LIC-04-0097, License Amendment Request (Lar), Core Operating Limits Report and Application of CASMO-4 Methodology.

License Amendment Request (Lar), Core Operating Limits Report and Application of CASMO-4 Methodology.

ML042530468
Person / Time
Site:	Fort Calhoun
Issue date:	09/07/2004
From:	Ridenoure R Omaha Public Power District
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
LIC-04-0097
Download: ML042530468 (88)

Text

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Omaha Public Power District 444 South 16th Street Mall Omaha NE 68102-2247 LIC-04-0097 September 7,2004 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001

REFERENCES:

1. Docket No. 50-285

2. Letter from NRC (S. Bloom) to OPPD (T. L. Patterson) dated December 16, 1994, Use of CASMO-3/SIMULATE-3 in [OPPD]

Topical Reports (NRC-94-346)

3. Letter from OPPD (R. T. Ridenoure) to NRC (Document Control Desk) dated January 23, 2004, Revision of Fort Calhoun Station Core Reload Methodology Topical Reports (LIC-04-0010)

SUBJECT:

Fort Calhoun Station, Unit No. 1, License Amendment Request (LAR), Core Operating Limits Report and Application of CASMO-4 Methodology In accordance with the provisions of Section 50.90 of Title 10 of the Code of Federal Regulations (10 CFR), the Omaha Public Power District (OPPD) is submitting a request for an amendment to the technical specifications (TSs) and licensing basis for Fort Calhoun Station, Unit No. 1 (FCS). The proposed amendment revises TS 5.9.5, Core Operating Limits Report, such that it will read consistent with Specification 5.6.5 of NUREG- 1432, Standard Technical Specifications-Combustion Engineering Plants. In addition, the list of core reload analysis methodologies contained in TS 5.9.5b used to determine the core operating limits is updated to move many of these references to OPPD core reload analysis methodology documents OPPD-NA-8301, 8302, and 8303. Several analytical method references that are no longer applicable to FCS are deleted from TS 5.9.5b; several references will remain, as they are not suitable for incorporation into the core reload analysis documents.

The changes to OPPD-NA-8301, 8302, and 8303 are primarily administrative in nature to incorporate the references removed from TS 5.9.5b and to remove P broprietary) and A (approved) designations since OPPD no longer considers these documents to be proprietary or topical reports. OPPD-NA-8302 is also revised to incorporate use of the Studsvik computer code CASMO-4 to support core reload analyses. The approved CASMO-3 code (Reference 2) has been used to evaluate core reloads at FCS for over 12 years. The newer CASMO-4 code streamlines the cross-section generation process and maintains current peaking factor uncertainties but does not provide a gain in margin.

Employment with Equal Opportunity

U. S. Nuclear Regulatory Commission LIC-04-0097 Page 2 In Reference 3, OPPD notified the NRC of a previous update to OPPD core reload analysis methodologies OPPD-NA-8301, 8302, and 8303 performed in accordance with Generic Letter 83-11, Supplement 1, Licensee Qualifications for Performing Safely Analyses, June 24, 1999. OPPD originally intended to incorporate the CASMO-4 methodology into OPPD-NA-8302 under the provisions of 10 CFR 50.59, as allowed by section 4.3.8 of NEI 96-07, Rev. 1, Guidelines for 10 CFR 50.59 Implementation, dated November 2000.

NEI 96-07, Rev. 1 was endorsed by the NRC in Regulatory Guide 1.187 Guidance for Implementation of 10 CFR 50.59, Changes, Tests, and Experiments dated November 2000. However, subsequent discussions with NRC staff revealed that a license amendment request (LAR) was the only acceptable means of incorporating CASMO-4 methodology. provides a description of the proposed change. Attachment 2 provides results from the CASMO-4/CASMO-3 benchmarking analysis. Attachment 3 provides the existing TS pages marked-up to show the proposed change. Attachment 4 provides the proposed TS pages. The aforementioned OPPD core reload analysis documents are submitted for NRC review and approval in Enclosures 1,2, and 3.

To assist the NRC in its review of Enclosures 1, 2, and 3, a draft version of the Core Operating Limits Report (COLR) revised to reflect the changes requested in this submittal is contained in Enclosure 4. The COLR will be finalized to list specific approved methodology documents and submitted to the NRC in accordance with TS 5.9.5 following NRC approval of this amendment request.

OPPD requests approval of the proposed license amendment by January 3 1,2005, with a 60-day implementation period. No commitments are made to the NRC in this letter.

In accordance with 10 CFR 50.91, a copy of this application, with attachments, is being provided to the designated State of Nebraska Official.

I declare under penalty of perjury that the foregoing is true and correct. (Executed on September 7,2004).

If you should have any questions regarding this submittal, please contact Tom Matthews at (402) 533-6938.

i"i" ice resihent

U. S. Nuclear Regulatory Commission LIC-04-0097 Page 3 Attachments: 1. Fort Calhoun Stations Evaluation

2. Markup of Technical Specification Pages

3. Proposed Technical Specification Pages

Enclosures:

1. OPPD-NA-8301, Reload Core Analysis Methodology Overview

2. OPPD-NA-8302, Neutronics Design Methods and Verification

3. OPPD-NA-8303, Transient and Accident Methods and Verification

4. TDB-VI, Core Operating Limits Report cc: Division Administrator - Public Health Assurance, State of Nebraska

LIC-04-0097 Page 1 Fort Calhoun Station's Evaluation For Amendment of Operating License 1 .O INTRODUCTION

2.0 DESCRIPTION

OF PROPOSED AMENDMENT

3.0 BACKGROUND

4.0 REGULATORY REQUIREMENTS AND GUIDANCE

6.0 REGULATORY ANALYSIS

7.0 NO SIGNIFICANT HAZARDS CONSIDERATION 8.0 ENVIRONMENTAL EVALUATION 9.0 PRECEDENT

10.0 REFERENCES

LIC-04-0097 Page 2 1.O INTRODUCTION As described below, the Omaha Public Power District (OPPD) proposes to revise Fort Calhoun Station, Unit No. 1 (FCS), Technical Specification (TS) 5.9.5, Core Operating Limits Report to be consistent with Specification 5.6.5 of NUREG-1432, Standard Technical Spectfications Combustion Engineering Plants. OPPD is also updating the TS 5.9.5b list of analytical methods used to determine core-operating limits by relocating many of these references to OPPD core reload analysis methodology documents OPPD-NA-8301, Core Analysis Methodology Overview (Reference 10.l), OPPD-NA-8302 Reload Core Analysis Methodology, Neutronics Design Methods and Verification (Reference 10.2), and OPPD-NA-8303 Reload Core Analysis Methodology, Transient and Accident Methods and Verification (Reference 10.3). Several analytical methods referenced in TS 5.9.513 that are no longer applicable to FCS are deleted.

The currently approved revisions of OPPD core reload analysis methodologies OPPD-NA-8301-P-A, OPPD-NA-8302-P-A, and OPPD-NA-8303-P-A use the Studsvik Scandpower codes MICBURN-3 (M-3) and CASMO-3 (C-3) for cross-section generation and SIMULATE-3 (S-3) for core simulation. The C-3 and S-3 codes have been used to evaluate core reloads at FCS for over twelve (12) years. OPPD proposes to allow the use of CASMO-4 (C-4) to support core reload analyses. The C-4 code streamlines the cross-section generation process while maintaining current peaking factor uncertainties. C-4 was benchmarked against C-3 and it was determined that C-4 does not reduce previously approved safety margins. In addition, the P and A designations, which stand for proprietary and approved respectively, are deleted from OPPD-NA-8301, 8302, and 8303, as the documents are no longer considered to be proprietary or topical reports.

OPPD core reload analysis methodologies OPPD-NA-8301, 8302, and 8303 are enclosed for NRC review and approval.

2.0 DESCRIPTION

OF PROPOSED AMENDMENT The description of the changes proposed by this amendment will be described in three general areas, which are: (1) TS 5.9.5 revisions necessary for consistency with Standard Technical Specifications (STS), 2) TS 5.9.5b analytical methods revisions, and 3) OPPD core reload analysis methodology revisions and use of C-4.

TS 5.9.5 Revisions Necessary for Consistency with Standard Technical Specifications The following changes are proposed to make FCS TS 5.9.5 consistent with Specification 5.6.5 of NUREG-1432.

1. The title of FCS TS 5.9.5, Core Operating Limit Report is changed to upper case followed by the abbreviation (COLR) to match Specification 5.6.5 of NUREG-1432.

2. FCS TS 5.9.5a currently states:

LIC-04-0097 Page 3 Core Operating Limits shall be established and documented in the Core Operating Limits Report (COLR) before each reload cycle or any remaining part of a reload cycle.

Specification 5.6.5a of NUREG-1432 states:

Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:

[ The individual specifications that address core operating limits must be referenced here. 1 Accordingly, FCS TS 5.9.5a is revised to be consistent with Specification 5.6.5a of NUREG-1432. The individual FCS TSs addressing core operating limits i.e.,

1.3(4), 1.3(8), 2.2.7, 2.2.8, 2.10.2, and 2.10.4 are listed as required by NUREG-1432.

FCS TS 5.9.5b currently states:

The analytical methods used to determine core operating limits shall be those previously reviewed and approved by the NRC as follows:

Specification 5.6.5b of NUREG-1432 states:

The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

Accordingly, FCS TS 5.9.5b is revised to be consistent with Specification 5.6.5b of NUREG- 1432.

3.0 The first sentence of FCS TS 5.9% currently states:

The core operating limits shall be determined so that all applicable limits of the safety analysis are met.

Specification 5 . 6 . 5 ~of NUREG-1432 states:

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

LIC-04-0097 Page 4 Accordingly, the first sentence of FCS TS 5.9% is revised to be nearly identical to Specification 5 . 6 . 5 ~of NUREG-1432. The only difference is that the term shutdown margin is incorporated followed by the abbreviation SDM, which is bracketed.

4.0 The second sentence of FCS TS 5 . 9 . 5 ~currently states:

The Core Operating Limits Report, including any mid-cycle revisions or supplements thereto, shall be provided upon issuance, for each reload cycle, to the NRC Document Control Desk with copies to the Region lV Administrator and Senior Resident Inspector.

Specification 5.6.5d of NUREG-1432 states:

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

The second sentence of FCS TS 5.952 will be designated FCS TS 5.9.5d and will be consistent with Specification 5.6.5d of NUREG-1432.

Thus, the changes listed above are administrative in nature to achieve consistency with STS.

TS 5.9.513 Analytical Methods Revisions

1. References 1, 2, and 3 of TS 5.9.5b are revised to remove the P and the A, which stand for proprietary and approved respectively. OPPD has reviewed these references and determined they no longer contain proprietary information nor is the A designation appropriate because these documents are not topical reports.

2. References 5, 11, 12, 15, 16, 17, 18, 21, and 23 are deleted from TS 5.9.5b and are relocated to OPPD-NA-8303, (Reference 10.3).

3. Reference 13, EMF-l961(P)(A) is deleted from TS 5.9.5b and relocated to OPPD-NA-8301, (Reference 10.1).

4. Reference 14, XN-NF-62 1(P)(A) is deleted. The reference applies to the departure from nucleate boiling (DNB) correlation for bi-metallic fuel. All of the fuel in the FCS core for Cycle 23 is high thermal performance (HTP) fuel provided by Framatome A.

5. Reference 19, ANF-84-73 Appendix B (P)(A) is deleted. Reference 19 is an old non-loss-of-coolant accident (LOCA) methodology that is no longer used at FCS.

LIC-04-0097 Page 5

6. Reference 20, EMF-84-093(P)(A) is deleted. The steam line break methodology is incorporated in EMF-23 1O(P)(A), which is relocated to OPPD-NA-8303 (Reference 10.3).

7. Reference 22, EMF-96-029(P)(A) is deleted from TS 5.9.5b and relocated to OPPD-NA-8302, (Reference 10.2).

8. The remaining references contained in TS 5.9.5b are renumbered to reflect the deletions noted above.

Thus, the changes listed above are administrative in nature to remove the proprietary and approved designations from OPPD core reload analysis methodologies, consolidate several non-OPPD core reload analysis methodologies by reference into OPPDs core reload analysis methodologies, and delete methodologies no longer applicable to FCS.

OPPD Core Reload Analysis Methodology Revisions and use of CASMO-4 OPPD core reload analysis methodologies OPPD-NA-8301, 8302, and 8303 are revised to remove the P and the A designations. These characters designated the documents as proprietary and approved respectively. OPPD has reviewed these documents and determined they do not contain proprietary information and that the A designation is unnecessary. As described above, many core reload analysis methodology references are also being relocated from TS 5.9.5b to OPPD-NA-8303, OPPD-NA-8302, or OPPD-NA-8301.

OPPD is also requesting NRC approval to utilize the CASMO-4 (C-4) computer code for the Cycle 23 core reload analysis to streamline the process of generating cross-sections and to reduce the potential for human error. OPPD does not gain margin with C-4 as shown below in Section 5.0. OPPD core reload analysis methodology document OPPD-NA-8302 is revised to incorporate the C-4 computer code.

3.0 BACKGROUND

The changes made to TS 5.9.5 are administrative in nature to achieve consistency with Standard Technical Specifications and update the list of references in TS 5.9.5b primarily by relocating many of them to OPPD core reload methodology documents.

OPPD no longer considers the enclosed core reload methodology documents to be proprietary and consequently, the P has been removed from them. The A designating NRC approval is removed as well based on the understanding that this designation normally applies to generic topical reports, which OPPDs core reload methodology documents are not.

LIC-04-0097 Page 6 Studsvik has developed a new cross-section generation code called CASMO-4 (C-4) since the initial development of OPPDs core reload methodologies. C-4, like C-3, is a two-dimensional neutron transport theory lattice physics code with depletion capability and the ability to generate cross-sections and discontinuity factors for boiling water reactor (BWR) and pressurized water reactor (PWR) diffusion theory core analysis.

OPPD will use the Studsvik C-4 code for cross-section generation for the Cycle 23 core reload analysis. The cross-sections generated by C-4 will be used in the S-3 model to generate physics parameters used for dispositioning or reanalyzing transient analyses.

C-4 incorporates the microscopic depletion of burnable absorbers into the main calculations, which replaces the M-3 auxiliary code. A new feature of C-4 is the use of the characteristics form for solving the transport equation, which allows for a true heterogeneous geometry in the two-dimensional calculation. Studsvik has also automated the cross-section generation process with C-4 so that complete nuclear data for S-3 can be generated in one execution, which results in reduced potential for human error in the model development process.

4.0 REGULATORY REQUIREMENTS AND GUIDANCE OPPD originally intended to incorporate the C-4 methodology into OPPD-NA-8302 under the provisions of 10 CFR 50.59, as allowed by section 4.3.8 of NEI 96-07, Rev. 1, Guidelinesfor 10 CFR 50.59 Implementation, dated November 2000. NEI 96-07, Rev. 1 was endorsed by the NRC in Regulatory Guide 1.187 Guidancefor Implementation of 10 CFR 50.59, Changes, Tests, and Experiments dated November 2000. However, subsequent discussions with NRC staff revealed that a license amendment request (LAR) was the only acceptable means of incorporating C-4 methodology.

The proposed technical specification changes are primarily administrative in nature to achieve consistency with Standard Technical Specifications or to consolidate the list of NRC approved analytical methods of TS 5.9.5b into OPPD core reload analysis documents. TS 5.9.5b requires NRC approval of the analytical methods used to determine core operating limits. As such, the OPPD core reload methodology documents, which are several of the analytical methods referenced in TS 5.9.5b are enclosed for NRC approval. The OPPD core reload methodology documents were revised to incorporate references removed from TS 5.9.5b, delete characters designating the documents as proprietary/approved, and incorporate a description of the C-4 computer code used for Cycle 23 core reload analysis.

FCS was licensed for construction prior to May 21, 1971, and at that time committed to the draft General Design Criteria (GDC), which are described in Appendix G of the Updated Safety Analysis Report (USAR).

In regards to the adoption of the C-4 computer code, this activity complies with FCS Design Criterion 6, Reactor Core Design, which is similar to 10 CFR 50, Appendix A, GDC 10, Reactor Design. FCS Design Criterion 6 states that the reactor core shall be designed to function throughout its design lifetime without exceeding acceptable fuel

LIC-04-0097 Page 7 damage limits, which have been stipulated and justified. The core design together with reliable process and decay heat removal systems, shall provide for this capability under all expected conditions of normal operation with appropriate margins for uncertainties and for transient situations which can be anticipated, including the effects of the loss of power to recirculation pumps, tripping out of a turbine generator set, isolation of the reactor from its primary heat sink, and loss of all off-site power.

This activity also complies with FCS Design Criterion 7, Suppression Of Power Oscillations, which is similar to 10 CFR 50, Appendix A, GDC 12, Suppression of Reactor Power Oscillations. FCS Design Criterion 7 states the core design, together with reliable controls, shall ensure that power oscillations which could cause damage in excess of acceptable fuel damage limits are not possible or can be readily suppressed.

This activity also complies with FCS Design Criterion 8, Overall Power Coefficient, which is similar to 10 CFR 50, Appendix A, GDC 11, Reactor Inherent Protection. FCS Design Criterion 8 states the reactor shall be designed so that the overall power coefficient in the power operating range shall not be positive.

5.0 TECHNICAL ANALYSIS

The changes proposed for TS 5.9.5 and the revisions to the OPPD core reload analysis methodologies are primarily administrative in nature. Section 2.0 provides a description of each proposed change and a justification as to why the change is considered administrative.

Regarding the adoption of the C-4 computer code for core reload analysis, OPPD has performed Cold Criticals benchmarking with C-4 (Reference 10.4) as well as a benchmarking against plant-specific data calculated with C-3 (Reference 10.5). This was done to fulfill the requirements of Generic Letter 83-1 1, Supplement 1 (Reference 10.7).

The benchmarking analysis (Reference 10.5) compared C-4/S-3 calculations with C-3/S-3 calculations. The following parameters were examined for Cycles 19 through 22:

critical boron concentration, control element assembly (CEA) worth, moderator temperature coefficient (MTC), and peaking factors (Fr and F,).

The results of the benchmarking analysis are shown in Attachment 2, Tables 1 through

11. These tables also include measured values. The critical boron concentrations, CEA worths, and MTCs generated with the C-4/S-3 model show good general agreement with those calculated with the C-3/S-3 model. Peaking factors Fr and F, were, on average, slightly more conservative with the C-4/S-3 model. These results demonstrate that no margin is gained by using C-4. Also, the peaking factor uncertainties are not changing with C-4. Thus, the newer version of CASMO does not reduce previously approved margins.

The same biases were used for the two models in generating the hot full power (HFP) critical boron concentrations and peaking factors shown in Attachment 2, Tables 4 through 11. OPPD has generated biases for the C-4/S-3 model (Reference 10.6), which

LIC-04-0097 Page 8 are equivalent to those currently used with the C-3B-3 model. These biases ensure equally accurate results. These new biases were used in the C-4/S-3 model in generating the CEA worths, hot zero power (HZP) critical boron concentrations, and MTCs shown in Attachment 2, Tables 1 through 3.

Calculations performed with C-4h-3 yield similar results as C-3/S-3. On average peaking factors were slightly higher or more conservative with the C-4/S-3 model. The results shown here demonstrate that the C-4/S-3 model is not providing margin gains.

Also, the uncertainties associated with the peaking factors are not changing; thus, C-4 does not reduce previously approved margins.

6.0 REGULATORY ANALYSIS

The changes proposed for TS 5.9.5 and the revisions to the OPPD core reload analysis methodologies are primarily administrative in nature. Section 2.0 provides a description of each proposed change and a justification as to why the change is considered administrative.

OPPD evaluated the C-4 computer code utilizing the guidance of Generic Letter 83-1 1, Supplement 1 (Reference 10.7 ) and determined that calculations performed with C-4 yield similar results to its predecessor C-3. On average peaking factors are slightly higher or more conservative with C-4. C-4 also does not change uncertainties associated with the peaking factors. Thus, C-4 does not reduce the previously approved margin of safety.

The changes do not alter, degrade, or prevent actions described or assumed in any accident analysis. They will not change any assumptions previously made in evaluating radiological consequences or affect any fission product barriers ,nor do they increase any challenges to safety systems. Therefore, the proposed change does not increase or have any impact on the consequences of events described and evaluated in Chapter 14 of the USAR.

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 Commissions 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.

7.0 NO SIGNIFICANT HAZARDS CONSIDERATION Omaha Public Power District (OPPD) has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, Issuance of amendment, as discussed below.

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

LIC-04-0097 Page 9 Response: No.

The proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated. The proposed changes are primarily administrative in nature to achieve consistency with Standard Technical Specifications and to update the list of NRC reviewed and approved analytical methods used to develop core operating limits. Several of the analytical methods are no longer applicable to Fort Calhoun Station, Unit No. 1, (FCS) and thus are deleted from the Technical Specifications (TSs). Many of the topical reports currently referenced in TS 5.9.5b are more suitably referenced in the OPPD core reload methodology documents where they have been relocated.

The OPPD core reload methodology documents remain referenced in TS 5.9.5b and as such are subject to NRC review and approval. The relocation of the topical reports referenced in TS 5.9.5b to OPPD core reload methodology documents is an administrative change. In addition to the incorporation of references currently found in TS 5.9.5b, OPPD core reload methodology documents OPPD-NA-8301, 8302, and 8303 are revised to remove characters designating them as proprietary, and approved. This is an administrative change, as OPPD no longer considers the documents to be proprietary or topical reports. OPPD core reload methodology documents OPPD-NA-8301, 8302, and 8303 are enclosed for NRC review and approval of the changes noted above and incorporation of the CASMO-4 (C-4) computer code, which is described below.

OPPD is adding the C-4 code to OPPD-NA-8302, Reload Core Analysis Methodology, Neutronics Design Methods and Verification and will use the code for nuclear design analysis. This will allow the use of the C-4 and SIMULATE-3 (S-3) methodology to perform all steady-state pressurized water reactor (PWR) core physics analyses. The probability of occurrence of an accident previously evaluated will not be increased by the proposed change in the particular codes used for physics calculations for nuclear design analysis. The results of nuclear design analyses are used as inputs to the analysis of accidents that are evaluated in the Updated Safety Analysis Report (USAR). These inputs do not alter the physical characteristics or modes of operation of any system, structure, or component involved in the initiation of an accident. Thus, there is no significant increase in the probability of an accident previously evaluated as a result of this change.

The consequences of an accident evaluated in the USAR are affected by the value of inputs to the transient safety analysis. An extensive benchmark of C-4 /S-3 predictions was performed with measured data using a variety of fuel designs and operating conditions in power reactors and critical experiments. The accuracy of C-4/S-3 is similar to, and sometimes better than, the accuracy of C-3/S-3.

Furthermore, there is always the potential for the value of the nuclear design

LIC-04-0097 Page 10 design parameters to change solely as a result of the new core reload fuel core loading pattern. Regardless of the source of a change, an assessment is always made of changes to the nuclear design parameters with respect to their effects on the consequences of accidents previously evaluated in the USAR. Refueling is an anticipated activity, which is described in the USAR. If increased consequences are anticipated, compensatory actions are implemented to neutralize any expected increase in consequences. These compensatory actions include, but are not limited to, crediting any existing margins in the analysis or redefining the operating envelope to avoid increased consequences. Thus, the nuclear design parameters are intermediate results and by themselves will not result in an increase in the consequence of an accident evaluated in the USAR.

Therefore, the use of the C-4/S-3 code package, which will perform the same functions as the C-3/S-3 codes with similar accuracy, does not significantly increase the consequences of an accident previously evaluated.

2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated. The proposed changes are primarily administrative in nature. The changes achieve consistency with Standard Technical Specifications, update the list of NRC reviewed and approved analytical methods used to develop core operating limits by deleting certain analytical methods no longer applicable to FCS and relocating many of the remainder to OPPD core reload analysis methodology documents, and make minor administrative changes to OPPD core reload analysis documents referenced in TS 5.9.5b. OPPD intends to utilize the C-4/S-3 code package for nuclear design analysis. The proposed amendment would add the C-4 code to OPPD core reload analysis methodology document OPPD-NA-8302.

The possibility for a new or different kind of accident evaluated previously in the USAR will not be created by the proposed administrative changes or the change to the particular codes used for physics calculations for nuclear design analyses.

The change involves adding the Studsvik C-4 code to OPPD core reload analysis methodology document OPPD-NA-8302. The C-4 code is an update to the C-3 code currently approved for use at FCS. The results of nuclear design analyses are used as inputs to the analysis of accidents that are evaluated in the USAR.

These inputs do not alter the physical characteristics or modes of operation of any system, structure or component involved in the initiation of an accident.

Therefore, these administrative changes and the addition of the C-4 code, which will perform the same functions, as the C-3 code with similar accuracy, does not

LIC-04-0097 Page 11 increase the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the proposed change involve a significant reduction in a margin of safety?

Response: No.

The proposed change does not involve a significant reduction in a margin of safety. The margin of safety as defined in the basis for any technical specification will not be reduced nor increased by the proposed administrative changes or the change to the codes used for physics calculations for nuclear design analyses.

The changes achieve consistency with Standard Technical Specifications, update the list of NRC approved analytical methods used to develop core operating limits by deleting certain analytical methods no longer applicable to FCS and relocating many of the remainder to OPPD core reload analysis methodology documents, and make minor administrative changes to OPPD core reload analysis documents referenced in TS 5.9.5b.

The change involves the addition of the Studsvik C-4 code to OPPD core reload analysis methodologies for nuclear design analysis. Extensive benchmarking of the C-4/S-3 computer codes has demonstrated that the values of those parameters used in the safety analysis are not significantly changed relative to the values obtained using the NRC approved C-3/S-3 computer codes. For any changes in the calculated values that do occur, the application of appropriate biases and uncertainties ensures that the current margin of safety is maintained. Specifically, use of these code specific biases and uncertainties in safety evaluations continues to provide the same statistical assurance that the values of the nuclear parameters used in the safety analysis are conservative with respect to the actual values on at least a 95/95 probability/confidence basis.

Based on the above, Omaha Public Power District concludes that the proposed amendment presents no significant hazards considerations under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of no significant hazards consideration is justified.

8.0 ENVIRONMENTAL EVALUATION Based on the above considerations, the proposed amendment does not involve and will not result in a condition which significantly alters the impact of FCS on the environment.

Thus, the proposed changes meet the eligibility criteria for categorical exclusion set forth in 10 CFR Part 51.22(~)(9),and pursuant to 10 CFR Part 51.22(b), no environmental assessment need be prepared.

LIC-04-0097 Page 12 9.0 PRECEDENT The NRC has approved C-4 for cross-section generation for Arizona Public Service Company (APS) for the Palo Verde power station, for Nuclear Management Company for the Prairie Island power station, and for Dominion Virginia Power for the North Anna and Surry power stations.

10.0 REFERENCES

1. OPPD-NA-8301, Reload Core Analysis Methodology Overview

2. OPPD-NA-8302, Reload Core Analysis Methodology, Neutronics Design Methods and Verification

3. OPPD-NA-8303, Reload Core Analysis Methodology, Transient and Accident Methods and Verijkation

4. OPPD Engineering Analysis EA-FC-02-027, Revision 0, B & W Cold Critical Experiments with CASMO-4

5. OPPD Engineering Analysis EA-FC 146, Revision 0, CASMO-4 Benchmarking

6. OPPD Engineering Analysis EA-FC-04-009, Revision 0, CASMO-4/SIMULATE-3 Bias Calculations

7. NRC Generic Letter 83- 11, Supplement 1, Licensee QualiJicationfor Performing Safety Analyses, June 24, 1999

LIC-04-0097 Page 1 CASMO-4KASMO-3 Benchmarking Analysis

LIC-04-0097 Attachment 2 Page 2 CASMO-4/CASMO-3 Benchmarking Analysis Table 1 CEA Group Worths from Startup Testing 3+4 1.28 1.27 -0.8 1.27 -0.8 Total 5.65 5.66 +0.2 5.63 -0.4 21 A 1.36 1.38 +I .5 1.36 0.0 B** 1.51 1.61 +6.6 1.62 +7.3 1 +3 1.37 1.41 +2.9 1.45 +5.8 2+4 1.38 1.40 +I .4 1.38 0.0 Total 5.62 5.80 +3.2 5.81 +3.4 22 A 1.36 1.38 +I .5 1.37 +0.7 B** 1S O 1.59 +6.0 1.59 +G.O 1+3 1.37 1.40 +2.2 1.41 +2.9 2+4 1.40 1.40 0.0 1.38 -1.4 Total 5.63 5.77 +2.5 5.75 +2.1 I 21 I 1327 I 1312 1 -15 I 1318 I -9 I

• Prediction - Measurement

LIC-04-0097 Attachment 2 Page 3 BOC HFP3 -8.3 -7.4 +0.9 -7.0 +I .3 EOC~ HFP -21.4 -20.6 +0.8 -21.1 +0.3 20 BOCHZP +1.1 +I .O -0.1 +I .4 +0.3 BOC HFP -6.9 -7.6 -0.7 -7.4 -0.5 EOC HFP NOT PERFORMED 21 BOC HZP -2.1 -2.1 0.0 -2.1 0.0 I I BOCHFP I -12.2 I -1 1.7 I +0.5 I -11.9 : 1 -1 I I EOCHFP I -20.3 1 -20.4 I -0.1 I -20.8 1 -0.5 I I 22 I BOCHZP I +1.0 I +0.3 I -0.7 I +0.5

• Prediction - Measurement Table 4 Cvcle 19 HFP Critical Boron Concentrations 2 1020 1014 -6 1004 -16 3 984 98 1 -3 972 -12 4 944 930 -14 922 -22 5 85 1 864 +I3 857 +6 6 778 787 +9 78 1 +3 7 682 70 1 +I9 696 +14 8 593 609 +I6 604 +11 I 9 I 502 I 512 I +lo I 509 I +7 I

• Prediction - Measurement 1 Beginning of Cycle (BOC) 2 Hot Zero Power (HZP) 3 Hot Full Power (HFP) 4 End of Cycle (EOC)

LIC-04-0097 Attachment 2 Page 4 2 1.701 1.637 -3.8 1.640 -3.6 2.061 1.983 -3.8 1.997 -3.1 3 1.690 1.650 -2.4 1.652 -2.2 2.052 1.977 -3.7 1.991 -3.0 4 1.677 1.660 -1.0 1.662 -0.9 2.030 1.975 -2.7 1.983 -2.3 5 1.667 1.667 0.0 1.670 +0.2 2.022 1.966 -2.8 1.975 -2.3 6 1.659 1.670 +0.7 1.674 +0.9 2.002 1.961 -2.0 1.969 -1.6 7 1.657 1.671 +0.8 1.675 +1.1 1.984 1.952 -1.6 1.966 -0.9 I 8 I 1.652 I 1.666 I +0.8 I 1.671 I +1.2 1 1.974 I 1.944 I -1.5 I 1.962 I -0.6 I 9 1.645 1.658 +0.8 1.664 +1.2 1 1.958 1.931 -1.4 1.954 -0.2 10 1.636 1.648 +0.7 1.655 +1.2 '1.939 1.921 -0.9 1.943 +0.2 I 11 1 1.624 I 1.636 I +0.7 I 1.643 I +1.2 1 1.972 I 1.907 I -3.3 I 1 1 9 m

• (Predicted - Measured) / (Measured) x 100%

LIC-04-0097 Attachment 2 Page 5 2 I 1041 I 1021 I -20 I 1009 I -32 I 3 968 96 1 -7 954 -14 4 924 901 -23 897 -27 5 1 865 I 844 I -21 I 840 1 ~ -25 I 6 805 790 -15 785 -20 7 752 739 -13 733 -19 8 695 688 -7 681 -14 634 I -17 I 628 I -23 I

• Prediction - Measurement Table 7 Cvcle 20 Peaking Factors - F, and F, 1 1.605 1.584 -1.3 1.585 -1.2 1.912 1.874 -2.0 1.880 -1.7 2 1.598 1.576 -1.4 1.574 -1.5 1.898 1.843 -2.9 1.851 -2.5 3 1.594 1.572 -1.4 1.570 -1.5 1.865 1.824 -2.2 1.826 -2.1 I 4 I 1.591 I 1.569 I -1.4 I 1.569 I -1.4 1 1.870 I 1.817 I -2.8 I 1.821 I -2.6 I 5 1.588 1.566 -1.4 1.568 -1.3 1.883 1.815 -3.6 1.819 -3.4 I

6 1.584 1.563 -1.3 1.566 -1.1 1.901 1.816 -4.5 1.821 -4.2 7 1.579 1.564 -0.9 1.567 -0.8 1.920 1.825 -4.9 1.832 -4.6 8 1.583 1.577 -0.4 1.579 -0.3 1.936 1.854 -4.2 1.861 -3.9 9 1.593 1.585 -0.5 1.587 -0.4 ' 1.929 1.861 -3.5 1.869 -3.1 10 1.609 1.593 -1.0 1.597 -0.7 1.909 1.855 -2.8 1.870 -2.0

• (Predicted - Measured) / (Measured) x 100%

LIC-04-0097 Attachment 2 Page 6 1 775 756 -19 74 I -34 2 740 720 -20 71 1 -29 3 70 1 674 -27 672 -29 4 654 628 -26 630 -24 5 608 586 -22 589 -19 6 568 549 -19 55 1 -17 7 535 517 -18 516 -19 8 I 503 I 486 I -17 I 484 I -19 I 9 460 454 -6 452 -8 10 407 414 +7 413 +6 12 1 261 1 266 I +5 I 270 I +9 1

• Prediction - Measurement 21 Peaking Factors - F, and F, Burnup Fr F, (GWD/MTU)

Meas. C-3E-3 Diff. C-4/S-3 Diff. Meas. C-3/S-3 Diff. C-4E-3 Diff.

Pred. (YO)* Pred. (YO)* Pred. (%)* Pred. (YO)*

1 0.25 8 1.605 1.572 -2.1 1.576 1 -1.8 I 1.930 I 1.864 -3.4 1.872 -3.0 9 1.616 1.597 -1.2 1.602 I -0.9 1 2.015 I 1.925 -4.5 1.938 -3.8 10 1.626 1.616 -0.6 1.622 -0.2 1.997 1.936 -3.1 1.950 -2.4 11 1.625 1.612 -0.8 1.618 -0.4 1.898 1.874 -1.3 1.886 -0.6 12 1.623 1.612 -0.7 1.613 -0.6 1.847 1.830 -0.9 1.834 -0.7

• (Predicted - Measured) / (Measured) x 100%

LIC-04-0097 Page 7

~

1.5 943 922 -2 1 91 1 -32 2.0 922 892 -30 885 -37 2.5 894 86 1 -33 856 -38 3.0 864 829 -35 827 -37

• Prediction - Measurement

• (Predicted -Measured) 1 (Measured) x 100%

LIC-04-0097 Page 1 MARKUP OF TECHNICAL SPECIFICATION PAGES

TECHNICAL SPECIFICATIONS 5.0 ADMINISTRATIVE CONTROLS 5.9 Reporting Requirements (Continued) 5.9.4 Unique Reporting Requirements

a. Annual Radioactive Effluent Release Report The Annual Radioactive Effluent Release Report covering the operation of the unit during the previous calendar year of operation shall be submitted before May 1 of each year. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit. The material provided shall be 1) consistent with the objectives outlined in the ODCM and PCP, and 2)in conformance with 10 CFR 50.36a.

and Section lV.B.l of Appendix I to 10 CFR 50.

b. Annual Radiological Environmental Operating Report The Annual Radiological Environmental Operating Report covering the operation of the unit during the previous calendar year shall be submitted before May 1 of each year. The report shall include summaries, interpretations, and analysis of trends of the results of the Radiological Environmental Monitoring Program for the reporting period. The material provided shall be consistent with the objectives outlined in (1) the ODCM and (2)Section IV.B.2, IV.B.3, and IV.C of Appendix I to 10 CFR 50.

c. NotUsed a.

b. The analytical methods used to dete previously

1. OPPD-NA-83014Ek4, "Reload Core Analysis Methodology Overview" approved version as specified in the COLR.

2. OPPD-NA-83024&4, "Neutronics Design Methods and Verification", approved version as specified in the COLR.

5.0 - Page 8 Amendment No. 9,2?,25,2~,46,75,8-&

iin 1 1 2 110 1 2 2 i n 1

,II"YLAJYA~',~JJYA 'L7 1/17 Y L " Y C 9 1 C& Q A.7j f l J I Y L J ' Y L Y 1 7 9 1&#&=9

' Y L I " Y 1 2 Y 338 h 7 h h

TECHNICAL SPECIFICATIONS 5.0 ADMINISTRATIVE CONTROLS 5.9 Reporting Requirements (Continued)

3. OPPD-NA-8303-PZ4, "Transient and Accident Methods and Verification",

approved version as specified in the COLR.

4. WCAP-12610-P-AY"VANTAGE + Fuel Assembly Report," April 1995 (Westinghouse Proprietary) as approved in the Safety Evaluation by the Office of Nuclear Reactor Regulation Related to Amendment No. 178 to Facility Operating License No. DPR-40, Omaha Public Power District, Fort Calhoun Station Unit No.

1, Docket No. 50-285, dated October 25, 1996.

5; XN-75-32(P)(A) Supplements 1,2,3, & 4,"ComputationalProcedure for Evaluating Fuel Rod Bowing," approved version as specified in the COLR.

XN-NF-82-06(P)(A) and Supplements 2,4, and 5, "Qualification of Exxon Nuclear Fuel for Extended Burnup," approved version as specified in the COLR.

XN-NF-85-92(P)(A), "Exxon Nuclear Uranium Dioxide/Gadolinia Irradiation Examination and Thermal Conductivity Results," approved version as specified in the COLR.

ANF-88-133(P)(A) and Supplement 1, "Qualification of Advanced Nuclear Fuels PWR Design Methodology for Rod Bumups of 62 GWd/MTU," approved version as specified in the COLR.

EMF-92-116(P)(A), "Generic Mechanical Design Criteria for PWR Fuel Designs,"

approved version as specified in the COLR.

5.0 - Page 9 Amendment No. 4 4 4 J 4 ? , ! 5 ? , ! ? ~ , 202

TECHNICAL SPECIFICATIONS 5.0 ADMINISTRATIVE CONTROLS 5.9.5 Reporting Requirements (Continued) 1A 1Q nr 4 31 r i l .

33 ria.

C.

d. -T eh shall be provided upon issuance7for each reload cycle? to the NRC n 7 5.0 - Page 10

,including any mid-cycle revisions or supplements Amendment No. 53,7 5,86,?3,99,~05, 152,155,1-, 203

LIC-04-0097 Page 1 Omaha Public Power District RELOAD CORE ANALYSIS METHODOLGY NEUTRONICS DESIGN METHODS AND VERIFICATION OPPD-NA-8302

LIC-04-0097 Page 1 PROPOSED TECHNICAL SPECIFICATION PAGES

TECHNICAL SPECIFICATIONS 5.0 ADMINISTRATIVE CONTROLS 5.9 Reporting Requirements (Continued) 5.9.4 Unique Reporting Requirements

a. Annual Radioactive Effluent Release Report The Annual Radioactive Effluent Release Report covering the operation of the unit during the previous calendar year of operation shall be submitted before May 1 of each year. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit. The material provided shall be

1) consistent with the objectives outlined in the ODCM and PCP, and 2) in conformance with 10 CFR 50.36a. and Section IV.B.l of Appendix I to 10 CFR 50.

b. Annual Radiological Environmental Operating Report The Annual Radiological Environmental Operating Report covering the operation of the unit during the previous calendar year shall be submitted before May 1 of each year.

The report shall include summaries, interpretations, and analysis of trends of the results of the Radiological Environmental Monitoring Program for the reporting period. The material provided shall be consistent with the objectives outlined in (1) the ODCM and (2)Section IV.B.2, IV.B.3, and 1V.C of Appendix I to 10 CFR 50.

c. Not Used 5.9.5 CORE OPERATING LIMITS REPORT (COLR) I

a. Core Operating Limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:

1.3(4) Thermal MargidLow Pressure Trip 1.3(8) Axial Power Distribution 2.2.7 Borated Water Sources - Shutdown 2.2.8 Borated Water Sources - Operating 2.10.2 Reactivity Control Systems and Core Physics Parameters Limits 2.10.4 Power Distribution Limits

b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents: I 5.0 - Page 8

TECHNICAL SPECIFICATIONS 5.0 ADMINISTRATIVE CONTROLS 5.9 Reporting Requirements (Continued)

1. OPPD-NA-8301, "Reload Core Analysis Methodology Overview approved version as specified in the COLR.

2. OPPD-NA-8302, "Neutronics Design Methods and Verification", approved version as specified in the COLR.

3. OPPD-NA-8303, "Transient and Accident Methods and Verification", approved version as specified in the COLR.

4. WCAP- 12610-P-A, "VANTAGE + Fuel Assembly Report," April 1995 (Westinghouse Proprietary) as approved in the Safety Evaluation by the Office of Nuclear Reactor Regulation Related to Amendment No. 178 to Facility Operating License No. DPR-40, Omaha Public Power District, Fort Calhoun Station Unit No. 1, Docket No. 50-285, dated October 25, 1996.

5. XN-75-32(P)(A) Supplements 1,2,3, & 4, "ComputationalProcedure for Evaluating Fuel Rod Bowing," approved version as specified in the COLR.

6. XN-NF-82-06(P)(A) and Supplements 2,4, and 5, "Qualification of Exxon Nuclear Fuel for Extended Bumup," approved version as specified in the COLR.

7. XN-NF-85-92(P)(A), "Exxon Nuclear Uranium Dioxide/Gadolinia Irradiation Examination and Thermal Conductivity Results," approved version as specified in the COLR.

8. ANF-88-133(P)(A) and Supplement 1, "Qualification of Advanced Nuclear Fuels PWR Design Methodology for Rod Burnups of 62 GWd/MTU," approved version as specified in the COLR.

9. EMF-92-116(P)(A), "Generic Mechanical Design Criteria for PWR Fuel Designs," approved version as specified in the COLR.

c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulics limits, Emergency Core Cooling System (ECCS) limits, nuclear limits such as shutdown margin (SDM), transient analysis limits, and accident analysis limits) of the safety analysis are met.

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

5.0 - Page 9 Amendment No.

1 A l l @ 1 4 1 174 1%

I 'A,& ' 7 A J ' Y I ' " Y A Y

LIC-04-0097 Page 1 Omaha Public Power District RELOAD CORE ANALYSIS METHODOLGY OVERVIEW OPPD-NA-8301

OMAHA PUBLIC POWER DISTRICT RELOAD CORE ANALYSIS METHODOLOGY OVERVIEW OPPD-NA-8301 Rev. 8 August 2004 Prepared By:

Matthew C. Rybenski

&"/

Date ReviewedBy: 6- C. 8-tr-oC Kevin CAHolthaus Date M a t h e m . Panicker Date Approved By: A . &&&&

Susan E. Baughn "

wDate Approved By: L 2 d .f W r/r 7 / 7 0 0 y Joe L. McManis Date SOJ OPPD-NA-8301 Rev. 8 i

TABLE OF CONTENTS Section Page

1. INTRODUCTION ................................................................................................................................................... 1 2 . FUEL MECHANICAL DESIGN ........................................................................................................................ 1 3 . NUCLEAR DESIGN ............................................................................................................................................... 1 3.1 FUEL MANAGEM ENT.. ................................................ ............ ....1 3.2 POWER DISTRIBUTION MEASUREMENT ......................................................................................................... 2 3.3 SAFETY RELATED PHYSICS DATA . ............................................................................................................. 2

4. THERMAL HYDRAULIC DESIGN ................................................................................................................. 2 5 . POSTULATED ACCIDENTS AND TRANSIENTS ..................................................................................... 2

6. SETPOINT GENERATION ................................................................................................................................. 2

7. CORE OPERATING LIMITS REPORT ......................................................................................................... 3 8 REFERENCES ......................................................................................................................................................... 4 OPPD-NA-8301 Rev . 8 11

OMAHA PUBLIC POWER DISTRICT RELOAD CORE ANALYSIS METHODOLOGY OVERVIEW Revision -

Date September 1983 June 1985 November 1986 April 1988 April 1991 January 1993 May 1994 December 2003 August 2004 OPPD-NA-8301 Rev. 8 111

OMAHA PUBLIC POWER DISTRICT RELOAD CORE ANALYSIS METHODOLOGY OVERVIEW

1. INTRODUCTION Omaha Public Power District (OPPD) has successfully performed reload core analysis for Fort Calhoun Station for many years and has (1) Established procedures and a personnel training and qualification program for this activity, (2) Performed comparison calculations to verify its ability to properly use reload analysis methods and codes, and (3) Established controls to ensure that reload analysis work is conducted under the Fort Calhoun Station Quality Assurance program.

Reload core analysis for Fort Calhoun Station is performed with a combination of efforts by OPPD and the fuel vendor using methods previously approved by the Nuclear Regulatory Commission (NRC). The current fuel vendor is Framatome ANP/Areva although future reload fuel may be supplied by another vendor. Any change in fuel vendor requires thorough evaluation of reload core analysis methodology to ensure proper analysis of the associated mixed cores and vendor-specific fuel designs. Operating experience with several fuel vendors has shown that it is important to maintain the flexibility to accommodate such changes, and OPPD intends to do so. The following sections discuss the division of responsibilities and the methodology used to perform reload core analysis.

2. FUEL MECHANICAL DESIGN Fuel assembly mechanical design and performance analysis are performed by the fuel vendor. Each fuel assembly used in mixed cores is evaluated by its vendor for acceptable operating performance. The fuel vendors use NRC-approved methods for performing these evaluations. Framatome ANP fuel mechanical design and performance analysis methods in current use are described in references contained in the Fort Calhoun Station Core Operating Limits Report (COLR).

3. NUCLEAR DESIGN The OPPD reload neutronics design methodology is discussed in Reference 1.

3.1 Fuel Management Reload core fuel management is performed by OPPD. Design goals are established to ensure reload cores can be bounded by safety analyses, maintain operating margins, minimize reactor vessel weld fluence, incorporate lessons learned from industry operating experience, and meet economic requirements.

OPPD-NA-8301 Rev. 8 1

3.2 Power Distribution Measurement Core power distribution measurement is performed using a self-powered fixed in-core detector system.

This and the associated measurement uncertainties are discussed in Reference 2.

3.3 Safety Related Physics Data Safety related reload core physics data are produced using the methodology discussed in Reference 1.

4. THERMAL HYDRAULIC DESIGN OPPD uses Framatome ANP methodology for reload core Departure from Nucleate Boiling (DNB) analyses documented in references contained in the COLR.

The Framatome XCOBRA -1IIC code is used to calculate the thermal hydraulic and DNB performance of the reload core limiting assembly based on linkage to the core thermal hydraulic boundary conditions obtained from transient calculations discussed in Reference 3. The Framatome High Thermal Performance (HTP) DNB correlation is used by XCOBRA-IIIC to calculate the limiting assembly DNB ratio. The time the minimum DNB ratio occurs during a transient as well as the value is determined. The minimum DNB ratio is then compared to the HTP DNB correlation limit (accounting for a mixed-core penalty, if appropriate) to confirm that the DNB Specified Acceptable Fuel Design Limit (SAFDL) is met.

5. POSTULATED ACCIDENTS AND TRANSIENTS The postulated accidents and transients are analyzed for reload cores using the methodology discussed in Reference 3.

6. SETPOINT GENERATION OPPD uses Framatome ANP methodology for reload core setpoint analyses (Reference 10). Each cycle the Limiting Safety System Settings (LSSS) and Limiting Conditions for Operation (LCO) setpoints are analyzed using statistical setpoint methodology along with the Framatome HTP DNB correlation to ensure the DNB and Fuel Centerline Melt (FCM) SAFDLs are met for the required range of conditions. This methodology is documented in references contained in the COLR.

OPPD-NA-8301 Rev. 8 2

7. CORE OPERATING LIMITS REPORT The Core Operating Limits Report (COLR) was incorporated into the Fort Calhoun Station Technical Specifications as part of the Cycle 14 reload license application (References 4 and 9. Subsequently, additional parameters were d d e d to the COLR (References 6 and 7). The COLR utilizes the guidance provided in Reference X The COLR values are generated with the methods described in References 1, 3, and 9.

The COLR consists of the following items:

Thermal MargidLow Pressure trip setpoints for 4 Pump Operation Axial Power Distribution trip setpoints Refueling Boron Concentration Limiting Conditions for Operation for Excore Monitoring of Linear Heat Rate (LHR)

Limiting Conditions for Operation for DNB Monitoring Power Dependent Insertion Limit (PDIL)

Unrodded Integrated Radial Peaking Factor (FR', and Core Power Limitations Allowable Peak Linear Heat Rate vs. Burnup Maximum Core Inlet Temperature Shutdown Margin with TCOLD > 2100F Most Negative Moderator Temperature Coefficient Minimum Boric Acid Storage Tank (BAST) Level vs. Stored BAST Concentration OPPD-NA-8301 Rev. 8 3

8. REFERENCES

1. OPPD-NA-8302, "Reload Core Analysis Methodology, Neutronics Design Methods and Verification," approved version as specified in the COLR

2. C E N P D 153-P, Revision 1-P-A, "INCAKECOR Power Peaking Uncertainty (Evaluation of Uncertainty in the Nuclear Power Peaking Measured by the Self-powered, Fixed In-core Detector System)," May 1980.

3. OPPD-NA-8303, "Reload Core Analysis Methodology, Transient and Accident Methods and Verification," approved version as specified in the COLR

4. NRC-92-105, Letter fromNRC (D. L. Wigginton) to OPPD (W. G. Gates), "Fort Calhoun Station, Unit No. 1 - Amendment No. 141 to Facility Operating License No. DPR-40 (TAC No. M82187),"

March 6, 1992.

5. NRC-92-138, Letter from NRC (D. L. Wigginton) to OPPD (W. G. Gates), "Correction to Amendment No. 141 -Fort Calhoun Station, Unit No. 1 (TAC No. M82187)," March 26, 1992.

6. NRC-95-184, Letter from NRC fl. Liu) to OPPD fl. L. Patterson), "Fort Calhoun Station, Unit No.

1 - Amendment No. 170 to Facility Operating License No. DPR-40 (TAC No. M92057),"

September 1, 1995.

7. NRC-99-106, Letter from NRC (L. R. Wharton) to OPPD (S. K. Gambhir), "Fort Calhoun Station, Unit No. 1 - Amendment No. 192 to Facility Operating License No. DPR-40 (TAC No. MA4651),"

July 27, 1999.

8. NRC Generic Letter 88- 16, "Removal of Cycle-Specific Parameter Limits from Technical Specifications," October 4, 1988.

9. OPPD-NA-830 1, "Reload Core Analysis Methodology Overview," approved version as specified in the COLR.

10. EMF- 1961(P)(A), "Statistical Setpoint/Transient Methodology for CE Reactors, Siemens Power Corporation," approved version as specified in the COLR.

OPPD-NA-8301 Rev. 8 4

OMAHA PUBLIC POWER DISTRICT RELOAD CORE ANALYSIS METHODOLOGY NEUTRONICS DESIGN METHODS AND VERIFICATION OPPD-NA-8302 Rev. 6 August 2004

OMAHA PUBLIC POWER DISTRICT RELOAD CORE ANALYSIS METHODOLOGY NEUTRONICS DESIGN METHODS AND VERIFICATION OPPD-NA-8302 Rev. 6 August 2004 Prepared By: @&

Matthew C. Rybensla Date Reviewed By: ;k;c C. 6l -ra-o+

Kevin C. Holthaus Date Approved By: JhuJsn e.+

Susan E. Baughn Date Approved By: d r L f 2 my Joe L. McManis Date

%f OPPD-NA-8302 Rev. 6 i

TABLE OF CONTENTS Section Page

1. INTRODUCTION ................................................................................................................................................... 1 2 . NEUTRONICS MODEL ....................................................................................................................................... 2 3 . NEUTRONICS DESIGN METHODS ............................................................................................................... 3

4. NEUTRONICS MODEL VERIFICATION ..................................................................................................... 4 5 . CONCLUSIONS .................................................................................................................................................... 11

6. REFERENCES ....................................................................................................................................................... 12 OPPD-NA-8302 Rev. 6 11

OMAHA PUBLIC POWER DISTRICT RELOAD CORE ANALYSIS METHODOLOGY NEUTRONICS DESIGN METHODS AND VERIFICATION Revision -

Date September 1983 November 1986 April 1988 January 1993 May 1994 December 2003 August 2004 OPPD-NA-8302 Rev. 6 111

OMAHA PUBLIC POWER DISTRICT RELOAD CORE ANALYSIS METHODOLOGY NEUTRONICS DESIGN METHODS AND VERIFICATION

1. INTRODUCTION This document describes the Omaha Public Power District (OPPD) Fort Calhoun Station neutronics design methods and model verification.

In the past OPPD has successfully developed and used approved neutronics models with computer code packages from Combustion Engineering and most recently Studsvik of America (now Studsvik Scandpower) (References 1, 2). The Studsvik CASM03/SIMULATE-3 system has been used and the CASMO-4/SIMULATE3 system is currently in use.

OPPD has previously obtained NRC approval to use Combustion Engineering neutronics design methods with CASMO-3/SIMULATE-3 (Reference 1). With the most recent change of fuel vendors to Framatome ANP/Areva in 2000, OPPD has adopted approved Framatome neutronics design methods (Reference 16) using OPPD's neutronics model for reload core analysis as necessary to support approved Framatome reload analysis methodology being used n other disciplines (References 3, 4). These disciplines include thermal hydraulic analysis, transient and accident analysis, fuel mechanical analysis, and setpoints analysis.

Section 2 of this document describes the neutronics model. Section 3 discusses the neutronics design methods used. Section 4 discusses neutronics model verification. Conclusions are given in Section 5.

Section 6 provides references.

OPPD-NA-8302 Rev. 6 1

2. NEUTRONICS MODEL OPPD uses the Studsvik Scandpower Inc. CASMO-WSIMULATE-3 (C-4/S-3) computer code system for Fort Calhoun Station reload core neutronics design analysis. The CASMO-3/SIMULATE-3 (C-3B-3) computer code system was previously approved (in Reference 1) for use and may be used in the future for cross section generation and neutronics parameter development. The Studsvik Scandpower code system is used for PWR and BWR core management by over 50 organizations including utilities and regulatory organizations in 11 countries. It has been used for over 70 PWRs and 70 BWRs to analyze over 500 fuel cycles for both reactor types. A number of U. S. utilities including OPPD have obtained NRC approval to use the system for safety related reload analysis. The system has proven its accuracy through comparisons with standard benchmarks and core operating measurements.

OPPD has established an approved C-3/S-3 model for Fort Calhoun using Studsvik modeling recommendations (Reference 1). Additionally, OPPD has developed a C-4/S3 model. These models are used to perform incore fuel management studies and calculations for core design, safety analysis input, and core operating and monitoring data.

Basically, the C-3/S3 and C 4 S - 3 code systems consist of two primary computer codes, CASMO-3 or CASMO-4 and SIMULATE-3. CASMO-3/CASM04 is used to calculate neutronics data for fuel assemblies and reflectors. This data is then compiled into a library. SIMULATE3 uses this library in calculations to predict the behavior of a core containing those fuel assemblies.

The C-3/S-3 code system requires three auxiliary programs: MICBURN-3, INTERPIN, and TABLES-3.

MICBURN-3 is used to calculate gadolina-bearing fuel pin cross-section data for input to CASMO-3.

INTERPIN is used to calculate he1 pin temperatures. Fuel pin temperature information is required by both CASMO-3 and SIMULATEr3. TABLES -3 is used to compile fuel assembly neutronics data produced by CASMO-3 into a library required by SIMULATE3.

The G4/S-3 code system requires Wo auxiliary programs. These are INTERPIN and CMSLINK.

INTERPIN is used to calculate fuel pin temperatures as explained above. CMS-LINK is used to compile fuel assembly neutronics data produced by CASMO-4 into a library required by SIMULATE3.

A description of the C-3E-3 and C-4/S-3 code systems is provided in References 5 through 11.

OPPD had been using approved versions of CASMO-3 and MICBURN-3 furnished by Framatome ANP for modeling fuel assemblies containing gadolinia as a burnable poison (References 12, 13, 14, 15). The use of MICBURN-3 is not necessary when using CASMO-4 as its capabilities are now built into the CASMO-4 code.

The computer codes in the OPPD neutronics model are controlled under the OPPD Quality Assurance Program.

OPPD-NA-8302 Rev. 6 2

3. NEUTRONICS DESIGN METHODS OPPD uses its neutronics model to provide neutronics data for reload core transient and accident safety analysis, setpoints analysis and fuel mechanical analysis. OPPD calculates this data in accordance with NRC approved methodology developed by the fuel vendors.

Core parameters used in safety analysis include the integrated radial peaking factor (FR), the 1 D peaking factor (FQ), axial power distributions, the moderator temperature coefficient of reactivity, the fuel temperature coefficient of reactivity, kinetics parameters, rod drop data, rod ejection data, scram worth, and reactivity insertion for the steam line break cooldown.

Neutronics model biases and uncertainties have been developed with previously approved OPPD methods such that values used in safety analyses conservatively bound the values anticipated during core operation.

The neutronics design methods used for a given analysis are dictated by the requirements of, and must be compatible with, the approved methodology for the particular analysis being performed. For example, the Fort Calhoun Station setpoint analysis is performed in accordance with Framatome ANP methodology referenced in the Fort Calhoun Station Core Operating Limits Report (COLR). Therefore, neutronics input to the setpoint analysis is prepared using guidance developed by Framatome ANP specifically for this purpose.

Mixed cores may require two separate analyses of the same event to be performed. For example, Cycle 20 contained both Westinghouse and Framatome ANP fuel. It was deemed necessary by safety analysis personnel to have each vendor perform a control rod ejection analysis or verify that the existing analysis was still valid. OPPD used its neutronics model to prepare the neutronics data for each analysis in accordance with the guidance provided by the respective vendor.

At the time of this writing, after three cycles of Framatome reloads, most of the reload safety analyses have been performed using Framatome methods referenced in the COLR. Therefore, neutronics inputs to these analyses are prepared using applicable guidance provided by Framatome.

A few infrequently performed or non-margin-limiting transient analyses of record are based on previously approved Westinghouse or Combustion Engineering methodology. These are discussed in Reference 4.

Should it become necessary to reanalyze any of these events, the approved methodology chosen (whether it be Framatome or otherwise) will dictate the neutronics design methods used. In any case, compatible neutronics design methods will be used to support the analysis.

In order to assure that neutronics analyses are correctly performed, OPPD personnel receive classroom training on vendor reload analysis methodology. Furthermore, when OPPD is performing a complex neutronics analysis for the first time, the analysis is formally reviewed by the vendor. This review is performed and documented to meet OPPD's Quality Assurance Program requirements.

OPPD-NA-8302 Rev. 6 3

4. NEUTRONICS MODEL VERIFICATION OPPD has demonstrated its ability to use the CASMO-3/SIMULATEr3 (C-3/S-3) and CASMO 4/SIMULAT&3 (C4S-3) computer code systems to accurately model the Fort Calhoun reactor core by performing operating parameter and cold critical benchmark comparisons.

OPPD has an ongoing neutronics model verification program. Core physics tests and core follow work provide measurement data for comparison to predictions of the following parameters: critical boron concentrations, moderator temperature coefficients, control rod worths, power distributions md peaking factors.

Summary results of these comparisons are provided on the following pages for the three most recently completed operating cycles as well as the startup for the current cycle in progress. These comparisons show good agreement between predictions and measurements and demonstrate the continuing accuracy of the OPPD neutronics model.

OPPD-NA-8302 Rev. 6 4

FCS Startup Testing Results for Cycles 19 to 22 CEA Worths C/o AD)

Measured by Rod Swap Method

• (Predicted - Measured)/(Measured)xlOO%

• • Reference bank measured via boration-dilution method.

Critical Boron Concentration (PPM)

Measurement C3iS-3 Difference C4iS-3 Difference (PPm) Prediction (ppm) (ppmp Prediction (pprn) (PPmY 19 1552 1512 -4 0 1553 +I 20 1672 1667 -5 1679 +7 1 I 21 1327 1312 -15 1318 -9 22 1569 1551 -18 1562 -7

• Prediction - Measurement OPPD-NA-8302 Rev. 6 5

Cycle Measurement C-31s-3 Prediction Difference C-4/S-3 Prediction Difference (pcm/"F) (pcml"F) (pcm/"F)* (pcm/"F) (pcrn/OF)*

-19 -8.3 -7.4 +0.9 -7.0 + I .3 BOC -21.4 -20.6 +0.8 -21.1 +0.3 EOC I& I 1 I I I I I I1 I II

-6.9 -7.6 -7.4 Notperformed ---

1 I I I II i II 1 Eg 1 1 I I I I LI

-12.2 -11.7 +0.5 -11.9

-20.3 -20.4 -0.1 -20.8 I I I I II i I I I I II

• Prediction - Measurement

• • EOC tests are performed before reaching actual end of cycle, when full power equilibrium critical boron concentration is about 300 PPM.

OPPD-NA-8302 Rev. 6 6

FCS Cycle 19 - Selected Data from Core Follow Reports and As-built Model

• Predicted - Measured

• (Predicted - Measured)/(Measured)xlOO%

OPPD-NA-8302 Rev. 6 7

FCS Cycle 20 - Selected Data from Core Follow Reports and As-built Model 100% Power All Rods

• Predicted - Measured 100% Power, All Rods Out unless otherwise noted.

• (Predicted - Measured)/(Measured)x 100%

OPPD-NA-8302 Rev. 6 8

FCS Cycle 21- Selected Data from Core Follow Reports and As-built Model 100% Power, All Rods Out unless otherwise noted.

• Predicted- Measured 100% Power All Rods Out unless otherwise noted.

• (Predicted - Measured)/(Measured)x 100%

OPPD-NA-8302 Rev. 6 9

FCS Cycle 22 - Selected Data from Core Follow Reports and As-built Model 100% Power, All Rods Out unless otherwise noted.

• Predicted - Measured 100% Power, All Rods Out unless otherwise noted.

• (Predicted - Measured)/(Measured)xlOO%

OPPD-NA-8302 Rev. 6 10

5. CONCLUSIONS OPPD has performed extensive benchmrking to incorporate the CASMO-3/SIMULATE-3 and CASMO LUSIMULATE.3 computer code systems into neutronics design methods for Fort Calhoun Station. This effort included comparisons of neutronics model predictions to measurements from previous Fort Calhoun Station operating cycles. The benchmarking results as well as successful experience in applications demonstrate the ability of OPPD to use its CASMO-3ISIMULAT53 or CASMOWSIMULATE-3 neutronics model to perform the neutronics design methods described in this document.

OPPD has continually demonstrated its ability to perform core neutronics calculations for reload core analysis as documented in References 1 and 2. Therefore, OPPD concludes that it can continue to perform steady-state reactor physics calculations with sufficient accuracy that the results may be used in licensing applications, reload design depletion analyses, reactor physics safety analyses, startup physics predictions, core physics data books, and reactor protection system and monitoring system setpoint analyses.

OPPD-NA-8302 Rev. 6 11

6. REFERENCES

1. NRC-94-346, Letter from NRC (S. D. B l o o 4 to OPPD (IL. . Patterson), Use of CASMO-3/SIMULATE-3 in Topical Reports OPPDNA-8301-P, Revision 6, Core Reload Analysis Methodology Overview, and OPPDNA-8302-P, Revision 4, Neutronics Design Methods and Verification - Fort Calhoun Station, Unit 1 (TAC Nos. M89455 and M89456), December 16, 1994.

2. NRC-93-0382, Letter from NRC (S. D. B l o o e to OPPD (T. L. Patterson), Safety Evaluation for OPPD-NA-8302-P, Revision 03, OPPD Nuclear Analysis, Reload Core Analysis Methodology, Neutronics Design Methods and Verification (TAC No. M85844), November 2, 1993.

3. OPPD-NA-830 1-P, Reload Core Analysis Methodology Overview, approved version as specified in the COLR.

4. OPPD-NA-8303-P, Reload Core Analysis Methodology, Transient and Accident Methods and Verification, approved version as specified in the COLR.

5. SIMULATE-3, Advanced Three-Dimensional Two-Group Reactor Analysis Code, Users Manual, StudsviWSOA-95/15, Revision 2, December 2001, Studsvik Scandpower.

6. TABLES -3, Library Preparation Code for SIMULATEr3, Users Manual, StudsviWSOA -95/16, Revision 0, October 1995, Studsvik of America, Inc.

7. CASMO-3, A Fuel Assembly Burnup Program, Users Manual, StudsviWSOA-94/9, Revision 0, November 1994, Studsvik of America, Inc.

8. MICBURN-3, Microscopic Bumup in Burnable Absorber Rods, Users Manual, Studsvik/NFA-89/12, Revision 0, November 1989, Studsvik AB.

9. INTERPIN-CS, Studsvik CMS Fuel Performance Code, Users Manual, StudsviWSOA -95/21, Revision 0, July 2000, Studsvik Scandpower.

10. CMS-LINK Users Manual, SOA-97/04, Revision 2, April 1999, Studsvik Scandpower.

11. CASMO-4, A Fuel Assembly Burnup Program, Users Manual, SSP-O1/400, Revision 2, November 2001, Studsvik Scandpower.

12. LIC-00-0110, Letter from OPPD (S. K. Gambhir) to NRC (Document Control Desk), Transmittal of Requested Information - Description of Core Reload Analysis Methodology, December 1, 2000.

13. NRC-01-019, Letter from NRC (L. R. Wharton) to OPPD (S. K. Gambhir), Fort Calhoun Station, Unit No. 1 - Issuance of Amendment (TAC No. MBOOS3), March 14, 2001.

14. LIC-01-0106, Letter from OPPD 6. K. Gambhir) to NRC (Document Control Desk), Fort Calhoun Station Unit No. 1 License Amendment Request, Additional Core Operating Limits Report (COLR) Analytical Methods, November 21,2001.

OPPD-NA-8302 Rev. 6 12

15. NRC-02-030, Letter from NRC (A. B. Wang) to OPPD Qi. T. Ridenoure), Fort Calhoun Station, Unit No. 1 - Issuance of Amendment RE: Addition of Topical Report References to TS 5.9.5, Core Operating Limits Report (TAC No. MB3449), March 4,2002.

16. EMF-96-029(P)(A) Volume 1, EMF-96-029(P)(A) Volume 2, EMF-96-029(P)(A) Attachment, Reactor Analysis System for PWRs, Volume 1, Volume 2, and Attachment, Framatome ANP, Inc., approved version as specified in the COLR.

OPPD-NA-8302 Rev. 6 13

LIC-04-0097 Page 1 Omaha Public Power District RELOAD CORE ANALYSIS METHODOLGY TRANSIENT AND ACCIDENT METHODS AND VERIFICATION OPPD-NA-8303

OMAHA PUBLIC POWER DISTRICT RELOAD CORE ANALYSIS METHODOLOGY TRANSIENT AND ACCIDENT METHODS AND VERIFICATION OPPD-NA-8303 Rev. 6 August 2004

OMAHA PUBLIC POWER DISTRICT RELOAD CORE ANALYSIS METHODOLOGY TRANSIENT AND ACCIDENT METHODS AND VERIFICATION OPPD-NA-8303 Rev. 6 August 2004 Prepared By: dlz/Cy Matthew C. Rybenslu Da{e Reviewed By: 8-1 2 Date Reviewed By: &/i+004 Mathew M. U c k e r Date Approved By: b . d b . W h SusanE.Baughn (I w Date Approved By: .w. -w Joe L. McManis YIlZ /zoo Date y

w-OPPD-NA-8303 Rev. 6 1

TABLE OF CONTENTS Section Page 1 . INTRODUCTION .................................................................................................................................................. 1 2 . COMPUTER MODELING ................................................................................................................................. 2 2.1 PLANT TRANSIENT MODELING USING SRELAPS ....................................................................... 2 2.2 CORE THERMAL-HYDRAULICS ANALYSIS USING XCOBRA -1IIC ......................................... 2 3 . USAR CHAPTER 14 EVENTS CONSIDERED IN RELOAD ANALYSIS .......................................... 3 3.1 USAR CHAPTER 14 EVENTS NOT CONSIDERED IN RELOAD CORE ANALYSIS ...............3 3.2 USAR CHAPTER 14 EVENTS CONSIDERED IN RELOAD CORE ANALYSIS ......................... 3 3.2.1 USAR Chapter 14 Events Analyzed by Fuel Vendor .......................................................................... 4 3.2.2 Margin-Limiting USAR Chapter 14 Events .......................................................................................... 4 3.2.3 USAR Chapter 14 Events Rarely Affected by Core Reload Design ................................................. 4 4 . RELOAD TECHNOLOGY IMPLEMENTATION ...................................................................................... 6 4.1 TRAINING....................................................................................................................................................... 6 4.2 CODE VERIFICATION AND VALIDATION ........................................................................................ 6 5 . CONCLUSIONS ..................................................................................................................................................... 7 6 . REFERENCES ........................................................................................................................................................ 8 OPPD-NA-8303 Rev . 6 11

OMAHA PUBLIC POWER DISTRICT RELOAD CORE ANALYSIS METHODOLOGY TRANSIENT AND ACCIDENT METHODS AND VERIFICATION Revision Date September 1983 November 1986 April 1988 March 1991 January 1993 December 2003 August 2004 OPPD-NA-8303 Rev. 6 111

OMAHA PUBLIC POWER DISTRICT RELOAD CORE ANALYSIS METHODOLOGY TRANSIENT AND ACCIDENT METHODS AND VERIFICATION

1. INTRODUCTION This document describes the Omaha Public Power District (OPPD) Fort Calhoun Station core thermak hydraulics, transient and accident analysis computer codes and methods for reload core analysis .

OPPD has previously obtained NRC approval to use Combustion Engineering core thermal-hydraulics and transient and accident analysis codes and methods for in-house Fort Calhoun Station reload analysis (Reference 1). With the most recent change of fuel vendors to Framatome ANP/Areva, OPPD has adopted and received NRC approval for the use of approved Framatome codes and methods (References 2 , 3 , 4 ) for the same purpose. This methodology is applicable to all NRC Standard Review Plan Chapter 15 events except loss of coolant accidents (LOCAs), but it does not include analysis of radiological consequences.

OPPD will use this methodology to analyze the limiting non-LOCA transients and accidents evaluated in Chapter 14 of the Fort Calhoun Station Updated Final Safety Analysis Report (USAR) (Reference 5 ) when reanalysis of an event is required.

OPPD does not perform detailed radiological consequences analyses of the Chapter 14 events. These are performed by qualified vendors. During the reload core analysis process OPPD verifies that the inputs and assumptions used in these analyses are valid for each reload core.

Section 2 of this document describes the computer codes used for core thermal-hydraulics analysis and modeling transients. Section 3 describes the overall approach used to perform a reload core analysis.

Section 4 discusses the process by which OPPD has adopted the Framatome methodology. Conclusions are given in Section 5 . Section 6 provides references.

OPPD-NA-8303 Rev. 6 1

2. COMPUTER MODELING 2.1 Plant Transient Modeling using S-RELAPS Analysis of plant system response during transients is performed with the S-RELAPS computer code (Reference 6). S-RELAPS is a general-purpose code that can be used for simulation of a wide variety of transients in pressurized water reactor systems. Reference 6 provides detailed descriptions of S-RELAPS and the various system models. The Fort Calhoun Station SRELAPS model 6 valid for transients discussed in Section 3 of this document with the exception of the Loss of Coolant Accident analysis.

2.2 Core ThennakHydraulics Analysis using XCOBRA-IIIC The thermal-hydraulics analysis of reload cores is performed using the XCOBRA-IIIC computer code (References 7, 8). XCOBRA-IIIC is used to perform thermal-hydraulic analyses o f nuclear reactor cores and fuel assemblies. It is used in the Framatom ANP methodology to assess thermal margin for operation of pressurized water reactors. The High Thermal Performance (HTP) Departure from Nucleate Boiling (DNB) correlation for critical heat ffux (Reference 9) is used to perform DNB ratio analyses for the Fort Calhoun Station Unit No, 1 reactor as approved in References 3 and 4. When applicable, Framatome ANP mixed-core methodology is used in the DNB analyses (Reference 19).

OPPD-NA-8303 Rev. 6 2

3. USAR CHAPTER 14 EVENTS CONSIDERED IN RELOAD ANALYSIS OPPD has reviewed the USAR Chapter 14 events to determine which events need to be considered in a reload analysis. Some events do not need to be considered in a reload analysis because the events are not affected by any reload core changes or are precluded by the Technical Specifications. The results of this review are discussed in the following sections.

3.1 USAR Chapter 14 Events Not Considered in Reload Core Analysis The following USAR Chapter 1 4 events are not evaluated in a reload core analysis.

Malpositioning of the Non-Trippable Control Element Assemblies CEAs) --- Insertion of the non-trippable CEAs during power operation is prohibited by the Fort Calhoun Station Technical Specifications. An inadvertent drop of a non-trippable CEA is included in the CEA Drop Analysis.

Idle Loop Startup Incident --- Part-loop operation is not permitted by the Fort Calhoun Technical Specifications.

Turbine-Generator Overspeed Incident --- This event is unrelated to any reload core changes.

Containment Pressure Analysis --- Containment pressure analysis is dependent upon the initial mass and energy contained in the primary or secondary system. These parameters do not change when the core is refueled.

Fuel Handling Accident -- The fuel handling accident analysis bounds the variations in fuel assemblies handled during normal reloads. However, raising fuel burnup or enrichrrent limits would require reconsideration ofthis analysis.

Gas Decay Tank Rupture --- This event is analyzed for radiological consequences and is independent of any reload core parameter.

Waste Liquid Incident --- This incident is analyzed for radiological consequences and is independent of any reload core parameter.

Control Room Habitability During Toxic Chemical Release Accident -- A toxic chemical release accident is independent of any reload core parameter.

Heavy Load Incident -- A heavy load incident is not affected by core refueling, and it is independent o f any reload core parameter.

3.2 USAR Chapter 14 Events Considered in Reload Core Analysis The USAR Chapter 14 events which must be considered are reviewed in order to determine whether they need to be evaluated for a given core reload. The criterion used to determine the events to be evaluated in reload core analyses is whether the changes in various neutronics, operations, or systems parameters adversely affect the safety analyses of these events. If these parameters change such that the previously reported results of a Chapter 14 event safety analysis are no longer conservative, then the event must be reanalyzed. If these parameters are conservative with respect to the values assumed in the analysis of record, the event is not required to be reanalyzed.

The USAR Chapter 14 events which must be considered are evaluated for each reload in a Disposition of Events analysis similar to that discussed in Reference 2. The purposes of the Disposition o f Events analysis are: (1) To evaluate the impact of changes to key parameters on the safety-related analyses that support the plants licensing basis, and (2) To determine the scope of analyses that need to be performed.

OPPD-NA-8303 Rev. 6 3

The Disposition of Events analysis evaluates changes in:

Plant configuration, operating conditions, Technical Specifications, and Reactor Protection System and other equipment setpoints.

Fuel design.

Neutronics parameters.

The Disposition of Events analysis for a reload classifies each event into one of the following categories.

Event must be reanalyzed as part of the reload analysis.

Event is bounded by another event. No reanalysis necessary.

Event is bounded by a previous analysis. No reanalysis necessary.

3.2.1 USAR Chapter 14 Events Analyzed by Fuel Vendor The following USAR Chapter 14 events are analyzed by nuclear fuel vendors. OPPD verifies that the inputs and assumptions used in these analyses are valid for each reload core.

Loss of Coolant Accident -- The LBLOCA analysis reported in USAR Chapter 14 was analyzed in accordance with the approved methodology given in Reference 10. The SBLOCA analysis reported in USAR Chapter 14 was analyzed in accordance with the approved methodology given in Reference 16. Future SBLOCA analyses could also use the approved methodology in Reference 11. The long-term core cooling analysis used the methodology described in Reference

12. The associated generation of hydrogen in containment was analyzed as described in USAR Chapter 14.

Control Element Assembly (CEA) Ejection Accident --- The CEA ejection accident analysis was performed using the approved methodologies in References 13 and 17. Future CEA ejection analyses could also use the approved methodology in Reference 2.

3.2.2 Margin-Limiting USAR Chapter 14 Events The following USAR Chapter 14 events are margin-limiting and are evaluated by OPPD in a reload core analysis. If it is found necessary to reanalyze any of these events, the analysis will be performed in accordance with the approved Framatome ANP non-LOCA transients and accidents methodology (References 2,3,4).

CEA Withdrawal Incident Boron Dilution Incident CEA Drop Incident Loss of Coolant Flow Incident Reactor Coolant Pump Seized Rotor Event Excess Load Increase Main Steam Line BreakAccident Reactor Coolant System Depressurization Incident 3.2.3 USAR Chapter 14 Events Rarely Affected by Core Reload Design The following USAR Chapter 14 events are evaluated by OPPD in a reload core analysis, but their consequences are rarely advesely affected by changes in core reload design. These events are not typically reanalyzed since their results are unaffected by reload changes or bounded by other events. Although analyses are not normally performed, each event is evaluated to ensure that the cnalysis of record is bounding. These events were analyzed using the approved methodology in Reference 18. If it is found necessary to reanalyze any of these events, the analysis will be performed in accordance with the approved Framatome ANP non-LOCA transients and accidents methodology (References 2,3,4).

Loss of Load --- The loss of load to both steam generators is not reanalyzed every reload because the relief capacity of the pressurizer safety valves does not change and the initial energy contained in the reactor coolant system will not change unless the rated power level is raised or the reactor OPPD-NA-8303 Rev. 6 4

coolant system temperature is significantly increased. The loss of load to one steam generator is not reanalyzed every reload because the asymmetric steam generator transient protection trip makes this a non-limiting event.

Malfunctions of the Feedwater System ---There are two events which are considered malfunctions of the feedwater system: (1) Loss of feedwater flow, and (2) Loss of feedwater heaters. The key input parameters for the loss of feedwater flow analysis include main steam safety valve and reactor coolant system power-operated relief valve setpoints, the auxiliary feedwater pump delay time, auxiliary feedwater temperature, decay heat generation, Eactivity feedback, and turbine valve closure time. This event is not reanalyzed every reload because there are generally no changes to these key parameters or to the relief capacity of the main steam or pressurizer safety valves, and the initial energy contained in the reactor coolant system will not change unless the rated power level is raised or the reactor coolant system temperature is significantly increased.

However, reanalysis of this event may be required should steam generator tube plugging exceed the value assumed in the analysis of record. The loss of feedwater heating incident is bounded by the excess load event, so the loss of feedwater heating incident is not required to be analyzed every re 1oa d .

Steam Generator Tube Rupture Accident --- This analysis primarily involves a calculation of the radiological consequences. These depend on a combination of fuel damage and subsequent primary coolant activity. Other key parameters affecting the radiological release are the cycling of the main steam safety valves and the atmospheric dump valve for the affected steam generator. A typical reload does not affect the key parameters for this event, and therefore, it is not necessary to reanalyze this event every reload. This event would require re-evaluation for high burnup fuel or a change in the heat transfer characteristics of the steam generators (such as if the number of plugged tubes in the steam generators increased beyond that analyzed or if new steam generators with different physical and operating characteristics were installed).

OPPD-NA-8303 Rev. 6 5

4. RELOAD TECHNOLOGY IMPLEMENTATION This section briefly discusses the training of OPPD personnel in computer codes and methodology. Also discussed in this section is the verification process that has been performed in order to demonstrate the fidelity of the SRELAPS and XCOBRA-IIIC codes for modeling Fort Calhoun Station system responses and OPPDs ability to correctly use the codes.

4.1 Training Classroom training on the generic version of the RELAP code for OPPD personnel was provided by Idaho National Engineering and Environmental Laboratory (INEEL) engineers. Framatome ANP engineers provided training on the use of S-RELAPS and XCOBRA-IIIC computer codes, as well as transients and setpoints methodology. These training sessions consisted of both classroom and hands-on training for both codes and methodology. When OPPD performs a transient or thermal-hydraulics analysis for the first time, the analysis is formally reviewed by Framatome.

4.2 Code Verification and Validation In order to demonstrate OPPDs ability to correctly use SRELAPS and XCOBRA-IIIC computer codes, verification and validation work has been performed by benchmarking them against independent safety analysis calculations. These benchmarking analyses were formally reviewed by Framatome and are documented in References 14 and 15.

OPPD-NA-8303 Rev. 6 6

5. CONCLUSIONS OPPD has demonstrated its ability to properly use Framatome codes and methods to analyze core thermal.

hydraulics and non-LOCA transients related to Fort Calhoun Station reload design and USAR Chapter 14 events. The application of these methods in performing core reload analysis has been described. OPPD has previously used Combustion Engineering methodology for similar purposes for many years. Therefore, OPPD concludes that it can perform reload core analysis using Framatorne methodology in accordance with References 2, 3 and 4. This conclusion has been further validated by the successful performance of analyses as found acceptable by the Framatome formal reviews.

OPPD-NA-8303 Rev. 6 7

6. REFERENCES

1. NRC-93-0301, Letter from NRC (S. D. Bloom) to OPPD (T. L. Patterson), Topical Report O P P D NA-8303, Transient and Accident Methods and Verification, Revision 4- Fort Calhoun Station, Unit 1 (TAC No. M85845), August 18, 1993.

2. EMF-23 10(P)(A), SRP Chapter 15 Non-LOCA Methodology for Pressurized Water Reactors, Framatome ANP, Inc., approved version as specified in the COLR.

3. NRC-01-019, Letter from NRC (L. R Wharton) to OPPD (S. K. Gambhir), Fort Calhoun Station, Unit No. 1 - Issuance of Amndment (TAC No. MB0083), March 14,2001.

4. NRC-02-030, Letter from NRC (A. B. Wang) to OPPD (R. T. Ridenoure), Fort Calhoun Station, Unit No. 1 - Issuance of Amendment RE: Addition of Topical Rport References to TS 5.9.5, Core Operating Limits Report (TAC No. MB3449), March 4, 2002.

5. Fort Calhoun Station Updated Safety Analysis Report.

6. EMF-2 lOO(P) Revision 4, S-RELAP5 Models and Correlations Code Manual, Framatome ANP Richland, Inc., May 2001.

7. EMF-CC-O93(P) Revision 4 XCOBRA-IIIC Theory and Users Manual, Framatome ANP, Inc.,

November 2003.

8. XN-NF-75-21 (P)(A), Revision 2, XCOBRA-IIIC: A Computer Code to Determine the Distribution of Coolant during Steady-State and Transient Core Operation, Exxon Nuclear Company, January 1986.

9. EMF-92-153(P)(A) and Supplement 1, HTP: Departure From Nucleate Boiling Correlation For High Thermal Performance Fuel, Siemens Power Corporation, approved version as specified in the COLR

10. EMF-2087(PXA), SEM/PWR-98: ECCS Evaluation Model for PWR LBLOCA Applications, Siemens Power Corporation, approved version as specified in COLR 1 1. EMF-2328(P)(A), PWR Small Break LOCA Evaluation Model, SRELAP5 Based, Framatome ANP, Inc., approved version as specified in the COLR.

12. WCAP-13027-P, Westinghouse ECCS Evaluation Model for Analysis of CE-NSSS, July 1991.

13. XN-NF-78-44(P)(A), A Generic Analysis of the Control Rod Ejection Transient For Pressurized Water Reactors, Exxon Nuclear Company, approved version as specified in the COLR.

14. EA-FC-02-036, Revision 0, S-RELAP5 Benchmarking, Qnaha Public Power District, June 26, 2003.

15. EA-FC-01-014, Revision 0, XCOBRA-IIIC Benchmarking, Omaha Public Power District, January 13, 2003.

16. XN-NF-82-49(P)(A), Supplement 1, Exxon Nuclear Company Evaluation Model Revised EXEM PWR Small Break Model, approved version as specified in the COLR

17. ANF 151(P)(A), ANF-RELAP Methodology for Pressurized Water Reactors: Analysis of Non-LOCA Chapter 15 Events, approved version as specified in the COLR.

18. OPPD-NA-8303-P-A, Revision 4, Transient and Accident Methods and Verification, January 1993.

19. XN-NF-82-2 1(P)(A), Application of Exxon Nuclear Company PWR Thermal Margin Methodology to Mixed Core Configurations, approved version as specified in the COLR.

OPPD-NA-8303 Rev. 6 8

LIC-04-0097 Page 1 TDB-VI CORE OPERATING LIMITS REPORT

Fort Calhoun Station Unit No. 1 TDB-VI TECHNICAL DATA BOOK

Title:

CORE OPERATING LIMIT REPORT FC-68 Number: EC 33044 Reason for Change: Update for Cycle 22 Requestor: Matthew Rybenski Preparer: Matthew Rybenski ISSUED: XX-XX-XX XXXX R29

FORT CALHOUN STATION TDB-VI TECHNICAL DATA BOOK PROCEDURE PAGE 1 OF 18 Fort Calhoun Station, Unit 1 Core Operatinq Limit Report Due to the critical aspects of the safety analysis inputs contained in this report, changes may not be made to this report without concurrence of the Nuclear Engineering Department.

R29

FORT CALHOUN STATION TDB-VI TECHNICAL DATA BOOK PROCEDURE PAGE 2 OF 18 TABLE OF CONTENTS Item Description Paqe 1.o Introduction 5 2.0 Core Operating Limits 5 3.0 TM/LP Limit 7 4.0 Maximum Core Inlet Temperature 7 5.0 Power Dependent Insertion Limit 8 6.0 Linear Heat Rate 8 7.0 Excore Monitoring of LHR 8 8.0 Peaking Factor (FRT)Limits 8 9.0 DNB Monitoring 8 10.0 FRTand Core Power Limitations 8 11.0 Refueling Boron Concentration 9 12.0 Axial Power Distribution 9 13.0 Shutdown Margin With TCoLD > 210 OF 9 14.0 Most Negative Moderator Temperature 9 Coefficient

FORT CALHOUN STATION TDB-VI TECHNICAL DATA BOOK PROCEDURE PAGE 3 0 F 18 LIST OF TABLES Table No. Paqe 1 TM/LP Coefficients 7 2 Refueling Boron Concentrations 9 R29

FORT CALHOUN STATION TDB-VI TECHNICAL DATA BOOK PROCEDURE PAGE 4 OF 18 LIST OF FIGURES Fiqure No. Title Paqe 1 Thermal Margin/Low Pressure for 4 Pump 10 Operation 2 Power Dependent Insertion Limit 11 3 Allowable Peak Linear Heat Rate vs. Burnup 12 4 Excore Monitoring of LHR 13 5 DNB Monitoring 14 6 FRTand Core Power Limitations 15 7 Axial Power Distribution LSSS for 4 Pump 16 Operation 8 Axial Power Distribution Limits for 4 Pump 17 Operation with lncores Inoperable 9 Minimum Boric Acid Storage Tank Level vs. 18 Stored Boric Acid Storage Tank Concentration

FORT CALHOUN STATION TDB-VI TECHNICAL DATA BOOK PROCEDURE PAGE 5 0 F 18 Core Operatinq Limit Report 1.O introduction This report provides the cycle-specific limits for operation of the Fort Calhoun Station Unit 1 for Cycle 22 operation. It includes limits for:

TM/LP LSSS for 4 Pump Operation (PvAR)

Core Inlet Temperature (TIN)

Power Dependent Insertion Limit (PDIL)

Allowable Peak Linear Heat Rate Excore Monitoring of LHR Integrated Radial Peaking Factor (FRT)

DNB Monitoring F Rversus

~ Power Trade-off Curve Refueling Boron Concentration Axial Power Distribution (APD)

Shutdown Margin With TCoLD > 210 OF Most Negative Moderator Temperature Coefficient These limits are applicable for the duration of the cycle. For subsequent cycles the limits will be reviewed and revised as necessary. In addition, this report includes a number of cycle-specific coefficients used in the generation of certain reactor protective system trip setpoints or allowable increases in radial peaking factors.

2.0 Core Operatinq Limits All values and limits in this TDB section apply to Cycle 22 operation. This cycle must be operated within the bounds of these limits and all others specified in the Technical Specifications. This report has been prepared in accordance with the requirements of Technical Specification 5.9.5. The values and limits presented within this TDB section have been derived using the NRC approved methodologies listed below:

OPPD-NA-8301, "Reload Core Analysis Methodology Overview," Rev. 8, dated August 2004. (TAC No. xxxxxx) I OPPD-NA-8302, "Reload Core Analysis Methodology, Neutronics Design Methods and Verification," Rev. 6, dated August 2004. (TAC No. xxxxxx) I OPPD-NA-8303, "Reload Core Analysis Methodology, Transient and Accident Methods and Verification," Rev. 6, dated August 2004. (TAC No. xxxxxx) I XN-75-32(P)(A) Supplements 1, 2, 3, & 4, "Computational Procedure for Evaluating Fuel Rod Bowing," October 1983.

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FORT CALHOUN STATION TDB-VI TECHNICAL DATA BOOK PROCEDURE PAGE 6 OF 18 XN-NF-82-06(P)(A) and Supplements 2,4, and 5, "Qualification of Exxon Nuclear Fuel for Extended Burnup," Revision 1, October 1986.

XN-NF-85-92(P)(A), "Exxon Nuclear Uranium Dioxide/Gadolinia Irradiation Examination and Thermal Conductivity Results," August 1985.

ANF-88-133(P)(A)and Supplement 1, "Qualification of Advanced Nuclear Fuels PWR Design Methodology for Rod Burnups of 62 GWd/MTU,"

December 1991.

EMF-92-116(P)(A), "Generic Mechanical Design Criteria for PWR Fuel Designs," Revision 0, February 1999.

XN-NF-78-44(P)(A), "A Generic Analysis of the Control Rod Ejection Transient for Pressurized Water Reactors," October 1983.

XN-NF-82-21(P)(A), "Application of Exxon Nuclear Company PWR Thermal Margin Methodology to Mixed Core Configurations," Revision 1, September 1983.

EMF-1961(P)(A), "Statistical Setpoinflransient Methodology for Combustion Engineering Type Reactors," Revision 0, July 2000.

ANF-89-151(P)(A), "ANF-RELAP Methodology for Pressurized Water Reactors: Analysis of Non-LOCA Chapter 15 Events," Revision 0, May 1992.

EMF-92-153(P)(A) and Supplement 1, "HTP: Departure from Nucleate Boiling Correlation for High Thermal Performance Fuel," March 1994.

XN-NF-82-49(P)(A), Supplement 1, "Exxon Nuclear Company Evaluation Model Revised EXEM PWR Small Break Model," Revision 1, December 1994.

EMF-2087(P)(A), "SEM/PWR-98: ECCS Evaluation Model for PWR LBLOCA Applications," Revision 0, June 1999.

EMF-2328(P)(A), "PWR Small Break LOCA Evaluation Model, S-RELAP5 Based,"

Framatome ANP, Inc., Revision 0, March 2001. I EMF-96-029(P)(A) Volume 1, EMF-96-029(P)(A) Volume 2, EMF-96-029(P)(A)

Attachment, "Reactor Analysis System for PWRs, Volume 1 - Methodology Description, Volume 2 - Benchmarking Results," Framatome ANP, Inc , January 1997.

EMF-231O(P)(A), "SRP Chapter 15 Non-LOCA Methodology for PressurizedWater Reactors. Framatome ANP. Inc.. Revision 0. Mav 2001.

'I R29

FORT CALHOUN STATION TDB-VI TECHNICAL DATA BOOK PROCEDURE PAGE 7 OF 18 3.0 TM/LP Limit The TM/LP coefficients are shown below:

Table 1 TMlLP Coefficients Coefficient Value a 29.6 P 20.63 Y -12372 The TM/LP setpoint is calculated by the PVAR equation, shown below and in Figure 1:

PvnR = 29.6 PF(B) AI(Y)B + 20.63TlN - 12372 PF(B) = 1.0 for B 2 100%

= -0.008(B)+1.8 for 50% < B c 100%

= 1.4 for B I 50%

A1(Y) = -0.6666(Y1)+ .OOO for Y1 I 0.00

= +0.3333(Y1) + I.OOO for Y1 > 0.00 Where:

B = High Auctioneered thermal (AT) or Nuclear Power, % of rated power Y = Axial Shape Index, asiu TIN = Core Inlet Temperature, OF P V A=~Reactor Coolant System Pressure, psia 4.0 Maximum Core Inlet Temperature The maximum core inlet temperature (TIN)shall not exceed 545 O F . This value includes instrumentation uncertainty of +2 OF (Ref: FCS Calculation FC06292, 6/9/95).

This limit is not applicable during either a thermal power ramp in excess of 5% of rated thermal power per minute or a thermal power step greater than 10% of rated thermal power.

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FORT CALHOUN STATION TDB-VI TECHNICAL DATA BOOK PROCEDURE PAGE 8 OF 18 5.0 Power Dependent Insertion Limit The power dependent insertion limit is defined in Figure 2.

6.0 Linear Heat Rate The allowable peak linear heat rate is shown in Figure 3.

7.0 Excore Monitorinq of LHR The allowable operation for power versus axial shape index for monitoring of LHR with excore detectors is shown in Figure 4.

8.0 Peakinq Factor Limits The maximum full power value for the integrated radial peaking factor (FRT)is 1.732.

9.0 DNB Monitoring The core operating limits for monitoring of DNB are provided in Figure 5. This figure provides the allowable power versus axial shape index for the cycle.

10.0 FRTand Core Power Limitations Core power limitations versus FRTare shown in Figure 6.

FORT CALHOUN STATION TDB-VI TECHNICAL DATA BOOK PROCEDURE PAGE 9 OF 18 11.O Refuelinq Boron Concentration The refueling boron concentration is required to ensure a shutdown margin of not less than 5%

with all CEAs withdrawn. The refueling boron concentration must be at least 1,900 ppm through the end of Cycle 21 operation and is valid until the beginning of core reload for Cycle 22.

Listed below in Table 2 are the refueling boron concentration values for cycle operations:

Table 2 Refuelinq Boron Concentrations Cycle Average Burnup Refueling Boron Concentration IMWDlMTU) 0 BOC 2,075 2 2,000 1,931 2 4,000 1,900 12.0 Axial Power Distribution The axial power trip is provided to ensure that excessive axial peaking will not cause fuel damage. The Axial Shape Index is determined from the axially split excore detectors. The setpoint functions, shown in Figure 7 ensure that neither a DNBR of less than the minimum DNBR safety limit nor a maximum linear heat rate of more than 22 kWlft (deposited in the fuel) will exist as a consequence of axial power maldistributions. Allowances have been made for instrumentation inaccuracies and uncertainties associated with the excore symmetric offset -

incore axial peaking relationship. Figure 8 combines the LHR LCO tent from Figure 4, the DNB LCO tent from Figure 5, and the APD LSSS tent from Figure 7 into one figure for a visual comparison of the different limits.

13.0 Shutdown Marqin With Tcold> 210 OF Whenever the reactor is in hot shutdown, hot standby or power operation conditions, the shutdown margin shall be 23.6% A u k . With the shutdown margin ~3.6%Auk, initiate and continue boration until the required shutdown margin is achieved.

14.0 Most Neqative Moderator TemDerature Coefficient The moderator temperature coefficient (MTC) shall be more positive than -3.05 x 10" AplOF, including uncertainties, at rated power.

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FORT CALHOUN STATION TDB-VI TECHNICAL DATA BOOK PROCEDURE PAGE I O O F 18 590 AS1 = 0.00 580 570 560 2400 psia 2250 psia 550 I 2100 psia 540 530 520 510 500 60 70 80 90 100 110 120 CORE POWER (% OF RATED)

PVAR = 29.6 PF(B)AI (Y)B + 20.63TlN - 12372 PF(B) = 1.0 B 2 100% A1 (Y) = -0.6666Y1 + 1.OOO Y1 IO.00

= -.008B+1.8 50% < B < 100% = +O.3333Y1 + 1.000 Y, > 0.00

= 1.4 B 5 50%

CYCLE 22 THERMAL MARGIN I LOW PRESSURE FIGURE COLR FOR 4 PUMP OPERATION I

FORT CALHOUN STATION TDB-VI TECHNICAL DATA BOOK PROCEDURE PAGE 11 OF18

-126 LONG TERM STEADY STATE

-100.8 INSERTION LIMIT 1104.51 e 194.51 a 182.51 3 -75.6 0

E

-126 (3 -50.4

- 100.8 -25.2 m

a -75.6 -0 3

0 lx (3 -50.4

-25.2

-0 c 126 100.8 13 21 125.21 113.21 12 BELOW F IL - EFER TO AOP-3

-126

-100.8 F

a -75.6 3

B (3 -50.4

-25.2

-0 0 20 40 60 80 100 REACTOR POWER (% OF ALLOWED POWER)

CYCLE 22 FIGURE COLR POWER DEPENDENT INSERTION LIMIT 2

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FORT CALHOUN STATION TDB-VI TECHNICAL DATA BOOK PROCEDURE PAGE 1 2 0 F 1 8 18 UNACCEPTABLE OPERATION 16 15.5 W l F T 14 ACCEPTABLE OPERATION 12 10 8

6 4

0 0 5000 10000 15000 20000 BURNUP (MWD/MTU)

CYCLE 22 ALLOWABLE PEAK LINEAR HEAT RATE FIGURE COLR VS.BURN P 3

110 -

(-0.12, I00) (0.10,100) 100 -

90-ol 80 -

W

- 70 -

e W

ol 0 60-0 n

W 50-40-L 0 30 -

I-z W

0 20 -

ol w

e 10 -

0-

-0.6 -0.4 -0.2 0.0 0.2 0.4 0.6 AXIAL SHAPE INDEX (ASIU)

CYCLE 22 FIGURE COLR EXCORE MONITORING OF LHR 4

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FORT CALHOUN STATION TDB-VI TECHNICAL DATA BOOK PROCEDURE PAGE 1 4 0 F 18 110 A I (-0.12,lOO) I I I (0.15,lOO) 100 ol 90 se 0 80 w

ol 0 70 0

n 60 -

5 0

II II 50 a

L 0 40 I-2 w

0 30 p1 w

e 20 10 0

-0.6 -0.4 -0.2 0.0 0.2 0.4 0.6 AXIAL SHAPE INDEX (ASIU)

CYCLE 22 FIGURE DNB MONITORING COLR 5 R29

FORT CALHOUN STATION TDB-VI TECHNICAL DATA BOOK PROCEDURE PAGE 1 5 0 F 18 110 100 90 80 70 60 50 40 30 20 10 0

1.70 1.75 1.80 1.85 1.90 PEAKING FACTOR CYCLE 22 FIGURE FRTAND CORE POWER LIMITATIONS COLR 6 I I R29

FORT CALHOUN STATION TDB-VI TECHNICAL DATA BOOK PROCEDURE PAGE 16 OF 18 150 140 130 120 p1 110 100 W

p1 90 0

0 80 n

70 60 50 c

40 0

p1 W 30 e

20 10 0

-0.8 -0.6 -0.4 -0.2 0.0 0.2 0.4 0.6 0.8 AXIAL SHAPE INDEX (ASIU)

CYCLE 22 AXIAL POWER DISTRIBUTION LSSS FIGURE COLR FOR 4 PUMP OPERATION 7

I FORT CALHOUN STATION TDB-VI TECHNICAL DATA BOOK PROCEDURE PAGE 1 7 0 F 18 p1 w

e I -0.8 -0.6 -0.4 -0.2 0.0 0.2 0.4 0.6 0.8 CYCLE 22 AXIAL POWER DISTRIBUTION LIMITS FOR 4 FIGURE COLR PUMP OPERATION WITH INCORES INOPERABLE 8

FORT CALHOUN STATION TDB-VI TECHNICAL DATA BOOK PROCEDURE PAGE 1 8 0 F 18 100 90 80 70 60 50 40 30 20 10 0

2.5 2.7 2.9 3.1 3.3 3.5 3.7 3.9 4.1 4.3 4.5 STORED BAST CONCENTRATION (WT % BORIC ACID)

A 1800 PPM X 1900 PPM +2000 PPM 4 2150 PPM +2300 PPM IN SIRWT IN SIRWT IN SIRWT IN SIRWT IN SIRWT CYCLE22 COLR I MINIMUM BAST LEVEL vs. STORED BAST CONCENTRATION I R29