plant system, equipment, component, or the processes used to operate the plant, the proposed change will not jeopardize or degrade the function or operation of any plant system or component governed by Technical Specifications.
The proposed two-loop and single-loop MCPR Safety Limits do not involve a significant reduction in the margin of safety as currently defined in the Bases of the applicable Technical Specification sections because the MCPR Safety Limits calculated for U2C14 preserve the required margin of safety.
Enclosure to PLA-6132 Page 12 of 14 The changes to the references in Section 5.6.5.b were made to properly reflect the NRC-approved methodology used to generate the U2C14 core operating limits. This approved methodology is used to demonstrate that all applicable criteria are met, thus, demonstrating that there is no reduction in the margin of safety.Therefore, the proposed changes do not involve a significant reduction in a margin of safety.Based upon-the above, PPL Susquehanna, LLC (PPL) concludes that the proposed amendment presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

===5.2 Applicable===

Regulatory Requirements/Criteria Title 10 of the Code of Federal Regulations (10 CFR) establishes the fundamental regulatory requirements with respect to reactivity control systems. Specifically, General Design Criterion 10 (GDC-10), "Reactor Design," in Appendix A, "General Design Criteria for Nuclear Power Plants," to 10 CFR Part 50 states, in part, that the reactor core and associated coolant, control, and protection systems shall be designed with appropriate margin to assure that specified acceptable fuel design limits are not exceeded.The proposed MCPR Safety Limit values in TS Section 2.1.1.2 will ensure that 99.9% of the fuel rods in the core are not expected to experience boiling transition.
This satisfies the requirements of GDC-10 regarding acceptable fuel design limits.NRC Generic Letter 88-16 (GL 88-16), "Removal of Cycle-Specific Parameter Limits from Technical Specifications," provides guidance on modifying cycle-specific parameter limits in TS. The proposed changes to TS Section 5.6.5.b are in compliance with the guidance specified in GL 88-16.In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner; (2) such activities will be conducted in compliance with the Commission's regulations; and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
Enclosure to PLA-6132 Page 13 of 14 6.0 ENVIRONMENTAL CONSIDERATION 10 CFR 51.22(c)(9) identifies certain licensing and regulatory actions which are eligible for categorical exclusion from the requirement to perform an environmental assessment.
A proposed amendment to an operating license for a facility does not require an environmental assessment if operation of the facility in accordance with the proposed amendment would not: (1) involve a significant hazards consideration; (2) result in a significant change in the types or significant increase in the amounts of any effluents that may be released offsite; or (3) result in a significant increase in individual or cumulative occupational radiation exposure.
PPL Susquehanna, LLC has evaluated the proposed changes and has determined that the proposed changes meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).
Accordingly, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment needs to be prepared in connection with issuance of the amendment.
The basis for this determination, using the above criteria, follows: BASIS As demonstrated in the "No Significant Hazards Consideration Evaluation," the proposed amendment does not involve a significant hazards consideration.
There is no significant change in the types or significant increase in the amounts of any effluents that may be released offsite. The proposed change does not involve any physical alteration of the plant (no new or different type of equipment will be installed) or change in methods governing normal plant operation.
There is no significant increase in individual or cumulative occupational radiation exposure.
The proposed change does not involve any physical alteration of the plant (no new or different type of equipment will be installed) or change in methods governing normal plant operation.
Enclosure to PLA-6132 Page 14 of 14 

==7.0 REFERENCES==

7.1. PLA-5702, B. L. Shriver (PPL) to USNRC, "Request for Additional Information Regarding Proposed Amendment No. 256 to Unit 1 License NPF-14: MCPR Safety Limits and Reference Changes," dated December 22, 2003.7.2. PLA-5793, B. T. McKinney (PPL) to USNRC, "Proposed Amendment to No. 233 to Unit 2 License NPF-22: MCPR Safety Limits and Reference Changes," dated September 8, 2004.7.3. PLA-5990, B. T. McKinney (PPL) to USNRC, "Proposed Amendment to No. 283 to Unit 1 License NPF-14: MCPR Safety Limits and Reference Changes," dated December 1, 2005.7.4. PLA-5931, B. T. McKinney (PPL) to USNRC, "Proposed License Amendment Numbers 279 for Unit 1 Operating License No. NPF-14 and 248 for Unit 2 Operating License No. NPF-22: ARTS/MELLLA Implementation," dated November 18, 2005.7.5. PLA-6076, B. T. McKinney (PPL) to USNRC, "Proposed License Amendment Numbers 285 for Unit 1 Operating License No. NPF-14 and 253 for Unit 2 Operating License No. NPF-22: Constant Pressure Power Uprate," dated October 11, 2006.
Attachment 1 to PLA-6132 Proposed Technical Specification Changes Unit 2 (Markups)
PPL Rev. 2 SLs 2.0 2.0 SAFETY LIMITS (SLs)2.1 SLs 2.1.1 Reactor Core SLs 2.1.1.1 With the reactor steam dome pressure < 785 psig or core flow< 10 million Ibm/hr: THERMAL POWER shall be < 25% RTP.2.1.1.2 With the reactor steam dome pressure _ 785 psig and core flow10 million Ibm/hr: zj , I, i MCPR shall be for two recirculation loop operation or >_'1"10r for single recirculation loop operation.
2.1.1.3 Reactor vessel water level shall be greater than the top of active irradiated fuel.2.1.2 Reactor Coolant System Pressure SL Reactor steam dome pressure shall be < 1325 psig.2.2 SL Violations With any SL violation, the following actions shall be completed within 2 hours: 2.2.1 Restore compliance with all SLs; and 2.2.2 Insert all insertable control rods.SUSUQUEHANNA 
-UNIT 2 TS / 2.0-1 Amendmnt 1,'1 Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 COLR (continued)
: 10. ANF-524(P)(A), "ANF Critical Power Methodology for Boiling Water Reactors," Advanced Nuclear Fuels Corporation.
: 11. ANF-913(P)(A), "COTRANSA2:
A Computer Program for Boiling Water Reactor Transient Analyses," Advanced Nuclear Fuels Corporation.
: 12. ANF-1358(P)(A). "The Loss of Feedwater Heating Transient in Boiling Water Reactors," Advanced Nuclear Fuels Corporation.
: 13. EMF-2209(P)(A), "SPCB Critical Power Correlation," Siemens Power Corporation.
-4.+ E "IF 197-(P)(A), "ANFB 10 CritiCCi P.wc.r ....elatio", men Poewcr 14. 46 EMF-CC-074(P)(A), "BWR Stability Analysis -Assessment of STAIF with Input from MICROBURN-B2," Siemens Power Corporation.
1 5, 4-6 NE-092-001 A, "Licensing Topical Report for Power Uprate With Increased Core Flow," Pennsylvania Power & Light Company.I (*. 4-t Caldon, Inc., "TOPICAL REPORT: Improving Thermal Power Accuracy and Plant Safety while Increasing Operating Power Level Using the LEFTM- System," Engineering Report -80P.17. 487 Caldon, Inc., "Supplement to Topical Report ER-80P: Basis for a Power Uprate with the LEFM4Tm or LEFM CheckPlusTm System," Engineering Report ER-160P.1. 49-. NEDO-32465-A, "BWROG Reactor Core Stability Detect and Suppress Solutions Licensing Basis Methodology for Reload Applications." 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 2 TS / 5.0-23 AmendmePt 1 1 &#xfd;2, 1JK Attachment 2 to PLA-6132 List of Regulatory Commitments (Unit 2)
Attachment 2 to PLA-6132 Page I of 1 LIST OF REGULATORY COMMITMENTS ReGULATORY COciMTMeN[itS thisseDbatelEtv.et Ther are no new commitments associated with this submittal.
NA Attachment 3 to PLA-6132 Proposed Changes to TS Bases Pages Unit 2 (MARKUP)
I-,1-L MeV. I Reactor Core SLs B 2.1.1 BASES APPLICABLE SAFETY ANALYSES The fuel cladding must not sustain damage as a result of normal operation and AOOs. The reactor core SLs are established to preclude violation of the fuel design criterion that an MCPR limit is to be established, such that at least 99.9% of the fuel rods in the core would not be expected to experience the onset of transition boiling.The Reactor Protection System setpoints (LCO 3.3.1.1, "Reactor Protection System (RPS) Instrumentation"), in combination with the other LCOs, are designed to prevent any anticipated combination of transient conditions for Reactor Coolant System water level, pressure, and THERMAL POWER level that would result in reaching the MCPR limit.2.1.1.1 Fuel Cladding Integrity 5 P( ,, The use of the-ANF-E-4Reference 
: 4) correlation is valid for critical power at pressures t461 sia and bundle mass fluxes > S x 106 71. 1 .. I Ib~hr- ft2f .Fo operation at low pressures or low flowsl the fuel cladding integrity SL is established by a limiting condition on core HERMAL-..... POWER, with the following basis: Provided that the water level in the vessel downcomer is maintained above the top of the active fuel, natural circulation is sufficient to ensure a minimum bundle flow for all fuel assemblies that have a relatively high power and potentially can approach a critical heat flux condition.
For the SPC Atrium 10 design, the minimum bundle flow is> 28 x 103 lb/hr. For Atrium-10 fuel design, the coolant minimum bundle flow and maximum area are such that the mass flux is always> .25 x 10' lb/hr-ft.
Full scale critical power test data taken from various SPC and GE fuel designs at pressures from 14.7 psia to 1400 psia indicate the fuel assembly critical power at 0.25 x 106 lb/hr-ft 2 is approximately 3.35 MWt. At 25% RTP, a bundle power of approximately 3.35 MWt corresponds to a bundle radial peaking factor of approximately 3.0, which is significantly higher than the expected peaking factor. Thus, a THERMAL POWER limit of 25% RTP for reactor pressures 
< 785 psig is conservative.
2 MCPR 2.1.1.The MCPR SL ensures sufficient conservatism in the operating MCPR limit that, in the event of an AOO from the limiting condition of operation, at least 99.9% of the fuel rods in the core would be expected to avoid boiling transition.
The margin between calculated boiling transition (i.e., MCPR = 1.00) and the MCPR SL is based on a detailed statistical procedure (continued)
SUSQUEHANNA 
-UNIT 2 TS I B 2.0-2 Revision 2 FF.. I V. I Reactor Core SLs B 2.1.1 BASES APPLICABLE SAFETY 2.1.1.2 MCPR (continued)
A NALYSES that considers the uncertainties in monitoring the core operating state. One specific uncertainty included in the SL is the uncertainty in the critical power correlation.
References 2, 4 and 5 describe the methodology used in determining the MCPR SL./Th ANF-B 1 critical power correlation is based on a significant body of practical test data. As long as the core pressure and flow are within the range of validity of the correlation (refer to Section B 2.1.1.1), the assumed reactor s pr- 3 conditions used in defining the SL introduce conservatism into the limit because bounding high radial power factors and bounding flat local peaking distributions are used to estimate the number of rods in boiling transition.
These conservatisms and the inherent accuracy of the A,&-l&#xfd;lcorrelation rovide a reasonable degree of assurance that during s.g tained operation at the MCPR SL there would be no transition boi ing in the core. If boiling transition were to occur, there is reason to believe that the integrity of the fuel would not be compromised.
I Significant test data accumulated by the NRC and private organizations indicate that the use of a boiling transition limitation to protect against cladding failure is a very conservative approach.
Much of the data indicate that BWR fuel can survive for an extended period of time in an environment of boiling transition.
SPC ATRIUM-10 fuel is monitored using the4NBQCritical Power Correlation.
The effects of channel bow on MCPR are explicitly included in the calculation of the MCPR SL. Explicit treatment of channel bow in the MCPR SL addresses the concerns of the NRC Bulletin No. 90-02 entitled"Loss of Thermal Margin Caused by Channel Box Bow." Monitoring required for compliance with the MCPR SL is specified in LCO 3.2.2, Minimum Critical Power Ratio.2.1.1.3 Reactor Vessel Water Level During MODES 1 and 2 the reactor vessel water level is required to be above the top of the active fuel to provide core cooling capability.
With fuel in the reactor vessel during periods when the reactor is shut down, consideration must be given to water level requirements due to the effect of decay heat. If the water level should drop below the top of the active irradiated fuel during this period, the ability to remove decay heat is reduced. This reduction in cooling capability could lead to elevated cladding temperatures and clad perforation in the event that the water level becomes < 2/3 of the core height.(continued)
SUSQUEHANNA 
-UNIT 2 TS / B 2.0-3 Revision 3 Reactor Core SLs B 2.1.1 BASES APPLICABLE SAFETY ANALYSES 2.1.1.3 Reactor Vessel Water Level (continued)
The reactor vessel water level SL has been established at the top of the active irradiated fuel to provide a point that can be monitored and to also provide adequate margin for effective action.SAFETY LIMITS The reactor core SLs are established to protect the integrity of the fuel clad barrier to the release of radioactive materials to the environs.
SL 2.1.1.1 and SL 2.1.1.2 ensure that the core operates within the fuel design criteria.SL 2.1.1.3 ensures that the reactor vessel water level is greater than the top of the active irradiated fuel in order to prevent elevated clad temperatures and resultant clad perforations.
APPLICABILITY SLs 2.1.1.1, 2.1.1.2, and 2.1.1.3 are applicable in all MODES.SAFETY LIMIT VIOLATIONS Exceeding an SL may cause fuel damage and create a potential for radioactive releases in excess of 10 CFR 100, "Reactor Site Criteria," limits (Ref. 3). Therefore, it is required to insert all insertable control rods and restore compliance with the SLs within 2 hours. The 2 hour Completion Time ensures that the operators take prompt remedial action and also ensures that the probability of an accident occurring during this period is minimal.REFERENCES
: 1. 10 CFR 50, Appendix A, GDC 10.2. ANFB 524 (P)(A), Revision 2, "Critical Power Methodology for Boiling Water Reactors," Supplement 1 Revision 2 and Supplement 2, November 1990.3. 10 CFR 100.4.5.ENIF 1997fP() ciin0 AF 0Ciia ao crlta, July 1998 and ,99,P) a)pplemcint 1 Rc-,sion r 0," ANFB 10 Critical Pewer Gerrclation:
High Local Peaking Results," Julyt&#xfd; 103 EMF-2158(P)(A), Rev. 0, "Siemens Power Corporation Methodology for Boiling Water Reactors:
Evaluation and Validation of CASMO-4 / MICROBURN-B2," October 1999.EMF -U0RV?(1qV Z , &#xfd;pCjh Cr I 4 , r"Ije SUSQUEHANNA 
-UNIT 2 TS I B 2.0-4 Revision 4 MCPR B 3.2.2 BASES SURVEILLANCE REQUIREMENTS SR 3.2.2.2 (continued)
Determining MCPR operating limits based on interpolation between scram insertion times is not permitted.
The average measured scram times and corresponding MCPR operating limit must be determined once within 72 hours after each set of scram time tests required by SR 3.1.4.1, SR 3.1.4.2, SR 3.1.4.3 and SR 3.1.4.4 because the effective scram times may change during the cycle. The 72 hour Completion Time is acceptable due to the relatively minor changes in average measured scram times expected during the fuel cycle.REFERENCES
: 1. NUREG-0562, June 1979.2. XN-NF-80-19(P)(A)
Volume 1 and Supplements 1 and 2, "Exxon Nuclear Methodology for Boiling Water Reactors," Exxon Nuclear Company, March 1983.3. XN-NF-80-19(P)(A)
Volume 3, Revision 2, "Exxon Nuclear Methodology for Boiling Water Reactors, THERMEX: Thermal Limits Methodology Summary Description," Exxon Nuclear Company, January 1987.4. ANF-913(P)(A)
Volume 1, Revision 1 and Volume 1 Supplements 2, 3, and 4, "COTRANSA2:
A Computer Program for Boiling Water Reactor Transient Analyses," Advanced Nuclear Fuels Corporation, August 1990.5. XN-NF-80-19 (P)(A), Volume 4, Revision 1, "Exxon Nuclear Methodology for Boiling Water Reactors:
Application of the ENC Methodology to BWR Reloads," Exxon Nuclear Company, June 1986.6. NE-092-001, Revision 1, "Susquehanna Steam Electric Station Units 1 & 2: Licensing Topical Report for Power Uprate with S 9- .i,., ,- Increased Core Flow," December.1992, and NRC Approval Letter: Letter from T. E. Murley (NRC) to R. G. Byram (PP&L)," Pb ri r,' I'. ?.wee. "Licensing Topical Report for Power Uprate With Increased Core Flow, Revision 0, Susquehanna Steam Electric Station, Units 1 , re and 2 (PLA-3788) (TAC Nos. M83426 and M83427)," ( Ct.r -orpor -,'k November 30, 1993.up7 F=IF 199:7, Revisien 0 (~ctobeir 19907) and Supplefrnont 1, F~ev~io 0 (anury.998), "ANFB 10 Critical Powcr Ccrrclation,"-nd eId NIRC SER dated 71,719g.(continued)
SUSQUEHANNA 
-UNIT 2 TS / B 3.2-8 Revision 3 rrL I_ \VZ V. I MCPR B 3.2.2 BASES Reference 
: 8. XN-NF-79-71(P)(A)
Revision 2, Supplements 1, 2, and 3, "Exxon (continued)
Nuclear Plant Transient Methodology for Boiling Water Reactors," March 1986.9. XN-NF-84-105(P)(A), Volume 1 and Volume 1 Supplements 1 and 2, "XCOBRA-T:
A Computer Code for BWR Transient Thermal-Hydraulic Core Analil February 1987.10. ANF-1358(P)(A)
Revision ,, "The Loss of Feedwater Heating Transient in Boiling Water Reactors " Advanced Nuclear Fuels Corporation, September 
.@E4. 5 11. Final Policy Statement on Technical Specifications Improvements, July 22,1993 (58 FR 39132).SUSQUEHANNA-UNIT 2 TS / B 3.2-9 Revision 3 Attachment 4 to PLA-6132 Preliminary Description of Unit 2 Cycle 14 Core Composition Attachment 4 to PLA-6132 Page 1 of 1 Preliminary Unit 2 Cycle 14 Core Composition Assembly Type Operational History, Number o~f Assemblies AREVA ATRIUMTM-10 Fresh 292 AREVA ATRIUM T M-10 Once-burned 292 AREVA ATRIUM T M-10 Twice-burned 180 Attachment 5 to PLA-6132 Preliminary Unit 2 Cycle 14 Core Loading Pattern to PLA-6 132 Page I of I Susquehanna Unit 2 Cycle 14 Fuel Cycle Design 31 33 35 37 39 41 43 45 41 49 51 53 55 57 59 30 21 24 27 25 27 24 27 25 26 24 28 24 28 21 21 29.2 22.3 0.0 22.7 0.0 21.7 0.0 23.1 0.0 23.5 0.0 21.2 0.0 32.6 42.0 28 24 27 24 27 24 2"7 25 27 24 27 24 28 28 21 21 22.3 0.0 22.8 0.0 23.0 0.0 23.6 0.0 22.9 0.0 20.4 0.0 0.0 30.1 42.1 26 27 24 27 24 27 25 25 24 27 24 28 28 24 21 21 0.0 22.8 0.0 22.8 0.0 23.4 22.9 22.2 0.0 22.9 0.0 0.0 21.7 38.5 42.6 24 25 27 24 26 25 27 24 27 24 27 23 28 23 21 21 22.9 0.0 22.9 0.0 21.7 0.0 21.5 0.0 22.4 0.0 22.8 0.0 17.1 40.0 42.7 22 27 24 27 25 26 24 27 23 27 24 28 28 23 22 22 0.0 22.8 0.0 21.6 0.0 22.5 0.0 23.2 0.0 20.7 0.0 0.0 17.6 36.1 40.4 20 24&#xfd; '- 27 25 27 24 26 24 26 24 27 24 28 23 21 22 21.6 0.0 23.5 0.0 22.7 0.0 23.4 0.0 22.9 0.0 19.3 0.0 18.8 40.7 40.4 18 27 25 25 24 27 24 27 24 27 24 28 28 23 21 22 0.0 23.6 22.9 21.4 0.0 23.3 0.0 23.2 0.0 22.8 0.0 0.0 19.4 41.6 42.0 16 25 *27 24 27 24 2E 24 25 28 24 28 23 21 22 23.1 0.0 22.3 0.0 23.4 0.0 22.9 23.8 0.0 21.3 0.0 18.2 40.4 41.6 14 26 24 27 24 27 24 27 28 28 28 23 21 21 0.0 23.0 0.0 22.2 0.0 22.9 0.0 0.0 0.0 0.0 21.1 38.4 41.3 12. 24 27 24 21 24 27 24 24 28 21 21 21 22 23.4 0.0 23.0 0.0 20.5 0.0 22.8 21.6 0.0 29.3 35.7 43.5 .41.8 10 28 24 28 23 28 24 28 28 23 21 21 0.0 20.4 0.0 23.2 0.0 19.2 0.0 0.0 21.3 37.1 42.7 8 24 28 28 28 28 28 28 23 21 21 21.2 0.0 0.0 0.0 0.0 0.0 0.0 18.1 37.3 43.1 6 28 28 24 23 23 23 23 21 21 22 0.0 0.0 21.8 17.2 17.8 18.5 19.3 41.0 42.4 41.9 4 21 21 21 21 22 21 22 22 Nuclear Fuel Type 32.5 30.3 38.8 39.0 35.9 40.7 36.3 41.7 BOC Exposure (Gwd/MTU)2 21 21 21 21 22 22 22 42.1 42.3 42.5 42.6 40.6 41.0 41.6 No. Per Fuel Type Description Cycle Loaded Quarter core 21 4.12B-13GV7 12 32 22 3.;90B-14GV7 12 13 23 4.12B-14GV7 13 15 24 4.12B-15GV8 13 45 25 3.90B-15GV7 13 13 26 3.674B-12GV7 14 1 27 4.057B-14GV8 14 35 28 4.241B-14GV8 14 3i Susquehanna Unit 2 Cycle 14 Lower Right Quarter Core Layout by Fuel Type Attachment 6 to PLA-6132 Preliminary Descriptions of Unit 2 Cycle 14 Fresh Fuel Description Attachment 6 to PLA-6132 Page 1 of 3 Preliminary Assembly Type 26 Reload Bundle Description (ATRIUM-10, 100mil Channel)Bundle Average Enrichment 
= 3.67%TAF 149.45" 144" 120" 96" 6" 0D" BAF Attachment 6 to PLA-6132 Page 2 of 3 Preliminary Assembly Type 27 Reload Bundle Description (ATRIUM-10, 100mil Channel)Bundle Average Enrichment 
= 4.06%TAF 149.45" 144" 132" 96" 6" BAF 0 0,7 mtuacnment o to PLA-6132 Page 3 of 3 Preliminary Assembly Type 28 Reload Bundle Description (ATRIUM-10, 100mil Channel)Bundle Average Enrichment 
= 4.24%TAF 149.45" 144" 138" 96" 42" 6" l0" BAF Attachment 7 to PLA-6132 Listing of AREVA NP, Inc.Approved Methodology and Applicable LCOs BWR Approved Topical Reports for Susquehanna Nuclear Plant Technical Specifications COLR References IAl u )~tI IIU Pn n 8 BWR Approved Topical Reports for Susquehanna Nuclear Plant Technical Specifications COLR References Report Applicable LCO Methodology 
/ Justification XN-NF-81-58(P)(A)
Revision 2 and Supplements 1 3.2.1 Provides an analytical capability to predict BWR fuel thermal and and 2, RODEX2 Fuel Rod Thermal-Mechanical Response 3.2.2 mechanical conditions for normal core operation and to establish Evaluation Model, Exxon Nuclear Company, March 1984. 3.2.3 initial conditions for power ramping, non-LOCA and LOCA analyses.XN-NF-85-67(P)(A)
Revision 1, Generic Mechanical 

====3.2.3 Describes====

the process used to develop linear heat generation rates Design for Exxon Nuclear Jet Pump BWR Reload Fuel, for fuel designs.Exxon Nuclear Company, September 1986.EMF-85-74(P)
Revision 0 Supplement I(P)(A) and 3.2.3 Extends the exposure limit of the RODEX2A code which is a version Supplement 2(P)(A), RODEX2A (BWR) Fuel Rod of RODEX2 that includes a fission gas release model specific to Thermal-Mechanical Evaluation Model, Siemens Power BWR fuel designs.Corporation, February 1998.ANF-89-98(P)(A)
Revision 1 and Supplement 1, Generic 3.2.3 Establishes a set of design criteria which assures that BWR fuel will Mechanical Design Criteria for BWR Fuel Designs, perform satisfactorily throughout its lifetime.Advanced Nuclear Fuels Corporation, May 1995.XN-NF-80-19(P)(A)
Volume 1 and Supplements 1 and 2, 3.1.1 Development of BWR core analysis methodology which comprises Exxon Nuclear Methodology for Boiling Water Reactors -3.2.1 codes for fuel neutronic parameters and assembly burnup Neutronic Methods for Design and Analysis, Exxon 3.2.2 calculations, reactor core simulation diffusion theory calculations, Nuclear Company, March 1983. 3.2.3 core and channel hydrodynamic stability predictions, and producing 3.3.2.1 Table 3.3.2.1-1 input for nuclear plant transients.
Subsequently approved codes or methodologies have superceded portions of this report. Applicable portions include CRDA, and methodology to determine neutronic reactivity parameters, void reactivity, Doppler reactivity, scram reactivity, delayed neutron fraction, and prompt neutron lifetime.XN-NF-80-19(P)(A)
Volume 4 Revision 1, Exxon Nuclear 3.2.1 Summarizes the types of BWR licensing analyses performed, Methodology for Boiling Water Reactors:
Application of 3.2.2 identifies the methodologies used.the ENC Methodology to BWR Reloads, Exxon Nuclear 3.2.3 Company, June 1986.EMF-2158(P)(A)
Revision 0, Siemens Power Corporation Methodology for Boiling Water Reactors:
Evaluation and Validation of CASMO-4/MICROBURN-B2, Siemens Power Corporation, October 1999.3.1.1 3.2.2 3.2.3 3.3.2.1 Table 3.3.2.1-1 Describes the reactor core simulator code MICROBURN-62 and the lattice physics code CASMO-4. >-J JG[0400.5 BWR Approved Topical Reports for Susquehanna 
.Nuclear Plant Technical Specifications COLR References Page A-, Report Applicable LCO Methodology 
/ Justification XN-NF-80-19(P)(A)
Volumes 2, 2A, 2B and 2C, Exxon Nuclear 3.2.1 Describes an evaluation model methodology for licensing analyses of Methodology for Boiling Water Reactors.
EXEM BWR ECCS postulated LOCAs in jet pump BWRs. The methodology was Evaluation Model, Exxon Nuclear Company, September 1982. developed to comply with 10 CFR 50.46 and.Appendix K criteria to 10 CFR 50.EMF-2361 (P)(A) Revision 0, EXEM BWR-2000 ECCS 3.2.1 Describes an upgraded evaluation model methodology for licensing Evaluation Model, Framatome ANP, May 2001. analyses of postulated LOCAs in jet pump BWRs. The methodology was developed to comply with 10 CFR 50.46 and Appendix K criteria to 10 CFR 50.EMF-2292(P)(A)
Revision 0, A TRIUM T M-1: Appendix K Spray. 3.2.1 Provides measured cladding temperatures from spray heat transfer Heat Transfer Coefficients, Siemens Power Corporation, tests to justify the use of Appendix K coefficients for ATRIUM-I 0 fuel September 2000. LOCA analyses.XN-NF-80-19(P)(A)
Volume 3 Revision 2, Exxon Nuclear 3.2.2 Provides overall methodology for determining a MCPR operating limit.Methodology for Boiling Water Reactors, THERMEX. Thermal Limits Methodology Summary Description, Exxon Nuclear Company, January 1987.XN-NF-84-105(P)(A)
Volume 1 and Volume.1 Supplements I and 2, XCOBRA-T., "A Computer Code for BWR Transient Thermal-Hydraulic Core Analysis, Exxon Nuclear Company, February 1987.ANF-524(P)(A)
Revision 2 and Supplements I and 2, ANF Critical Power Methodology for Boiling Water Reactors, Advanced Nuclear Fuels Corporation, November 1990.3.2.2 Provides a capability to perform analyses of transient heat transfer behavior in BWR assemblies.

====3.2.2 Provides====
a methodology for the determination of thermal margins, specifically the MCPR safety limit.ANF-913(P)(A)
Volume 1 Revision 1 and Volume 1 3.2.2 Provides a c Supplements 2, 3 and 4, COTRANSA2:
A Computer Program for Boiling Water Reactor Transient Analyses, Advanced Nuclear Fuels Corporation, August 1990.somputer program for analyzing BWR system transients.-V6>,~JG604005 BWR Approved Topical Reports for Susquehanna Nuclear Plant Technical Specifications COLR References Page A-2 Report Applicable LCO Methodology 
/ Justification ANF-1358(P)(A)
Revision 3, The Loss of Feedwater Heating 3.2.2 Presents a generic methodology for evaluating the loss of feedwater Transient in Boiling Water Reactors, Advanced Nuclear Fuels heating event.Corporation, September 2005.EMF-2209(P)(A)
Revision 2, SPCB Critical Power Correlation, 3.2.2 Presents a critical power correlation for use with the ATRILUMTM-10*
Siemens Power Corporation, September 2003. fuel designs. This correlation is used during core design and monitoring and in the BWR-2000 LOCA methodology.
EMF-CC-074(P)(A)
Volume 4 Revision 0, BWR Stability 

====3.4.1 Provides====
a computer program for performing stability analysis.Analysis -Assessment of S TA IF with Input from MICROBURN-B2, Siemens Power Corporation, August 2000.* ATRIUM is a trademark of Framatome ANP.JG104005 Attachment 8 to PLA-6132 MCPR Safety Limit Methodology I
Attachment 8 to PLA-6132 Page I of 1 MCPR Safety Limit Methodology
-----------------1 FRAMATOME SLMCPR To Be Supported I L I Plant, Fuel &CPR Correlation Uncertainties ANF-524 (P)(A), Rev. 2 & Supplement 1, Rev. 2 E.MF-2209(P)(A), Rev.2 EMF-2158 (P)(A), Rev. 0 I Steady State I ~Core T/H t e (XCOBRA) Bundle Flow Statistical Analysis ANF-524 (P)(A), Rev. 2 VSLHGR of Rods in Boiling Transition I I S Conservative 
/Power Profiles (MICROBURN-B2 Radials, CASMO-4) Axials, EMF-2158 (P) (A), Rev. 0 Locals I INo~Increase I Yes SLMCPR I j Input SLMCPR is Supported t---------- 
-----------------
j}}