PLA-5793, Proposed Amendment No. 233 to Unit 2 License NPF-22: MCPR Safety Limits and Reference Changes

Proposed Amendment No. 233 to Unit 2 License NPF-22: MCPR Safety Limits and Reference Changes

ML042590562
Person / Time
Site:	Susquehanna
Issue date:	09/08/2004
From:	Mckinney B Susquehanna
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
PLA-5793
Download: ML042590562 (53)

Text

Britt T. McKinney Vice President-Nuclear Site Operations PPL Susquehanna, LLC 769 Salem Boulevard Berwick, PA 18603 Tel. 570.542.3149 Fax 570.542.1504 btmckinney~pplweb.com

%I II.

I Is IHa5--pa mC; 4,0 SEP 0 8 2004 U. S. Nuclear Regulatory Commission Attn.: Document Control Desk Mail Stop OPI-17 Washington, DC 20555 SUSQUEHANNA STEAM ELECTRIC STATION PROPOSED AMENDMENT NO. 233 TO UNIT 2 LICENSE NPF-22: MCPR SAFETY LIMITS AND REFERENCE CHANGES PLA-5793 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 13 (U2C13) Minimum Critical Power Ratio (MCPR) Safety Limits for both two-loop and single-loop operation. Additionally, Section 5.6.5.b is revised to include 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. to this letter contains the applicable pages of the Susquehanna SES Unit 2 Technical Specifications, marked to show the proposed change. contains the "camera ready" version of the revised Technical Specification pages. is included which identifies that there are no regulatory commitments associated with this change. contains the applicable pages of the Susquehanna SES Unit 2 Technical Specifications Bases, marked to show the proposed change (Provided for Information).

-AuoI Document Control Desk PLA-5793 has been provided as a description of the U2C13 core composition to assist in your review. contains a table correlating potential safety limits to the calculated number of pins in boiling transition. contains the relationship between Framatome-ANP (FANP) references and Technical Specification LCO's.

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 2005 to support the startup of U2C 13 operation. Therefore, we request NRC complete its review of this change by January 31, 2005 with the changes effective upon startup following the Unit 2 13'h 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:

B. T. McKinney

Enclosure:

PPL Susquehanna Evaluation of the Proposed Changes Attachments: - Proposed Technical Specification Changes Unit 2, (Mark-ups) - Proposed Technical Specification Pages Unit 2, (Camera Ready) - List of Regulatory Commitments - Proposed Technical Specification Bases Changes Unit 2, (Mark-ups) - Description of U2C13 Core Composition - Correlation of Potential Safety Limit to Calculated Pins in Boiling Transition - Relationship between FANP references and TS LCO's Document Control Desk PLA-5793 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-5793 PPL SUSQUEHANNA EVALUATION OF PROPOSED CHANGES UNIT 2 CYCLE 13 MCPR SAFETY LIMIT AND COLR REFERENCES

1.

DESCRIPTION

2.

PROPOSED CHANGE

3.

BACKGROUND

4.

TECHNICAL ANALYSIS

5.

REGULATORY ANALYSIS 5.1 No Significant Hazards Consideration 5.2 Applicable Regulatory Requirements/Criteria

6.

ENVIRONMENTAL CONSIDERATIONS

7.

REFERENCES

Enclosure to PLA-5793 Page 1 of 10 PPL EVALUATION

Subject:

Unit 2 Cycle 13 MCPR Safety Limit and COLR

References:

TS Sections 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 13 (U2C13) Minimum Critical Power Ratio (MCPR) Safety Limits for both two-loop and single-loop operation. The change to Section 2.1.1.2 is necessary because, as a result of U2C 13 cycle specific calculations, the two-loop and single-loop operation MCPR Safety Limits decreased 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 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 replaces PPL's analytical methods with analytical methods developed by Framatome-ANP's (FANP). FANP's analytical methods will be used to develop the core operating limits documented in the COLR.

The changes are described in detail in Section 4.0.

The requested approval date (January 31, 2005) will allow time for the Core Operating Limits Report to be prepared and reviewed by the Plant Operation Review Committee (PORC) prior to the outage scheduled for the Spring of 2005.

2.0 PROPOSED CHANGE

S 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 1.10 (two-loop operation) and 1.11 (single loop operation) to 1.09 (two-loop operation) and 1.10 (single loop operation) to reflect results of the cycle specific MCPR Safety Limit analysis for Unit 2 Cycle 13.

Enclosure to PLA-5793 Page 2 of 10 2.2 TS 5.6.5.b Core Operating Limits Report (COLR) references are revised to delete PPL's analytical methods and add the FANP's NRC approved analytical methods that are not already contained in Section 5.6.5.b. The references were reordered to correspond to the list provided as Attachment 7.

In summary, the proposed changes would revise the Susquehanna Unit 2 Technical Specifications (TS) Sections 2.1.1.2 and 5.6.5.b. TS Section 2.1.1.2 is revised to reflect the Unit 2 Cycle 13 (U2C 13) MCPR Safety Limit for both two-loop and single-loop operation. TS Section 5.6.5.b is revised to remove references applicable to PPL's analytical methods and add FANP's analytical methods. The TS Bases changes corresponding to the proposed TS changes are included for information.

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. This change to Section 2.1.1.2 reflects the U2C13 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.

The proposed Safety Limit MCPR values (two-loop and single-loop) were calculated using FANP's NRC approved analytical methods with the ANFB-10 critical power correlation for ATRIUMTNI-1 0 fuel assuming a rated core thermal power of 3489 MWt.

The proposed Safety Limit MCPR values (two-loop and single-loop) 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 core loading for U2C13 is operated in a safe manner. Additional analyses are performed

Enclosure to PLA-5793 Page 3 of 10 (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 proposed here to generate the MCPR operating limits in the U2C13 COLR. The MCPR operating limits 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 to confirm the NRC acceptance criteria are met.

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.

References pertaining to PPL's analytical methods will no longer be used and were removed. FANP's corresponding analytical methods were added.

The references in Section 5.6.5.b were reordered to be consistent with Attachment 7. is a list of FANP methodology references that provides the relationship between the references and the applicable Technical Specification Limiting Condition for Operation. The attachment also identifies, for each reference, a description and justification for its use.

4.0 TECHNICAL ANALYSIS

4.1 MCPR SAFETY LIMIT CHANGE This Technical Specification change decreases the U2C12 MCPR Safety Limits from 1.10 two-loop and 1.11 single loop to 1.09 two-loop and 1.10 single loop for U2C13. The following two changes that impact the MCPR Safety Limit are:

1. Incorporation of smaller power distribution uncertainties in the MCPR Safety Limit analysis that are consistent with the NRC approved CASMO-4/MICROBURN-B2 methodology used in the POWERPLEX-III Core Monitoring System for U2C 13.

2. Increase in the amount of channel bow assumed in the MCPR Safety Limit analysis.

Enclosure to PLA-5793 Page 4 of 10 Power Distribution Uncertainties The NRC approved MCPR Safety Limit methodology referenced in T.S. 5.6.5.b uses radial and local power distribution uncertainties that are based on NRC approved statistical methods and code system benchmarks. For the previous Unit 2 Cycle 12 MCPR Safety Limit, radial and local power distribution uncertainties were based on the NRC approved CASMO-3/MICROBURN-B code system that is implemented within the POWERPLEX-Il core monitoring system. The Unit 2 Cycle 13 MCPR Safety Limit radial and local power distribution uncertainties are based on the NRC approved CASMO-4/MICROBURN-B2 code system that is implemented within the POWERPLEX-III core monitoring system. The POWERPLEX-III core monitoring system will be used for U2C13. Radial and local power distribution uncertainties based on the CASMO-4/MICROBURN-B2 code system are smaller than the corresponding uncertainties based on the CASMO-3/MICROBURN-B code system. This change will tend to decrease the MCPR Safety Limit.

Channel Bow 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, FANP issued Supplement 1 to their CPR Methodology, ANF-524(P)(A), "ANF Critical Power Methodology for Boiling Water Reactors." 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 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 Unit 1 Cycle 13 operation (completed February 2004), and subsequent channel bow measurements and analysis of discharged bundles, the amount of channel bow assumed in the Unit 2 Cycle 13 MCPR Safety Limit calculation was increased from that assumed for U2C12. The channel bow assumed in the Unit 2 Cycle 13 MCPR Safety Limit calculation accounts for the U1C13 observations and was chosen to bound the mean channel bow expected during Unit 2 Cycle 13 operation.

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 U2C13 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,

Enclosure to PLA-5793 Page 5 of 10 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 in the Bases of the applicable Technical Specification sections, because the MCPR Safety Limits calculated for U2C13 preserve the required margin of safety, and NRC approved methodology is used to demonstrate all applicable criteria are met.

Operator performance and procedures are unaffected by these proposed changes since the changes 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 U2C13 transient analysis results) may be different from the U2C12 limits. Following use of the methodology to analyze the Unit 2 Cycle 13 core design and future Unit 2 reloads, the reload cycle specific results are incorporated into the FSAR via a licensing document change notice.

4.2 CHANGES TO COLR REFERENCES References pertaining to the PPL's analytical methods are removed and FANP's analytical methods not already incorporated in Section 5.6.5.b are added. These changes are necessary since the Unit 2 Cycle 13 core operating limits are developed using FANP's analytical methods and not the removed PPL analytical methods.

The references in Section 5.6.5.b were reordered to be consistent with Attachment 7. is a list of FANP methodology references that provides the relationship between the references and the applicable Technical Specification Limiting Condition for Operation. The attachment also identifies, for each reference, a description and justification for its use.

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 13.

The proposed change to the MCPR Safety Limits are developed in accordance with NRC approved methods and 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 Limits do not involve a significant reduction in the margin of safety as currently defined in the Bases of the applicable Technical Specification sections. The MCPR Safety Limits calculated for U2C13 preserve the required margin of safety.

Enclosure to PLA-5793 Page 6 of 10 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 U2C13 COLR. Thus, the MCPR operating limits assure that the MCPR Safety Limits will not be exceeded during normal operation or anticipated operational occurrences. The required protection for the fuel rod cladding will be provided and this 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 by the utilization of these new MCPR Safety Limits.

5.0 REGULATORY SAFETY ANALYSIS 5.1 NO SIGNIFICANT HAZARDS CONSIDERATION 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:

The proposed changes to the Unit 2 Technical Specifications are:

TS 2.1.1.2 The Minimum Critical Power Ratio (MCPR) Safety Limits (two-loop operation and single-loop operation) are revised from 1.10 (two-loop operation) and 1.11 (single loop operation) to 1.09 (two-loop operation) and 1.10 (single loop operation) to reflect results of the cycle specific MCPR Safety Limit analysis for Unit 2 Cycle 13.

TS 5.6.5.b Core Operating Limits Report (COLR) references are revised to delete PPL's analytical methods and add the FANP's NRC approved analytical methods that are not already contained in Section 5.6.5.b. The references were reordered to correspond to the list provided as Attachment 7.

Enclosure to PLA-5793 Page 7 of 10 The TS Bases changes corresponding to the proposed TS changes are included for information.

1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

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. Further, the U2C13 MCPR Safety Limits are generated using NRC approved methodology and meet the applicable acceptance criteria. In addition, the effects of channel bow were conservatively addressed by increasing the amount of channel bow assumed in the MCPR SL calculation. Thus, this proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated.

Prior to the startup of U2C13, 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 proposed herein to generate the MCPR operating limits in the U2C13 COLR. These limits could be different from those specified for the U2C12 COLR. The COLR operating limits thus assure that the MCPR Safety Limit will not be exceeded during normal operation or anticipated operational occurrences. Postulated accidents are also analyzed to confirm NRC acceptance criteria are met.

The changes to the references in Section 5.6.5.b were made to properly reflect the NRC approved methodology used to generate the U2C13 core operating limits. The use of this approved methodology does not increase the probability or consequences of an accident previously evaluated.

Therefore, this proposed amendment does not involve a significant increase in the probability 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 change to the MCPR Safety Limits does not directly or indirectly affect any plant system, equipment, or component and therefore does not affect the failure modes of

Enclosure to PLA-5793 Page 8 of 10 any of these items. Thus, the proposed changes do 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 U2C13 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 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 U2C13 preserve the required margin of safety.

The changes to the references in Section 5.6.5.b were made to properly reflect the NRC approved methodology used to generate the U2C13 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

Enclosure to PLA-5793 Page 9 of 10 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.

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.

Enclosure to PLA-5793 Page 10 of 10 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.

7.0 REFERENCES

None.

to PLA-5793 Proposed Unit 2 Technical Specification Changes (Markups)

PL &V SLs Z.0 2.0 SAFETY LIMITS (SLs) 2.1 SLs 2.1.1 Reactor Core 2.1.1.1 2.1.1.2 2.1.1.3 SLs VWih the reactor steam dome pressure < 785 psig or core flow

< 10 million Ibm/hr.

THERMAL POWER shall be s 25% RTP.

VVith the reactor steam dome pressure 2 785 psig and core flow

Ž 10 million Ibm/hr:

MCPR shall be Ž4140for ulation loop operation or 444-for single recirculation loop operation.

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 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />s:

2.2.1 Restore compliance with all SLs; and 2.2.2 Insert all insertable control rods.

SUSUQUEHANNA - UNIT 2 TS / 2.0-1 Amendm nt 1 V1

PeLA 99ez.

Reporting Requirements 5.6 5.6 Reporting Requirements 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 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 LEFMvrm system is used as the feedwater flow measurement input into the core thermal power calculation.

r>. S 0et 1 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, "App!ication of Reactor Analysis Methods for i

d

) \\

~Design and Analysis'../

2.

\\<N-NF-80-19(P)(A), -Exxon Nuclear Methodology for Boilinatr actors" Exxon Nuclear Company, Inc.

3.

XN-N 5-67(P)(A), "Generic Mechanical Design foron Nuclear Jet Pump B Reload Fuel,' Exxon Nuclear Compa

, Inc.

4.

ANF-524(P)(A,Advanced Nuclear Fuels C oration Critical Power Methodology for iling Water Reactors".

5.

NE-092-001A, "Licens Topical Re rt for Power Uprate With Increased Core Flow," Pe sylvan'Power & Light Company.

6.

ANF-89-98(P)(A) "Generic M anical Design Criteria for BWR Fuel Designs," Advanced Nucl Fue Corporation.

7.

ANF-91-048(P)(A),

anced Nuclea els Corporation Methodology for Boiling Water R tors EXEM BWR luation Model."

8.

XN-NF-79-71(A) *Exxon Nuclear Plant Tran nt Methodology for Boiling Wa Reactors."

9.

EMF-7 (P)(A) "ANFB-10 Critical Power Correlation.

10.

don, Inc., 'TOPICAL REPORT: Improving Thermal Powe curacy and Plant Safety while Increasing Operating Power Level Usinge LEFTMIVIm System,' Engineering Report - 80P.

(continued)

SUSQUEHANNA - UNIT 2A TS /5.0-22 A;end 9, ;e;
o

Idet-A Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 COLR (continued)

1.

aldon, Inc., 'Supplement to Topical Report ER-8AP: Basis fr a

rate with the LEFMm or LEFM CheckPluNB S

m ering

~Re-1 60P.-

eMF-85aDEX 2A (WR) eraechanical cmergenicyo l

~~Evaluation Model.'

13-EMF-2158(P)(A), "Sieme nro Mtodology for Boliling Water ReactCs) ElIin and Valimitsiuha SM 4Microbunt B2,ly Siernens P ortion.

% 14 EM,07(P)(A), Volume 4, 'BWR Stability Analysis Assess rf l

/

t_

IF with nput from MICROBURN-132."

1

c.

Th coe=lcbe iis(~.

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 cyde to the NRC.

5.6.6 EDG Failures Report If an individual emergency diesel generator (EDG) experiences four or more valid failures in the last 25 demands, these failures and any nonvalid failures experienced by that EDG in that time period shall be reported within 30 days.

Reports on EDG failures shall include the information recommended In Regulatory Guide 1.9, Revision 3, Regulatory Position C.4.

5.6.7 PAM Report When a report is required by Condition B or F of LCO 3.3.3.1, 'Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the Inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.

SUSQUEHANNA - UNIT 2 TS / 5.0-23 Ampnlment 1V9 jAA, l

INSERT 1:

1.

XN-NF-81-58(P)(A), URODEX2 Fuel Rod Thermal-Mechanical Response Evaluation Model," Exxon Nuclear Company.

2.

XN-NF-85-67(P)(A), "Generic Mechanical Design for Exxon Nuclear Jet pump BWR Reload Fuel," Exxon Nuclear Company.

3.

EMF-85-74(Pi(A), "RODEX2A (BWR) Fuel Rod Thermal-Mechanical Evaluation Model," Siemens Power Corporation.

4.

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

Advanced Nuclear Fuels Corporation.

5.

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

Exxon Nuclear Company.

6.

EMF-2158(P)(A), "Siemens Power Corporation Methodology for Boiling Water Reactors: Evaluation and Validation of CASMO-4/MICROBURN-B2," Siemens Power Corporation.

7.

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

8.

EMF-2361 (P)(A), -EXEM BWR-2000 ECCS Evaluation Model," Framatome ANP.

9.

EMF-2292(P)(A), "ATRIUMT'-10: Appendix K Spray Heat Transfer Coefficients,"

Siemens Power Corporation.

10.

XN-NF-84-105(P)(A), "XCOBRA-T: A Computer Code for BWR Transient Thermal-Hydraulic Core Analysis," Exxon Nuclear Company.

11.

ANF-524(P)(A), "ANF Critical Power Methodology for Boiling Water Reactors,"

Advanced Nuclear Fuels Corporation.

12. ANF-913(P)(A), "COTRANSA2: A Computer Program for Boiling Water Reactor Transient Analyses," Advanced Nuclear Fuels Corporation.

13. ANF-1358(P)(A), 'The Loss of Feedwater Heating Transient in Boiling Water Reactors," Advanced Nuclear Fuels Corporation.

14.

EMF-2209(P)(A), "SPCB Critical Power Correlation," Siemens Power Corporation.

15.

EMF-1997(P)(A), "ANFB-10 Critical Power Correlation", Siemens Power Corporation.

16.

EMF-CC-074(P)(A), "BWR Stability Analysis - Assessment of STAIF with Input from MICROBURN-B2," Siemens Power Corporation.

17.

NE-092-001A, uLicensing Topical Report for Power Uprate With Increased Core Flow," Pennsylvania Power & Light Company.

18.

Caldon, Inc., "TOPICAL REPORT: Improving Thermal Power Accuracy and Plant Safety while Increasing Operating Power Level Using the LEFTMNtM System,"

Engineering Report - 80P.

19.

Caldon, Inc., "Supplement to Topical Report ER-80P: Basis for a Power Uprate with the LEFMvh1' or LEFM CheckPluswm System," Engineering Report ER-1 60P.

to PLA-5793 Proposed Unit 2 Technical Specification Changes (Camera Ready)

PPL Rev. 0 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 flow

> 10 million Ibm/hr:

MCPR shall be 2 1.09 for two recirculation loop operation or 2 1.10 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 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />s:

2.2.1 Restore compliance with all SLs; and 2.2.2 Insert all insertable control rods.

SUSUQUEHANNA - UNIT 2 TS /2.0-1 Amendmrint 1/1 1,A4, fB4

PPL Rev.

Reporting Requirements 5.6 5.6 Reporting Requirements 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 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 LEFMTM 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.

XN-NF-81-58(P)(A), "RODEX2 Fuel Rod Thermal-Mechanical Response Evaluation Model," Exxon Nuclear Company.

2.

XN-NF-85-67(P)(A), "Generic Mechanical Design for Exxon Nuclear Jet pump BWR Reload Fuel," Exxon Nuclear Company.

3.

EMF-85-74(P)(A), "RODEX2A (BWR) Fuel Rod Thermal-Mechanical Evaluation Model," Siemens Power Corporation.

4.

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

5.

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

6.

EMF-2158(P)(A), "Siemens Power Corporation Methodology for Boiling Water Reactors: Evaluation and Validation of CASMO-4/MICROBURN-B2," Siemens Power Corporation.

7.

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

8.

EMF-2361 (P)(A), "EXEM BWR-2000 ECCS Evaluation Model,"

Framatome ANP.

9.

EMF-2292(P)(A), "ATRIUMTM-10: Appendix K Spray Heat Transfer Coefficients," Siemens Power Corporation.

(continued)

SUSQUEHANNA - UNIT 2 TS / 5.0-22 A endment 10 1091, 1, 1 4

PPL Rev.

Reporting Requirements 5.6 5.6 Reporting Requirements

10.

XN-NF-84-105(P)(A), "XCOBRA-T: A Computer Code for BWR Transient Thermal-Hydraulic Core Analysis," Exxon Nuclear Company.

11.

ANF-524(P)(A), "ANF Critical Power Methodology for Boiling Water Reactors," Advanced Nuclear Fuels Corporation.

12.

ANF-913(P)(A), "COTRANSA2: A Computer Program for Boiling Water Reactor Transient Analyses," Advanced Nuclear Fuels Corporation.

13.

ANF-1358(P)(A), "The Loss of Feedwater Heating Transient in Boiling Water Reactors," Advanced Nuclear Fuels Corporation.

14.

EMF-2209(P)(A), "SPCB Critical Power Correlation," Siemens Power Corporation.

15.

EMF-1997(P)(A), "ANFB-10 Critical Power Correlation", Siemens Power Corporation.

16.

EMF-CC-074(P)(A), "BWR Stability Analysis - Assessment of STAIF with Input from MICROBURN-B2," Siemens Power Corporation.

17.

NE-092-001 A, "Licensing Topical Report for Power Uprate With Increased Core Flow," Pennsylvania Power & Light Company.

18.

Caldon, Inc., 'TOPICAL REPORT: Improving Thermal Power Accuracy and Plant Safety while Increasing Operating Power Level Using the LEFTMqTm System," Engineering Report - 80P.

19.

Caldon, Inc., "Supplement to Topical Report ER-80P: Basis for a Power Uprate with the LEFMV' or LEFM CheckPlusT 'l System," Engineering Report ER-1 60P.

(continued)

SUSQUEHANNA - UNIT 2 TS / 5.0-23 Arpendmentr 1 9 4,3, 1§4,XOF

PPL Rev.

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 COLR (continued)

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.

5.6.6 EDG Failures Report If an individual emergency diesel generator (EDG) experiences four or more valid failures in the last 25 demands, these failures and any nonvalid failures experienced by that EDG in that time period shall be reported within 30 days.

Reports on EDG failures shall include the information recommended in Regulatory Guide 1.9, Revision 3, Regulatory Position C.4.

5.6.7 PAM Report When a report is required by Condition B or F of LCO 3.3.3.1, NPost Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.

SUSQUEHANNA - UNIT 2 TS / 5.0-23a Ampnoment 1 '9 1,

4,1-§Q to PLA-5793 List of Regulatory Commitments to PLA-5793 Page 1 of 1 REGULATORY COMMITMENTS

.Due Date/Event There are no new commitments associated with this submittal.

NA to PLA-5793 Proposed Unit 2 Technical Specification Bases (Changes)

Markups for Information Only

PPL Rev.0I.

Reactor Core SLs B 2.1.1 BASES APPLICABLE SAFETY ANALYSES 2.1.1.2 MCPR (continued) 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 and 4 describe the methodology used in determining the MCPR SL.

The ANFB-10 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 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 ANFB-1 0 correlation provide a reasonable degree of assurance that during sustained operation at the MCPR SL there would be no transition boiling 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 I

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 the ANFB-10 Critical 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.

I 2.1.1.3 Reactor Vessel Water Level During MODES I 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 / B2.0-3 RevisionZ3g

PPL Rev. 2R t Reactor Core SLs B 2.1.1 BASES APPLICABLE 2.1.1.3 Reactor Vessel Water Level (continued)

SAFETY ANALYSES 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 Exceeding an SL may cause fuel damage and create a potential for VIOLATIONS 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 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. The 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 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.

10CFR100.

4.

EMF-1 997(P)(A), Revision 0, uANFB-1 0 Critical Power Correlation,'

July 1998 and EMF-1997(P)(A) Supplement 1 Revision 0," ANFB-10 Critical Power Correlation: High Local Peaking Results," July 1998.

TvNsecr TISP: 2.1.1-1 SUSQUEHANNA - UNIT 2 TS / B 2.0-4 Revision ;r

eCL 2 t

SIM B 3.1.1 I

BASES ACTIONS E.1. E.2. E.3. E.4. and E.5 (continued) assumed to be isolated to mitigate radioactivity releases.

This may be performed as an administrative check, by examining logs or other information, to determine if the components are out of service for maintenance or other reasons. It is not necessary to perform the Surveillances as needed to demonstrate the OPERABILITY of the components.

If. however, any required component is inoperable, then it must be restored to OPERABLE status.

In this case. SRs may need to be performed to restore the component to OPERABLE status.

Action must continue until all required components are OPERABLE.

SURVEILLANCE REQUIREMENTS SR 3.1.1.1 SDM must be verified to be within limits to ensure that the reactor can be made subcritical from any initial operating condition.

Adequate SDM is demonstrated by testing before or during the first startup after fuel movement. control rod replacement, or shuffling within the reactor pressure vessel.

Control rod replacement refers to the decoupling and removal of a control rod from a core location, and subsequent replacement with a new control rod or a control rod from another cbre location.

Since core reactivity will vary during the cycle as a function of fuel depletion and poison burnup. the beginning of cycle (BOC) test must also account for changes in core reactivity during the cycle.

Therefore, to obtain the SDM. the initial measured value must be increased by an adder. "R". which is the difference between the calculated value of maximum core reactivity during the operating cycle and the calculated BOC core reactivity.

If the value-of "R" is zero (that is, BOC is the most reactive point in the cycle). no correction to the BOC measured value is required (Ref.7).' For the SON demonstrations that rely solely on calculation of the highest worth control rod, additional margin (0.10% Ak/k) must be added-to the SDM limit of 0.28% Ak/k to account for uncertainties in the calculation.

The SDM may be demonstrated during an in sequence control rod withdrawal, in which the highest worth control rod is analytically determined, or during local criticals. where (continued)

SUSQUEHANNA - UNIT 2 B 3.1-5 Revision ?,t

PPL Rev. Ar :L SDM B 3.1.1 BASES SURVEILLANCE REQUIREMENTS SR 3.1.1.1 (continued) the highest worth control rod is determined by analysis or testing.

Local critical tests require the withdrawal of control rods in a sequence that is not in conformance with BPWS. This testing would therefore require re-programming or bypassing of the rod worth minimizer to allow the withdrawal of control rods not in conformance with BPWS, and therefore additional requirements must be met (see LCO 3.10.7, "Control Rod Testing-Operating").

The Frequency of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after reaching criticality.is allowed to provide a reasonable amount of time to perform the required calculations and have appropriate verification.

During MODE 5, adequate SDM is required to ensure that the reactor does not reach criticality during control rod withdrawals. An evaluation of each planned in-vessel fuel movement during fuel loading (including shuffling fuel within thebcore) is required to ensure adequate SDM is maintained during refueling. This evaluation ensures that the intermediati loading patterns are bounded by the safety analyses for the final, core loading pattern. For example, bounding analyses that demonstrate adequate SDM for the most reactive configurations during the refueling may be performed to demonstrate acceptability of the entire fuel movement sequence. These bounding analyses include additional margins to the associated uncertainties. Spiral offload/reload sequences inherently satisfy the SR, provided the fuel assemblies are reloaded in the same configuration analyzed for the new cycle. Removing fuel from the core will always result in an increase in SDM.

REFERENCES Tunesect 3'3.1

1.

10 CFR 50, Appendix A, GDC 26.

2.

FSAR, Section 15.

3.

-PLNF 90 001 A, "Applieation of Reaotor Analycic Methods for S 3wn Design and Analysis," Seetiens 2.2 and 2.8, July 1992, 6tipplernent 1 lA, August ag 995, Supplemeng;t 2-A, juy

._ 9 ae--

-Cuppemet 3 A, Mareh 2001..

4.

FSAR, Section 15.4.1.1..

(continued)

SUSQUEHANNA - UNIT 2 TS / B 3.1-6 RevisionA, Z

PPL Rev. 2i-SDM B 3.1.1 BASES REFERENCES (continued)

5.

Final Policy Statement on Technical Specfications Improvements, July 22, 1993 (58 FR 39132).

I PeleA~t

6.

FSAR, Section 4.3.

P-NF 90 001 A, OApplieationof eRcator Analysis Mthodsfor BWR Design and Analysis, etion 2.4, July 1992, Supplemcnt 1-A, d

August 1995, Supplemcnt 2 A, July 1 996, and Supplement 3 A, I

SUSQUEHANNA - UNIT 2 TS / B 3.1-7 Revision.4, Z.

PPL Rev. Of-Control Rod Scram Times B 3.1.4 BASES REFERENC (continued)

ES

4.

FSAR, Section 15.0

5..L-_-90-e01

-A, Appl t

aflti Ansis Methods foi-I De/igr and Analysis, Ceetien 4.1.2, July 1992, and Supplmn 1 A r~~~Ags 195 Suppxire lement 2 A,- ul 19, Supleen a-A"",}S"u'

~i

6.

Final Policy Statement on Technical Specifications Improvements, July 22, 1993 (58 FR 39132).

FS.r R, Sec 4o n

7.

Letter from R.F. Janecek (BWROG) to R.W. Starostecki (NRC),

"BWR Owners Group Revised Reactivity Control System Technical Specifications, BWROG-8754, Septemberj7, 1987.

SUSQUEHANNA - UNIT 2 TS / B3.1-28 Revision~t Z-

PPL Rev..t Rod Pattern Control B 3.1.6 BASES ACTIONS (continued)

B.1 and B.2 If nine or more OPERABLE control rods are out of sequence, the control rod pattern significantly deviates from the prescribed sequence. Control rod withdrawal should be suspended immediately to prevent the potential for further deviation from the prescribed sequence. Control rod insertion to correct control rods withdrawn beyond their allowed position is allowed since, in general, insertion of control rods has-less impact on control rod worth than' withdrawals have. Required Action B.1 is modified by a Note which allows the RWM to be bypassed to allow the affected control rods to be returned to their correct position. LCO 3.3.2.1 requires verification of control rod movement by a qualified member of the technical staff.

When nine or more OPERABLE control rods are not in compliance with BPWS, the reactor mode switch must be placed in the shutdown position within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. With the mode switch in shutdown, the reactor is shut down, and as such, does not meet the applicability requirements of this LCO. The allowed Completion Time of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> is reasonable to allow insertion of control rods to restore compliance, and is appropriate relative to the low probability of a CRDA occurring with the control rods out of sequence.

SURVEILLANCE SR 3.1.6.1 REQUIREMENTS The control rod pattern is verified to be in compliance with the BPWS at a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Frequency to ensure the assumptions of the CRDA analyses are met. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Frequency was developed considering that the primary check on compliance with the BPWS is performed by the RWM (LCO 3.3.2.1), which provides control rod blocks to enforce the required sequence and is required to be OPERABLE when operating at s 10% RTP.

REFERENCES Inv)sect I s 6II o

3.- I

1. } *L

-<V 00 001 A, mApplieatien ! feleetr Analysis Metheds fe.F 5

WfR Design end Analysis," Geetien 2.Q, July 1992, Supplement 1 4A August 1995, Supplement 2 A, July 1996, end Supplemcnt 3-A, I AmMereh QIA6-I (continued)

SUSQUEHANNA - UNIT 2 TS / B3.1-37 Revision I, Z

PPL Rev.Zj.

APLHGR B 3.2.1 BASES ACTIONS B.1 If the APLHGR cannot be restored to within its required limits within the associated Completion Time, the plant must be brought to a MODE or other specified condition in which the LCO does not apply. To achieve this status, THERMAL POWER must be reduced to < 25% RTP within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. The allowed Completion Time is reasonable, based on operating experience, to reduce THERMAL POWER to < 25% RTP in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.2.1.1 REQUIREMENTS APLHGRs are required to be initially calculated within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after THERMAL POWER is > 25% RTP and then every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.thereafter.

Additionally, APLHGRs must be calculated prior to exceeding 50% RTP unless performed in the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. APLHGRs are compared to the specified limits in the COLR to ensure that the reactor Is operating within the assumptions of the safety analysis. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Frequency is based on both engineering judgment and recognition of the slowness of

• changes in power distribution during normal operation. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> allowance after THERMAL POWER' 25% RTP is achieved is acceptable given the large inherent margin to operating limits at low power levels and because the APLHGRs must be calculated prior to exceeding 50% RTP.

REFERENCES

1.

Not Used I

2.

Not Used-l

3. {aAF91 048(P)(A), "Advaneed NuoleaF Fuels Croain-(Methedelegy efo B3oiing-a Rceoter SXEMBREvlaio-=

JeE' T -au 3,2. 1-n 1O3.

4.

ANF-CC-33(P)(A) Supplement 2, NHUXY: A Generalized Multirod Heatup Code with 1 OCFR50 Appendix K Heatup Option,' January 1991.

5.

XN-CC-33(P)(A) Revision 1, HUXY: A Generalized Multirod Heatup Code with 1 OCFR50 Appendix K Heatup Option Users Manual,' November 1975.

(continued)

SUSQUEHANNA - UNIT 2 TS / B3.2-3 Revision,* 7-

PPL Rev.21 MCPR B 3.2.2 B 3.2 POWER DISTRIBUTION LIMITS B 3.2.2 MINIMUM CRITICAL POWER RATIO (MCPR)

BASES BACKGROUND MCPR is a ratio of the fuel assembly power that would result in the onset of boiling transition to the actual fuel assembly power. The MCPR Safety Limit (SL) is set such that 99.9% of the fuel rods avoid boiling transition if the limit is not violated (refer to the Bases for SL 2.1.1.2). The operating limit MCPR is established to ensure that no fuel damage results during anticipated operational occurrences (AOOs). Although fuel damage does not necessarily occur if a fuel rod actually experienced boiling transition (Ref. 1), the critical power at which boiling transition Is calculated to occur has been adopted as a fuel design criterion.

The onset of transition boiling is a phenomenon that Is readily detected during the testing of various fuel bundle designs. Based on these experimental data, correlations have been developed to predict critical bundle power (i.e., the bundle power level at the onset of transition boiling) for a given set of plant parameters (e.g., reactor vessel pressure, flow, and subcooling). Because plant operating conditions and bundle power levels are monitored and determined relatively easily, monitoring the MCPR is a convenient way of ensuring that fuel failures due to inadequate cooling do not occur.

APPLICABLE The analytical methods and assumptions used in evaluating the A0Os to SAFETY ANALYSES establish the operating limit MCPR are presented in References 2 throughpt To ensure that the MCPR SL is not exceeded during any transient event that occurs with moderate frequency, limiting transients have been analyzed to determine the largest reduction in critical power ratio (CPR). The types of transients evaluated are loss of flow, increase in pressure and power, positive reactivity insertion, and coolant temperature decrease. The limiting transient yields the largest change in CPR (ACPR). When the largest ACPR is added to the MCPR SL, the required operating limit MCPR is obtained.

The MCPR operating limits derived from the transient analysis are dependent on the operating core flow and power state to ensure adherence to fuel design limits during the worst transient that occurs with moderate frequency These analyses may also consider other (continued)

SUSQUEHANNA - UNIT 2 TS I B 3.2-5 RevisionZ3

PPL Rev.O-s MCPR B 3.2.2 BASES APPLICABLE combinations of plant conditions (i.e., control rod scram speed, bypass SAFETY ANALYSES valve performance, EOC-RPT, cycle exposure, etc.). Flow dependent (continued) n a

of slow flow runout transients.

The MCPR satisfies Criterion 2 of the NRC Policy Statement (Ref~a.

LCO The MCPR operating limits specified in the COLR are the result of the Design Basis Accident (DBA) and transient analysis. The operating limit MCPR is determined by the larger of the flow dependent MCPR and power dependent MCPR limits.

APPLICABILITY The MCPR operating limits are primarily derived from transient analyses that are assumed to occur at high power levels. Below 25% RTP, the reactor is operating at a minimum recirculation pump speed and the moderator void ratio is small. Surveillance of thermal limits below 25% RTP is unnecessary due to the large inherent margin that ensures 4that the MCPR SL is not exceeded even if a limiting transient occurs.

• Studies of the variation of limiting transient behavior have been performed over the range of power and flow conditions. These studies encompass the range of key actual plant parameter values important to typically limiting transients. The results of these studies demonstrate that a margin is expected between performance and the MCPR requirements, and that margins increase as power is reduced to 25% RTP. This trend is expected to continue to the 5% to 15% power range when entry into MODE 2 occurs. When in MODE 2, the intermediate range monitor provides rapid scram initiation for any significant power increase transient, which effectively eliminates any MCPR compliance concem.

Therefore, at THERMAL POWER levels < 25% RTP, the reactor is operating with substantial margin to the MCPR limits and this LCO is not required.

ACTIONS A.1 If any MCPR is outside the required limits, an assumption regarding an initial condition of the design basis transient analyses may not be met.

Therefore, prompt action should be taken to restore the MCPR(s) to within the required limits such that the plant remains operating within (continued)

SUSQUEHANNA - UNIT 2 TS / B3.2-6 Revision Z3

PPL Rev./ t-MCPR B 3.2.2 BASES ACTIONS A.1 (continued) analyzed conditions. The 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Completion Time is normally sufficient to restore the MCPR(s) to within its limits and Is acceptable based on the low probability of a transient or DBA occurring simultaneously with the MCPR out of specification.

B.1 If the MCPR cannot be restored to within its required limits within the associated Completion Time, the plant must be brought to a MODE or other specified condition in which the LCO does not apply. To achieve this status, THERMAL POWER must be reduced to < 25% RTP within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. The allowed Completion Time Is reasonable, based on operating experience, to reduce THERMAL POWER to < 25% RTP in an orderly manner and without challenging plant systems.

SURVEILLANCE REQUIREMENTS SR 3.2.2.1 The MCPR is required to be initially calculated within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after THERMAL POWER is 2 25% RTP and then every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter.

Additionally, MCPR must be calculated prior to exceeding 50% RTP unless performed in the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. MCPR is compared to the specified limits in the COLR to ensure that the reactor is operating within the assumptions of the safety analysis. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Frequency is based on both engineering judgment and recognition of the slowness of changes in power distribution during normal operation. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> allowance after THERMAL POWER Ž 25% RTP is achieved is acceptable given the large inherent margin to operating limits at low power levels and because the MCPR must be calculated prior to exceeding 50% RTP.

-I NS E 9, SR 3.2.2.2_

f fusethe transient analysis takes cred it for conservatism in the sa ime pa ce, it must be demonstrated that the specii me

/ s cnsistent w os used in the transient anals. 2.

.2 T BI 3

-determines the scram faction which is a e of the actual

) scram time compardwt th-e

~ f tme. The COLR contains a table of scram ti~mef

)hsetisd on the LCO 3.1.4, 'Control Rod Scram Times' and t alstic scram tin sed in the transient analysis. The MC oerating limit is then determin sed on an interpolati tween the applicable limits for scram times o 3.1.4, Rod Scram Timest and realistic scram time analyses using (continued)

SUSQUEHANNA - UNIT 2 TS / B3.2-7 RevislonkZ-

PPL Rev.z1t MCPR B 3.2.2 BASES SURVEILLANCE SR 3.2.2.2 REQUIREMENTS iued) me fraction. The scram time fraction and correspondin I

chratngesinthescrm tierfrintd once wethin 7d the of scra tie tsts equ S4.

, SR 3.1.4.3 and SR 3.1.4.4 because the chnedrn h cycle. TiZta~mlto Time sacpal~lLrltvl fanges in the scram time frcineptddunghe1 l_

REFERENCES jiJ.SERT TSB 3.-2.2d

1.

NUREG-0562, June 1979.

2.

L -NF9909001 A, !Applicatiin of Reactor Analysis Methods for BWAr Design And Analysis, July 1992, Supplement 1 A, March 2001.

3.

PL NF °7 001 A, "Qualification of Steady State core Physc Me athods adefa BWR Pesi path' ApMz 28t

-1988.Q_

4.

-PL fl 9 005 A, "Qualification of Transicnt Analysis Mcthods for

-WR DBesign and Analysis," July 1992, including Supplemcnte 1

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 Increased Core Flow,' December 1992, and NRC Approval Letter: Letter from T. E. Murley (NRC) to R. G. Byram (PP&L),

'Licensing Topical Report for Power Uprate With Increased Core Flow, Revision 0, Susquehanna Steam Electric Station, Units 1 and 2 (PLA-3788) (TAC Nos. M83426 and M83427),"

November 30, 1993.

7.

EMF-1 997, Revision 0 (October 1997) and Supplement 1, Revision 0 (January 1998), 'ANFB-10 Critical Power Correlation,"'

and associated NRC SER dated 7/17/98.

8.

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

March 1986.

(continued)

SUSQUEHANNA-UNIT 2 TS / B 3.2-8 RevisionZ.3

PPL Rev.Oj0 MCPR B 3.2.2 BASES Reference (continued) ert TSB 3.2z.

9.

XN-NF-84-1 05(P)(A), Volume 1 and Volume 1 Supplements 1 and 2, 'XCOBRA-T: A Computer Code for BWR Transient Thermal-Hydraulic Core Analysis," February 1987.

I I X

Final Policy Statement on Technical Specifications Improvements, July 22,1993 (58 FR 39132).

2.

Ins I

SUSQUEHANNA - UNIT 2 TS / B3.2-9 Revislon~e3

PPL Rev.,f Control Rod Testing-Operating B 3.10.7 BASES SURVEILLANCE SR 3.10.7.2 REQUIREMENTS (continued)

When the RWM provides conformance to the special test sequence, the test sequence must be verified to be correctly loaded into the RWM prior to control rod movement. This Surveillance demonstrates compliance with SR 3.3.2.1.8, thereby demonstrating that the RWM is OPERABLE. A Note has been added to indicate that this Surveillance does not need to be performed if SR 3.10.7.1 is satisfied.

REFERENCE 1 USE 5

T TSf3 3.)o.7-1

1. FSAR 15.4.9

2.

1 l

Dosign and Analysis," July 1992, Supplement 1 A, August 1995, I

Suppement 2 A, july 1996, and Supplement 3 A, March 2001.

I-CS Iko An SUSQUEHANNA - UNIT 2 TS / B3.10-33 Revisionsr 7

PPL Rev.Z0.1 SDM Test-Refueling B 3.1 0.8 BASES SURVEILLANCE SR 3.10.8.4 REQUIREMENTS (continued)

Periodic verification of the administrative controls established by this LCO will ensure that the reactor is operated within the bounds of the safety analysis. The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Frequency is intended to provide appropriate assurance that each operating shift is aware of and verfies compliance with these Special Operations LCO requirements.

SR 3.10.8.5 Coupling verification is performed to ensure the control rod is connected to the control rod drive mechanism and will perform its intended function when necessary. The verification Is required to be performed any time a control rod is withdrawn to the "full out" notch position, or prior to declaring the control rod OPERABLE after work on the control rod or CRD System that could affect coupling. This Frequency is acceptable, considering the low probability that a control rod will become uncoupled when it is not being moved as well as operating experience related to uncoupling events.

SR 3.10.8.6 CRD charging water header pressure verification is performed to ensure the motive force is available to scram the control rods in the event of a scram signal. A minimum accumulator pressure is specified, below which the capability of the accumulator to perform its intended function becomes degraded and the accumulator is considered inoperable. The minimum accumulator pressure of 940 psig is well below the expected pressure of 1100 psig. The 7 day Frequency has been shown to be acceptable through operating experience and takes into account indications available in the control room.

REFERENCE 1 PL-NF 910 001 A, "Applicatien ef Rcator Analysis Methd3 for BWl E

Design and Analysis," July 1992, Supplement 1 A, Augu3t 1995, IiB1SE-TT-1 Supplement2 A, July 1996, and Supplement 0 A, Mareh 2001.

SUSQUEHANNA - UNIT 2 TS / B 3.10-39 Revislon..V ?_

INSERT TSB2.1.1-1:

5.

EMF-2158(P)(A), "Siemens Power Corporation Methodology for Boiling Water Reactors: Evaluation and Validation of CASMO-4/MICROBURN-B2,"

Siemens Power Corporation.

INSERT TSB3.1.1-1:

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.

INSERT TSB3.1.6-1:

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.

INSERT TSB 3.2.1-1 EMF-2361 (P)(A), "EXEM BWR-2000 ECCS Evaluation Model," Framatome ANP.

INSERT TSB 3.2.2-1

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-80-1 9(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.

INSERT TSB 3.2.2-2

10. ANF-1 358(P)(A) Revision 1, 'The Loss of Feedwater Heating Transient in Boiling Water Reactors," Advanced Nuclear Fuels Corporation, September 1992.

INSERT TSB 3.2.2-3 Because the transient analysis takes credit for conservatism in the scram time performance, it must be demonstrated that the specific scram time is consistent with those used in the transient analysis. SR 3.2.2.2 compares the average measured scram times to the assumed scram times documented in the COLR. The COLR contains a table of scram times based on the LCO 3.1.4, "Control Rod Scram Times and the realistic scram times, both of which are used in the transient analysis. If the average measured scram times are greater than the realistic scram times then the MCPR operating limits corresponding to the Maximum Allowable Average Scram Insertion Time must be implemented. 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 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> 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 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Completion Time is acceptable due to the relatively minor changes in average measured scram times expected during the fuel cycle.

INSERT TSB3.10.7-1:

XN-NF-80-1 9(P)(A) Volume 1 and Supplements 1 and 2, "Exxon Nuclear Methodology for Boiling Water Reactors," Exxon Nuclear Company, March 1983.

INSERT TSB3.10.8-1:

XN-NF-80-1 9(P)(A) Volume 1 and Supplements 1 and 2, "Exxon Nuclear Methodology for Boiling Water Reactors," Exxon Nuclear Company, March 1983.

I to PLA-5793 Unit 2 Cycle 13 Core Composition to PLA-5793 Page 1 of I Unit 2 Cycle 13 Core Composition Assembly Type Operational History Number of-Assemblies:

FANP ATRIUM'-i0 Fresh 292 FANP ATRIUMTM-10 Once-burned 284 FANP ATRIUMTM-10 Twice-burned 188 to PLA-5793 Correlation of Potential Safety Limit to Calculated Pins in Boiling Transition to PLA-5793 Page 1 of 1 U2C13 MCPR SL Results Percentage Fuel Rods in Boiling Transition Operating Condition MCPR SL-of Pins in BT*

Two Loop 1.09 0.0826 1.08 0.1189 Single Loop 1.11 0.0678 1.10 0.0897 1.09 0.1135

• Total number of fuel pins in core = 69524.

to PLA-5793 Relationship between FANP References and SSES Technical Specification LCO's

BWR Approved Topical Reports for Susquehanna Nuclear Plant Technical Specifications COLR References Attachment Page A-1 i

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 BVvR 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 I 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 I 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 bumup 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 3.1.1 Describes the reactor core simulator code MICROBURN-B2 and the Methodology for Boiling Water Reactors: Evaluation and 3.2.2 lattice physics code CASMO4.

Validation of CASMO-4IMICROBURN-B2, Siemens 3.2.3 Power Corporation, October 1999.

3.3.2.1 Table 3.3.2.1-1 JGtO%

BWR Approved Topical Reports for Susquehanna Nuclear Plant Technical Soecifications COLR References Attachment' Page A-2 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.

ANF-91-048(P)(A), Advanced Nuclear Fuels Corporation 3.2.1 Describes an upgraded evaluation model methodology for licensing Methodology for Boiling Water Reactors EXEM BWNR analyses of postulated LOCAs in jet pump BWRs. The methodology Evaluation Model, Advanced Nuclear Fuels Corporation, was developed to comply with 10 CFR 50.46 and Appendix K criteria to January 1993.

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, ATRIUMm-10: 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-10 fuel September 2000.

LOCA analyses.

XN-NF-80-1 9(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 1 3.2.2 Provides a capability to perform analyses of transient heat transfer and 2, XCOBRA-T: A Computer Code for BWR Transient behavior in BWR assemblies.

Thermal-Hydraulic Core Analysis, Exxon Nuclear Company, February 1987.

ANF-524(P)(A) Revision 2 and Supplements I and 2, ANF 3.2.2 Provides a methodology for the determination of thermal margins, Critical Power Methodology for Boiling Water Reactors, specifically the MCPR safety limit.

Advanced Nuclear Fuels Corporation, November 1990.

ANF-913(P)(A) Volume I Revision I and Volume 1 3.2.2 Provides a computer program for analyzing BWR system transients.

Supplements 2, 3 and 4, COTRANSA2: A Computer Program for Boiling Water Reactor Transient Analyses, Advanced Nuclear Fuels Corporation, August 1990.

JGSKYM

BWR Approved Topical Reports for Susquehanna Nuclear Plant Technical Specifications COLR References Attachment

-A Page A-3 Report Applicable LCO Methodology / Justification ANF-1358(P)(A) Revision 1, 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 1992.

EMF-2209(P)(A) Revision 1, SPCB Critical Power Correlation, 3.2.2 Presents a critical power correlation for use with the ATRIUMu-1 O' Siemens Power Corporation, July 2000.

fuel designs. This correlation is used in the BWR-2000 LOCA methodology.

EMF-1997(P)(A) Revision 0, ANFB-10 Critical Power 3.2.2 Presents a critical power correlation for use with the ATRIUM-10 fuel Correlation, Siemens Power Corporation, July 1998; and designs. This correlation Is used during core design and monitoring EMF-1997(P) Supplement 1(P)(A) Revision 0, ANFB-10 Critical Power Correlation: High Local Peaking Results, Siemens Power Corporation, July 1998.

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 STAIF with Input from MICROBURN-B2, Siemens Power Corporation, August 2000.

ATRIUM is a trademark of Framatome ANP.

4Gb 400