PLA-5990, Proposed Amendment No. 283 to Unit 1 MCPR Safety Limits and Reference Changes PLA-5990

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Proposed Amendment No. 283 to Unit 1 MCPR Safety Limits and Reference Changes PLA-5990
ML053470314
Person / Time
Site: Susquehanna Talen Energy icon.png
Issue date: 12/01/2005
From: Mckinney B
Susquehanna
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
PLA-5990
Download: ML053470314 (57)
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Text

% t Britt T. McKinney PPL Susquehanna, LLC Sr. Vice President & Chief Nuclear Officer 769 Salem Boulevard I "

Berwick, PA 18603 Tel. 570.542.3149 Fax 570.542.1504 btmckinneyO pplweb.com PP DEC 01 2005 U. S. Nuclear Regulatory Commission Attn.: Document Control Desk Mail Stop OPI-17 Washington, DC 20555 SUSQUEHANNA STEAM ELECTRIC STATION PROPOSED AMENDMENT NO. 283 TO UNIT 1 LICENSE NPF-14: MCPR SAFETY LIMITS AND REFERENCE CHANGES PLA-5990 Docket No. 50-387 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 1.

The purpose of this letter is to propose changes to the Susquehanna Steam Electric Station Unit 1 Technical Specifications. Included is a revision to Section 2.1.1.2 which reflects the Unit 1 Cycle 15 (UIC 15) Minimum Critical Power Ratio (MCPR) Safety Limits for single-loop operation. (Note that there is no change required to the MCPR Safety Limit for two-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.

Attachment I to this letter contains the applicable pages of the Susquehanna SES Unit 1 Technical Specifications, marked to show the proposed change. is included which identifies that there are no regulatory commitments associated with this change.

A0O(

Document Control Desk PLA-5990 Attachment 3 contains the applicable pages of the Susquehanna SES Unit 1 Technical Specifications Bases, marked to show the proposed changes (Provided for Information). has been provided as a description of the UlC15 core composition to assist in your review. provides the preliminary UlC15 Core Loading Pattern. contains preliminary descriptions of the reload bundles utilized in UIC15. contains the relationship between Framatome-ANP (FANP) references and Technical Specification LCO's. 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 2006 to support the startup of U ICl15 operation. Therefore, we request NRC complete its review of this change by February 17, 2006 with the changes effective upon startup following the Unit 1 14th 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: 22bi~/-(IL B. T. McKinney

Document Control Desk PLA-5990

Enclosure:

PPL Susquehanna Evaluation of the Proposed Changes Attachments: - Proposed Technical Specification Changes Unit 1, (Mark-ups) - List of Regulatory Commitments - Proposed Technical Specification Bases Changes Unit 1, (Mark-ups) - Description of U1C15 Core Composition - Preliminary UlC15 Core Loading Pattern - Preliminary descriptions of Reload Bundles for UlC15 - Relationship between FANP references and TS LCO's - Diagram of the NRC approved MCPR Safety Limit Methodology cc: NRC Region I Mr. B. A. Bickett, NRC Sr. Resident Inspector Mr. R. V. Guzman, NRC Project Manager Mr. R. Janati, DEPIBRP

ENCLOSURE to PLA-5990 PPL Evaluation UNIT 1 CYCLE 15 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-5990 Page I of 12 PPL EVALUATION

Subject:

Unit 1 Cycle 15 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-14 for PPL Susquehanna, LLC (PPL), Susquehanna Steam Electric Station Unit 1 (SSES).

The proposed changes would revise the Susquehanna Unit 1 Technical Specifications (TS) Section 2.1.1.2 to reflect the Unit 1 Cycle 15 (U1C15) Minimum Critical Power Ratio (MCPR) Safety Limit for single-loop operation. The change to Section 2.1.1.2 is necessary because, as a result of UlC15 cycle specific calculations, the single-loop operation MCPR Safety Limit needs to be increased. The proposed changes also would revise Susquehanna Unit 1 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 Framatome-ANP's (FANP) NRC approved analytical methods.

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 (February 17, 2006) 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 2006 Unit 1 outage.

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 Limit, single-loop operation, is revised from 1.10 to 1.12 to reflect results of the cycle specific MCPR Safety Limit analysis for Unit 1 Cycle 15.

Enclosure to PLA-5990 Page 2 of 12 2.2 TS 5.6.5.b Core Operating Limits Report (COLR) references are revised to delete PPL's analytical methods and add 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 I 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 1 Cycle 15 (U1C15) MCPR Safety Limit for 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 1 Technical Specifications) were established. The change to Section 2.1.1.2 reflects the change from the previous Unit 1 MCPR Safety Limits to the UlC15 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.

FANP calculated both two-loop and single-loop Safety Limit MCPR values for Unit I Cycle 15 using NRC approved analytical methods with the SPCB critical power correlation for ATRIUMTM-10 fuel. Input to the UlC15 MCPR Safety Limit analysis, provided by PPL, assumed the rated core thermal power of 3489 MWt. The Safety Limit

Enclosure to PLA-5990 Page 3 of 12 MCPR values assure that at least 99.9% of the fuel rods are expected to avoid boiling transition during normal operation or anticipated operational occurrences. The calculation results demonstrated that the two-loop Safety Limit MCPR did not change from the value contained in the current Unit 1 TS, and the current value of 1.09 remains applicable for Unit 1 Cycle 15. However, the single-loop Safety Limit MCPR increased compared to the current Unit 1 TS value (1.10 to 1.12).

The MCPR Safety Limit analysis is the first in a series of analyses that assure the new core loading for U1C15 is operated in a safe manner. Prior to the startup of UlC15, 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 UlCI5 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 UlCI5 and the results shown to be within the NRC approved criteria.

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 the PPL's analytical methods were removed and FANP's NRC approved analytical methods not already incorporated in Section 5.6.5.b were added.

These changes are necessary since the Unit 1 Cycle 15 core operating limits were developed using FANP's NRC approved analytical methods.

The references in Section 5.6.5.b were reordered to be consistent with Attachment 7. provides the relationship between the references generated by FANP in Section 5.6.5.b and the applicable Technical Specification Limiting Condition for Operation.

4) ob Enclosure to PLA-5990 Page 4 of 12

4.0 TECHNICAL ANALYSIS

4.1 MCPR SAFETY LIMIT CHANGE This Technical Specification change increases the single-loop MCPR Safety Limit from 1.10 (current Unit 1 TS) to 1.12 (Unit 1 Cycle 15). The two-loop MCPR Safety Limit (1.09) did not change as a result of the Unit 1 Cycle 15 analysis and remains applicable.

The single-loop MCPR Safety Limit change occurs due to cycle-to-cycle variation and changing the critical power correlation from ANFB-10 to SPCB. A more detailed description of the reasons for the 0.02 increase in the single-loop MCPR Safety Limit is provided below in sections "Cycle-to-Cycle Variation" and "SPCB Critical Power Correlation." In addition, due to recent control cell interference issues due to channel bow on Unit 1, the channel bow assumptions for the Unit 1 Cycle 15 MCPR Safety Limit are discussed below ("Channel Bow" section). The channel bow assumptions used in the Unit 1 Cycle 15 MCPR Safety Limit did not contribute to the increase in the single-loop MCPR Safety Limit.

Cycle-to-Cycle Variation The preliminary Unit 1 Cycle 15 core consists of a full core of FANP's ATRIUMTm-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 1 Cycle 15 is split into three different assembly types as described in Attachment 6. The descriptions of the Unit 1 Cycle 14 core loading and reload fuel assemblies 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) 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 1 MCPR Safety Limit is based on FANP's NRC approved ANFB-l0 Critical Power Correlation, EMF-1997(P)(A), which is referenced in TS 5.6.5.b.

Beginning with Unit 1 Cycle 15, PPL plans on using FANP's NRC approved SPCB Critical Power Correlation, EMF-2209 (P)(A), for MCPR Safety Limit determination, reload licensing analyses, and MCPR monitoring. SPCB will be included in TS 5.6.5.b per Section 4.2.

Enclosure to PLA-5990 Page 5 of 12 Single-loop safety analyses including the MCPR Safety Limit are performed at the highest rod-line that is within the analyzed operating domain for single-loop operation; 76% of rated thermal power and 52 Mlbm/hr core flow. The core flow value of 52 Mlbmn/hr is the maximum that can be achieved due to recirculation pump speed limitations in single-loop operation specified in TS 3.4.1. Due to a flow discontinuity in the ANFB-10 Critical Power Correlation near 52 Mlbmlbr core flow, the previous Unit 1 MCPR Safety Limit calculation was performed at a core flow of 55 Mlbmlhr. FANP had previously performed sensitivities for ANFB-1O at various single-loop core flows and demonstrated that 55 Mlbm/hr core flow provides a conservative single-loop MCPR Safety Limit. The SPCB Critical Power Correlation does not have a flow discontinuity at 52 Mlbmlhr. Therefore, FANP performed the UlC15 single-loop MCPR Safety Limit at 52 Mlbm/hr core flow and also at 55 Mlbmlhr core flow. The results of the analyses are provided in the following table:

Percentage of Pins in Boiling Transition at Single-Loop Conditions for SPCB

% of Pins in Boiling  % of Pins in Boiling Transition Transition 52 Mlbm/hr Core Flow 55 Mlbm/hr Core Flow 1.10 MCPR SL 0.198 Calculation Not Performed

1. 1 MCPR SL 0.131 0.095 1.12 MCPR SL 0.087 Calculation Not Performed The preceding table shows that changing the core flow from 55 to 52 Mlbm/hr increases the MCPR Safety Limit by 0.01 for the SPCB Critical Power Correlation. Analyzing core flows less than 52 Mlbm/hr is not necessary since a smaller core flow at 76% rated thermal power is outside the analyzed operating domain, and a reduction of both core flow and core thermal power would result in a smaller MCPR Safety Limit.

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 1 operating cycles. The channel bow assumptions used for Unit 1 Cycle 15 are consistent with the current Unit 1 assumptions. Therefore, the change in the single-loop MCPR Safety Limit is not the result of a change in channel bow assumptions.

Enclosure to PLA-5990 Page 6 of 12 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 I 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 1 operating cycles which may indicate that fuel channel bow is larger than the FANP nominal database, PPL requested FANP to increase the amount of channel bow assumed in the Unit 1 Cycle 15 MCPR Safety Limit calculation. PPL requested that FANP use a mean channel bow for the highly exposed Unit 1 Cycle 15 fuel assemblies of 122.6 mils (from an initial value of 61.3 mils) which is consistent with mean channel bow assumed for the current Unit 1 operating cycle following the recently completed Unit 1 maintenance outage. As part of the continuous validation of safety analyses assumptions, PPL will confirm that the actual Unit 1 Cycle 15 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 1 Cycle 15 operation and a potential channel measurement and re-channeling campaign during the refueling and inspection outage preceding Unit 1 Cycle 15 operation. PPL will continue to monitor fuel channel performance in conformance with PPL's fuel channel monitoring program.

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 UlC15 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 in the Bases of the applicable Technical Specification sections, because the MCPR Safety Limits calculated for UlC15 preserve the required margin of safety.

Enclosure to PLA-5990 Page 7 of 12 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 U1C15 transient analysis results) may be different from the current Unit 1 limits. Following use of the methodology to analyze the Unit I Cycle 15 core design and future Unit 1 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 References pertaining to PPL's analytical methods are removed and FANP's NRC approved analytical methods not already incorporated in Section 5.6.5.b are added. These changes are necessary since the Unit 1 Cycle 15 core operating limits are developed using FANP's NRC approved analytical methods and will allow the use of FANP's NRC approved analytical methods for future reloads.

The references in Section 5.6.5.b are reordered to be consistent with Attachment 7. provides the relationship between the references generated by FANP 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 FANP 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 I Cycle 15.

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 UlC15 preserve the required margin of safety.

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

R A 0,

Enclosure to PLA-5990 Page 8 of 12 Limit values proposed herein to generate the MCPR operating limits in the UIC15 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.

5.0 REGULATORY SAFETY ANALYSIS 5.1 NO SIGNIFICANT HAZARDS CONSIDERATION The proposed changes would revise the following:

TS 2.1.1.2 The Minimum Critical Power Ratio (MCPR) Safety Limit, single-loop operation, is revised from 1.10 to 1.12 to reflect results of the cycle specific MCPR Safety Limit analysis for Unit 1 Cycle 15.

TS 5.6.5.b Core Operating Limits Report (COLR) references are revised to delete PPL's analytical methods and add 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.

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:

Enclosure to PLA-5990 Page 9 of 12

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 single-loop MCPR Safety Limit does not directly or indirectly affect any plant system, equipment, component, or change the processes used to operate the plant. Further, the proposed U1C15 MCPR Safety Limit was generated using NRC approved methodology and meets 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 UlC15, 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 UlCl5 COLR. These limits could be different from those specified for the current Unit 1 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 prior to the startup of UlC15 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 UlC15 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 change to the single-loop MCPR Safety Limit does 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.

£ Enclosure to PLA-5990 Page 10 of 12 The changes to the references in Section 5.6.5.b were made to properly reflect the NRC approved methodology used to generate the UlC15 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 single-loop MCPR Safety Limit 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 UlC15 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 U1C15 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.

E Enclosure to PLA-5990 Page 11 of 12 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.

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:

J -

Enclosure to PLA-5990 Page 12 of 12 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.

7.0 REFERENCES

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.

J Attachment 1 to PLA-5990 Proposed Technical Specification Changes (Markups)

(Unit 1)

OT PPL Rev. 2 Safety Limits (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

2.1.1.2 With the reactor steam dome pressure 2785 psig and core flow 210 million Ibm/hr:

MCPR shall be 2 1.09 for two recirculation loop operation or Ž 449for 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.

SUSQUEHANNA - UNIT 1 TS /2.0-1 Amendment IX,4;1`61

PPL Rev. 3 Reporting Requirements 5.6 p 5.6 Renortina Requlrements 5.6.5 COLR (continued)

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

Future revisions of approved analytical methods listed inthis Technical Specification that are currently referenced to 102% of rated thermal power (3510 MWt) shall include reference that the licensed RTP is actually 3489 MWL The revisions shall document that the licensing analysis performed at 3510 MWt bounds operation at the RTP of 3489 MWt so long as the LEFM/*'T system is used as the feedwater flow measurement input into the core thermal power calculation.

The approved analytical methods are described in the following ASE1RT 1 if NL-NNF-90-001-A, "Application of Reactor Analysis Methods for W)

{ Designand Analysis."/

l 2. XN-NF8 P)(A), "Exxon Nuclear Methlodology for Boili tr Reactors," N uclear Company, Inc.

p0 3. XN-NF-85-67(P)(A) nerc Mechanical De for Exxon Nuclear Jet Pump BWR Reload Fue, on Nudear pany, Inc.

4. ANF-524(P)(A), "Advanced Nu uels Corporation Critical Power Methodology for Boiling Wastea r "
5. NE-092-001A, "Uce gTopical Report oower Uprate With Increased Core ," Pennsylvania Power & t Company.
6. ANF-89 (A), "Generic Mechanical Design Critea r BWR Fuel Des' s Advanced Nuclear Fuels Corporation.

(continued) p SUSQUEHANNA - UNIT 1 TS / 5.0-22 ,,ge ,Ten t i76 9216

X r PPL Rev. 3 Reporting Requirements 5.6 P 5.6 Reporting Requirements 5.6.5 COLR (co ANF-91-048(P)(A), "Advanced Nuclear Fuels Corporation Methodolo for Boiling Water Reactors EXEM BWR Evaluation Model.'

8.. X F-79-71 (P)(A), "Exxon Nuclear Plant Transient Method for Boilin. ater Reactors."

9. EMF-1997 A, "ANFB-10 Critical Power Correlatio C 10. Caldon, Inc.,"TO CAL REPORT: Improving rmalPower Accuracy and Plant Safet Increasing Operating wer Level Using the LEFM/Th System," En eeing Report-8
11. Caldon, Inc., "Supplement toopical eport ER-80P: Basis for a Power Uprate with the LEFM/ EFM CheckPlusm System,",

Engineering Report ER -160P.

12. EMF-85-74(P)(A), "ROD (BWR) I Rod Thermal-Mechanical Evaluation Model."
13. EMF-CC-074 (P)(A, olume 4, "BWR StabilityA lysis: Assessment of STAIF with Int from MICROBURN-B2."
14. EMF-2158 A), "Siemens Power Corporation Methodo for Boiling Water R ctors: Evaluation and Validation of CAS IMlCROBURN-B2."

s15. DO32465-A, BWROG Reactor Core Stability Detect and Supprs olutions Uicensing Basis Methodology for Reload Applications."

C. The core p t a a i (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 supplerments, shall be provided upon issuance for each reload cycleto the NRC.

(continued)

SUSQUEHANNA - UNIT I TS / 5.0-23 Amendment 8A 18p, 10, 1,X4, ?6Q,,915 2 f6, 217

I r INSERT 1:

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. EMF-2361 (P)(A), "EXEM BWR-2000 ECCS Evaluation Model," Framatome ANP.
8. EMF-2292(P)(A), "ATRIUM'-10: Appendix K Spray Heat Transfer Coefficients,"

Siemens Power Corporation.

9. XN-NF-84-105(P)(A), 'XCOBRA-T: A Computer Code for BWR Transient Thermal-Hydraulic Core Analysis," Exxon Nuclear Company.
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.
14. EMF-CC-074(P)(A), "BWR Stability Analysis - Assessment of STAIF with Input from MICROBURN-B2," Siemens Power Corporation.
15. NE-092-001 A, -Licensing Topical Report for Power Uprate With Increased Core Flow," Pennsylvania Power & Light Company.
16. Caldon, Inc., "TOPICAL REPORT: Improving Thermal Power Accuracy and Plant Safety while Increasing Operating Power Level Using the LEFTM@M System,"

Engineering Report - 80P.

IC 7

17. Caldon, Inc., "Supplement to Topical Report ER-80P: Basis for a Power Uprate with the LEFMV41 or LEFM CheckPlusSystem," Engineering Report ER-160P.
18. NEDO-32465-A, UBWROG Reactor Core Stability Detect and Suppress Solutions Licensing Basis Methodology for Reload Applications."

tb Attachment 2 to PLA-5990 List of Regulatory Commitments (Unit 1)

Attachment 2 to PLA-5990 Page 1 of 1 LIST OF REGULATORY COMMITMENTS REGULATORY COMMITMENTS Due Date/Event There are no new commitments associated with this submittal. NA to PLA-5990 Changes to TS Bases Pages (Markup)

(Unit 1)

If N PPL Rev. I Reactor Core SLs B 2.1.1 a BASES BACKGROUND Operation above the boundary of the nucleate boiling regime (continued) could result in excessive cladding temperature because of the onset of transition boiling and the resultant sharp reduction in heat transfer coefficient Inside the steam film, high cladding temperatures are reached, and a cladding water (zirconium water) reaction may take place. This chemical reaction results in oxidation of the fuel cladding to a structurally weaker form. This weaker form may lose its integrity, resulting in an uncontrolled release of activity to the reactor coolant.

APPLICABLE The fuel cladding must not sustain damage as a result of normal SAFETY ANALYSES 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 Systenm setpoints ( CO 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 S limit 2.1;1.1 Fuel Cladding Inteqrgvd.

S-Pc )The use of the ANFB (Reference 4) correlation is valid for critical power calculations at pressures A-57tpsia and bundle mass fluxes

>-r-1tx 10 Ib/hr-ft2. For operatio at low pressures or low flows, the

__---gel cladding integrity SL is establishd by a limiting condition on core THERMAL POWER, with the foll ng 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. I (continued)

SUSQUEHANNA - UNIT I TS /B 2.0-2 Revision 2'

f PPL Rev. 1 Reactor Core SLs B 2.1.1 BASES APPLICABLE 2.1.1.1 Fuel Cladding Inteartv (continued)

SAFETY ANALYSES For the SPC ATRIUM-10 design, the minimum bundle flow is >

28 x 103 lb/hr. For the ATRIUM-10 fuel design, the coolant minimum bundle flow and maximum area are such that the mass flux is always > .25 x 106 Ib/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-fe 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.

2.1.1.2 MCPR 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 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.

i~~AF- c critical power correlation is based on a significant body of GTe practical test data. As long as the core pressure and flow are within the range of validity of the correlations (referto 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 herent accuracv of rlation provide a reasonable degree of assurance that during sustained operation at the MCPR SL there would be no transition boiling in the core.

(continued)

NTIT/203Rvso SUSQUEHANNA~~

- ~ ~ ~

SUSQUEHANNA - UNIT I TS/ B2.0-3 R~evision 3

A PPL Rev. 1 Reactor Core SLs B 2.1.1 p BASES APPLICABLE 2.1.1.2 MCPR (continued)

SAFETY ANALYSES If boiling transition were to occur, there is reason to believe that the integrity of the fuel would not be compromised.

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 ANFE 1Critical Power I 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 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 i and 2 the reactor vesser 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 The reactor vessel water level SL has been established at the top of the active irradiated fuel to provide a point that can be S (continued)

NTIT/2O4Rvso SUSQUEHANNA~~ ~ -

SUSQUEHANNA - UNIT I TS / B2.0-4 Revision 2

PPL Rev. 1 Reactor Core SLs B 2.1.1 p BASES APPLICABLE 2.1.1.3 Reactor Vessel Water Level (continued)

SAFETY ANALYSES 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 dad 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.

2. ANF 524 (P)(A), Revision 2, "Critical Power Methodology for Boiling Water Reactors," Supplement I Revision 2 and Supplement 2, November 1990.
3. 10 CFR 100.
4. EMF-197 (9)(A),Re9v1ion 0, !A4FB-10 Critical Power

..L1JsE.1T TS132.1. 1-I Carelafien," July 10998 and EMF 1907 (P)(A) Supplement 1 Revision 0, "ANFB-1B Critical Power Correlation: High LoeaI-aking Results," July 1498.

5. EMF-2158(P)(A), Revision 0, "Siemens Power Corporation Methodology for Boiling Water Reactors: Evaluation and Validation of CASMO-4/Micobum-B2, October 1999.

. SUSQUEHANNA

-UNIT I TS I 2.0 (continued)

Reiio SUSQUEHANNA - UNIT I TS/ B2.0-5 Revision 2

SDM B 3.1.1 S 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 SR 3.1.1.1 REQUIREMENTS 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 core 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 9

the most reactive point in the cycle), no correction to the BOC measured value is required (Ref.?>). For the SUM demonstrations that rely solely on calculation of the highest worth control rod, additional margin (0.10% Wk/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 1 8 3.1-5 Revision 0

.r r PPL Rev. 0 SDM B 3.1.1 BASES SURVEILLANCE SR 3.1.1.1 (continued)

- REQUIREMENTS 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 the core) is required to ensure adequate SDM is maintained during p refueling. This evaluation ensures that the intermediate 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 1. 10 CFR 50, Appendix A, GDC 26.

2. FSAR, Section 15.
3. PL-NF-00 -A, *4plieatien of Rea Methods foDvBWR n T at 3.. - I

~N SE Imsevc TS83-1.1-1 4

_ _ _ - seigand_An

_ _ _ _ _ _-_ys*_-*evin

_ _ _ mavnd

_ 2_

_ _ _ 2.w

_ _ _ _ _ . ._.y

_ Supemn,!A Atigust 1095, Supplerncnt 2 A,jul' 10096 and Supplement a A,_ .. _,

March 2001.

4. FSAR, Section 15.4.1.1.

SUSQUEHANNA - UNIT 1 TS / B3.1-6 Revision I

I PPL Rev. 0 SDM B 3.1.1 S

BASES REFERENCES 5. Final Policy Statement on Technical Specifications Improvements, (continued) July 22, 1993 (58 FR 39132).

6. FSAR, Section 4.3.

')fL reh -- 1 * -0g

' a A, A.plri Ad Ie Retio t) f -LereW g . BWR Design and Analysis," Seetiwi 24.,

190, A.nJulyis eupdsc or July 19X2, Supplemnent 1-A.

at- August 1996, Supplomei# 2 A, July 1896, and Supplement 3 A, 1vrp (i SUSQUEHANNA - UNIT 1 TS /B 3.1-7 Revision 1

PPL Rev. 1 Control Rod Scram limes B3.1.4 BASES REFERENCES (continued) 4. FSAR, Section 15.0.

\ _ .. ...

5.rL NF 000 AApplieability of Rcector Analysis Metho~ frd W

- 5Design an nlsi, ;eebn 4.1.2, July 1992, anid S~upplement 1-A, August 1Me5, Supplzmcnt 2-A, July 1096, and Supplement SA, FsRVR, 5 FCYXOA1 .5'4.7 -I 4a-4 8i

6. Final Policy Statement on Technical Specifications Improvements, July 22, 1993 (58 FR 39132).
7. Letter from R. F. Janecek (BWROG) to R.W. Starostecki (NRC),

"BWR Owners Group Revised Reactivity Control System Technical Specifications," BWROG-8754, September 17, 1987.

0 SUSQUEHANNA - UNIT 1 TS / 8 3.1-28 Revision 1

PPL Rev. 1 Rod Pattern Control B 3.1.6 0 BASES ACTIONS B.1 and B.2 (continued) 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 considerirg 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 10% RTP.

REFERENCES 1. FPL 001 A, "Application of Raoctor Analysis Mcthods for UF0-

,_>,-BR Besign end AnalysisN CSeeien 2.8, July 1002, Eupplen . .

sEmr 1 A 0n A3.1l-a dS )

0 2. "Modifications to the Requirements for Control Rod Drop Accident Mitigating System," BWR Owners Group, July 1986.

3. NUREG-0979, Section 4.2.1.3.2, April 1983.
4. NUREG-0800, Section 15.4.9, Revision 2, July 1981.
5. 10CFR 100.11.
6. NEDO-21778-A, "Transient Pressure Rises Affected Fracture Toughness Requirements for Boiling Water Reactors,"

December 1978.

7. ASME, Boiler and Pressure Vessel Code.
8. Final Policy Statement on Technical Specifications Improvements, July 22, 1993 (58 FR 39132).
9. NEDO 33091-A, Revision 2, "Improved BPWS Control Rod Insertion Process," April 2003. I 0

SUSQUEHANNA- UNIT 1 TS / B3.1-38 Revision 1

I PPL Rev. 0 APLHGR B 3.2.1 BASES ACTIONS S (continued)

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 c 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 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, 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 a 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.

2. Not used. I
3. AF 0 04~P)(), Adyeonced Nueleef Fuels Gerporatiz

~ Mchocl~c~ fcrDofing MFMer- Reeeters EXEM 8WR EYkatc

.LNS ERZ Tsg3 . 1- I

4. ANF-CC-33(P)(A) Supplement 2, "HUXY: A Generalized Multirod Heatup Code with 10CFR50 Appendix K Heatup Option,"

January 1991.

5. XN-CC-33(P)(A) Revision 1, "HUXY: A Generalized Multirod Heatup Code with 10CFR50 Appendix K Heatup Option Users Manual," November 1975.

(continued)

SUSQUEHANNA- UNIT I TS / B3.2-3 Revision I

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 dpmage 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 SAFETY ANALYSES the AOOs to establish the operating limit MCPR are presented erences 2 through X. To ensure that the MCPR SL is I

-not exceed 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 S (continued)

SUSQUEHANNA - UNIT 1 TS / B 3.2-5 ' Revisi-on 1

  • -A MCPR B 3.2.2 BASES D

APPLICABLE state to ensure adherence to fuel design limits during SAFETY ANALYSES the worst transient that occurs with moderate frequency.

(continued) These analyses may also consider other combinations of plant conditions (i.e.. control rod scram speed. bypass valve performance. EOC-RPT. cycle exposure. etc.). Flow dependent MCPR limits ns_.QfIjl>N_ ow runout transients.uin th ehR~lg nl;. 2-f Refcet-c I The MCPR satisfies Criterion 2 of the NRC Policy Statement I (Ref. X LCO The MCPR operating limits specified in the COLR are'the result of the Design Basis Accident (OBA) 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 that 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 (continued) s SUSQUEHANNA - UNIT 1 TS / B 3.2-6 Revision 1

PPL Rev. 0 MCPR B 3.2.2

REQUIREMENTS 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 distnbution 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 MCPR must be calculated prior to exceeding 50% RTP.

SR 3.22.2 cause the transient analysis takes credit for conservatism in the scm tim rformrance, it must be demonstrated that the specific scram is consiste those used Inthe transient analysis. SR 3.2.2 etemines I1JSERY TS 3..2-S the scram tim ction which isa measure of the actual time compared with theymed scram time. The COL ntains a table of scram time fractions ba4n te ICC 3.14 rol Rod Scram Times \

and the realistic scram times in the tament analysis. The MCPR operating limit is then determined b on an Interpolation between the applicable limits for scram times CO .O4,sControl. Rod Scram Times I

-and realistic scram time an s using the. S timi fraction. The scram time fraction an rresponding MCPR oper limit must be determined once in 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 e tests required by .1.4.1, SR 3.1.4.2, SR 3.1.4.3 and SR 3.1.4 te tuse leffectiv ram 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 /> eion Time is acceptable due to the relatively minor changes in the cram time fraction expected during the fuel cycle".

REFERENCES 1. NUREG-0562, June 1979.

2. N-NF-96%-Gel -A, W~ppleatain of Reactor Analysis Method fo 111 BWR Design and Analysis," july 19X2, Supplement 1 A, I)sEvr TS%3.L.A-) I t -August 1995, S3upplement 2 A, July 1 996, and Supplemet a -A,-

(continued)

S.

SUSQUEHANNA - UNIT 1 TS / B3.2-8 Revision I

7 MCPR B 3.2.2 S

dgh 5

BASES REFERENCES 3PELNF 37 001 A, "Qualification of,Steady State Cu-(continuued) rPhysics Metods for BWR Design and Adalysis".

rollvAprl 2 L, t9tu.

L-2 4. PL NF 89 0O5 A, "Qualification of Tranzicnt Analyui -

INsEiur TS5B3.2'L lMethods for BWR Dcsign and Analysis." July 1992.

includinmg Sup Iemcets44 '1'and 4.2, 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 1997 (P)EA) Re-vision 0. ANF8 lo Critical Poster '11 Tl E T I 31 3 I'-3 eui-leitiW1n." July 1993. alid EMF-1997 (P)(A)

TSupplemont 1 Revision 0, "ANFB 10 Critical Power 8.

Corrohtior 1 High Local

. Peaking Results." July 1990.

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

Water Reactors." March 1986.

9. XN-NF-84-105(P)(A). Volume 1 and Volume 1 Sup lements I 1 and 2, "XCOBRA-T: A Computer Code for 8WR Transient Thermal-Hydraulic Core Analysis," February 1987.

Final Policy Statement on Technical Specifications

'Improvements. July 22. 1993 (58 FR 39132).

TImseltT TSz93.i.?-' I SUSQUEHANNA - UNIT 1 TS / B 3.2-9 Revision 1

I PPL Rev. 0 S 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. FSAR 15.4.9 INSERT s

D. SUSQUEHANNA - UNIT 1 TS / B 3.10-32 Revision 1

PPL Rev. 0 W SDM Test-Refueling B 3.10.8 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 verifies I 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 veiffication 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. 4NF 00c0o01 A, "Applicatien flaeaeter Amnalysi Methe fr-e Dl.F.lR E!igniand Analysis," July 1992 and SuppemeFnt 1 A, August 1499, XSupplemert 2 A, du..1..9, and SupplemerA 3A, March 2001.

jxs T-1 T583.1.8-1 SUSQUEHANNA - UNIT 1 TS / B 3.10-38 Revision I

INSERT TSB2.1.1-1:

4. EMF-2209(P)(A), Revision 1, "SPCB Critical Power Correlation," Siemens Power Corporation, July 2000.

INSERT TSB3.1.1-1:

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 TSB3.1.6-1:

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 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.

INSERT TSB 3.2.2-2

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.

INSERT TSB 3.2.2-3

7. EMF-2209(P)(A), Revision 1, "SPCB Critical Power Correlation," Siemens Power Corporation, July 2000.

INSERT TSB 3.2.2-4

10. ANF-1358(P)(A) Revision 1, "The Loss of Feedwater Heating Transient in Boiling Water Reactors," Advanced Nuclear Fuels Corporation, September 1992.

INSERT TSB 3.2.2-5 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 Attachment 4 to PLA-5990 Preliminary Unit 1 Cycle 15 Core Composition

I Attachment 4 to PLA-5990 Page 1 of 1 Preliminary Unit 1 Cycle 15 Core Composition FANP ATRIUWM T -10 Fresh 284 FANP ATRIUMTm-10 Once-burned 280 FANPATRIUMTm-10 Twice-burned 200

I I Attachment 5 to PLA-5990 Preliminary Unit 1 Cycle 15 Core Loading Pattern

PRELIMINARY SUSQUEHANNA UNIT 1 CYCLE 15 CORE LOADING PATTERN 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 233 24 25 26 27 28 29 30 J 61 1 U . I I I I-1I I roo I- -4 4 -4 I I I I I I I _ _ - _ .

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II I I I I I t TI I I ITI I I IIII Tl I I II I I I 00 02 04 06 08 10 12 14 16 18 20 22 24 26 28 30 32 34 36 38 40 42 44 46 48 50 52 54 56 5860 GE x

I3SQA-12 12GdV1Gd7 TWICE BURNED (3.97) MSUS1-15 15GdZ FRESH (4.11) rSQA-12 13Gd6/12GdZ TWICE BURNED (3.75) [I!SUS1-15 14GdZ FRESH (3.91)

[MSQA-13 14GdZ ONCE BURNED (4.12) [ SUS1-15 16GdZ FRESH (4.09)

[BSQA-13 14GdZ ONCE BURNED (3.90) 11/15=205 h:Wog'dw~clS'ulclS oorenmp rmvO.2ds to PLA-5990 Preliminary Unit 1 Cycle 15 Fresh Fuel Description

Preliminary Assembly Type 60 Reload Bundle Description (ATRIUM-10, 100mil Channel)

Bundle Average Enrichment = 3.91%

TAF - Nat Lattice 149.45" End Type 134.

4.09 W/ 138"-

U-235 Lattice I3Gd6 Type 133 126" Lattice Type 132 96" Lattice Type 131 6"

Lattice TyDe 130 BAF - V" G:%NFENFEPKGW1C15VNFE115001-R(ANQPlots'IBUNOLE6O.PPT

Preliminary Assembly Type 61 Reload Bundle Description (ATRIUM-10, 100mil Channel)

Bundle Average Enrichment = 4.09%

TAF- 149.45" Lattice Type 134 138" Lattice Type 138 126" Lattice Type 137 96" Lattice Type 136 78" Lattice Type 135 Lattice 6of Type 130 BAF- off G.NNFEWNFEPKGWI1CI5\NFE1 15001_ROWNQPlotrsSBUNDLE61.PPT

i Preliminary Assembly Type 62 Reload Bundle Description (ATRIUM-10, 100mil Channel)

Bundle Average Enrichment = 4.11%

TAF-C l149.45" TAF Nat Lattice End Type 134 4M Wt 138" U-235 Latc 1G5 Type'142 126" 4.45 wlo U-235 Lattice 15Gd8 Type 141 4.34 w1o 96 U-235 Lattice 12Gd6S/ Type 140 3Gd5 78" 4.64 W/Io Lattice U-235 ye3 15Gd8 Tp 3 Nat Lattice 6" A End Type 130 0" BAF- I I G.\NFElNFEPKGU1Ci6WFE115001_RO\40PIoSts'UNDLE62.PPT

Attachment 7 to PLA-5990 Listing of Framatome-ANP's Approved Methodology and Applicable LCOs

BWR Approved Topical Reports for Susquehanna Attachment Nuclear Plant Technical Specifications COLR References Page A-1 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 Therrnal-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 1(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 afission 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 aset 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 1and 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-80-1 9(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 CASMO-4.

Validation of CASMO-4/MICROBURN-B2, Siemens 3.2.3 Power Corporation, October 1999. 3.3.2.1 Table 3.3.2.1-1 J0104X

BWR Approved Topical Reports for Susquehanna Attachment Nuclear Plant Technical Specifications COLR References Page A-2 Report Applicable LCO Methodology / Justification XN-NF-80-1 9(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 10CFR 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, ATRIUM'-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-80-19(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-84-1 05(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 1 and 2, ANF 3.2.2 Provides a methodology for the determination of thermaF margins, Critical Power Methodology for Boiling Water Reactors, specifically the MCPR safety limit.

Advanced Nuclear Fuels Corporation, November 1990.

ANF-913(P)(A) Volume 1 Revision 1 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.

13WR Approved Topical Reports for Susquehanna Attachment Nuclear Plant Technical Specifications COLR References Page A-3 Report Applicable LCO Methodology / Justification ANF-1 358(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 ATRIUM'M-10*

Siemens Power Corporation, July 2000. fuel designs. This correlation isused 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 STAIF with Input from MICROBURN-B2, Siemens Power Corporation, August 2000.

Attachment 8 to PLA-5990 MCPR Safety Limit Methodology

j i MCPR SafetY Limit Methodolocv fl - - - - - - - - - -

ANF-524 (PXA), Rev. 2 & Supplement 1,Rev. 2 EMF-2209(PXA), Rev. 1 EMF-2158 (PXA), Rev. 0 EMF-2158 (P) (A), Rev. 0 I

I I

I I

I I

I L- - -_ - - _ _ - - -_ - _ _ _ - - J g:\nfe~techspec.submrnls\u2c13dig2.rovised.mcpr.safeJimLt.method.ppt 01/20/2005

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