ML15253A668

NRR E-mail Capture - (TAC Nos. MF5734/MF5735) Revisions to Draft RAIs for License Amendment Request Spent Fuel Storage Pool Criticality Methodology and Proposed Change TS 4.3.1,

ML15253A668
Person / Time
Site:	Dresden
Issue date:	09/02/2015
From:	Haskell R Plant Licensing Branch III
To:	Byam T Exelon Corp
References
TAC MF5734, TAC MF5735
Download: ML15253A668 (9)

Text

1 NRR-PMDAPEm Resource From:

Haskell, Russell Sent:

Wednesday, September 02, 2015 8:46 AM To:

Byam, Timothy A:(GenCo-Nuc)

Cc:

Mathews, Mitchel A:(GenCo-Nuc); Nicely, Ken M.:(GenCo-Nuc); Brown, Eva; Pulvirenti, April; Krepel, Scott; Wood, Kent; Jackson, Christopher; Obodoako, Aloysius; Yoder, Matthew; Kulesa, Gloria; Tate, Travis

Subject:

(TAC Nos. MF5734/MF5735) Revisions to draft RAI's for License Amendment Request RE:

Spent Fuel Storage Pool Criticality Methodology and Proposed Change TS 4.3.1, Attachments:

Revised Draft RAIs for Dresden SFP LAR.pdf Mr. Tim Byam:

By application dated December 30, 2014, Exelon Generation Company, LLC (EGC, the licensee) submitted a license amendment request to modify the DNPS unit 2 and 3 SFP criticality safety analysis (CSA) methodology and to the DNPS Technical Specification (TS) 4.3.1, "Criticality.", NRC Agencywide Documents Access and Management System (ADAMS) Accession No. ML14364A100, supplemented by NRC letter dated April 24, 2015 (ML15105A550) and EGC letter dated May 8, 2015 (ML15128A305).

On August 19, 2015, the NRC and EGC had a teleconference to discuss preliminary questions (i.e., draft requests for additional information [RAIs]) related to the Dresden SFP criticality analysis. The purpose of the teleconference was to have several of the staff questions clarified which resulted in several revisions (see attached).

Additionally, as summarized in a phone conversation we had with you yesterday, the NRC will follow-up with a formal transmittal of the RAIs. The transmittal will establish an expectation for EGC to respond to the NRC regarding all RAIs by October 15, 2015, COB. This timeline will facilitate the completion of the staffs safety evaluation in response to the licensees submittal. Furthermore, the staff has informed you that we are not aware these RAIs include any vender-specific proprietary content; these questions will docketed and added to ADAMS for public review.

Thank you.

The NRC appreciates EGCs participation (incl. support personnel) during the August 19th teleconference.

(Below is a listing of attendees.)

NRC staff:

Kent Wood - Reactor Safety Branch (SRXB); Scott Krepel - Reactor Safety Branch (SRXB); Aloysius Obodoako - Steam Generator Tube Integrity & Chemical Engineering Branch (ESGB); Matthew Yoder - Steam Generator Tube Integrity & Chemical Engineering Branch (ESGB); Eva Brown - Sr. Project Manager (DORL);

April Pulvirenti - Project Manager (DORL); Russ Haskell - Project Manager (DORL)

Licensee staff:

Timothy Byam (Exelon); David Phegley (Exelon); Jill Fisher (Exelon); Jeff Dunlap (Exelon); Mitchell Mathews (Exelon); Bret Brickner (Holtec)

Russell S. Haskell II United States Nuclear Regulatory Commission (NRC)

Reactor Systems Engineer NRR/DORL/LPL 3-2

2 P.O.C.:Dresden Nuclear Power Stations 2,3 Russell.Haskell@NRC.Gov; (301) 415-1129; Office: O-7D21; Mail-Stop O-8G9A From: Byam, Timothy A:(GenCo-Nuc) [1]

Sent: Wednesday, August 12, 2015 10:21 AM To: Haskell, Russell Cc: Mathews, Mitchel A:(GenCo-Nuc) ; Nicely, Ken M.:(GenCo-Nuc)

Subject:

[External_Sender] RE: [EXTERNAL] (TAC Nos. MF5734/MF5735) Draft RAI's for License Amendment Request RE:

Spent Fuel Storage Pool Criticality Methodology and Proposed Change TS 4.3.1,

Russ, Exelon and Holtec have reviewed the subject draft RAIs. Based on that review we have determined that a clarification call would be helpful to ensure we provide the information that the reviewer is looking for. We will be available for a clarification call any time after Friday August 14, 2015. Please let me know when the NRC is available for a call and we will be prepared to support that. Thanks, Tim Byam From: Haskell, Russell [2]

Sent: Wednesday, July 29, 2015 2:44 PM To: Byam, Timothy A:(GenCo-Nuc)

Cc: Mathews, Mitchel A:(GenCo-Nuc); Nicely, Ken M.:(GenCo-Nuc)

Subject:

[EXTERNAL] (TAC Nos. MF5734/MF5735) Draft RAI's for License Amendment Request RE: Spent Fuel Storage Pool Criticality Methodology and Proposed Change TS 4.3.1, Mr. Byam:

On December 30, 2014, Exelon Generation Company, LLC (EGC, the licensee) submitted a license amendment request for the Dresden Nuclear Power Station (DNPS), Units 2 and 3, (RS-14-310), NRC Agencywide Documents Access and Management System (ADAMS) Accession No. ML14364A100, supplemented by NRC letter dated April 24, 2015 (ML15105A550) and EGC letter dated May 8, 2015 (RS 119) ML15128A305. The license amendment request proposes to change the criticality safety analysis (CSA) methodology in the unit 2 and 3 spent fuel pools, and would change the DNPS Technical Specification (TS) 4.3.1, "Criticality."

The purpose of this email is to inform EGC of the attached preliminary questions (i.e., draft requests for additional information [RAIs]), which are being provided for you to determine if any clarifications are needed, prior to the NRCs formal issuance (docketing) of these RAIs.

As necessary, I will coordinate a teleconference between NRC staff and EGC (et al.), to address any clarifications or any questions/comments EGC may have regarding these RAIs.

To aid in staff scheduling for both the NRC and EGC, your response for the need for a teleconference is appreciated by August 17, 2015 COB.

Thank you.

Russell S. Haskell II United States Nuclear Regulatory Commission (NRC)

Reactor Systems Engineer NRR/DORL/LPL 3-2 P.O.C.:Dresden Nuclear Power Stations 2,3 Russell.Haskell@NRC.Gov; (301) 415-1129; Office: O-7D21; Mail-Stop O-8G9A

3 This Email message and any attachment may contain information that is proprietary, legally privileged, confidential and/or subject to copyright belonging to Exelon Corporation or its affiliates ("Exelon"). This Email is intended solely for the use of the person(s) to which it is addressed. If you are not an intended recipient, or the employee or agent responsible for delivery of this Email to the intended recipient(s), you are hereby notified that any dissemination, distribution or copying of this Email is strictly prohibited. If you have received this message in error, please immediately notify the sender and permanently delete this Email and any copies.

Exelon policies expressly prohibit employees from making defamatory or offensive statements and infringing any copyright or any other legal right by Email communication. Exelon will not accept any liability in respect of such communications. -EXCIP

Hearing Identifier:

NRR_PMDA Email Number:

2372 Mail Envelope Properties (34725d316e2248858fd701ec2002424b)

Subject:

(TAC Nos. MF5734/MF5735) Revisions to draft RAI's for License Amendment Request RE: Spent Fuel Storage Pool Criticality Methodology and Proposed Change TS 4.3.1, Sent Date:

9/2/2015 8:46:28 AM Received Date:

9/2/2015 8:46:30 AM From:

Haskell, Russell Created By:

Russell.Haskell@nrc.gov Recipients:

"Mathews, Mitchel A:(GenCo-Nuc)" <Mitchel.Mathews@exeloncorp.com>

Tracking Status: None "Nicely, Ken M.:(GenCo-Nuc)" <ken.nicely@exeloncorp.com>

Tracking Status: None "Brown, Eva" <Eva.Brown@nrc.gov>

Tracking Status: None "Pulvirenti, April" <April.Pulvirenti@nrc.gov>

Tracking Status: None "Krepel, Scott" <Scott.Krepel@nrc.gov>

Tracking Status: None "Wood, Kent" <Kent.Wood@nrc.gov>

Tracking Status: None "Jackson, Christopher" <Christopher.Jackson@nrc.gov>

Tracking Status: None "Obodoako, Aloysius" <Aloysius.Obodoako@nrc.gov>

Tracking Status: None "Yoder, Matthew" <Matthew.Yoder@nrc.gov>

Tracking Status: None "Kulesa, Gloria" <Gloria.Kulesa@nrc.gov>

Tracking Status: None "Tate, Travis" <Travis.Tate@nrc.gov>

Tracking Status: None "Byam, Timothy A:(GenCo-Nuc)" <timothy.byam@exeloncorp.com>

Tracking Status: None Post Office:

HQPWMSMRS03.nrc.gov Files Size Date & Time MESSAGE 6189 9/2/2015 8:46:30 AM Revised Draft RAIs for Dresden SFP LAR.pdf 146482 Options Priority:

Standard Return Notification:

No Reply Requested:

No Sensitivity:

Normal Expiration Date:

Recipients Received:

ATTACHMENT REQUEST FOR ADDITIONAL INFORMATION SPENT FUEL POOL SURVEILLANCE COUPON PROGRAM EXELON GENERATION COMPANY, LLC DRESDEN NUCLEAR POWER STATION, UNITS 1 AND 2 DOCKET NOS. 50-237 AND 50-249 Section 50.68 to Title 10 of the Code of Federal Regulations (10 CFR 50.68) states that the k-effective (Keff) of the spent fuel storage racks loaded with fuel of the maximum fuel assembly reactivity must not exceed 0.95, at a 95 percent probability, 95 percent confidence level, if flooded with unborated water. The proprietary version of Holtec International Report No. HI-2146153, Licensing Report of the Criticality Analysis of the Dresden Units 2 and 3 SFP for ATRIUM 10XM Fuel Design (HI-2146153), Attachment 3 of the licensees submittal dated December 30, 2014 (ADAMS Accession No. ML14364A100), documents a criticality analysis performed to demonstrate that this regulatory limit is met. The staff has identified some instances where it is not clear if the reactivity impact due to specific conditions was adequately addressed in the criticality analysis. The potential reactivity impacts may be positive, therefore the staff requires additional information to verify the regulatory limit will not be challenged by these potential impacts.

1. In the licensees submittal, Evaluation of Proposed Changes (Attachment 1, pg. 4) it states that blisters have been identified as part of the licensees BORAL monitoring program. Blisters displace water in the spent fuel pool (SFP), reducing moderation of neutrons, resulting in a harder neutron spectrum. Since the absorption cross section of boron-10 decreases for neutrons with higher energies, the neutron attenuation effectiveness of the BORAL may be reduced. Therefore, the reactivity may be higher in the SFP when blisters exist on the BORAL.

Provide information demonstrating that any blisters on BORAL installed in the SFP, being credited for sub-criticality, will not result in the regulatory limit being challenged.

2. The neutron-absorbing core of a BORAL panel is composed of a mixture of Al and boron carbide. The two constituents initially exist in powdered form and are mixed together prior to heating and rolling. The resulting product is not homogeneous, since the boron carbide particles remain discrete. Section 3.b.ii of DSS-ISG-2010-01, Interim Staff Guidance, Staff Guidance Regarding the Nuclear Criticality Safety Analysis for Spent Fuel Pools Introduction (ADAMS Accession No. ML110620086) indicates that the effect of neutron streaming or boron-10 self-shielding effects should be considered when establishing the efficiency of the neutron-absorbing material credited in criticality

Draft RAIs (revised) 2 analyses. The BORAL material appears to be modeled in the HI-2146153 analysis as a homogeneous material. If that is the case, given the relatively small margin to the regulatory limit demonstrated by this analysis; Provide the technical justification for concluding that the neutron absorption of the BORAL material will be bounded by the neutron absorption as modeled in the criticality analysis.

3. In the HI-2146153 analysis the BORAL appears to be modeled as a homogeneous mixture that blends the BORAL core and the cladding into a single material. The thickness is then set equal to the minimum BORAL panel thickness (including cladding) with the remaining space between the rack wall and the sheathing filled with water. An increase in the BORAL thickness would displace the water in the gap between the BORAL and the rack/sheathing walls, which may or may not be less reactive. The results of the rack manufacturing tolerance calculations suggest that a reduction in moderator between the fuel and the BORAL will increase reactivity.

Confirm if this is an accurate description of the modeling of the BORAL, and provide information demonstrating that this approach is appropriately conservative with respect to the calculated Keff.

4. Based on the Updated Final Safety Analysis Report (UFSAR) for Dresden, the BORAL panels are enclosed by two square tubes, with the thickest tube on the outside. Based on the figures given in HI-2146153, the rack cells appear to be modeled as a thick tube with BORAL panels attached to the outside surface using a thin sheathing. The material modeled at the corners of the cells, where the individual cells should be welded together, appears to be water. The geometry differences of the model as compared to the physical configuration in the SFP may result in an unevaluated reactivity deviation.

Discuss how the model used in the criticality analysis differs from the physical configuration of the BORAL installed in the SFP, and why any differences bound the physical configuration.

5. In the HI-2146153 analysis the manufacturing tolerance associated with the SFP storage cell inside diameter is an assumed value. This value is used to evaluate the reactivity impact of the SFP storage cell inside diameter manufacturing tolerance. It appears that the absolute value for the cell inside diameter manufacturing tolerance is the same as the manufacturing tolerance for the SFP cell pitch. However, the SFP cell pitch would be related to the dimensions of the outside tube, not the inside tube, so it is not clear what the basis for the assumed value is.

Provide information demonstrating that the assumed value for the SFP storage cell inside diameter manufacturing tolerance is appropriate and acceptable.

Draft RAIs (revised) 3

6. The minimum boron-10 areal density is one of the most important parameters affecting the neutron absorption effectiveness of the BORAL material.

Provide the basis for the minimum boron-10 areal density used in the HI-2146153 analysis and how it bounds the BORAL material installed in the SFP. In particular, discuss the following:

a) Whether this value is based on a pre-defined acceptance criterion or measured values; b) Whether this value is validated based on the lower end of the 95/95 range for the measurements associated with each BORAL panel; and c) Whether there are BORAL panels with a minimum as-built boron-10 areal density near the minimum value used in the criticality analysis.

7. In Section 2.3.1.2 of the HI-2146153 analysis, the first step in the screening calculations documented is to remove all legacy fuel lattices with a low U [uranium]-235 enrichment from consideration. This screen-out criterion is based on the fact that they will be bounded by more recent fuel designs with much higher lattice average enrichments. The methodology being used to evaluate the reactivity of the stored fuel is a peak reactivity methodology, which means that the limiting reactivity for a lattice is a function of both U-235 enrichment and Gd [gadolinium] content. The screen-out criterion does not appear to consider the Gd loading of the legacy fuel lattices. As a result, the staff was unable to determine if there may be lower enrichment fuel with low Gd loadings that may have a reactivity that is comparable to the higher enrichment fuel with higher Gd loadings.

Provide additional justification to support the conclusion that the unanalyzed legacy fuel will be bounded by this analysis.

8. In the HI-2146153 analysis, the calculations to identify the design basis lattice are performed at four different sets of core operating parameters. However, the individual parameters that make up each set do not appear to have been independently confirmed to be at their limiting values.

Provide information demonstrating that the final set of core operating parameters used in the design basis calculations will bound all possible operating scenarios.

9. In the HI-2146153 analysis, the accident conditions analyzed do not appear to consider all potential SFP configurations. In particular, if a Boral panel is lost for some reason, it may occur in one of the cells adjacent to the interface between rack modules, resulting in more than two face adjacent fuel assemblies with no Boral between them. This represents a new configuration with a reactivity impact that may not adequately be evaluated by use of the interface bias.

Draft RAIs (revised) 4 Provide information demonstrating that these SFP configurations would not result in a challenge to the regulatory limit.

10. In Table 7.2 of the HI-2146153 analysis, the biases and uncertainties applied to the calculations are from the normal condition calculations. The normal condition has a lower reactivity than the limiting accident condition, so the potential reactivity impact of changes in the SFP configuration may be larger for the latter condition.

Given the limited remaining margin to the regulatory limit, provide justification for considering the reactivity impact of the various biases and uncertainties to be essentially the same for the normal and the limiting abnormal/accident configurations.

11. Sections 2.3.11.1.1 and 2.3.11.1.2 of the HI-2146153 analysis explain that fuel rod growth, cladding creep, and crud buildup do not need to be evaluated because these factors are not expected to be significant at the peak reactivity burnup of the design basis lattice. Changes to the fuel rod geometry as a result of irradiation may result in a positive reactivity impact.

Provide information regarding the expected fuel rod growth, cladding creep, and crud buildup at this burnup, and explain why the reactivity impact would not be significant.

12. Section 2.3.11.1.3 of the HI-2146153 analysis does not describe clearly how the Monte Carlo N-Particle Transport Code, Version 5 (MCNP) geometry is changed to evaluate bowing of fuel rods. The text states that the reactivity impact of this geometry change to the fuel in the SFP is evaluated using the depletion related fuel rod pitch positive tolerance provided in Table 5.1(h). This appears to refer to the Fuel rod pitch exposed entry.

Confirm this inference, and describe how the MCNP model was altered for Case 2.3.11.1.3.2. Also, discuss how this tolerance was determined and how it bounds any expected ATRIUM 10XM fuel rod bowing.

13. Section 2.3.11.2 of the HI-2146153 analysis does not describe clearly how the geometry is changed to evaluate fuel channel bulging and bowing. The text refers to the channel outer exposed width tolerance, but it is not clear if the outer exposed width is changed by varying the channel inner width or the channel wall thickness.

Describe how the MCNP model was altered for Case 2.3.11.2.1. Also, discuss how the channel bowing tolerance was determined and how it bounds any expected ATRIUM 10XM channel bulging/bowing.

14. Section 2.3.16 of the HI-2146153 analysis dispositions all reconstituted fuel assemblies currently stored in the SFP based on the fact that none of them would exceed the criteria for initial screening of fuel discussed in the first paragraph of Section 2.3.1.2. It also

Draft RAIs (revised) 5 states that all future reconstituted fuel will need to be evaluated to determine if they are bounded by this analysis. The geometry and composition changes due to fuel reconstitution will have a reactivity impact that needs to be evaluated using an appropriate methodology.

Provide details on how reconstituted fuel would be evaluated, including any relevant assumptions, uncertainties, and/or biases.

15. Section 2.3.15.4 of the HI-2146153 analysis explains that the Missing BORAL Panel analysis is intended to cover the potential that a BORAL panel was inadvertently not installed during construction of the rack. A missing BORAL panel would increase the local reactivity in the SFP for the area surrounding the missing BORAL. If this is expected to have occurred, then it would be part of the normal condition for the SFP and it is inappropriate to consider this as an accident condition.

Provide information demonstrating how inadvertent non-installment of a BORAL panel was precluded from occurring.

16. Table 2.1(a) of the HI-2146153 analysis indicates that the analysis includes Uranium Oxide (UO2) and Mixed Oxide (MOX) fuel. The report does not discuss MOX fuel at Dresden at any other point. The higher plutonium content in MOX fuel may result in unique reactivity impacts that are not considered for UO2 fuel.

Clarify if this analysis is intended to cover MOX fuel, and if so, provide further information about the MOX fuel that this analysis is intended to cover.

17. Provide the number of coupons currently available in each Dresden SFP and discuss whether the ambient conditions of the coupon tree (i.e., radiation exposure, flow, temperature) are bounding or representative of the neutron-absorbing material in SFP racks. Also, discuss whether coupons are returned to the SFP after inspection. If so, provide the technical justification for this approach.

18. The Dresden Technical Specifications (TS) 4.3.1.1(a) requires a Keff of < 0.95. The NRC staff notes the calculated Keff found in the SFP criticality safety analysis (CSA) is determined, in part, from the minimum Boron-10 areal density of Boral provided in the CSA. The calculated Keff is used to demonstrate compliance with the Keff TS requirement. In Section 9.1.2.3.1 of the licensees UFSAR, and in information provided in the licensees supplement dated May 8, 2015, it is stated that the corrosion sampling program (coupon surveillance program) is used as an essential long-term monitoring program to ensure the Boron-10 areal density of the Boral remains at or above its minimum credited value during both normal operating conditions and design basis events.

Discuss what controls the licensee will implement to ensure that the actual Boron-10 areal density of the Boral remains at or above its minimum credited value and that the regulatory requirement to maintain the TS value of Keff < 0.95 continues to be met.