ML18002A351
| ML18002A351 | |
| Person / Time | |
|---|---|
| Site: | Nuclear Energy Institute |
| Issue date: | 12/21/2017 |
| From: | Holtzman B Nuclear Energy Institute |
| To: | John Wise Rules, Announcements, and Directives Branch, Renewals and Materials Branch |
| References | |
| 82FR49233 00004, NRC-2016-0238, NUREG-2214 | |
| Download: ML18002A351 (24) | |
Text
Page 1 of 1 As of: 12/27/17 3:18 PM Received: December 21, 2017 Status: Pending_Post PUBLIC SUBMISSION Tracking No. lkl-90h9-uio7 Comments Due: December 26, 2017 Submission Type: Web Docket: NRC-2016-0238 Managing Aging Processes in Storage Comment On: NRC-2016-0238-0001 Managing Aging Processes in Storage Report; Request for Comment on Draft NUREG Document: NRC-2016-0238-DRAFT-0003 Comment on FR Doc# 2017-22983 Submitter Information /IJ/Jt/f-)~17
~~r~ //~33 Name: Benjamin Holtzman Submitter's Representative: Anya Barry Organization: Nuclear Energy Institute General Comment See attached file( s)
Attachments 12-21 NRC- NEI MAPS Letter v7 12-21 NRC- NEI Comments to Draft MAPS-Rev 5 Attachment SUNSI Review Complete Template= ADM - 013 E-RIDS= ADM-03 Add= J. [VJ~ {;)fl()II-)
https://www.fdms.gov/fdms/getcontent?objectld=0900006482d50e24&format=xrnl&showorig=false 12/27/2017
BENJAMIN HOLTZMAN Senior Projed Manager, Fuel & Decommissioning 1201 F Street, NW, Suite 1100 Washington, DC 20004
~I NUCLEAR ENERGY INSTITUTE P: 202.739.8031 bah@nei.org nei.org December 21, 2017 Mr. John Wise Division of Spent Fuel Management Office of Nuclear Material Safety and Safeguards U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 Submitted via Regulations.gov
Subject:
Submittal of NEI comments on draft NUREG-2214, "Managing Aging Processes in Storage (MAPS)
Report," 82 Federal Register 49233, 10/24/2017 (Docket ID: NRC-2016-0238)
Project Number: 689
Dear Mr. Wise:
On behalf of the nuclear energy industry, the Nuclear Energy Institute (NEI)1 appreciates the opportunity to provide comments on the U.S. Nuclear Regulatory Commission's (NRC) draft report NUREG-2214, "Managing Aging Processes in Storage (MAPS) Report." Industry very much appreciates NRC's efforts to develop NUREG-2214 and continue the collaborative dialogue that has informed staff's efforts through our interactions on the detailed comments submitted herein.
These comments were developed by the industry to provide input that NRC may use to modify NUREG-2214 so that it will provide clearer guidance. Some of industry's most notable concerns are summarized below. The complete set of industry comments are delineated in the attachment to this letter.
One of industry's primary concerns is that NUREG-2214 includes guidance that focuses on types of corrosion that do not occur in an Independent Spent Fuel Storage Installation (ISFSI). Environmental or other factors prevent these types of corrosion from occurring in ISFSis (e.g., comments #15-17, 20, 22, and 23).
Another concern is whether there should be an aging management program (AMP) for the fuel stored in the dry storage system (e.g., comments #105 and 106). It is suggested that the NRC focus should be on the aging of the dry storage system itself, which will ensure that the fuel remains safely stored.
1 The Nuclear Energy Institute (NEI) is the organization responsible for establishing unified industry policy on matters affecting the nuclear energy industry, including the regulatory aspects of generic operational and technical issues. NEI's members include all entities licensed to operate commercial nuclear power plants in the United States, nuclear plant designers, major architect/engineering firms, fuel cycle facilities, nuclear materials licensees, and other organizations and entities involved in the nuclear energy industry.
NUCLEAR . CLEAN AIR ENERGY
Mr. John Wise December 21, 2017 Page 2 The next concern of industry's is that the proposed implementation of the AMPs include specifications that are contrary to recent rulemaking at 82 Federal Register 57819 Renewal of Certificate of Compliance No. 1004 -
TN Americas LLC, Standardized NUHOMS Horizontal Modular Storage System (e.g., comments #150, 152, and 162). Moreover, some aspects of these AMPs are unnecessary and inconsistent with ALARA principles.
For example, the inclusion of some of the radiation monitoring requirements in NUREG-2214 would increase exposure to station employees but provide no safety assurance benefit (e.g., comment# 146). Finally some of the AMP acceptance criteria are too prescriptive and would result in undue burden without commensurate safety benefit (e.g., comment# 149).
In addition to the above concerns we have one overall question; how will NUREG-2214 keep from becoming out of date given its many references to Interim Staff Guidance (ISG) when NUREG-2215 will be incorporating several of the ISGs directly into NUREG-2215 (e.g., comment# 1)?
Thank you again for the opportunity to comment on NUREG-2214. We look forward to working with the NRC on the implementation of this important component of a comprehensive regulatory framework. This guidance will facilitate sustained assurance of spent nuclear fuel in dry storage systems while improving efficiency in the preparation and review of licensing documents involving the dry storage of spent nuclear fuel.
NUREG-2214 is one of three pillars that will form the foundation for an adaptive aging management program.
These programs will allow industry to address potential aging effects that have not yet been observed or identified by collecting, evaluating, and sharing information regarding future operating experience. This approach is already embodied in Revision 1 to NUREG-1927 "Standard Review Plan for Renewal of Specific Licenses and Certificates of Compliance for Dry Storage of Spent Nuclear Fuel," which NRC finalized in June 2016 (ML16179A148). For a learning approach to aging management to be fully realized, industry preparers of renewal applications must have tools equivalent to those provided to NRC reviewers by NUREG-1927. NRC endorsement of Revision 2 to NEI 14-03 "Format, Content and Implementation Guidance for Dry Cask Storage Operations-Based Aging Management" would make this possible. Industry's December 21, 2016 submittal of NEI 14-03 (ML16356A210) addresses staff feedback and should be suitable for endorsement in conjunction with the finalization of NUREG-2214.
In submitting these comments, NEI would also like to express our appreciation to Mr. Paul Plante of Maine Yankee Atomic Power Company who led the industry review team. If you have any questions, please contact me or Paul at pplante@3yankees.com or (207) 882-1320.
Mr. John Wise December 21, 2017 Page 3 Sincerely, Benjamin Holtzman Attachment c: Mike Layton, NMSS, NRC Tony Hsia, NMSS, NRC Meraj Rahimi, NMSS, NRC May Ma, NMSS, NRC
NEI Comments to Draft NUREG NUREG-2214 MAPS Report Comment# Line #, Table Comment Page#
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For example: Since Interim Staff Guidance (ISG's) will be incorporated into the Lines 20 and new, combined NUREG-2215 currently issued in draft and sunset 21, Section 3.6 as documents, how will this NU REG keep from being out of date 1 General & Table 6-6, with its many references to ISG's.
Element 2, Preventive Actions Suggest re-wording " contains one acceptable method" to 2 1-2 21
" describe acceptable methods."
3 1-2 35 Suggest adding NU HOMS HD (CoC 1030)
The definition of high burnup fuel embedded in the definition of 4 2-2 Table 2-1 Zirconium-based alloys seems odd . The burnup of the fuel is not a material. Suggest deleting and defining separately in the text.
Suggest including Metamic-HT as it is a different material than the 5 2-1 Table 2-1 Metamic mentioned in the table with a different purpose, usage, and composition.
6 2-2 Table 2-1 Add definition of Metal Matrix Composite (MMC) to the list.
Recommend expansion of Fully Encased (steel) (FE) to include neutron shielding and gamma shielding materials within sealed or welded steel enclosures such as the Transfer Cask Body and Shield 7 2-3 Table 2-2 Doors, and VCC Shield Plug and Lid. Currently some of these components are identified as embedded-NS or embedded-lead, which is not as good a description .
For "OD," it is overly conservative to assume outdoor air for transfer casks subject to indoor air because the TCs are generally 8 2-3 Table 2-2 not exposed to precipitation, wind and salt-laden air for extended periods of time (if at all for many vertical systems). It would be more appropriate for a TC to be assigned the "SH" environment.
Shrinkage cracking also occurs from sharp corner geometries in 9 2-7 Table 2-3 the design, such as on outlet vents. In these instances, once the crack forms, it is stress is re lieved and the crack does not grow.
It seems inappropriate to include wet corrosion and blistering as an aging mechanism for dry storage systems. The described phenomenon occurs during loading and drying, not storage 10 2-8 Table 2-3 operation, so it's not a mechanism that's possible in a dry storage cask over the PEO. Blistering, if it occurs, does not affect functionality of the material.
"Loss of criticality control" is poor wording because it connotes a 11 2-9 Table 2-4 complete loss of criticality control or even a critical condition.
Suggest using "Reduction in neutron attenuation" instead
" Loss of bond" and "Loss of material" should include reference to 12 2-9 Table 2-4 these effects on coatings .
Is it necessary to include "None"? Isn't it obvious that that's a 13 2-9 Table 2-4 choice?
NEI Comments to Draft NUREG NUREG-2214 MAPS Report Comment# Line#, Table Comment Page#
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Differential settlement is not an aging mechanism associated with 14 3-5 Table 3-5 Air-Outdoor environment or Sheltered environment.
It is hard to imagine what kind of environment would qualify as an aggressive chemical attack environment that would be associated 15 3-5 Table 3-5 with Outdoor Air for concrete. This tab le should delete this environment.
Microbiological degradation of concrete is not a common mode of degradation and generally occurs in environments that are not typical of ISFSls, such as seawater service and sewage treatment.
Since OE indicates that it is not occurring in concrete in nuclear 16 3-5 Table 3-5 plant environments, there is insufficient justification for including it in MAPS. Even if it were to occur, it would produce similar results to aggressive chemical attack (it is indeed a form of that phenomenon) which is included. It should be classified as not credible.
Salt scaling cannot happen below grade and hence the GW environment should be excluded. The "freeze line" NRC must be referring to is what is most commonly known as the "frost line" 17 3-5 Table 3-5 which is by definition, the depth below which soil does not freeze in the winter. Sa lt scaling, if it occurs, would only be above ground. Salts and thawing agents are generally prohibited from ISFSls in plants subject to these conditions.
18 3-8 Line 18 Suggest adding a parenthetical definition of " deliquescence."
Are there temperature-humidity combinations where general corrosion becomes life-limiting? Change wording to " ....... to outdoor and sheltered environments are potentially present, but 19 3-8 Line 35 general corrosion, although plausible, will not propagate at rates sufficient to affect component intended function . Therefore aging management of general corrosion is not required during the 60 year time frame."
Steel exposed to an embedded concrete environment should be protected and general corrosion is not expected. Only when the concrete becomes faulted and allows the rebar to be exposed to 20 3-9 Line 23-25 other environments does the possibility of corrosion occur. This document should not assume that the concrete will become comprom ised and the rebar exposed . Aging management of concrete should prevent this type of corrosion from occurring.
21 3-10 Line 40 Define the term "electroless."
NEI Comments to Draft NUREG NUREG-2214 MAPS Report Comment# Line#, Table Comment Page#
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Steel exposed to an embedded concrete environment should be protected and pitting and crevice corrosion is not expected . Only when the concrete becomes faulted and allows the rebar to be Line 15-16, exposed to other environments does the possibility of this 22 3-10 21-23, corrosion occur. This document should not assume that the concrete will become compromised and the rebar exposed. Aging management of concrete should prevent this type of corrosion from occurring.
This section discusses MIC, but does not justify the inclusion of steel embedded in concrete environments as a credible environment. It assumes that the concrete exposed to soil is in a sufficiently degraded condition to allow rebar to be exposed to microbes (if they are present and are of an aggressive type). We 3-11 &
23 Section 3.2.1.4 have not seen this kind of damage in nuclear plants which 3-12 indicates that this scenario is highly unlikely and does not warrant inclusion . This document should not assume that the concrete will become compromised and the rebar exposed. There are more likely credible environments that will result in pad inspections without confusing the issue with improbable scenarios.
What is " significant" creep? Undefined term casts doubt. Either 24 3-14 Line 10 define or delete . Delete "significant".
The initiation of sec requires the presence of stress, environment and susceptible material. Absent anyone of these, there is no SCC.
What stresses are considered? Not stated here. Should be 25 3-19 Line 4 included with a comment on the magnitude relative to sec susceptibility. Page 3-19 line 1 change to: " ..... are known to be precursors to SCC in the presence of weld residual stresses of sufficient magnitude to initiate SCC."
What is meant by "indoor/outdoor environments"? Doesn't the 26 3-22 1 outdoor part ofthis statement contradict the previous section?
Shouldn't "Indoor Environment" be listed in Table 3-1?
27 3-22 28 Typo ...Should be "subcomponents" .
In this section, thermal aging of 17-4 PH in a helium environment is deemed a credible aging mechanism based on experiences in the reactor coolant system which is a significantly different environment. The ASME temperature limits for this alloy seems to 28 3-25 Line 1-27 be a reasonable approach to take, so the recommended action from the users should entail demonstrating that it stays below this temperature limit or that the material meets the necessary material requirements if modelling shows it does not meet these temperature limits.
Wording question? Maybe better to state "could occur in spite of 29 3-27 Line 35 the passive oxide layer on the surface of aluminum materials."
NEI Comments to Draft NUREG NUREG-2214 MAPS Report Comment# Line #, Table Comment Page#
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The corrosion damage calculated (0.6mm over 60 years), which is certainly conservative, is hardly concerning from a general 30 3-36 28 corrosion standpoint and shows why copper is used all over the world in exposed application . General corrosion should not be credible. It should be the same as pitting and crevice corrosion .
Line 3, Sent ence should read as follows : " ... materials. Hydrogen in the 31 3-49 Section 3.3 water reduces the energy ...II Line 4, Please clarify what is meant by "possible relocation of shielding 32 3-49 Section 3.3 materials," Is this meant to be dislocation or cracking?
Lines 15 and Lines should begin with "EPRI."
33 3-51 17, Section 3.3 .2 Lines 32 and Lines should begin with "NRC."
34 3-51 35, Section 3.3.2 Lines 1 and 3, Lines should begin with "NRC."
35 3-52 Section 3.3.2 Line 44, Typo - The word "verticle" should be "vertical."
36 3-54 Section 3.4 On page 3-54, line 23 and 24, it states that for borated stain less steel, "boron depletion is not considered to be credible, and therefore, aging management is not required during the 60-year timeframe." The next paragraph (i.e., Line 25) changes the language to "boron depletion in borated stainless steel is not generally considered to be a credible aging mechanism ."
Line 23-25, 37 3-54 The addition of "generally" is incorrect. It appears that the third Section 3.4.1.1 paragraph in Section 3.4.1.1. is simply trying to say that any past boron depletion materials in the original application basis be treated in the renewal application (vs. ignoring it). The term "generally" may have been included based on past calculations by some vendors to determine the impact of potential boron depletion. The term should be depleted as these calculations were, and are still not necessary for this impact.
This section does not appear to address Metamic-HT which is a 38 3-55 Section 3.4.2 different MMC than the Baral or Metamic described .
NEI Comments to Draft NUREG NUREG-2214 MAPS Report Comment# Line#, Table Comment Page#
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The following statement begins, "It is important to note that, because only a trace amount of water will be left in a dry storage cask after dehydration and helium backfill, the occurrence of wet corrosion and blistering will be minimal in a dry cask environment during the period of extended operation." This continues on page Line 43, 39 3-56 3-57, line 3 which states: "corrosion and blistering will be Section 3.4.1.1 minimal." This should also state that since DCS system temperature decays over time and there is a very small, finite source of water in a dried and sealed cask, the progression of any corrosion and blistering will be minimal during the period of extended operation .
Boron depletion, states that the generic evaluation does not Line 22-26, identify Boron depletion as a significant aging mechanism, but 40 3-57 Section 3.4.2.4 depletion analyses (thus aging management work) is still needed?
This seems inconsistent.
Line 11, Reference is not used in the text.
41 3-59 Section 3.4.3 Lines 26 and Lines should begin with " EPRI."
42 3-59 29, Section 3.4.3 Line 34, Reference is not used in the text.
43 3-59 Section 3.4.3 Line 1, Reference is not used in the text.
44 3-60 Section 3.4.3 Line 15, Line should begin with "NRC."
45 3-60 Section 3.4.3 Line 27, 35, There does not appear to be a " NRC 2012" reference .
46 3-62 Section 3.5.1.1 One of the main ways to protect against freeze-thaw damage is to specify air-entrained concrete (your reference ACI 2008c). Utilities Line 36-38 47 3-62 should be allowed to take credit for properly spec'd air entrained Section 3.5.1.1 concrete if they are in a freeze-thaw zone as an alternative to conducting aging management for this degradation mode.
You r reference Sindelar 2011 concludes "The requisite conditions of refreshed water and the overall kinetics ofthe coupled Line 35-37 processes involved in calcium leaching renders it an insignificant Section 3.5.1.8 degradation mechanism for the concrete in the pad or cask of the 48 3-69 "Leaching of DCSS for EST." This is consistent with industry observations. We Calcium believe that while it is observed, it does not require aging Hydroxide" management. Also, this reference does not appear to be a "Draft Report" as stated in the reference section on page 3-83. The IAEA reference also supports this position .
NEI Comments to Draft NUREG NUREG-2214 MAPS Report Comment# Line#, Table Comment Page#
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It is for precisely this reason that this can be excluded from further consideration. ISFSI pads are constructed on pad bedding using high quality fill materials and not on polluted soils.
Line 41-43, Furthermore, there are embedded concretes elsewhere in the 49 3-72 Sect. 3.5.1.12 nuclear plant that are subjected to much more severe conditions (i.e. circulating water systems). The absence of OE for MIC on concrete in these areas further illustrates that this mechanism is not expected in this kind of service.
Of the 11 fuel cladding degradation mechanisms, 9 were determined not to be credible during a 60 year storage so 3.6.1 timeframe. It is not clear if the studies referenced considered effects of pinhole or hairline cracks in the cladding.
For hydride reorientation, dissolved hydrogen is associated with the high temperatures experienced during vacuum drying. Does 51 3-86 3.6.1.1 this apply to systems that dry HBU fuel with forced helium dehydration?
Line 44, Maximum dissolved hydrogen at 400 C is 200 ppm, but isn't this 52 3-86 Section 3.6.1.1 material dependent? Suggest qualifying the 200 ppm.
The statement is made that cladding with hydrides " ...has been Line 8, shown to have reduced ductility under pinch-load stresses ... ". Is 53 3-87 Section 3.6.1.1 this still true for cladding with fuel in it? Did the references test with fuel in the cladding?
Section discusses potential cladding failures when fuel is subject t o pinch-load stresses. It' s not apparent when a pinch-load can 54 3-87 3. 6.1.1 occur during normal storage, or how fuel pellet-cladding bounding is considered .
Sentence beginning on line 19 - Says HRO is driven by hoop stress which is determined by peak cladding temperature. This is overly Line 19, pessimistic. The hydrides precipitate and potentia lly reorient as 55 3-87 Section 3.6.1.1 the cladding cools over time. The stress at the time the hydrides precipitate will be less than the peak stress at the time of peak cladding temperature.
Mentions the negative characteristic of RXA cladding in that it is more susceptible to HRO due to larger fraction of grain Line 20, 56 3-88 boundaries in the radial direction . The paragraph should include a Section 3.6.1.1 counter to that in that RXA has lower hydride concentrations, so there are fa r fewer hydrides available to reorient.
Discusses the impact of cooling rate on HRO and concludes the slow coo ling rate under actual dry storage cond itions will not inhibit HRO. However, a key phenomenon that is ignored is the Line 24, effect of annealing. Due to the slow cooling rate, the cladding 57 3-88 Section remains at somewhat higher temperature for long periods. This 3.6.1.1 time at temperature provides some annealing and repairing of irradiation defects. This needs to be included in the cooling rate discussion in this paragraph.
NEI Comments to Draft NUREG NUREG-2214 MAPS Report Comment# Line#, Table Comment Page#
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Line 31, Editorial -add "to" after "expected" 58 3-91 Section 3.6.1.2 Line 17, Editorial- add a space between "350" and "degrees" 59 3-98 Section
~
3.6.1.9 With regards to the statement that sec of the cladding is not credible and aging management is not required during the 60 year 60 3-99 3.6.1.9 storage timeframe, is this applicable to all cladding types? The evidence provided seems to focus on Zr-2 and Zr-4 versus the advanced alloys.
With regards to radiation embrittlement, the conclusion that embrittlement is not considered credible is based on a cumulative fluence (10A22 n/cm2) not expected during storage. However, in the second paragraph states that embrittlement of cladding is 61 3-99 3.6.1.10 observed in reactor due to cumulative fast neutron exposure on the order of lQA22 n/cm2. Thus, as worded it appears that the cladding has already reached the necessary fluence for embrittlement before getting into dry storage.
Add NU HOMS HD System to the evaluated Storage system design-62 4-1 Table 4-1
-NRC Docket Number 72-1030.
HSM described, but heat shield not mentioned. It is part of the 63 4-6 Section 4.2.3 HSM, and is shown on page 4-41 Listed in this Table are the AMR results of the subcomponents of all the NU HOMS DSCs certified in Coe 1004. However, the Report does not list the DSC Type associated with a specific subcomponent. For example. The first and second row lists the AMR results for Guide Sleeves and Over sleeves. It would be useful to add in the first Column "Spacer Disc Type DSC Basket" instead of " DSC Basket" . The Standardized NU HOMS System 64 4-10 Table 4-2 provides for 2 alternate basket designs" spacer disc" and "Tube" type basket design. Further, the next line item lists the AMR resu lts for Aluminum Plate. This subcomponent is only present in high heat load DSCs and not present in the earlier DSC designs.
Hence, here also, it would add to the clarity of this Table if it is annotated that this subcomponent is only present in high heat load DSC designs. This is a generic change suggested for the entire Table 4-2.
Listed in this Table are the AMR results of the subcomponents of all the NUHOMS HSMs certified in Coe 1004. However, the report does not list the HSM Type associated w ith a specific 65 4-33 Table 4-4 subcomponent. For example, the DSC Support Structure for HSM Model 80 is quite different from that provided for HSM-H or HSM-HS. Hence, it wou ld be useful to annotate each row of Table 4-4 by adding the HSM Type in Column 1.
NEI Comments to Draft NUREG NUREG-2214 MAPS Report Comment# Line#, Table Comment Page#
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Listed in this Table are the AMR results of the subcomponents of all the NUHOMS Transfer Casks. The Standardized NUHOMS System provides for 5 different TC Types: Standardized TC, OS197, OS197H, OS200 and OS197L. Each of these TCs has a unique 66 4-69 Table 4-6 design with materials of construction which differ from one TC Type to another. To avoid any confusion, it would be useful to annotate each row of Table 4-6 by adding the TC Type in Column 1.
This section is limited to the HI-STORM 100 and its associated variants. Other Holtec dry cask storage systems, that have 67 Section 4.3 separate Certificates (HI-STORM FW and HI-STORM UMAX) have been placed into service and utilize larger capacity canisters.
68 4-91 Line 43 Fix typo ("HI-TORM 100")
Recommend changing, "Baral and METAMIC" to "Either Baral or 69 4-94 Line 9 METAM IC" to make it clear that a given basket is comprised of only one type of neutron absorber.
Recommend noting that the pocket enclosure for neutron 70 4-94 Line 11 absorbers does not apply to METAM IC-HT (MPC-68M) configuration.
The bolted on shield lid atop the MPC lid is not a standard configuration. In fact it was used one time at one site. MAPS 71 4-94 Lines 23-24 should state this is not present on all MPCs. There is no further consideration for this shield lid in Table 4-7, or how if present can affect aging management of the MPC lid.
It's not apparent why the Transfer Cask is being included for a 72 4-99 Section 4.3.5 document that is focused on extended storage.
Boron depletion (under subsections for Metamic-HT, Baral and Metamic) is not an apparent aging mechanism for dry storage as 73 4-104 Table 4-7 there is practica lly no thermal neutron flux under the conditions for normal storage. Stating that a TLAA may be required shou ld not be necessary.
With regards to the items combined under "concrete shield", for those items that are fully encased in steel, including the reaction with aggregates as an aging effect should be a 'no', since these do 74 4-110 Table 4-8 not have the susceptibility mechanisms (for example large aggregate surface area for reaction or exposure to moisture) described in Section 3.5.1.3. Likewise these components do not relay on the concrete for structural strength.
Should aging management of external vent screens be included 75 4-111 Table 4-8 with gamma shields, or can the condition of the vent screens be used as an indication of gamma shield aging?
NEI Comments to Draft NUREG NUREG-2214 MAPS Report Comment# Line#, Table Comment Page#
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In the HI-STAR, the safety function of the neutron shield is only shielding. Thus, aging effects should not include structural related items (fracture toughness and loss of ductility) . These aging effects refer to MAPS Section 3.3 .1. 2 which does not describe 76 4-132 Table 4-9 either of these. Additionally, these effects are covered under thermal aging, which describe the breakdown of polymers under elevated temperatures. It would seem that over an extended storage period, elevated temperatures should not be a concern.
If aging management of the transfer cask is deemed necessary, 77 4-140 Table 4-10 then should lead shield mat erial gapping or slumping be included as an aging mechanism?
Should pool lid seals be included with the other related HI-TRAC 78 4-142 Table 4-10 components considered for aging effects A note should be added to the section stating that not all HI-79 Section 4.3 STORM systems include all the SSCs provided in Tables 4-7 through 4-10 Incorrectly identifies Structural Lid as exposed to a helium environment. The Structural Lid is installed above and encases 80 4-176 Table 4-12 the Shield Lid. Environment between the two lids would be indoor air.
Leaves out Shield Lid, though shield lid support ring is identified.
4-177 The Shield Lid (top) would be encased by the stainless steel 81 Table 4-12 Structural Lid and the interior surface exposed to the internal helium environment. Add immediately following Spacer Ring.
Port Covers are installed in the shield lid above the quick 4-177 disconnect valved couplings and only underside of port covers 82 thru Table 4-12 would be potentially exposed to a helium environment. The top 4-178 welded side of the Port Covers would be exposed to the encased indoor air environment.
Reinforcing Steel environment is air-outdoor. These components 83 4-185 Table 4-13 are not exposed to groundwater, so this environment shou ld be deleted.
Inner Shell main safety functions are as a gamma shielding component and a heat transfer component, not structural, so 84 4-186 Table 4-13 intended safety function should be identified as "SH, TH ", or "SH, TH, SR" not "SH, SR" .
Outlet Vent hardware is defined as a structural component, but 4-188 its main intended safety function is thermal (external protection 85 thru Table 4-13 of the outlet vent from entry and blockage of foreign materia ls),
4-189 so intended safety function should be "TH, SH", not "SR".
4-188 The listing does not include the Inlet Vent Hardware, so Outlet 86 thru Table 4-13 Vent Hardware should be revised to be " Inlet and Outlet Vent 4-189 Hardware" .
NEI Comments to Draft NUREG NUREG-2214 MAPS Report Comment# Line#, Table Comment Page#
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Outlet Vent hardware is defined as a structura l component, but 4-188 its main intended safety function is thermal (external protection 87 thru Table 4-13 of the outlet vent from entry and blockage of foreign materials),
4-189 so intended safety function should be "TH, SH", not "SR".
The neutron shield materials contained in the Shield Plug should 88 4-191 Table 4-13 be defined as "Fully Encased (FE) (Steel)" rather than "Embedded (steel)".
Why is Neutron Shield (Cask Body) identified as requiring a TLAA/AMP for Thermal Aging, whereas it is not identified as 89 4-195 Table 4-14 required for the Steel MAGNASTOR Transfer Cask, but is for the Stainless Steel MAGNASTOR Transfer Cask? Please clarify or correct.
4-194 NAC defines the Transfer Cask body and shield door neutron 90 thru Table 4-14 shielding and gamma shielding as Fully Encased (FE) (Steel) as 4-195 these components are fully encased in steel components.
. Shield Door Rails, which are coated steel, are identified as requiring aging management by the Transfer Cask AMP for 91 4-197 Table 4-14 Galvanic Corrosion for loss of material when the doors rails are not connected to other non-carbon steel materials. Please clarify or delete.
Fuel Basket Support Disk for all MPC TSCs is 17-4 stainless steel, identical to NAC-UMS PWR support disks. However, MPC TSC table requires aging management for thermal aging for the 92 4-206 Table 4-15 stainless-steel support disks although it is not required for the UMS stainless-steel support disks, only the steel support disks of the BWR fuel basket assembly. Please correct to delete the aging management requirement or clarify discrepancy.
Inner Shell main safety functions are as a gamma shielding component and a heat transfer component, not structural, so 93 4-215 Table 4-16 intended safety function should be identified as "SH, TH", or "SH, TH, SR", not "SR" .
Table 4-16, Lid Radiation Damage {both cracking and loss of strength) refers to 94 4-220 Assembly, TLAA or AMP but 3.5.1.9 says radiation damage to concrete is not Concrete cred ible. Aging management should therefore be "No ".
Why is Neutron Shield (Cask Body) identified as requiring a TLAA/AMP for Thermal Aging, whereas it is not identified as 95 4-224 Table 4-17 required for the Steel MAGNASTOR Transfer Cask, but is for the Stainless Steel MAGNASTOR Transfer Cask? Please clarify or correct.
NAC defines the Transfer Cask body neutron shielding and lead gamma shielding as Fully Encased (FE) (Steel) as these 96 4-223 Table 4-17 components are fully encased in steel components. Correct
" Gamma Shielding" to " Neutron Shielding" for NS-4-FR materials.
NEI Comments to Draft NUREG NUREG-2214 MAPS Report Comment# Line#, Table Comment Page#
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Shield Door Rails, which are coated steel, are identified as requiring aging management by the Transfer Cask AMP for 97 4-226 Table 4-17 Galvanic Corrosion for loss of material when the doors rails are not connected to other non-carbon steel materials. Please clarify or delete.
Under Damaged Fuel Can Screen also add Wiper to the list as the 98 4-239 Table 4-18 wiper extends out further than the screens and acts as the CO boundary between the DFC lid and DFC collar.
Table 4-19, Lid Radiation Damage (both cracking and loss of strength) refers to 99 4-246 Assembly, TLAA or AMP but 3.5.1.9 says radiation damage to concrete is not Concrete credible. Aging management should therefore be "No".
The Lid Assembly incorporates a concrete neutron shield embedded in concrete. It is not feasible for Reaction with 4-245 Aggregates aging effects to be managed by an External Surfaces 100 thru Table 4-19 Monitoring of Metallic Component AMP. If aging management is 4-247 required, a TLAA would be required. Please correct or clarify.
Also, the environment should be Fully Encased (FE) (steel) .
The Lid Anchor (standard and alternate configuration) steel components that are embedded in concrete cannot be 4-247 monitored for aging management for General Corrosion or thru 101 Table 4-19 Pitting and Crevice Corrosion by an External Surfaces Monitoring 4-248 of Metallic Component AMP. If aging management is required, a TLAA or the Reinforced Concrete Structures AMP would be appropriate. Please correct or clarify.
Inner Shell main safety functions are as a gamma shielding 4-250 component and a heat transfer component, not structural, so 102 thru Table 4-19 intended safety function should be identified as "SH, TH ", not 4-251 "SR, SH, TH ".
103 4-259 41 Replace "of the" with "adjacent."
Differential settlement is not an aging mechanism associated with 104 4-292 Table 4-24 Air-Outdoor environment.
From a fundamental standpoint, should there even be an AMP for fuel? It is the fuel we are trying to protect in the dry storage system. Rather than focusing on degradation of the fuel, instead, 105 4-297 Section 4.8 we should ensure that any impacts from accidents and aging of the dry storage system do not create an impact on the fuel that would create unacceptable consequences.
How does this integrate with the revised definition of 106 4-297 Section 4.8 RETRIEVABILITY which allows for canister-based retrievability to meet this requirement?
Line reads " ...neutron absorber rods and burnable poison rods ... "
Line 18, 107 4-297 These are the same thing. Maybe meant to say "control rods and Section 4.8.2 burnable poison rods ."
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Line should read as follows: ... rods . The fuel rods are hollow Line 29, 108 4-297 cladding tubes fabricated from Zircaloy-2 filled with uranium Section 4.8.2 dioxide pellets ."
Line 34, Line should read as follows: ... channels). The channels ..." Deleted 109 4-297 Section 4.8.2 text is repeated in previous paragraph.
Line 36, Line should read as follows : ... levels. Both the upper and lower tie 110 4-297 II Section 4.8.2 plates ...
Table 4-25, For Fuel rod cladding, Hydride-induced embrittlement Aging 111 4-300 Aging Mechanism, page 3-89, line 35 requires and AMP.
Management Hydride-induced embrittlement should be "Hydride reorientation" per the title of Section 3.6.1.1. Also, Table 4-25 says no aging management, but Section 3.6.1.1 provides 2 112 4-300 Table 4-25 approaches for aging management to address hydride reorientation: A defense in depth with consequence analysis or using results from a demo.
Table 4-25, For Guide tubes (PWR) or water channels (BWR), the reference for 113 4-300 Technical Basis Radiation embrittlement should be 3.6.2.5 not 3.6.1.10.
(Section)
Table 4-25, For Guide tubes (PWR) or water channels (BWR), the reference for 114 4-300 Technical Basis Fatigue should be 3.6.2.6 not 3.6.1.11.
(Section)
Table 4-25, For Spacer grids, Zirconium-based alloy, the reference for 115 4-301 Technical Basis Radiation embrittlement shou ld be 3.6.2.5 not 3.6.1.10.
(Section)
Table 4-25, For Spacer grids, Zirconium-based alloy, the reference for Fatigue 116 4-301 Technical Basis shou ld be 3.6.2.6 not 3.6.1.11.
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Table 4-25, For Spacer grids, lnconel, the reference for Radiation 117 4-301 Technical Basis embrittlement should be 3.6.2.5 not 3.6.1.10.
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Table 4-25, For Spacer grids, lnconel, the refe rence for Fatigue should be 118 4-301 Technical Basis 3.6.2.6 not 3.6.1.11.
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Table 4-25, For Lower and upper end fittings, Stain less steel, the reference for 119 4-301 Technical Basis Radiation embrittlement should be 3.6.2.5 not 3.6.1.10.
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Table 4-25, For Lower and upper end fittings, lnconel, the reference for 120 4-301 Technical Basis Fatigue should be 3.6.2.6 not 3.6. 1. 11.
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Table 4-25, For Fuel channel (BWR), the reference for Radiation 121 4-302 Technical Basis embrittlement should be 3.6.2.5 not 3.6.1.10.
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Table 4-25, For Fuel channel (BWR), the reference for Fatigue should be 122 4-302 Technical Basis 3.6.2.6 not 3.6.1.11.
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Table 4-25, For Poison rod assemblies (PWR), the reference for Radiation 123 4-302 Technical Basis embrittlement should be 3.6.2.5 not 3.6.1.10.
(Section)
Table 4-25, For Poison rod assemblies (PWR), the reference for Fatigue should 124 4-302 Technical Basis be 3.6.2.6 not 3.6.1.11.
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Lines 6 and 8, Lines should begin with "FuelSolutions."
125 4-303 Section 4.9 Line 16, Line should begin with " Holtec International."
126 4-303 Section 4.9 Lines 21, 22 & Lines should begin with "NAC International."
127 4-303 24, Section 4.9 Lines 30 and Lines should begin with "NRC."
128 4-303 32, Section 4.9 After "design basis," insert "described in, or incorporated by 129 5-1 4 reference in the ISFSI or cask FSAR."
Suggest considering breaking section 6.7 into two separate AMPS.
One for readily accessible, metallic, external surfaces exposed to outdoor atmospheres and one for sheltered, external metallic 130 6-1 Table 6-1 surfaces exposed to outdoor atmospheres. There will be different inspection programs, repairs, inspection frequencies, etc. for these two cases.
The scope of this bullet should be revised to "Known areas of the canister to which temporary supports or attachments ... . This Table 6-2, wou ld be based on document reviews and evidence of grind 131 6-6 Element 1, marks on the canister (though all grind marks do not necessarily 3rd bullet indicate the presence of a location where temporary attachments were used .
Regarding this item: Effort should be made to identify and prioritize examinations of areas on canisters that have two or more of the above attributes (e.g., canister surface that is cold relative to average surface temperature and also has a weld or Table 6-2 132 6-6 weld heat affected zone). Why prioritize any area that is not near Element 1 a weld? What does " prioritize" mean relative to coverage area? Is canister removal from overpack expected? The actions and scope should be commensurate with the safety significance and this needs to be conveyed in the document.
Table 6-2 The phrase "adequate cleaning" needs to be qualified. There is Element 4, evidence of some canisters requiring cleaning to remove heavy 133 6-7 Volumetric deposits of pollen, but there is evidence of other canisters having Inspection bare metal surfaces that require no cleaning.
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Table 6-2 The phrase " For accessible areas where adequate cleaning should Element 4, be performed ..." , "accessible" should be qualified as this is 134 6-7 Volumetric accessible by remote VT.
Inspection Table 6-2 Please change the first phrase to "For sites conduct ing a canister 135 6-8 Element 4, inspection,"
"Sample Size" Use of ASME Section XI acceptance criteria for RCS piping is inappropriate for this application involving low pressure canisters of ductile material. AMP as it is not based on the canister design .
Table 6-2, There is a significant difference in response to defects between 136 6-10 Element 6, the high temperature and pressure of RCS piping and the low 2nd and 3rd temperatures and pressures of dry storage. Acceptance criteria have been identified in EPRl -3002008193 that are appropriate for this service and they should be for use in this AMP.
This removal of iron deposits and rust stains should be reserved Table 6-2, for welds and their associated heat affected zones. While this 137 6-10 Element 6, 4th section implies this, it is not clearly stated and the section should be revised accordingly.
What is the basis for the 1mm criteria? This appears to be the first use of this criteria and the basis for it has not been identified and Table 6-2, justified for this application. In lieu of using this, acceptance 138 6-10 Element 6, criteria have been identified in EPRl -3002008193 that are 2nd bullet appropriate for this service and they should be for use in this AMP.
The guidance is not clear on how indications may be dispositioned . The language provided indicates that these are not actua lly criteria for acceptance, rather criteria for doing additional Table 6-2 139 6-10 evaluations. The referenced standards do not specifically address Element 6 the question of confinement integrity. What are examples of indications that can be accepted for continued service? What indications cannot be accepted?
This blanket statement: No indications of localized corrosion pits, etching, crevice corrosion, SCC, red-orange-colored corrosion products emanating from crevice locations, or red-orange-colored corrosion products in the vicinity of canister fabrication welds, Table 6-2 140 6-10 closure welds, and welds associated with temporary attachments Element 6 during canister fabrication. May prevent any inspection from being acceptable without further evaluation. Consider using acceptance criteria provided in referenced document EPRI-3002008193.
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Canisters with localized corrosion or SCC that do not meet the prescribed evaluation criteria are not permitted to Table 6-2 141 6-11 remain in service without an engineering analysis or mitigation Element 7 actions. - Is this referring to the acceptance criteria in element 6 or to some other prescribed evaluation criteria?
The phrase "a re not permitted to remain in service" does not properly recognize the nature of passive dry storage systems .
Table 6-2, Suggest rewording this last sentence as follows: "Canisters with 142 6-12 Element 7, localized corrosion or sec that do not meet the prescribed last acceptance criteria shall be entered into the licensee's corrective action program as a condition adverse to quality to allow for appropriate evaluations and follow-up."
This section, the one that follows, and the end of Table 6-4 are the only places in MAPS that identifies crack growth rates. With the exception of the Kosaki reference, these crack growth rates are
" apparent" crack growth rates that are empirically derived from Table 6-2, operating data . The nature of these numbers is well explained in Element 10, EPRI 3002002528 which is cited in the next paragraph. This data is 143 6-12 "Operating useful, but appropriate qualifications should be identified where Power they are cited to ensure that this fact is understood. A follow-up Reactors" EPRI report, EPRI 3002002785 "EPRI Public Flaw Growth and Flaw Tolerance Assessment for Dry Cask Storage Canisters" provides a more in-depth treatment of this area and may be a useful reference.
Ground water monitoring as part of concrete aging management-This would be conducted when a pad is in scope and there is reason to believe that the ground water is conducive to applicable Table 6-3, concrete degradation . If past characterization shows that the Element 1, ground water is below the cited limits, then new additional 144 6-18 Section 2 and characterization, new groundwater monitoring wells, or routine elsewhere in monitoring requirements should NOT be required. Ground water this AMP monitoring has shown little change over many years of trending when there are no new contributors to a site and this is not an insignificant expense.
Radiation surveys as part of concrete aging management-Radiation surveys are initially conducted per Tech. Spec.
Table 6-3, requirements. Usually, this means the cask is surveyed and Element 1, verified to meet requirements before it is allowed into storage.
Section 3 &
145 6-18 The need for follow-up surveys, as part of aging management, is Footnote 1 and unnecessary. Dry cask storage systems are routinely performed elsewhere in and evaluated IAW site approved procedures. This approach is this AMP more conservative and robust than an extra and unnecessary aging management driven survey process. This is not ALARA.
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Item 3 is not related to aging management. Sites are already required to perform periodic monitoring of boundary doses which are cited as sufficient to detect failed systems and abnormal conditions. Sufficient radiation monitoring will already be Table 6-3, performed during execution of other aging management activities 146 6-18 Element l, above and beyond surveys required by DSS and ISFSI TS Item 3 requirements . A radiation monitoring requirement embedded in an AMP is not only not substantiated but it will cause undue work and exposure to station employees for no gained benefit (Not in keeping with ALARA principals). All other references to radiation surveys should be removed from Table 6-3.
Leaching of calci um silicate and efflorescence as part of concrete aging management This statement should say "due to excessive Table 6-3, leaching of calcium hydroxide". As discussed in NU REG CR-7116, Element 1, 4th NUREG CR-7153 and IAEA TECDOC-1025, calcium leaching is not 6-18 bullet from t he considered a significant degradation mechanism and would not 147 Also 6- bottom and result in a significant impact to the safety functions of the 20 elsewhere in concrete. There are many instances of leaching of calcium this AMP carbonate on dry storage canisters and most are self-limiting and
& Element 3 benign. Only the more excessive ones with corresponding affected concrete needs to be tracked, evaluated and potentially remediated.
The first paragraph should be deleted. This paragraph is not substantiated by the MAPs document and is not related to or required to mitigate any aging management issues. TS monitoring requirements (via temperature monitoring or inlet/outlet Table 6-3, 148 6-19 inspections) are not changed during license renewal and will Element 2 continue to be required on a more frequent basis than prescribed here. Any abnormal conditions identified by this already required monitoring will be corrected by licensee's corrective action program .
Table 6-3 references 100% inspection of all concrete structures.
This overly prescriptive and should be an inspection based on sampling a few systems at 100% and a gross inspection of all Table 6-3 other structures. Detailed inspection of a few systems will be 149 6-21 Element 4, sufficient to identify initiation of issues that would warrant "Sample Size" increased inspections per licensee's CAP program (extent of condition evaluation) . Performing 100% surface exam of all systems will result in undue work and exposure with no measureable benefit.
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These types of inspections are conducted as required by the certificate or part 72 license for the period of initial operation (during the annual inspections). Therefore, the timing of the initial inspections required by the AMP should be after the period of Table 6-3, initial operation, which is consistent with NRC language on prior 150 6-22 Element 4, renewals. NRC has allowed up to 300 days after the effective date "Timing" of the renewal to implement the AMP. Initial inspection timing is driven by the AMP implementing documents. What is specified here is contrary to recent rulemaking at 82 Federal Register 57819 renewing certificate of compliance No. 1004 - TN Americas LLC, Standardized NUHOMS" Horizontal Modular Storage System .
Too Prescriptive-While it may be important for canister walls with controlled thickness on the walls, it is not necessary to perform Table 6-4, visual inspections of coated carbon steel surfaces to VT-3 and this Element 4, will indeed cause hardship on ISFSI Only sites who do not have 151 6-30 "Readily qualified personnel for this. There are numerous examples of rust Accessible spots in exposed carbon steel and this is readily detected by Surfaces" average inspections. Most coatings are NQ and do not perform a safety function on metallic dry storage components and do not require this level of inspection.
These types of inspections are conducted as required by the certificate or part 72 license for the period of initial operation (during the annual inspections) . Therefore, the timing of the initial inspections required by the AMP should be after the period of Table 6-4, initial operation, which is consistent with NRC language on prior 152 6-31 Element 4, renewals. NRC has allowed up to 300 days after the effective date "Timing" of the renewal to implement the AMP. Initial inspection timing is driven by the AMP implementing documents. What is specified here is contrary to recent rulemaking at 82 Federal Register 57819 renewing certificate of compliance No. 1004 - TN Americas LLC, Standardized NU HOMS* Horizontal Modular Storage System.
Table 6-2 does not include a statement similar to this one: The extent of inspection coverage should be specified Table 6-4 153 6-31 and demonstrated to sufficiently characterize the condition of the Element 4 metallic components. It is unclear why this is applicable in Table 6-4 and not in Table 6-2.
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Too prescriptive- These acceptance criteria are impossibly restrictive . For instance, minor coating failures for structures that have been in service for 10+ years are common place. As most coatings on external surfaces are NO, they have no impact on Table 6-4, safety. Minor superficial rust in the areas of these coating failures Element 6, 154 6-32 is also evident. The first bullet here is fine as is the 4th bullet. The "Acceptance 2nd and 3'd bullets however, are unduly restrictive . This is Criteria" important because of the corrective actions section. Failure to meet bullets 2 and 3 are not necessarily conditions adverse to quality and do not require apparent cause evaluations and root cause evaluations.
Too prescriptive- This section provides details that are in all licensee's corrective action programs and is unnecessary and may be counterproductive. For instance, based on the above Table 6-4, comment, failure to meet bullets 2 and 3 are not necessarily Element 7, conditions adverse to quality and do not require apparent cause 155 6-33 "Corrective evaluations and root cause evaluations. While such actions may Actions" be warranted on canister shells, they are not necessary on large, thick coated carbon steel structures, where unsatisfactory amounts of corrosion will be readily visible at inspection opportunities.
Table 6-4, The operating experience cited for this AMP has nothing to do Element 10, with the issues addressed by this AMP. There is Op-Ex in the AMID 156 6-33 "Operating database that applies to this AMP .
Experience" This AMP is not necessary. The passive ventilation systems of dry storage systems are subject to daily tech. spec. verification and the structures associated with forming the passive ventilation 157 6-35 Section 6.8 system are addressed in other AMPS. By creating this AMP, a source of confusion, conflicting requirements, and unnecessary administrative requirements are also being created. We recommend deleting this AMP .
This entire section should be deleted. Validation of ventilation acceptability is already required by DSS TS and is not specifically related to or required to mitigate any aging management condition . DSS TS already have a monitoring requirement 158 6-35 Section 6.8 sufficient to detect abnormal conditions. Any abnormal condition will then be entered into the licensee's CAP. This section will do nothing but provide additional administrative burden on the licensee.
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The statement "Temperature monitoring is performed with qua lified and cal ibrated measurement devices or sensors that are Table 6-5, maintained in accordance w ith the site QA program" is not Element 4, 159 6-37 generally true. As these sensors have no impact on safety and are "Operating generally very rel iable and not subject to significa nt inaccuracy, Experience" they are not in scope and not subject to these requirement at most ISFSls.
Table 6-5, Same comment as above ... this equipment is generally not Element 9, calibrated . RTD's and T/C either operate or fail...fai lure is obvious 160 6-40 "Operating based on the circuitry for these systems.
Experience Bolted cask sea l leakage monitoring systems have a TS t o ensure the pressure is not decreasing and to ensure t he system is funct ioning properly. An AMP is not necessary and is not specifica lly related to or required to mitigate any aging Section 6.9 ma nagement condition. DSS TS already have a monitoring 161 6-43 and Table 6-6 requirement sufficient to detect abnormal condit ions. Any abnormal condition will the n be entered into the licensee's CAP.
Th is section will do nothing but provid e additional administrative burden on th e licensee.
These types of inspections are conducted as required by the certificate or part 72 license for the period of initia l operation (during the annual inspections). Therefore, the timing of the initial inspections required by the AMP shou ld be after the period of Table 6-6, initial operation, which is consistent with NRC language on prior 162 6-46 Element 4, renewals . NRC has allowed up to 300 days after the effective date "Timing" of the renewal to implement the AMP. Initial inspection timing is driven by the AMP implementing documents. What is specified here is contrary to recent rulemaking at 82 Federa l Register 57819 renewing certificate of compliance No. 1004 - TN Americas LLC, Standardized NUHOMs* Horizontal Mod ular Storage System .
Some discuss ion should be added here t hat TCs wo uld genera lly be considered t ools and scoped out for aging management if t hey 163 6-53 Section 6.10 we re not class ified as important to safety. They are not in continuous service and play no role in sto rage operations at t he ISFSI.
Table 6-7, " No coating defects" is an impossible acceptance criteria for Element 6, transfer casks. Coating defects should be identified, but do not 164 6-56 "Acceptance necessarily need to be repair unless this is required by the Criteria" corrective action program .
165 6-57 Table 6-7 Appears to be something missing in the second reference.
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Table 6-8, There should be a statement in the Scope that this AMP does not Element 1, apply if the HBU fuel is canned or otherwise contained .
166 6-60 "Scope of the Program" Says the scope is to provide a description of the design bases characteristics ofthe HBU fuel. You do not "design" HBU fuel. You Table 6-8 design the DSS. Suggest changing "design bases characteristics" to 167 6-60 Item 1 something like "characteristics and properties assumed in the DSS design". This comment applies throughout Table 6-8 HBU Fuel AMP .
Table 6-8 Nominal burnups in the HDRP are 50-55 GWD/MTU assembly 168 6-60 Item 1 average (not 53-58).
Table 6-8 Change "is to be licensed" to "is licensed" 169 6-60 Item 1 Last paragraph in Item 1 says to justify the surrogate demonstration program is applicable by demonstrating it is bounding for the specific licensee. Suggest adding language to Table 6-8 justify use of the demonstration program if it is not bounding, 170 6-60 Item 1 similar to language from ISG e.g. " ... that the demonstration fuel is reasonably characteristic of the stored fuel and the added burn -up will not change the results determined by the demonstration."
Table 6-8, Do these three bullets/acceptance criteria make sense? Do they Element 6, have a safety basis?
171 6-61 "Acceptance Criteria"