RS-16-198, Request for License Amendment Regarding High Burnup Atrium-10 Partial Length Fuel Rods

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Request for License Amendment Regarding High Burnup Atrium-10 Partial Length Fuel Rods
ML16274A237
Person / Time
Site: LaSalle  Constellation icon.png
Issue date: 09/30/2016
From: Gullott D
Exelon Generation Co
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
IR 2016002, RS-16-198
Download: ML16274A237 (28)
v • d • e


Text

4300 Winfield Road Warrenville, IL 60555 630 657 2000 Office RS-16-198 10 CFR 50.90 September 30, 2016 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 LaSalle County Station, Units 1 and 2 Facility Operating License Nos. NPF-11 and NPF-18 NRC Docket Nos. 50-373 and 50-374

Subject:

Request for License Amendment Regarding High Burnup Atrium-10 Partial Length Fuel Rods

Reference:

Letter from B. Dickson (U.S. NRC) to B. C. Hanson (Exelon Generation Company, LLC), "LaSalle County Station, Units 1 and 2 - NRC Integrated Inspection Report 05000373/2016002; 05000374/2016002," dated August 4, 2016 In accordance with 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit," Exelon Generation Company, LLC (EGC) requests an amendment to Facility Operating License Nos. NPF-11 and NPF-18 for LaSalle County Station (LSCS), Units 1 and 2, respectively. The proposed change revises the LSCS licensing basis to allow movement of irradiated Atrium-10 fuel bundles containing part length rods that have been in operation above 62,000 MWD/MTU, which is the current rod average burnup limit specified in Footnote 11 of NRC Regulatory Guide 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors."

In the Reference, the NRC issued a Green finding and Severity Level IV non-cited violation because, on September 23, 2015, EGC made changes to the plant as described in the Updated Final Safety Analysis Report, in accordance with 10 CFR 50.59, "Changes, tests, and experiments," paragraph (c), and did not perform an accurate written evaluation which provided the bases for determining that these changes did not require a license amendment.

Specifically, the NRC determined that EGC did not provide an accurate written evaluation supporting the determination that exceeding the peak burnup limit of 62,000 MWD/MTU for fuel, provided that the maximum linear heat generation rate did not exceed 6.3 kw/ft peak rod average power for burnups exceeding 54 GWd/MTU, did not require a license amendment. The proposed change resolves the Green finding and Severity Level IV non-cited violation by seeking NRC approval.

September 30, 2016 U.S. Nuclear Regulatory Commission Page2 This request is subdivided as follows.

  • Attachment 1 provides a description and evaluation of the proposed change.

The proposed change has been reviewed by the LSCS Plant Operations Review Committee in accordance with the requirements of the EGC Quality Assurance Program.

EGC requests approval of the proposed change by February 6, 2017, to support the upcoming refueling outage. Once approved, the amendment will be implemented within 30 days.

In accordance with 10 CFR 50.91, "Notice for public comment; State consultation,"

paragraph (b), EGC is notifying the State of Illinois of this application for license amendment by transmitting a copy of this letter and its attachments to the designated State Official.

There are no regulatory commitments contained in this letter. Should you have any questions concerning this letter, please contact Mr. Kenneth M. Nicely at (630) 657-2803.

I declare under penalty of perjury that the foregoing is true and correct. Executed on the 30th day of September 2016.

Respectfully, D//111&3 David M. Gullatt Manager - Licensing Attachments:

1. Evaluation of Proposed Change
2. Design Analysis L-003067, "Re-Analysis of Fuel Handling Accident (FHA) Using Alternative Source Terms," Revision 2D cc: NRC Regional Administrator, Region Ill NRC Senior Resident Inspector- LaSalle County Station Illinois Emergency Management Agency - Division of Nuclear Safety

ATTACHMENT 1 Evaluation of Proposed Change 1.0

SUMMARY

DESCRIPTION 2.0 DETAILED DESCRIPTION

3.0 TECHNICAL EVALUATION

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria 4.2 No Significant Hazards Consideration 4.3 Conclusions

5.0 ENVIRONMENTAL CONSIDERATION

6.0 REFERENCES

Page 1

ATTACHMENT 1 Evaluation of Proposed Change 1.0

SUMMARY

DESCRIPTION In accordance with 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit," Exelon Generation Company, LLC (EGC) requests an amendment to Facility Operating License Nos. NPF-11 and NPF-18 for LaSalle County Station (LSCS), Units 1 and 2, respectively. The proposed change revises the LSCS licensing basis to allow movement of irradiated Atrium-10 fuel bundles containing part length rods that have been in operation above 62,000 MWD/MTU, which is the current rod average burnup limit specified in Footnote 11 of NRC Regulatory Guide 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors" (i.e., Reference 1, to which LSCS is committed).

2.0 DETAILED DESCRIPTION In Reference 2, the NRC issued amendments 197 and 184 to the facility operating licenses for LSCS, Units 1 and 2, respectively. The amendments support the application of alternative source term (AST) methodology with respect to the loss-of-coolant accident (LOCA) and the fuel handling accident (FHA). EGC's AST analyses for LSCS were performed following the guidance in NRC Regulatory Guide 1.183; Standard Review Plan 15.0.1, "Radiological Consequence Analyses Using Alternative Source Terms" (i.e., Reference 3); and 10 CFR 50.67, "Accident source term."

Section 3 of Reference 1 provides an AST that is acceptable to the NRC, and states that once approved, the AST assumptions or parameters specified in these positions become part of the facility's design basis. After the NRC has approved an implementation of an AST, subsequent changes to the AST require NRC review under 10 CFR 50.67.

EGC's AST analyses used the core inventory release fractions for the gap release and early in-vessel damage phases for the DBA LOCA that are listed in Table 1 of Reference 1. For the FHA event, EGC's analyses used the fractions of the core inventory assumed to be in the gap for the various radionuclides in Table 3 of Reference 1. These release fractions were used in conjunction with the fission product inventory calculated with the maximum core radial peaking factor.

Table 3 is annotated with the following footnote:

11 The release fractions listed here have been determined to be acceptable for use with currently approved LWR fuel with a peak burnup up to 62,000 MWD/MTU provided that the maximum linear heat generation rate does not exceed 6.3 kw/ft peak rod average power for burnups exceeding 54 GWD/MTU. As an alternative, fission gas release calculations performed using NRC-approved methodologies may be considered on a case-by-case basis. To be acceptable, these calculations must use a projected power history that will bound the limiting projected plant-specific power history for the specific fuel load. For the BWR rod drop accident and the PWR rod ejection accident, the gap fractions are assumed to be 10% for iodines and noble gases.

On April 14, 2014, Issue Report 1647125 was initiated under EGC's Corrective Action Program to document a concern that 19 fuel bundles in the Unit 1 Cycle 16 reactor core contain at least one part length rod that exceeds the 62,000 MWD/MTU burnup limit specified in Table 3 of Page 2

ATTACHMENT 1 Evaluation of Proposed Change Reference 1. At that time, it was projected that the burnup limit would be reached in September 2015. The fuel bundles containing part length rods were removed from the reactor core during the refueling outage in the first quarter of 2016.

Subsequent reviews identified that some part length rods in the Unit 2 Cycle 16 reactor core were also expected to exceed the 62,000 MWD/MTU burnup limit. The impact of this issue on Unit 2 was documented in Issue Report 2537519. The affected fuel bundles in the Unit 2 Cycle 16 reactor core are currently expected to reach the burnup limit in November 2016.

The NRC's review of these Issue Reports is discussed in Reference 4. In summary, the NRC's review identified a Green finding and a Severity Level IV non-cited violation of 10 CFR 50.59.

The NRC concluded that a change to the plant as described in the UFSAR was made, pursuant to 10 CFR 50.59(c), and EGC did not perform an accurate written evaluation which provided the bases for determining that these changes did not require a license amendment. Specifically, EGC did not provide an accurate written evaluation supporting the determination that exceeding the peak burnup limit of 62,000 MWD/MTU for fuel, provided that the maximum linear heat generation rate did not exceed 6.3 kw/ft peak rod average power for burnups exceeding 54,000 MWD/MTU, did not require a license amendment.

The proposed change resolves the Green finding and Severity Level IV non-cited violation by requesting NRC approval to revise the LSCS licensing basis to allow movement of irradiated Atrium-10 fuel bundles containing part length rods that have been in operation above 62,000 MWD/MTU.

For Unit 1, the fuel bundles containing part length rods were removed from the reactor core and placed in the spent fuel storage pool in the first quarter of 2016. However, approval of the proposed change would support future movement of those fuel bundles in the spent fuel storage pool since a fuel handling accident involving fuel bundles containing high burnup part length rods could occur. Administrative controls have been put in place to prevent movement of the affected Unit 1 fuel bundles, and these controls will remain in place until NRC approval of the proposed change is received.

For Unit 2, the affected fuel bundles are currently expected to reach the burnup limit in November 2016. Continued plant operation past this date is acceptable because for the LOCA analysis, release fractions are applied uniformly to the entire core and the core average exposure is not projected to exceed the 62,000 MWD/MTU burnup limit. However, since a fuel handling accident is assumed to occur during movement of irradiated fuel, approval of the proposed change would support movement of the affected Unit 2 fuel bundles.

3.0 TECHNICAL EVALUATION

The fuel bundles containing part length rods that have exceeded, or are projected to exceed, the 62,000 MWD/MTU burnup limit are Atrium-10 fuel bundles. The Atrium-10 part length rod is a fuel rod that is shorter than a full length rod, and is located in the bottom portion of the reactor core. It has less uranium content and, as a result of the axial power distribution due to voiding in the top of the core, it experiences higher rod-average exposure. However, it is very similar to the corresponding segment of the full length rod. Essentially the segment is the same as a full length rod, except it has the top portion of uranium removed. Using this equivalence aspect and Page 3

ATTACHMENT 1 Evaluation of Proposed Change the gap fraction calculation procedure, a comparative analysis was performed to demonstrate that a part length rod is bounded by a full length rod.

The associated design analysis is provided in Attachment 2. The scope of the design analysis is limited to the Atrium-10 fuel design at LSCS. The methodology associated with this design analysis is the AST as defined in Reference 1. Reference 1 provides gap release fractions for non-LOCA events in Table 3. These gap release fractions were developed prior to July 2000 and do not explicitly address part length rods. For LSCS, the fuel handling accident is the only non-LOCA event that uses AST methodology in the radiological consequence analyses.

A calculation procedure available to assess the gap inventory is specified in ANSI/ANS-5.4-2011, "Method for Calculating the Fractional Release of Volatile Fission Products from Oxide Fuel," (i.e., Reference 5) to assess the gap inventory. In that calculation procedure, the fractional release of fission gases is computed in a radial and axial (nodal) domain that represents the uranium oxide fuel pellet stack. The effective release fraction is then computed by a volume and power weighted average of the spatial dependent release fractions. The design analysis used this approach to demonstrate that the gap release fraction of the part length rod is bounded by the full length rod, given operational characteristics inherent to the part length rods.

The design analysis demonstrates that the power and burnup for an Atrium-10 part length rod is bounded by the power and burnup in the same axial portion of several neighboring full length rods. Therefore, since the full length rod operating characteristics bound the part length rod, and since the power and burnup of the full length rods comply with the limits specified in Footnote 11 of Reference 1, the Atrium-10 part length rods may operate beyond the 62,000 MWD/MTU burnup limit and meet the intent of Reference 1 (i.e., there is no increase in the non-LOCA gap release fraction required), with an adjustment to the Atrium-10 core damage fraction.

The core damage fraction was adjusted to provide conservatism and to make the Atrium-10 portion of the current analysis of record consistent with this approach. The Atrium-10 fuel assembly at LSCS has eight part length rods that have 90-inch active fuel length and 83 full length rods that have 149-inch active fuel length. Therefore, a part length rod is equivalent to 60.4 percent of the fueled mass of a full length rod. Using this basis, the total equivalent full length rods in the bundle is 87.83 and the revised core damage fraction is 0.002325 versus the current analysis value of 0.002244. This increase of 3.6 percent effectively assumes that the part length rods contain the same source term inventory as the full length rods. However, the limiting GE14 core damage fraction of 0.002578 remains bounding over Atrium-10. Therefore, there is no change in the dose consequences of the analysis of record. With a radial peaking factor of 1.7, the release fraction becomes 0.00395. The margin between the Atrium-10 and limiting dose consequences is therefore approximately 10 percent.

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria In accordance with 10 CFR 50.67, a licensee may revise its current accident source term by re-evaluating the consequences of DBAs with the AST. In Reference 2, the NRC Page 4

ATTACHMENT 1 Evaluation of Proposed Change approved the allocation of AST methodology at LSCS with respect to the LOCA and FHA. The guidance associated with the implementation of an AST is provided in NRC Regulatory Guide 1.183 (i.e., Reference 1), which states that subsequent changes to the AST require NRC review under 10 CFR 50.67.

Fundamental to the definition of an AST according to Reference 1 are gap release fractions. Table 3 of Reference 1 provides gap release fractions for various volatile fission product isotopes and isotope groups, to be applied to non-LOCA accidents.

Footnote 11 of Table 3 limits the peak burnup to 62,000 MWD/MTU provided that the maximum linear heat generation rate does not exceed 6.3 kw/ft peak rod average power for burnups exceeding 54,000 MWD/MTU. As an alternative, fission gas release calculations performed using NRC-approved methodologies may be considered on a case-by-case basis. To be acceptable, these calculations must use a projected power history that will bound the limiting projected plant-specific power history for the specific fuel load. For the BWR rod drop accident and the PWR rod ejection accident, the gap fractions are assumed to be 10% for iodines and noble gases.

Based on the review of the above requirements, EGC has determined that the proposed change does not require any exemptions or relief from regulatory requirements, and does not affect conformance with any of the above noted regulatory requirements or criteria. EGC's design analysis demonstrates that the radiological consequences of a FHA are not affected by the proposed change.

4.2 No Significant Hazards Consideration In accordance with 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit," Exelon Generation Company, LLC (EGC) requests an amendment to Facility Operating License Nos. NPF-11 and NPF-18 for LaSalle County Station (LSCS), Units 1 and 2, respectively. The proposed change revises the LSCS licensing basis to allow movement of irradiated Atrium-10 fuel bundles containing part length rods that have been in operation above 62,000 MWD/MTU, which is the current rod average burnup limit specified in Footnote 11 of NRC Regulatory Guide 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors."

According to 10 CFR 50.92, "Issuance of amendment," paragraph (c), a proposed amendment to an operating license involves no significant hazards consideration if operation of the facility in accordance with the proposed amendment would not:

(1) Involve a significant increase in the probability or consequences of any accident previously evaluated; or (2) Create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) Involve a significant reduction in a margin of safety.

Page 5

ATTACHMENT 1 Evaluation of Proposed Change EGC has evaluated the proposed change, using the criteria in 10 CFR 50.92, and has determined that the proposed change does not involve a significant hazards consideration. The following information is provided to support a finding of no significant hazards consideration.

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

Response: No The proposed change revises the LSCS licensing basis to allow movement of irradiated Atrium-10 fuel bundles containing part length rods that have been in operation above 62,000 MWD/MTU. The proposed change does not involve any physical changes to the plant design and is not an initiator of an accident. The proposed change does not adversely affect accident initiators or precursors, and does not alter the design assumptions, conditions, or configuration of the plant or the manner in which the plant is operated or maintained. In addition, the proposed change does not affect the probability of a fuel handling accident because the method and frequency of fuel movement activities are not changing.

An analysis has been performed that demonstrates that the power and burnup for an Atrium-10 part length rod is bounded by the power and burnup in the same axial portion of neighboring full length rods. Therefore, since the full length rod operating characteristics bound the part length rod, and since the power and burnup of the full length rods comply with the limits specified in Footnote 11 of NRC Regulatory Guide 1.183, the Atrium-10 part length rods may operate beyond the 62,000 MWD/MTU burnup limit and meet the intent of NRC Regulatory Guide 1.183 with an adjustment to the postulated core damage fraction. However, there is no change in the dose consequences of the analysis of record since the limiting GE14 core damage fraction of 0.002578 remains bounding over Atrium-10.

Therefore, the proposed change does not involve a significant increase in the probability or 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 revises the LSCS licensing basis to allow movement of irradiated Atrium-10 fuel bundles containing part length rods that have been in operation above 62,000 MWD/MTU. The proposed change does not introduce any changes or mechanisms that create the possibility of a new or different kind of accident. The proposed change does not install any new or different type of equipment, and installed equipment is not being operated in a new or different manner. No new effects on existing equipment are created nor are any new malfunctions introduced.

Page 6

ATTACHMENT 1 Evaluation of Proposed Change Therefore, the proposed change does not create 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 revises the LSCS licensing basis to allow movement of irradiated Atrium-10 fuel bundles containing part length rods that have been in operation above 62,000 MWD/MTU. An analysis has been performed that demonstrates that the power and burnup for an Atrium-10 part length rod is bounded by the power and burnup in the same axial portion of neighboring full length rods. There is no change in the dose consequences of the fuel handling accident analysis of record since the limiting GE14 remains bounding over Atrium-10. The margin of safety, as defined by 10 CFR 50.67 and NRC Regulatory Guide 1.183, has been maintained.

Therefore, the proposed change does not involve a significant reduction in a margin of safety.

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

4.3 Conclusions In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or the health and safety of the public.

5.0 ENVIRONMENTAL CONSIDERATION

EGC has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, "Standards for Protection Against Radiation." However, the proposed amendment does not involve: (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22, "Criterion for categorical exclusion; identification of licensing and regulatory actions eligible for categorical exclusion or otherwise not requiring environmental review,"

paragraph (c)(9). Therefore, pursuant to 10 CFR 51.22, paragraph (b), no environmental impact statement or environmental assessment needs to be prepared in connection with the proposed amendment.

Page 7

ATTACHMENT 1 Evaluation of Proposed Change

6.0 REFERENCES

1. NRC Regulatory Guide 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," dated July 2000
2. Letter from C. Gratton (U.S. NRC) to M. J. Pacilio (Exelon Nuclear), "LaSalle County Station, Units 1 and 2 - Issuance of Amendments Re: Application of Alternative Source Term (TAC Nos. ME0068 and ME0069)," dated September 6, 2010
3. NRC Standard Review Plan 15.0.1, "Radiological Consequence Analyses Using Alternative Source Terms," Revision 0, dated July 2000
4. Letter from B. Dickson (U.S. NRC) to B. C. Hanson (Exelon Generation Company, LLC),

"LaSalle County Station, Units 1 and 2 - NRC Integrated Inspection Report 05000373/2016002; 05000374/2016002," dated August 4, 2016

5. ANSI/ANS-5.4-2011, "Method for Calculating the Fractional Release of Volatile Fission Products from Oxide Fuel" Page 8

ATTACHMENT 2 Design Analysis L-003067, "Re-Analysis of Fuel Handling Accident (FHA)

Using Alternative Source Terms," Revision 2D

CC-AA-309-1001 Revision 8 Page 30 of 67 ATTACHMENT 1 Design Analysis Cover Sheet Page 1of5 Design Analysis J Last Page No.

  • H-14 Analysis No.: 1 L-003067 Revision: 2 20 Major D Minor [8l

Title:

3 Re-Analysis of Fuel Handling Accident (FHA) Using Alternative Source Terms EC/ECR No.:

  • 406676 Revision: 5 001 Station(s): 1 LAS Component(s): 1 Unit No.:* 1&2 NA Discipline: 9 MEDC, NUDC R01 & R02/ FHA, AST, Descrip. Code/Keyword: 10 OBA, Dose Safety/QA Class: 11 SR System Code: 12 VC,VG,VS,ZZ Structure: 13 NA CONTROLLED DOCUMENT REFERENCES 15 Document No.: From/To Document No.: From/To NF-AB-110-2210 From Drawing 5054 77 4 From UFSAR CH 15 Table 15.7-21 To Is this Design Analysis Safeguards Information? 16 Yes D No [8l If yes, see SY-AA-101-106 Does this Design Analysis contain Unverified Assumptions? 11 Yes D No [8l If yes, ATl/AR#:

This Design Analysis SUPERCEDES: 19 L-003067 Rev. 2C in its entirety.

Description of Revision (list changed pages when all pages of original analysis were not changed): 1*

Revision addresses Atrium-10 part-length rods exposure impacts on the FHA dose consequences and addresses IR 2691476 related to a violation of 10 CFR 50.59. Affected pages are 5, 6, and addition of a new Attachment H.

Preparer: 20 Shane Gardner

...!~ tv >>---- 2016-09-27 14:51-04:00 Print Name Sien Name Date Method of Review: 21 Detailed Review [8l Alternate Calcula~sj,ttached) D Testing D Reviewer: 22 Greg Heasley ":!!_ ~~ 9/27/16 Print Name SlgSme Date Review Notes: 23 Independent review [8l Peer review D An independent review of the evaluation was performed in accordance with CC-AA-309 Revision 11 and T&RM CC-AA-309-1001 Revision 8. All review comments were resolved satisfactorily. Based on the review, the numeric analysis and results support the conclusion.

(For External Analyses Only}

External Approver: 2

  • NA Print Name Sign Name Date Exelon Reviewer: 25 NA Print Name Sien Name Date Independent 3rc1 Party Review Reqd? 2
  • Yes D No [8l Exelon Approver: 21 John Massari 9/27/2016 Pnnt Name Date

(

L-003067 Rev. 20 Page 5 of 21 vall.l!es tmllzed. and .a evalua~on o the aoooptablllry ot oonsiderarloo o an addmana~

Secondary COllltainment penetration lbeiing !left open (oonwieration of any such penauallons !Is remoYed in 1his Revi!OOn 2).

Revision 2 of this caf.oulation:

  • removed the d'efilnlhl'on of recenrly 11*1adlatt'!!d fuel.
  • added lhe requirement for Seconljary Concs*nmenit lnnegritY and Standby Gas Treatment System operal>illfy, wJllh fillratiOR ayeilatlilily fOl&oWilng an ii)$'$Ul'l'IEld Secondary Contahnmenl drawdolM!'I period..
  • r emoved poirnt and diffu:se 1;1rea pa1hwa)I$ li;\$$0r;:iabed wiNl not lh$'1ring Seoonclary Contalnmenl lntagrl~
  • ep;plied reYised 1./0 v.cil~es,
  • added CRAF operability wllh l~l<e fl~ter crecli'I after 20 mll"lll.ltes aind ll'e(:i1Tcul\allon fll,er credit after 4 hOtlrs ([in the same manner as Is uised for the AS'T lloiss of Cooling Accident), and
  • edded seYerar minor ,cleirificatiol'lS.
2. METHOD OF ANALYSIS AND ACCEPTANCE CRtTfRIA Analyses of rradlologi~ conseQUences resulting fmm a des,;gn basis FHA are pell'formed 1JSing the ~idence r~ ap.plication Qr AST to ltlis event in RG 1.183. and the approYed AST VaJJues 'for FHA p:rtMded In :the Ref. 13 "Ttansmittal Qf Design lnfoR!l'll;;Ulon".

An:iilyse.s oftadiatron lran:spon ~n(ll dose assessrnEml are iPe.rfotmed U$ing RADYRAJO v. 3.03.

RADTRAD Is a slmJpltft.ed*model of B!Qronucllda Ira~mt and Bemoval !ltd Qotie IEtitimallon deYeloped for the NRC 1;11nd ~n~ed by the NrRC ;es an accepleble meqh(>(JotQgy for reane.ty.sls QI the radlologlcal oonsequences or design !basis accidents. The technlca basis for the AADTRAD code i$ dtl(:ul'li'lellllted in NUREGJCR-6604 (Ref. 3). The melthodologieg sigl1iiffce1n to ttJJ5 analyiis are *he dose 0011seq1Jeooe unatysls (NUREG Sec00n 2.3) and tie Radloacllve Oeeay Calcul\alions (NU REG Section 2.4). Trus veltironi of RADTRAD has been ps:~a1ified for safety ma*ed design anarysls by URS pa its 10CFR50 Appendix B Oualhy A:ssuraMJO program.

2.1. FueJ Source Term Model AJ. per ~r. 13, the fue~ sourir:.e term is bued on the ~eactor core sou:r:oe terms descrtiboo In Attael'lment A These Get.Jree 1err'r\$ are b~nding f<:ir LSCS fuel cycle i::vestgl1$ as documented ir.

Anactimen* A, whieh repeats the rellBvaril s.oUll'ce tetm ln6ormatlon ftom Ref. 20 for cot1va:nienoe.

The fraction .of the core fue* (~n lhe 764 fuel tNJndle com) dama"Qed Is per the R-elf. 9, Ret. 10,

.cilld Ref. 111 GESTAR ,u 1imiltng case of damaging 172 l'iLlel pens (based 001 a He~ Mast

0 design: i.e.* the "NF50() mast" ln Ref. 10) ftom GE12 or GE14 10x10 fuel bundte array.s with. illle equivalanl oi .87.33 pins per bun&.e. and wi1h all oHhe damegMI fuel assumed kt have a r iting Radial Peaking Fooror (PF) *of 1.7 (per Ref. 13). Th.is an.alysis is fa*Tan assembly and mast drop 1

from a 34 reel :maximum llileigtrt from *the ll'efueliing, pfs,tforrn trt1er !he reacbor WEiii onto the reacklr 1

core. btlundlna In fterms od' *ft.1111 dal'l'lage 1Poten11a1. Based en ruel damage assassrrmeait$ in ref~ enoes u. 14, emt 15~ , C>Wn in labie 1 befow, this bolJJlcb all currenl!ly uwa and hlstarical roo1 types (indu an~ 7-,.7 array fuel 1hal may have beElfll used ea in the ~clDr life, now S\lfflclently deiceyed any releases wi.11 be bouooed by n~ *o ther :Jue' ~pes coosidered below).

and Attachment H

L-003067 Rev. 20 Page 6 of 21 Table 1: B oundlna Fuel Bundle T*voe Evaluation Radial Asaumed Oamaged Damaged Fuel Peaking I

Bundle Type Array i IPin& In Faliled Core Factor Core Fraction Bundle fqns fraetlon (Pf, wlHiPF

_;E-Varlous sx.a 62 1.24 0.(102618 1.5 0.003927 s:.ANP Atrium*9B 9x'9 72 131 0.002381 I

1"5 O.Q035'72 Alnum-10 1GJC10 156 - 1.7 BI.83 0.002325 0.003952

~IE11&GE~J 9JC9 74 140 1).002476 1.-fj 0.00'3714 -

GIE1.2&0E14" 10x10 87.33 172 0.002578 1.7 1 0.004382

  • Bound*n11 As&amblv tvoe, wtth RadlaR P&ak.ilna Factor commen:&ural8 with f1,1lt core a,oollcatton.

Wilh a 1.7 radial pee.king factor, the as&O<:iated pO'Her of lhe d:11maged fuel= 3559 Mwth

  • 0.0()2578. ~.7 = ~5.597 MWlh.

This accident artal)'Sis evaluales !hie movemenn of fuel ttist has decayed a minimum or 24 lhourPiil since iloooupied part of a or:itlcal raar::tor cor-e.

2.2. Gap Actlvljy This ca~lation - applleabre to fuel who&e bur1111up and power lll1'liit;s. are bounded by those spec:med mRG 1. 1SJ. 'fooa!"Kl!e 11.

  • 111;W;"S applieatiOl't or th& gap actirJity ft.actions t.ar Non-1Loss of Coolan' A.cclderit (LOCA) events pe ,of RG 1.183, which are as follows:

5% Gf lhe nobl~ gases (eJ<ciudlnig K'l'-85) (Atrium-10 part-10% o the Kr-85 length rods are 5% of lhe lodlrte lnventtol)I (exciudit'llg t-131) evaluated in 8% of ~e 1*131 Attachment H) 12% of ttie Alka!i melal jnvennoiy The nob3e 9alii allbd iodine inventories ere ptrovfded fQr RADTRAD U$e in Ollie Re&ea&e*Fraction

.aml Timing (~rfl") ffle *~salle ast ilha.rfr 1ila inclooed as A.ttachmen~ 0. Becausio RADTRAD does not low ~()'!' appli:eatloo o~ Isotope spedftc releE1se fractions, the Nuclide lnform~tioo File

("nlf"') rue "l:scs ast source terms kif fha.n~r. rncl!Jd~ as Attachmient C, Is modified to aocommotlab!t the dinterential gap activities amClngi lhfl halogen (lm131) ainel nle gas {Kr-65}

gap fracticms diele1ed by RG 1.183 (Ref. 2} shown above. Thell'efOfe. the ini~ial activity Of

Isotope 1*131 and Kr-85 am multiplied by HI arid 2.0, ,respedlvely, In ~his "nlf' rille In order eo accommodaue the respec'lfve 10% and 8% r-OOease fradions cfirected' by r,egulatOI)' guidance

{Ref, 2).

2. 3. Pool Deconitemln* tlo.n Factor (DF)

Anadllrnenl E prlY\lide-.s assessmflnit$ of worst-case waUtr coverage and *tuea damage or FHAs over the reactor well ealld lhe spent fueO pool * .and demonsrrates th:irt Ute drop ov~ the reactor wot! ie mw.a limiting and ttiere-1Qra lhe bQuridmg case. This i!S due 'to the graalet number of fuel ll'C<ls dam~ed for the TNctor we" drqp (117 for file spent fuel pool drop w.. 172 for 1he reactor INflll drop*fos Om bQunding 10-.X~ Ofillel a:rray, or a ~of 68.0%). and I

  • e tad thait !he fower 'ltlan

ATTACHMENT H - ATRIUM-10 PART-LENGTH ROD EVALUATION L-003067 Rev. 2D Page H-1 of H-14 PURPOSE/OVERVIEW IR 1647125 identified a concern with excessive exposure in part-length rods for LaSalle 1 Cycle 16. Subsequently IR 2537519 was written to document the condition as it relates to Alternative Source Terms (AST) Regulatory Guide (RG) 1.183 footnote limitation of 62 GWD/MTU for non-LOCA gap release fractions specified in Table 3 of the RG. The other RG 1.183 footnote limitation of 6.3 kW/ft for rod average burnup exceeding 54 GWD/MTU is not a concern for part-length rods as documented in Reference 1. Therefore, a minor revision is necessary to resolve the impact of a few part-length rods that exceed 62 GWD/MTU rod average burnup. This minor revision can be incorporated into future major revision of the design analysis by inserting this attachment in its entirety as a new attachment.

SCOPE The scope of this design analysis is limited to LaSalle station. The results and limitations of this design analysis revision are not applicable to other stations. The data evaluated supports LaSalle 1 Cycle 16, but this analysis can be utilized for other operating cycles, including at LaSalle 2, provided the fuel rod characteristics used to support the conclusions herein remain bounding.

CONCLUSIONS The impact on the AST FHA analysis of record due to operation of Atrium-10 part-length rods in excess of 62 GWD/MTU has been evaluated. A comparative analysis was performed to understand how the part-length rod compares to the full-length rod, which are co-located in the fuel bundle and experience similar nuclear operating characteristics. It was concluded that power and burnup for an Atrium-10 part-length rod is bounded by the power and burnup in the same axial portion of several neighboring full-length rods. Therefore, since the full-length rod operating characteristics bound the part-length rod, and since the power and burnup of the full-length rods comply with the RG 1.183 footnote 11 limits, it can be concluded that Atrium-10 part-length rods may operate beyond 62 GWD/MTU burnup and meet the intent of RG 1.183 (i.e. there is no increase in the non-LOCA gap release fraction required). This conclusion required an adjustment to the Atrium-10 core damage fraction. The discussion in the original analysis of record and the UFSAR (Table 15.7-21) should be updated to reflect that Atrium-10 part-length rods are treated separately with respect to the RG 1.183 footnote 11 limits.

METHODOLOGY The methodology associated with this design analysis is the Alternative Source Term (AST) as defined in RG 1.183. A part of this methodology is the gap release fractions for non-LOCA events like FHA as specified in Table 3 of RG 1.183. These fractions were developed prior to July 2000 and do not explicitly address part-length rods.

A calculation procedure available to assess the gap inventory is specified in ANSI/ANS-5.4-2011, Method for Calculating the Fractional Release of Volatile Fission Products from Oxide Fuel. In that calculation procedure the fractional release of fission gases is computed in a radial and axial (nodal) domain that represents the uranium oxide fuel pellet stack. The effective release fraction is then computed by a volume and power weighted average of the spatial dependent release fractions. This design analysis revision will rely on this approach to

ATTACHMENT H-ATRIUM-10 PART-LENGTH ROD EVALUATION L-003067 Rev. 20 Page H-2 of H-14 demonstrate that the gap release fraction of the part-length rod is bounded by the full-length rod, given operational characteristics inherent to the part-length rods.

The Atrium-10 part-length rod is a fuel rod that is shorter than a full-length rod and is located in the bottom portion of the core. It has less uranium content and, as a result of the axial power distribution due to voiding in the top of the core, it experiences higher rod-average exposure.

However, it is very similar to the corresponding segment of the full-length rod. Essentially the segment is the same as a full-length rod, except it has the top portion of uranium removed.

Using this equivalence aspect and the gap fraction calculation procedure a comparative analysis is performed to demonstrate that a part-length rod is bounded by a full-length rod.

Consider the following depiction of a full and part-length rod in Figure H-1.

Figure H Diagram of Full and Part-Length Fuel Rod (Not to scale; Example burnup profile from LaSalle data) 27 26 25 ~

Nodes 17 to 25 24 23 22 - ......

21 20 19 """"'"\...'1 18 17 16 ~

'L 15 "";l

~

QI 14 iii 13 ~ FLR

12 11 -B- PLR Nodes 2 10 to 16 9 Nodes 1 8 to 16 7 6

5 4 .....I' 3

~

2 -

1 ~

0 0.00 20.00 40.00 60.00 80.00 FLR PLR Bumup (GWD/ST)

Figure H-1 shows a depiction of the Atrium-10 full and part-length rods in use at LaSalle. The burnup profile shows a typical comparison of the two rods. The part-length rod has a slightly lower burnup than the corresponding portion of the full-length rod. This behavior is partly due to the decrease in neutron multiplication in the upper portion of the part-length rod as a result of the missing fueled region above node 16. This behavior forms the basis for this evaluation.

Actual plant fuel rod power and exposure data will be evaluated to confirm this behavior exists.

Once it is concluded that the part-length rod is bounded by the corresponding section of the

ATTACHMENT H - ATRIUM-10 PART-LENGTH ROD EVALUATION L-003067 Rev. 2D Page H-3 of H-14 full-length rod then it can be concluded that the corresponding full-length rod gap fission product release fraction also bounds the part-length rod. Again, as stated above, this approach can be taken because the overall fission gas release fraction is a power and spatial average over the entire rod. In other words, based on the physical process of fission gas release, if the power and burnup of the portion of fuel in the full length rod is higher than the part-length rod it is reasonable to expect that the fission gas release from that portion of fuel is larger than part-length rod.

Should the evaluation of the fuel rod data indicate that part-length rods are not bounded by the full-length rods, then the dose consequence analysis would require modification accordingly.

Since the current analysis considers an Atrium-10 core damage fraction of 0.002244 based on 156 failed rods and 91 rods per bundle an adjustment will be made to be more consistent with this full-length rod equivalency methodology used in this minor revision. The damage of 156 rods is not affected by this revision.

INPUTS Inputs to this analysis are taken from References 1, 2, and 3 and generally correspond to Atrium-10 fuel design and operating data. The inputs are referenced throughout this minor revision.

ASSUMPTIONS

1. It is assumed that the data provided Reference 1 is a bounding representation of all part-length rods in LaSalle Cycles 14 through 16. This assumption is justified by Reference 1, which states the data was extracted from the bundle (34C217) projected to host the part-length rod with the highest average exposure at end of cycle 16.

NUMERIC ANALYSIS AND RESULTS LaSalle fuel rod data was provided in Reference 1. This data corresponds to the rod histories for the LaSalle 1 Cycle 16 fuel bundle with the highest part-length rod exposure at end of cycle. This bundle was chosen to be representative of and bounding over all Atrium-10 part-length rods through LaSalle 1 Cycle 16. This bundle can also apply to other cycles and LaSalle 2, provided the fuel rod characteristics remain bounding.

The data provided in Reference 1 was reformatted and plotted to allow for comparisons of part-length rods to full-length rods, which is shown in the Appendix to this minor revision. Fuel rod design layout and axial distributions are taken from Reference 2. A review of the data shows that each part-length rod is bounded by several full-length rods for the average of nodes 2 through 16. It was determined that for all cycle depletion steps each part-length rod has at least one neighbor that is more limiting in exposure, while in most situations a part-length rod will be bounded by several neighboring rods. With respect to power, part-length rods (9,2) and (2,2) have a few instances where the part-length rod has a slightly higher power. However, the difference is between approximately 0.2 and 0.3 kW/ft, which is small and not expected to have an adverse effect on the fission gas release fraction. Note, the observation that the part-length rods are bounded by some neighboring full-length rods applies to both burnup and power. The data shows that the part-length rod is not the leading rod in terms of power and burnup in the assembly.

ATTACHMENT H - ATRIUM-10 PART-LENGTH ROD EVALUATION L-003067 Rev. 2D Page H-4 of H-14 Based on the analysis of the fuel rod data the following conclusions are made:

1. Full-length rod power and burnup in the axial region corresponding to the part-length rod (i.e. nodes 2 through 16) is typically higher than part-length rods.
2. Since the full-length rod power and burnup is more limiting, it can be used to satisfy the Regulatory Guide 1.183 footnote 11 rod power and burnup restrictions.
3. Since the fuel rod power and burnup restrictions in Regulatory Guide 1.183 footnote 11 are controlled by core design (Reference 3) and since full-length rods have been shown to bound the part-length rods, it can be concluded that this FHA analysis meets the intent of Regulatory Guide 1.183. No increase in the Regulatory Guide 1.183 Table 3 non-LOCA gap release fractions is required for part-length rods provided they are bounded by a full length rod as described above.

To provide conservatism and make the Atrium-10 portion of the current analysis of record consistent with this approach, the core damage fraction will be modified. The Atrium-10 fuel assembly at LaSalle has 8 part-length rods that have 90 inch active fuel length and 83 full-length rods that have 149 inch active fuel length (Reference 2). Therefore, a part-length rod is equivalent to 60.4% of the fueled mass of a full-length rod. Using this basis, the total equivalent full length rods in the bundle is 87.83 (8*0.604+83) and the revised core damage fraction is 0.002325 (156/87.83/764) vs. the current analysis value of 0.002244. This increase of 3.6% effectively assumes that the PLRs contain the same source term inventory as the FLRs. However, the limiting GE14 core damage fraction of 0.002578 remains bounding over Atrium-10. Therefore, there is no change in the final dose consequences of the analysis of record. With the radial peaking factor of 1.7, the release fraction becomes 0.00395. The margin between the Atrium-10 and limiting dose consequences is therefore now about 10%

(0.004382/0.003952-1).

IDENTIFICATION OF COMPUTER PROGRAMS There are no computer programs utilized in this design analysis. Only Personal Productivity Software (PPS) Excel was used for data manipulation, reformatting and display.

REFERENCES

1. NF TODI NF151446, Rev. 0, LaSalle Unit 1 Cycle 16 Limiting Bundle Edits Supporting AST Evaluation.
2. LaSalle UFSAR
a. Section 4.2.2.3, Rev. 21, Fuel Bundle.
b. Table 4.2-4(d), Rev. 17, Data for the AREVA Atrium-10 Fuel Design.
c. Figure 4.2-3d, Rev. 15, Fuel Bundle FANP Atrium-10 Type..
3. NF-AB-110-2210, Rev 16, Core Loading Pattern Development.

ATTACHMENT H - ATRIUM-10 PART-LENGTH ROD EVALUATION L-003067 Rev. 2D Page H-5 of H-14 APPENDIX - FUEL ROD DATA The following figures show data corresponding to the LaSalle 1 Cycle 16 bundle with the limiting part-length rods. This bundle is evaluated as being representative and bounding with respect to all the LaSalle 1 Cycle 16 part-length rods that are projected to exceed a rod-average burnup of 62 GWD/MTU. This data was provided in Reference 1. A pin map is provided in Figure H-10 and a description of the burnup steps is provided in Figure H-11.

Note that all burnup information is presented in the figures below in GWD/ST. A burnup of 62 GWD/MTU is equal to 56.245 GWD/ST.

In the pin exposure/power plots, the part-length rods are indicated by thick solid outline. The pin exposure/power plots are showing the behavior of the rods at approximately the beginning, middle and end of the cycle. Pin powers are expressed in kW/ft.

Figure H LaSalle 1 Cycle 14 Pin Exposure Plots Step 4 = ~BOC, Step 8 = ~MOC, Step 16 = ~EOC STEP: 4 NODE: 2-16 AVE STEP: 8 NODE: 2-16 AVE STEP: 16 NODE: 2-16 AVE 1 2 3 4 5 6 7 8 9 10 1 2 3 4 5 6 7 8 9 10 1 2 3 4 5 6 7 8 9 10 1 8.6 8.9 8.9 8.8 8.7 8.4 8.6 9.0 8.9 8.6 1 16.1 16.7 17.0 16.9 16.7 16.2 16.5 17.1 16.8 16.1 1 23.2 24.4 25.2 25.7 25.5 24.9 25.6 26.5 26.4 25.6 2 8.9 8.7 7.0 3.4 6.6 6.6 3.4 7.1 8.8 8.9 2 16.7 16.5 13.8 8.2 12.9 13.0 8.3 14.0 16.7 16.8 2 24.4 25.1 22.1 16.3 21.3 21.5 17.0 23.5 27.3 27.1 3 8.9 7.0 3.2 5.7 6.3 6.3 5.6 4.0 7.1 9.0 3 17.0 13.8 7.5 11.5 12.4 12.4 11.5 9.0 14.1 17.1 3 25.2 22.1 15.2 19.8 21.0 21.4 20.8 18.3 24.4 28.4 4 8.8 3.4 5.7 6.3 7.3 7.4 3.4 5.5 3.4 8.6 4 16.9 8.2 11.5 12.5 14.2 14.5 8.6 11.5 8.4 16.6 4 25.7 16.3 19.8 21.4 24.4 25.3 19.1 21.8 18.7 28.3 5 8.7 6.6 6.3 7.3 3.6 6.6 8.6 5 16.7 12.9 12.4 14.2 9.0 13.3 16.5 5 25.5 21.3 21.0 24.4 20.3 24.2 28.3 6 8.4 6.6 6.3 7.4 7.3 7.0 8.7 6 16.2 13.0 12.4 14.5 14.7 14.1 16.8 6 24.9 21.5 21.4 25.3 26.2 25.2 28.7 7 8.6 3.4 5.6 3.4 6.9 3.7 8.4 7 16.5 8.3 11.5 8.6 14.1 9.3 16.3 7 25.6 17.0 20.8 19.1 25.6 20.6 28.1 8 9.0 7.1 4.0 5.5 3.6 7.3 6.9 3.5 7.4 9.2 8 17.1 14.0 9.0 11.5 9.0 14.7 14.1 8.9 14.8 17.6 8 26.5 23.5 18.3 21.8 20.3 26.2 25.6 20.1 26.4 29.7 9 8.9 8.8 7.1 3.4 6.6 7.0 3.7 7.4 8.7 9.1 9 16.8 16.7 14.1 8.4 13.3 14.1 9.3 14.8 16.6 17.1 9 26.4 27.3 24.4 18.7 24.2 25.2 20.6 26.4 28.5 28.5 10 8.6 8.9 9.0 8.6 8.6 8.7 8.4 9.2 9.1 8.7 10 16.1 16.8 17.1 16.6 16.5 16.8 16.3 17.6 17.1 16.4 10 25.6 27.1 28.4 28.3 28.3 28.7 28.1 29.7 28.5 27.1 FLR-to-PLR Neighbor Ratios 1 2 3 4 5 6 7 8 9 10 1 2 3 4 5 6 7 8 9 10 1 2 3 4 5 6 7 8 9 10 1 0.99 1.02 1.03 1.35 1.33 1.29 1.02 1.02 0.98 1 0.97 1.01 1.02 1.31 1.29 1.25 1.02 1.01 0.97 1 0.93 0.97 1.01 1.21 1.20 1.17 0.97 0.97 0.94 2 1.02 1.00 0.81 0.52 1.00 1.00 0.81 1.00 1.02 2 1.01 1.00 0.83 0.64 1.00 1.00 0.84 1.00 1.01 2 0.97 1.00 0.88 0.76 1.00 1.01 0.86 1.00 1.00 3 1.03 0.81 0.37 0.87 0.96 0.96 0.46 0.81 1.03 3 1.02 0.83 0.46 0.89 0.96 0.96 0.54 0.84 1.03 3 1.01 0.88 0.61 0.93 0.99 1.01 0.67 0.90 1.04 4 1.35 0.52 0.87 4 1.31 0.64 0.89 4 1.21 0.76 0.93 5 1.33 1.00 0.96 0.51 0.94 1.22 5 1.29 1.00 0.96 0.64 0.95 1.18 5 1.20 1.00 0.99 0.81 0.96 1.12 6 1.29 1.00 0.96 1.04 1.00 1.24 6 1.25 1.00 0.96 1.04 1.00 1.19 6 1.17 1.01 1.01 1.04 1.00 1.14 7 0.99 0.52 1.20 7 1.00 0.66 1.16 7 1.02 0.82 1.12 8 1.02 0.81 0.46 0.51 1.04 0.99 0.41 0.86 1.06 8 1.02 0.84 0.54 0.64 1.04 1.00 0.54 0.89 1.06 8 0.97 0.86 0.67 0.81 1.04 1.02 0.71 0.93 1.04 9 1.02 1.00 0.81 0.94 1.00 0.52 0.86 1.00 1.05 9 1.01 1.00 0.84 0.95 1.00 0.66 0.89 1.00 1.03 9 0.97 1.00 0.90 0.96 1.00 0.82 0.93 1.00 1.00 10 0.98 1.02 1.03 1.22 1.24 1.20 1.06 1.05 1.01 10 0.97 1.01 1.03 1.18 1.19 1.16 1.06 1.03 0.98 10 0.94 1.00 1.04 1.12 1.14 1.12 1.04 1.00 0.95

ATTACHMENT H - ATRIUM-10 PART-LENGTH ROD EVALUATION L-003067 Rev. 2D Page H-6 of H-14 Figure H LaSalle 1 Cycle 15 Pin Exposure Plots Step 17 = ~BOC, Step 24 = ~MOC, Step 33 = ~EOC STEP: 17 NODE: 2-16 AVE STEP: 24 NODE: 2-16 AVE STEP: 33 NODE: 2-16 AVE 1 2 3 4 5 6 7 8 9 10 1 2 3 4 5 6 7 8 9 10 1 2 3 4 5 6 7 8 9 10 1 23.3 24.5 25.3 25.8 25.6 25.0 25.6 26.6 26.5 25.7 1 33.3 34.9 36.5 37.4 37.2 36.4 37.3 38.3 37.7 36.5 1 40.0 41.7 43.7 45.0 44.8 43.9 45.0 46.0 45.1 43.8 2 24.5 25.1 22.2 16.3 21.3 21.6 17.0 23.6 27.4 27.2 2 34.9 36.3 33.1 27.2 32.0 32.4 28.4 35.2 39.4 38.5 2 41.7 43.6 40.5 34.7 39.4 39.9 36.2 43.0 47.4 46.0 3 25.3 22.2 15.3 19.9 21.1 21.5 20.9 18.4 24.5 28.5 3 36.5 33.1 25.9 30.3 31.6 32.3 32.1 30.0 36.4 40.6 3 43.7 40.5 33.4 37.7 39.1 40.0 40.0 38.1 44.5 48.7 4 25.8 16.3 19.9 21.5 24.4 25.4 19.2 21.9 18.7 28.4 4 37.4 27.2 30.3 32.2 36.0 37.5 31.7 33.8 31.0 40.9 4 45.0 34.7 37.7 39.8 44.1 45.8 40.3 42.2 39.4 49.3 5 25.6 21.3 21.1 24.4 20.4 24.3 28.3 5 37.2 32.0 31.6 36.0 33.3 36.4 40.8 5 44.8 39.4 39.1 44.1 42.1 44.7 49.3 6 25.0 21.6 21.5 25.4 26.3 25.3 28.8 6 36.4 32.4 32.3 37.5 38.9 37.5 41.3 6 43.9 39.9 40.0 45.8 47.4 45.9 49.8 7 25.6 17.0 20.9 19.2 25.7 20.7 28.2 7 37.3 28.4 32.1 31.7 38.2 33.3 40.5 7 45.0 36.2 40.0 40.3 46.8 41.9 48.8 8 26.6 23.6 18.4 21.9 20.4 26.3 25.7 20.2 26.5 29.8 8 38.3 35.2 30.0 33.8 33.3 38.9 38.2 32.9 38.9 42.1 8 46.0 43.0 38.1 42.2 42.1 47.4 46.8 41.7 47.4 50.4 9 26.5 27.4 24.5 18.7 24.3 25.3 20.7 26.5 28.6 28.6 9 37.7 39.4 36.4 31.0 36.4 37.5 33.3 38.9 40.8 40.2 9 45.1 47.4 44.5 39.4 44.7 45.9 41.9 47.4 49.0 48.0 10 25.7 27.2 28.5 28.4 28.3 28.8 28.2 29.8 28.6 27.2 10 36.5 38.5 40.6 40.9 40.8 41.3 40.5 42.1 40.2 38.3 10 43.8 46.0 48.7 49.3 49.3 49.8 48.8 50.4 48.0 46.0 FLR-to-PLR Neighbor Ratios 1 2 3 4 5 6 7 8 9 10 1 2 3 4 5 6 7 8 9 10 1 2 3 4 5 6 7 8 9 10 1 0.93 0.97 1.01 1.21 1.20 1.17 0.97 0.97 0.94 1 0.92 0.96 1.00 1.17 1.16 1.14 0.97 0.96 0.93 1 0.92 0.96 1.00 1.14 1.14 1.12 0.97 0.95 0.92 2 0.97 1.00 0.88 0.77 1.00 1.01 0.86 1.00 0.99 2 0.96 1.00 0.91 0.85 1.00 1.01 0.89 1.00 0.98 2 0.96 1.00 0.93 0.88 1.00 1.01 0.91 1.00 0.97 3 1.01 0.88 0.61 0.93 0.99 1.01 0.67 0.90 1.04 3 1.00 0.91 0.71 0.95 0.99 1.01 0.76 0.92 1.03 3 1.00 0.93 0.77 0.96 0.99 1.02 0.80 0.94 1.03 4 1.21 0.77 0.93 4 1.17 0.85 0.95 4 1.14 0.88 0.96 5 1.20 1.00 0.99 0.81 0.96 1.12 5 1.16 1.00 0.99 0.89 0.97 1.09 5 1.14 1.00 0.99 0.92 0.98 1.07 6 1.17 1.01 1.01 1.04 1.00 1.14 6 1.14 1.01 1.01 1.04 1.00 1.10 6 1.12 1.01 1.02 1.03 1.00 1.09 7 1.02 0.82 1.12 7 1.02 0.89 1.08 7 1.02 0.91 1.06 8 0.97 0.86 0.67 0.81 1.04 1.02 0.71 0.93 1.04 8 0.97 0.89 0.76 0.89 1.04 1.02 0.81 0.95 1.03 8 0.97 0.91 0.80 0.92 1.03 1.02 0.85 0.97 1.03 9 0.97 1.00 0.90 0.96 1.00 0.82 0.93 1.00 1.00 9 0.96 1.00 0.92 0.97 1.00 0.89 0.95 1.00 0.98 9 0.95 1.00 0.94 0.98 1.00 0.91 0.97 1.00 0.98 10 0.94 0.99 1.04 1.12 1.14 1.12 1.04 1.00 0.95 10 0.93 0.98 1.03 1.09 1.10 1.08 1.03 0.98 0.94 10 0.92 0.97 1.03 1.07 1.09 1.06 1.03 0.98 0.94

ATTACHMENT H - ATRIUM-10 PART-LENGTH ROD EVALUATION L-003067 Rev. 2D Page H-7 of H-14 Figure H LaSalle 1 Cycle 16 Pin Exposure Plots Step 34 = ~BOC, Step 41 = ~MOC, Step 50 = ~EOC STEP: 34 NODE: 2-16 AVE STEP: 41 NODE: 2-16 AVE STEP: 50 NODE: 2-16 AVE 1 2 3 4 5 6 7 8 9 10 1 2 3 4 5 6 7 8 9 10 1 2 3 4 5 6 7 8 9 10 1 39.9 41.6 43.6 44.9 44.7 43.8 45.0 45.9 45.1 43.8 1 45.3 47.1 49.4 51.0 50.7 49.7 50.8 51.5 50.3 48.8 1 49.5 51.3 53.7 55.5 55.2 54.1 55.2 55.7 54.2 52.7 2 41.6 43.5 40.4 34.6 39.3 39.9 36.2 43.0 47.4 46.0 2 47.1 49.3 46.1 40.4 45.0 45.5 41.9 48.6 52.8 51.0 2 51.3 53.6 50.5 44.8 49.2 49.8 46.3 52.8 56.9 54.8 3 43.6 40.4 33.3 37.7 39.1 40.0 40.0 38.1 44.5 48.7 3 49.4 46.1 39.0 43.2 44.6 45.5 45.5 43.7 50.0 53.9 3 53.7 50.5 43.4 47.5 48.8 49.7 49.8 48.0 54.1 57.9 4 44.9 34.6 37.7 39.8 44.1 45.8 40.3 42.2 39.4 49.3 4 51.0 40.4 43.2 45.3 49.6 51.4 46.0 47.8 45.0 54.6 4 55.5 44.8 47.5 49.4 53.8 55.6 50.3 52.1 49.2 58.7 5 44.7 39.3 39.1 44.1 42.1 44.8 49.3 5 50.7 45.0 44.6 49.6 47.8 50.2 54.6 5 55.2 49.2 48.8 53.8 52.1 54.3 58.7 6 43.8 39.9 40.0 45.8 47.4 45.9 49.8 6 49.7 45.5 45.5 51.4 52.9 51.3 55.1 6 54.1 49.8 49.7 55.6 57.0 55.4 59.1 7 45.0 36.2 40.0 40.3 46.8 42.0 48.9 7 50.8 41.9 45.5 46.0 52.2 47.5 54.1 7 55.2 46.3 49.8 50.3 56.3 51.6 58.0 8 45.9 43.0 38.1 42.2 42.1 47.4 46.8 41.7 47.4 50.4 8 51.5 48.6 43.7 47.8 47.8 52.9 52.2 47.3 52.7 55.6 8 55.7 52.8 48.0 52.1 52.1 57.0 56.3 51.4 56.7 59.4 9 45.1 47.4 44.5 39.4 44.8 45.9 42.0 47.4 49.0 48.0 9 50.3 52.8 50.0 45.0 50.2 51.3 47.5 52.7 54.1 52.9 9 54.2 56.9 54.1 49.2 54.3 55.4 51.6 56.7 57.9 56.6 10 43.8 46.0 48.7 49.3 49.3 49.8 48.9 50.4 48.0 46.0 10 48.8 51.0 53.9 54.6 54.6 55.1 54.1 55.6 52.9 50.9 10 52.7 54.8 57.9 58.7 58.7 59.1 58.0 59.4 56.6 54.6 FLR-to-PLR Neighbor Ratios 1 2 3 4 5 6 7 8 9 10 1 2 3 4 5 6 7 8 9 10 1 2 3 4 5 6 7 8 9 10 1 0.92 0.96 1.00 1.14 1.14 1.11 0.97 0.95 0.92 1 0.92 0.96 1.00 1.13 1.13 1.11 0.98 0.95 0.92 1 0.92 0.96 1.00 1.13 1.12 1.10 0.98 0.95 0.93 2 0.96 1.00 0.93 0.88 1.00 1.01 0.91 1.00 0.97 2 0.96 1.00 0.94 0.90 1.00 1.01 0.92 1.00 0.97 2 0.96 1.00 0.94 0.91 1.00 1.01 0.93 1.00 0.96 3 1.00 0.93 0.77 0.96 0.99 1.02 0.80 0.94 1.03 3 1.00 0.94 0.79 0.96 0.99 1.01 0.83 0.95 1.02 3 1.00 0.94 0.81 0.96 0.99 1.01 0.84 0.95 1.02 4 1.14 0.88 0.96 4 1.13 0.90 0.96 4 1.13 0.91 0.96 5 1.14 1.00 0.99 0.92 0.98 1.07 5 1.13 1.00 0.99 0.93 0.98 1.07 5 1.12 1.00 0.99 0.94 0.98 1.06 6 1.11 1.01 1.02 1.03 1.00 1.09 6 1.11 1.01 1.01 1.03 1.00 1.07 6 1.10 1.01 1.01 1.03 1.00 1.07 7 1.02 0.91 1.06 7 1.02 0.93 1.05 7 1.02 0.93 1.05 8 0.97 0.91 0.80 0.92 1.03 1.02 0.85 0.97 1.03 8 0.98 0.92 0.83 0.93 1.03 1.02 0.87 0.97 1.03 8 0.98 0.93 0.84 0.94 1.03 1.02 0.89 0.98 1.03 9 0.95 1.00 0.94 0.98 1.00 0.91 0.97 1.00 0.98 9 0.95 1.00 0.95 0.98 1.00 0.93 0.97 1.00 0.98 9 0.95 1.00 0.95 0.98 1.00 0.93 0.98 1.00 0.98 10 0.92 0.97 1.03 1.07 1.09 1.06 1.03 0.98 0.94 10 0.92 0.97 1.02 1.07 1.07 1.05 1.03 0.98 0.94 10 0.93 0.96 1.02 1.06 1.07 1.05 1.03 0.98 0.94

ATTACHMENT H - ATRIUM-10 PART-LENGTH ROD EVALUATION L-003067 Rev. 2D Page H-8 of H-14 Figure H LaSalle 1 Cycle 14 Pin Power Plots Step 4 = ~BOC, Step 8 = ~MOC, Step 16 = ~EOC STEP: 4 NODE: 2-16 AVE STEP: 8 NODE: 2-16 AVE STEP: 16 NODE: 2-16 AVE 1 2 3 4 5 6 7 8 9 10 1 2 3 4 5 6 7 8 9 10 1 2 3 4 5 6 7 8 9 10 1 7.4 7.5 7.6 7.6 7.5 7.3 7.4 7.7 7.7 7.5 1 7.1 7.3 7.6 7.7 7.6 7.4 7.5 7.6 7.4 7.0 1 5.8 5.7 5.9 6.0 6.0 5.8 6.0 5.9 5.6 5.3 2 7.5 7.3 6.2 3.7 5.8 5.8 3.7 6.3 7.5 7.7 2 7.3 7.4 6.6 5.2 6.2 6.3 5.3 6.7 7.5 7.3 2 5.7 5.6 5.2 4.5 4.9 5.0 4.7 5.3 5.6 5.4 3 7.5 6.2 3.4 5.2 5.5 5.6 5.2 4.1 6.3 7.8 3 7.6 6.6 4.7 5.8 5.9 6.0 6.0 5.3 6.8 7.6 3 5.8 5.2 4.2 4.6 4.7 4.8 5.0 4.8 5.3 5.6 4 7.5 3.7 5.1 5.5 6.3 6.4 3.8 5.2 3.8 7.5 4 7.7 5.2 5.8 6.0 6.6 6.9 5.6 6.3 5.4 7.6 4 6.0 4.4 4.6 4.7 5.1 5.4 5.0 5.2 4.8 5.7 5 7.4 5.7 5.5 6.3 4.0 6.0 7.5 5 7.6 6.2 5.9 6.6 5.9 6.7 7.6 5 5.9 4.9 4.7 5.1 5.1 5.3 5.7 6 7.2 5.8 5.5 6.4 6.5 6.3 7.6 6 7.4 6.3 6.0 6.9 7.1 6.9 7.6 6 5.8 5.0 4.8 5.4 5.5 5.3 5.7 7 7.3 3.7 5.1 3.7 6.3 4.1 7.4 7 7.6 5.3 6.0 5.6 7.0 6.0 7.5 7 5.9 4.6 4.9 4.9 5.5 4.9 5.6 8 7.6 6.2 4.0 5.2 3.9 6.5 6.3 3.9 6.6 7.9 8 7.6 6.7 5.3 6.3 5.9 7.2 7.0 5.8 7.2 7.8 8 5.8 5.2 4.7 5.1 5.1 5.5 5.4 4.9 5.4 5.6 9 7.5 7.3 6.2 3.7 5.9 6.2 4.1 6.6 7.4 7.8 9 7.4 7.5 6.8 5.4 6.7 6.9 6.0 7.2 7.5 7.4 9 5.5 5.5 5.2 4.7 5.2 5.3 4.9 5.4 5.4 5.3 10 7.3 7.5 7.6 7.4 7.4 7.5 7.3 7.8 7.7 7.5 10 7.0 7.3 7.6 7.6 7.6 7.6 7.5 7.8 7.4 7.1 10 5.1 5.2 5.5 5.6 5.6 5.6 5.5 5.5 5.2 5.0 FLR-to-PLR Neighbor Ratios 1 2 3 4 5 6 7 8 9 10 1 2 3 4 5 6 7 8 9 10 1 2 3 4 5 6 7 8 9 10 1 1.01 1.03 1.04 1.31 1.30 1.26 1.03 1.03 1.00 1 0.96 0.99 1.02 1.24 1.22 1.19 1.02 0.99 0.94 1 1.03 1.02 1.05 1.22 1.20 1.18 1.05 1.00 0.94 2 1.02 1.00 0.84 0.64 1.00 1.00 0.84 1.00 1.03 2 0.99 1.00 0.90 0.83 1.00 1.01 0.90 1.00 0.98 2 1.02 1.00 0.93 0.90 1.00 1.01 0.95 1.00 0.96 3 1.03 0.84 0.47 0.89 0.95 0.96 0.55 0.85 1.04 3 1.02 0.90 0.64 0.92 0.95 0.97 0.71 0.91 1.02 3 1.04 0.92 0.76 0.94 0.95 0.98 0.85 0.95 1.00 4 1.31 0.64 0.90 4 1.24 0.83 0.92 4 1.22 0.90 0.94 5 1.29 1.00 0.95 0.63 0.95 1.19 5 1.22 1.00 0.95 0.86 0.97 1.10 5 1.20 1.00 0.95 0.96 0.99 1.06 6 1.25 1.00 0.96 1.03 1.00 1.20 6 1.19 1.01 0.97 1.04 1.00 1.11 6 1.17 1.01 0.98 1.04 1.00 1.06 7 1.00 0.65 1.18 7 1.02 0.87 1.09 7 1.02 0.92 1.04 8 1.03 0.85 0.55 0.63 1.04 1.00 0.53 0.89 1.07 8 1.02 0.90 0.71 0.86 1.04 1.02 0.77 0.96 1.04 8 1.05 0.95 0.86 0.96 1.04 1.03 0.91 1.00 1.04 9 1.02 1.00 0.85 0.95 1.00 0.66 0.89 1.00 1.05 9 0.99 1.00 0.91 0.97 1.00 0.87 0.96 1.00 0.99 9 0.99 1.00 0.95 0.99 1.00 0.93 1.00 1.00 0.97 10 1.00 1.02 1.04 1.18 1.20 1.17 1.06 1.04 1.02 10 0.94 0.98 1.02 1.10 1.11 1.09 1.04 0.99 0.95 10 0.93 0.95 1.00 1.05 1.06 1.04 1.02 0.97 0.93

ATTACHMENT H - ATRIUM-10 PART-LENGTH ROD EVALUATION L-003067 Rev. 2D Page H-9 of H-14 Figure H LaSalle 1 Cycle 15 Pin Power Plots Step 17 = ~BOC, Step 24 = ~MOC, Step 33 = ~EOC STEP: 17 NODE: 2-16 AVE STEP: 24 NODE: 2-16 AVE STEP: 33 NODE: 2-16 AVE 1 2 3 4 5 6 7 8 9 10 1 2 3 4 5 6 7 8 9 10 1 2 3 4 5 6 7 8 9 10 1 5.4 5.5 5.7 5.9 5.8 5.8 5.9 5.9 5.7 5.4 1 5.5 5.5 5.7 5.9 5.8 5.7 5.8 5.7 5.3 5.0 1 4.6 4.5 4.6 4.7 4.7 4.6 4.7 4.6 4.3 4.1 2 5.5 5.6 5.3 4.6 5.2 5.3 4.9 5.6 5.9 5.6 2 5.5 5.5 5.3 4.7 5.1 5.1 4.8 5.3 5.4 5.0 2 4.5 4.4 4.3 3.9 4.2 4.2 3.9 4.3 4.2 4.0 3 5.6 5.3 4.5 5.0 5.1 5.3 5.5 5.3 5.8 6.0 3 5.7 5.3 4.6 4.9 4.9 5.0 5.1 4.9 5.2 5.2 3 4.6 4.3 3.8 4.1 4.0 4.1 4.2 4.0 4.2 4.2 4 5.8 4.6 5.0 5.2 5.8 6.1 5.6 5.8 5.4 6.2 4 5.9 4.7 4.9 5.0 5.2 5.4 5.0 5.3 4.8 5.3 4 4.8 3.9 4.1 4.1 4.2 4.3 4.0 4.3 3.9 4.2 5 5.8 5.1 5.1 5.7 5.8 5.9 6.2 5 5.8 5.1 4.9 5.2 5.0 5.2 5.3 5 4.7 4.2 4.1 4.2 4.0 4.2 4.2 6 5.7 5.2 5.3 6.1 6.3 6.0 6.2 6 5.7 5.1 5.0 5.4 5.4 5.2 5.3 6 4.7 4.2 4.1 4.3 4.3 4.2 4.2 7 5.8 4.9 5.5 5.6 6.2 5.6 6.1 7 5.8 4.8 5.1 5.0 5.4 4.8 5.2 7 4.7 3.9 4.2 4.0 4.3 3.8 4.1 8 5.9 5.6 5.3 5.9 5.8 6.3 6.3 5.7 6.2 6.2 8 5.6 5.3 4.9 5.3 5.0 5.4 5.4 4.9 5.3 5.2 8 4.6 4.3 4.0 4.3 4.0 4.3 4.3 3.9 4.1 4.1 9 5.7 6.0 5.9 5.5 6.1 6.1 5.7 6.2 6.2 5.9 9 5.3 5.4 5.3 4.8 5.3 5.3 4.9 5.3 5.1 4.9 9 4.3 4.2 4.2 3.8 4.2 4.1 3.8 4.1 4.0 3.9 10 5.5 5.8 6.2 6.4 6.4 6.4 6.3 6.4 5.9 5.6 10 5.0 5.0 5.3 5.4 5.4 5.3 5.2 5.2 4.9 4.8 10 4.1 4.1 4.2 4.2 4.2 4.2 4.1 4.1 3.9 4.0 FLR-to-PLR Neighbor Ratios 1 2 3 4 5 6 7 8 9 10 1 2 3 4 5 6 7 8 9 10 1 2 3 4 5 6 7 8 9 10 1 0.97 0.98 1.02 1.13 1.13 1.11 0.99 0.95 0.90 1 1.01 1.01 1.04 1.15 1.14 1.12 1.06 0.98 0.93 1 1.04 1.03 1.05 1.14 1.13 1.11 1.08 1.01 0.98 2 0.98 1.00 0.95 0.90 1.00 1.02 0.95 1.00 0.95 2 1.00 1.00 0.96 0.93 1.00 1.01 0.99 1.00 0.93 2 1.03 1.00 0.97 0.93 1.00 1.00 1.01 1.00 0.95 3 1.02 0.95 0.81 0.97 0.99 1.02 0.88 0.98 1.01 3 1.03 0.96 0.84 0.97 0.96 0.98 0.92 0.98 0.97 3 1.05 0.97 0.86 0.98 0.97 0.98 0.94 0.99 0.98 4 1.13 0.90 0.98 4 1.14 0.93 0.97 4 1.14 0.93 0.97 5 1.12 1.00 1.00 0.96 0.99 1.03 5 1.13 1.00 0.97 0.96 1.00 1.01 5 1.13 1.00 0.97 0.96 1.00 1.01 6 1.11 1.02 1.03 1.04 1.00 1.03 6 1.11 1.01 0.98 1.03 1.00 1.01 6 1.11 1.00 0.98 1.03 1.00 1.01 7 1.04 0.93 1.02 7 1.03 0.92 0.99 7 1.02 0.92 0.99 8 0.98 0.94 0.88 0.95 1.04 1.03 0.92 1.00 1.01 8 1.05 0.99 0.91 0.96 1.03 1.02 0.96 1.03 1.01 8 1.08 1.01 0.94 0.96 1.03 1.03 0.97 1.03 1.02 9 0.95 1.00 0.98 0.99 1.00 0.93 1.01 1.00 0.95 9 0.98 1.00 0.98 1.00 1.00 0.93 1.03 1.00 0.95 9 1.01 1.00 0.99 1.00 1.00 0.92 1.03 1.00 0.98 10 0.91 0.96 1.03 1.04 1.05 1.03 1.03 0.96 0.91 10 0.93 0.94 0.98 1.02 1.02 1.00 1.02 0.96 0.94 10 0.98 0.96 0.99 1.01 1.01 1.00 1.02 0.98 0.99

ATTACHMENT H - ATRIUM-10 PART-LENGTH ROD EVALUATION L-003067 Rev. 2D Page H-10 of H-14 Figure H LaSalle 1 Cycle 16 Pin Power Plots Step 34 = ~BOC, Step 41 = ~MOC, Step 50 = ~EOC STEP: 34 NODE: 2-16 AVE STEP: 41 NODE: 2-16 AVE STEP: 50 NODE: 2-16 AVE 1 2 3 4 5 6 7 8 9 10 1 2 3 4 5 6 7 8 9 10 1 2 3 4 5 6 7 8 9 10 1 2.8 2.9 3.1 3.2 3.3 3.3 3.4 3.4 3.2 3.2 1 2.4 2.6 2.8 2.9 3.0 3.0 3.1 3.1 3.0 3.1 1 1.7 1.8 1.9 2.1 2.1 2.1 2.2 2.2 2.2 2.3 2 2.6 2.7 2.7 2.5 2.8 2.9 2.7 3.0 3.1 3.0 2 2.3 2.4 2.5 2.4 2.6 2.7 2.6 2.9 2.9 2.9 2 1.6 1.6 1.7 1.6 1.8 1.9 1.8 2.0 2.0 2.1 3 2.6 2.5 2.3 2.6 2.6 2.7 2.8 2.8 3.0 3.1 3 2.2 2.3 2.2 2.4 2.5 2.6 2.7 2.6 2.8 2.9 3 1.5 1.6 1.5 1.7 1.7 1.8 1.9 1.8 2.0 2.1 4 2.6 2.2 2.4 2.5 2.7 2.8 2.7 2.9 2.7 3.1 4 2.2 2.0 2.3 2.3 2.5 2.6 2.5 2.7 2.6 2.9 4 1.5 1.4 1.6 1.6 1.7 1.8 1.7 1.9 1.8 2.1 5 2.5 2.3 2.4 2.6 2.7 2.9 3.0 5 2.1 2.1 2.2 2.3 2.5 2.7 2.9 5 1.4 1.4 1.5 1.6 1.7 1.9 2.0 6 2.4 2.3 2.3 2.6 2.8 2.8 3.0 6 2.1 2.1 2.1 2.3 2.6 2.6 2.8 6 1.4 1.4 1.5 1.6 1.8 1.8 2.0 7 2.3 2.1 2.3 2.3 2.7 2.5 2.8 7 2.0 1.9 2.1 2.1 2.5 2.4 2.7 7 1.3 1.3 1.4 1.4 1.7 1.6 1.9 8 2.2 2.2 2.2 2.4 2.3 2.5 2.6 2.4 2.7 2.8 8 1.9 2.0 1.9 2.2 2.1 2.3 2.4 2.3 2.5 2.6 8 1.3 1.3 1.3 1.5 1.4 1.6 1.6 1.5 1.7 1.8 9 2.0 2.1 2.2 2.1 2.3 2.4 2.2 2.5 2.5 2.6 9 1.7 1.8 1.9 1.9 2.1 2.1 2.0 2.3 2.3 2.4 9 1.2 1.2 1.3 1.3 1.4 1.4 1.4 1.6 1.6 1.7 10 1.9 1.9 2.1 2.2 2.2 2.3 2.3 2.3 2.3 2.5 10 1.6 1.7 1.8 1.9 2.0 2.0 2.1 2.1 2.2 2.4 10 1.1 1.2 1.2 1.3 1.3 1.4 1.4 1.5 1.5 1.7 FLR-to-PLR Neighbor Ratios 1 2 3 4 5 6 7 8 9 10 1 2 3 4 5 6 7 8 9 10 1 2 3 4 5 6 7 8 9 10 1 1.03 1.07 1.15 1.16 1.17 1.18 1.09 1.04 1.03 1 1.01 1.07 1.15 1.12 1.15 1.16 1.08 1.05 1.07 1 1.04 1.10 1.18 1.13 1.15 1.17 1.09 1.08 1.13 2 0.96 1.00 1.02 0.91 1.00 1.03 0.99 1.00 0.98 2 0.95 1.00 1.04 0.90 1.00 1.03 0.99 1.00 1.01 2 0.96 1.00 1.05 0.90 1.00 1.03 0.99 1.00 1.04 3 0.96 0.94 0.87 0.92 0.94 0.98 0.90 0.97 1.00 3 0.93 0.95 0.90 0.92 0.95 0.98 0.91 0.98 1.01 3 0.93 0.95 0.91 0.92 0.94 0.98 0.91 0.97 1.03 4 1.09 0.95 1.04 4 1.05 0.96 1.07 4 1.04 0.96 1.08 5 1.06 1.00 1.02 0.96 1.02 1.07 5 1.01 1.00 1.03 0.95 1.03 1.08 5 1.00 1.00 1.04 0.95 1.03 1.11 6 1.02 0.98 1.00 1.00 1.00 1.05 6 0.97 0.98 1.01 0.99 1.00 1.06 6 0.96 0.97 1.02 0.99 1.00 1.08 7 0.97 0.90 1.01 7 0.96 0.90 1.02 7 0.96 0.90 1.04 8 1.05 1.05 1.02 0.98 1.07 1.10 0.98 1.07 1.11 8 1.03 1.06 1.05 0.97 1.08 1.11 0.97 1.07 1.13 8 1.03 1.06 1.08 0.98 1.09 1.12 0.97 1.07 1.16 9 0.95 1.00 1.03 0.98 1.00 0.95 1.00 1.00 1.03 9 0.94 1.00 1.05 0.97 1.00 0.96 0.99 1.00 1.06 9 0.96 1.00 1.06 0.97 1.00 0.96 0.98 1.00 1.09 10 0.88 0.91 0.97 0.93 0.95 0.96 0.94 0.93 0.99 10 0.89 0.92 0.98 0.92 0.95 0.97 0.93 0.94 1.03 10 0.92 0.94 1.00 0.92 0.95 0.98 0.93 0.96 1.07

ATTACHMENT H-ATRIUM-10 PART-LENGTH ROD EVALUATION L-003067 Rev. 20 Page H-11 of H-14 Figure H LaSalle 1 Cycles 14 thru 16 Representative Exposure Histories Lasalle 1 PLR to FLR Comparisons, Rod Segment Lasalle 1 PLR to FLR Comparisons, Rod Segment Average Exposures Over Nodes 2-16 Average Exposures OVer Nodes 2-16 6

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ATTACHMENT H-ATRIUM-10 PART-LENGTH ROD EVALUATION L-003067 Rev. 20 Page H-12 of H-14 Figure H LaSalle 1 Cycles 14 thru 16 Representative Power Histories Lasalle 1 PLR to FLR Comparisons, Rod Segment Average Powers Lasalle 1 PLR to FLR Comparisons, Rod Segment Average Powers Over Nodes 2-16 Over Nodes 2-16 90 . - - - - - - - - - - - - - - - - - - - - - - - - - - 10.0 ~---------------------

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ATTACHMENT H - ATRIUM-10 PART-LENGTH ROD EVALUATION L-003067 Rev. 2D Page H-13 of H-14 Figure H LaSalle Atrium-10 Pin Layout This geometry in this figure is derived from Reference 2.c Moderate fuel rod burnup Leading FLR burnup 1 2 3 4 5 6 7 8 9 10 1

2 P P P 3

4 5 P 6 P 7

8 9 P P P 10

ATTACHMENT H - ATRIUM-10 PART-LENGTH ROD EVALUATION L-003067 Rev. 2D Page H-14 of H-14 Figure H-11 Burnup Step Map to Cycle Exposure CYCLE CYCLE CYCLE BURNUP EXPOSURE BURNUP EXPOSURE BURNUP EXPOSURE STEP CYCLE (GWD/ST) STEP CYCLE (GWD/ST) STEP CYCLE (GWD/ST) 1 14 0.06 17 15 0.057 34 16 0.036 2 14 1.78 18 15 1.826 35 16 1.732 3 14 1.801 19 15 1.846 36 16 1.775 4 14 4.381 20 15 4.129 37 16 4.299 5 14 4.402 21 15 4.149 38 16 4.342 6 14 6.504 22 15 6.388 39 16 6.33 7 14 6.549 23 15 6.432 40 16 6.373 8 14 8.77 24 15 8.831 41 16 8.735 9 14 8.792 25 15 8.874 42 16 8.778 10 14 10.07 26 15 9.971 43 16 10.178 11 14 10.115 27 15 9.99 44 16 10.221 12 14 12.552 28 15 12.436 45 16 12.7 13 14 12.597 29 15 12.478 46 16 12.701 14 14 15.057 30 15 14.714 47 16 14.05 15 14 15.078 31 15 14.779 48 16 14.051 16 14 16.393 32 15 15.866 49 16 15.53 33 15 16.012 50 16 16.369

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