Text

Non-proprietary Rev 4 to HI-982083, Licensing Rept for Spent Fuel Rack Installation at Byron & Braidwood Nuclear Stations
	ML20204H994
Person / Time
Site:	Byron, Braidwood
Issue date:	03/03/1999
From:	Bullard C, Gupta V; HOLTEC INTERNATIONAL
To:	;
Shared Package
ML20137D740	List: ML20204H966; ML20204H978; ML20204H994;
References
	HI-982083, HI-982083-R04, HI-982083-R4, NUDOCS 9903290232
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ATTACHMENT F

' HOLTEC INTERNATIONAL LICENSING REPORT (Non-proprietary)

HI-982083 i

9903290232 990323 PDR ADOCK 05000454 P PDR __

F-1 J

Holtec Center,555 Lincoln Drive West, Marlton, NJ 08053

{ (( Telephone (609) 797-0900 Fax (609) 797-0909 l

I LICENSING REPORT for SPENT FUEL RACK INSTALLATION i at BYRON AND BRAIDWOOD NUCLEAR STATIONS 1

Holtec Report HI-982083 (Non-Proprietary Version)

Report Category: A l Prepared for Commonwealth Edison Co.

Purchase Order No. 367585 Holtec Project 80944 I

COMPANY PRIVATE This document version has all proprietary information removed and has replaced those sections, figures, and tables with highlighting and/or notes to designate the removal of such information. This document is to be used only in connection with the performance of work by Holtec Internationai or its designated subcontractors. Reproduction, publication or presentation, in whole or in part, for any other purpose by an i party other than the Client is expressly forbidden.

q L_) _] L____ k_._j Holtic C?nt r,555 Lincoln Drive Wrst, Mrriton, NJ 08053

*P" HOLTEC "a (S S) 7S7- S Fax (609) 797 - 090!

INTERNATlONAL QA AND ADMINISTRATIVEINFORMATION LOC (To Be Filled in By the Principal Author of the Document and Placed After the Title Page)

CATEGORY: O Generic Document No Hi 982083 S Project Specifi.c Holtec Project No 80944 in accordance with the Holtec Quality Assurance Manual, and associated Holtec Quality Procedures (HQPs), this document is categorized as a :

O Calculation Package

2 Technical Report (Per HQP (Per HQP 3.2) 3.2)

(Such as a Licensing report)

Design Criterion O Design O Specification Document (Per HQP 3.4)

O Other (Specify):

The formatting of the contents of this document is in accordance with the instructions of HQP 3.2 or 3.4 except as noted below:

This document is labelled :

l 0 Nonproprietary Z Holtec Proprietary O Privileged intellectual Property (PIP)

Documents labelled Privileged intellectual Property contains extremely valuable intellectual / commercial property of Holtec International. They can not be released to external organizations or entites without explicit approval of a company corporate officer. The recipient of Holtec's proprietary or Privileged Intellectual Property (PIP) document bears full and undivided responsibility to safeguard it against loss or duplication.

Revisions to the calculation Packages may be made by adding supplements to the document and replacing the

" Table of Contents", the " Review and Certification" page and the " Revision Log".

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m AFFIDAVIT I'URSUANT TO 10CFR2.790 I, Michael P. McNamara, being duly sworn, depose and state as follows:

(1) I am the: Director of Nuclear Projects for Holtec International and have been delegated the function of reviewing the information described in paragraph (2) which:is sought to be withheld, and have been authorized to apply for its withholding; (2) The information sought to be withheld is contained in the document entitled

" Licensing Report for Spent Fuel Rack Installation at Byron and Braidwood Nuclear Stations", Hu.c Report HI-982083.

. i (3) In making this application for withholding of proprietary mformation of which it l" is the owner, Holtec International relies upon the exemption from disclosure set forth in the Freedom of Information Act ("FOIA"), 5 USC Sec. 552(b)(4) and the Trade Secrets Act,18 USC Sec.1905, and NRC regulations 10CFR Part 9.17(a)(4),'2.790(a)(4), and 2.790(b)(1) for " trade secrets and commercial or l financial information obtained from a. person and privileged or confidential" (Exemption 4). The material for which exemption from disclosure is here sought is all " confidential commercial information", and some portions also I qualify under the narrower definition of " trade secret", within the meanings assigned to those tenns for purposes of FOIA Exemption 4 in, respectively, Critical Mass Enerev Project v. Nuclear Regulatory Commission, 975F2d871 (DC Cir. 1992), and Public Citizen Health Research Group v. FDA, 1 704F2d1280 (DC Cir.1983). J (4) Some examples of categories of information which fit into the definition of proprietary information are:

a. Information that discloses a process, method, or apparatus, including supporting data and analyses, where prevention of its use by Hohec's competitors without license from Holtec International constitutes a competitive economic advantage over other companies; 1

~

l

AFFIDAVIT PURSUANT TO 10CFR2.790

b. Information which, if used by a competitor, would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing of a similar product.

c. -Information which reveals cost or price information,- production, capacities, budget levels, or commercial strategies of Holtec International, i its customers, or its' suppliers;

d. Information which reveals aspects of past, present, or future Holtec International customer-funded development plans and programs of potential commercial value to Holtec International;

e. Information which discloses patentable subject matter for which it may be 4 desirable to obtain patent protection. )

i The information sought to be withheld is considered to be proprietary for the reasons set forth in paragraphs 4.a,4.b,4.d, and 4.e, above. ]

l i

(5) The information sought to be withheld is being submitted to the NRC in 1 confidence. The information (including that compiled from many sources) is of

{

a sort customarily held in confidence by Holtec International, and is in fact so )

held. The information sought to be withheld has, to the best of my knowledge and belief, consistently been held in confidence by Holtec International. No public disclosure has been made, and it is not available in public sources. All disclosures to third parties, including any required transmittals to the NRC, have i been made, or must be made, pursuant to regulatory provisions or proprietary j agreements which provide for maintenance of the information in confidence. Its initial designation as proprietary information, and the subsequent steps taken to 1 prevent its unauthorized disclosure, are as set forth in paragraphs (6) and (7) j followmg. .

1 (6) Initial approval of proprietary treatment of a document is made by the manager I of the originating component, the person most likely to be acquai'ited with the value and sensitivity of the information in relation to industry knowledge.

Access to such documents within Holtec International is limited on a "need to ,

i 2

l 1

1

AFFIDAVIT PURSUANT TO 10CFR2.790 know" basis.

(7) The procedure for approval of external release of such a document typically requires review by the staff manager, project manager, principal scientist or other equivalent authority, by the manager of the cognizant marketing function (or his designee), and by the Legal Operation, for technical content, competitive effect, and determination of the accuracy of the proprietary designation.

Disclosures outside Holtec International 'are limited to regulatory bodies, customers, and potential customers, and their agents, suppliers, and licensees, and others with a legitimate need for the information, and then only in accordance with appropriate regulatory provisions or proprietary agreements.

(8) The information classified as. proprietary was developed and compiled by Holtec International at a significant cost to Holtec International. This information is classified as proprietary because it contains detailed historical data and analytical results'not available elsewhere. This information would provide other parties, including competitors, with information from Holtec International's technical database and the results of evaluations performed using codes developed by Holtec International. Release of this information would improve a competitor's position without the competitor having to expend similar resources for the development of the database. A substantial effort has been expended by Holtec International to develop this information.

(9)- Public disclosure of the information sought to be withheld is likely to cause substantial harm to Holtec International's competitive position and foreclose or reduce the availability of profit-making opportunities. The information is part of Holtec International's comprehensive spent fuel storage technology base, and its commercial value extends beyond the original development cost. The value of the technclogy base goes beyond the extensive physical data'oase and analytical methodology, and includes development of the expertise to determine and apply the appropriate evaluation process.

The research, development, engineering, and analytical costs comprise a substantial investment of time and money by Holtec International.

3

n AFFIDAVIT PURSUANT TO 10CFR2.790 The precise' value of the expertise to devise an evaluation process and apply the correct' analytical methodology is difficult to quantify, but it clearly is substantial.

Holtec'. International's competitive advantage will be lost if its competitors are able to,use the results of the Holtec International experience to normalize or verify their own process or if they are able to claim an equivalent understanding by demonstrating that they can arrive at the same or similar conclusions.

The value of this information to Holtec International would be lost if the {

information were disclosed to the public. Making such information available to j competitors without their having been required to undertake a similar I expenditure of resources would unfairly provide competitors with a windfall, and deprive Holtec' International of the opportunity to exercise its competitive advantage to seek an adequate return on its large investment in developing these very valuable analytical tools. l STATE OF NEW JERSEY )

) ss: ;

COUNTY 0F BURLINGTON )

Michael P. McNamara, being duly sworn, deposes and says:

That he has read the foregoing affidavit and the matters stated therein are true and correct to the best of his knowledge, information, and belief. I Executed at Marlton, New Jersey, this 15th day of February 1999.

l Michael P. McNamara Holtec International l

Subscribed and sworn before me this

day of A"*1, ,1999.

i i 7

A 1Ah ,

4 '

JERSEY NOTARY PUBLIO OF MyCommission Expives AprH 25,2000

)

ATTACHMENT A BYRON STATION, UNITS 1 AND 2 BRAIDWOOD STATION, UNITS 1 AND 2 DESCRIPTION AND SAFETY ANALYSIS FOR PROPOSED CHANGES A.

SUMMARY

OF PROPOSED CHANGES in accordance with 10 CFR 50.90, Commonwealth Edison (Comed) Company is requesting a change to Appendix A, Technical Specifications (TS) of Facility Operating License Nos.

NPF-72, NPF-77, NPF-37 and NPF-66, for Braidwood Station, Units 1 and 2, and Byron Station, Units 1 and 2, respectively.

The proposed changes to Technical Specifications (TS) Sections 3.7.15, " Spent Fuel Pool Boron Concentration"; 3.7.16, " Spent Fuel Assembly Storage"; 4.3.1, " Criticality"; and 4.3.3,

" Capacity"; support installation of new Boral high-density spent fuel pool storage racks at the Byron and Braidwood Stations. The proposed changes reflect our plan to remove all 23 of the existing spent fuel storage racks at each station and replace them with 24 new spent fuel storage racks. Byron Station plans to install the new spent fuel storage racks starting in January 2000. Braidwood Station plans to install the new spent fuel storage racks starting in January 2001.

During the installation of the new Holtec spent fuel pool storage racks, both Holtec and the existing Joseph Oat spent fuel pool storage racks will be in the spent fuel pool at the same time. At the completion of installation, only Holtec spent fuel pool storage racks will remain in the spent fuel pool. The proposed changes to the TS address the requirements for both the new Holtec racks, during and after installation, and the existing Joseph Oat racks, during the Holtec rack installation.

A complete description of the proposed changes is given in Section E of this Attachment A.

The _ marked-up TS pages are provided in Attachments B-1 and B-2 for Braidwood and Byron Stations, respectively.

B. DESCRIPTION OF CURRENT REQUIREMENTS TS Limiting Condition for Operation (LCO) 3.7.15, "Sperd Fuel Pool Boron Concentration,"

and LCO 3.7.16, " Spent Fuel Assembly Storage," specify requirements for storage of fuel assemblies in the spent fuel pool. The following are the specific requirements.

LCO 3.7.15 currently states: "The spent fuel pool boron concentration shall be 2 2000 ppm." This requirement is applicable whenever fuel assemblies are stored in the spent fuel pool. For a condition where the spent fuel pool boron concentration is not within the limit, Required Action, A.1, is to immediately suspend movement of fuel assemblies in the spent fuel pool. Required Action, A.2, is to immediately initiate action to restore spent fuel pool boron concentration to within limit.

A-1 J

TS Surveillance Requirement (SR) 3.7.15.1 specifies that the spent fuel pool boron concentration be verified to be within limits every 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

LCO 3.7.16.a. places limitations on fuel stored in Region 1 racks. Region 1 fuel assemblies shall, "Have an initial nominal enrichment of s 4.7 weight percent U-235 or satisfy a minimum number of Integral Fuel Bumable Absorbers (IFBAs) for higher enrichments up to 5.0 weight percent U-235 to permit storage in any cell location." LCO 3.7.16.b. places limitations on fuel stored in Region 2 racks. Region 2 fuel assemblies shall, "Have a combination of initial enrichment, bumup, and decay time within the Acceptable Bumup Domain of Figure 3.7.16-1, 3.7.16-2, or 3.7.16-3, as applicable for that storage configuration." LCO 3.7.16.c.. places restrictions on assembly placement within and between adjacent racks and currently reads,

" Comply with the Interface Requirements within and between adjacent racks."

. When the requirements of the LCO are not met, Required Action, A.1, specifies to immediately initiate action to move the noncomplying fuel assembly to a location which restores compliance.

SR 3.7.16.1 states to, " Verify by administrative means the initial nominal enrichment of the fuel assembly is s 4.7 weight percent U-235 or a minimum number of IFBAs is met," prior to storing a fuel assembly in Region 1.

SR 3.7.16.2 states to, " Verify by administrative means the combination of initial enrichment, bumup, and decay time of the fuel assembly is within the Acceptable Bumup Domain of Figure 3.7.16-1,3.7.16-2, or 3.7.16-3," prior to storing a fuel assembly in Region 2.

SR 3.7.16.3 states to " Verify by administrative means the interface requirements within and between adjacent racks are met."

Figure 3.7.16-1 establishes the Region 2 rack "All Cell configuration Bumup Credit Requirements." This curve put limitations on assembly storage in the Region 2 racks based on a function of initial enrichment vs. assembly bumup. Figure 3.7.16-2 establishes the Region 2 rack "3-out-of-4 Checkerboard Configuration Burnup Credit Requirements." This curve put limitations on assembly storage in the Region 2 racks based on a function of initial enrichment vs. assembly burnup for a 3-out-of-4 checkerboard configuration. Figure 3.7.16-3 ~

establishes the Region 2 rack "2-out-of-4 Checkerboard Configuration Burnup Credit Requirements." This curve put limitations on assembly storage in the Region 2 racks based on a function of initial enrichment versus assembly burnup for a 2-out-of-4 checkerboard

, configuration.

TS Design Features, Section 4.3.1, addresses criticality requirements of the spent fuel pool storage racks. The following are the specific requirements.

Subsection 4.3.1.a. states that Qe spent fuel storage racks are designed and shall be maintained with: " Fuel assemblies having a maximum U-235 enrichment of 5.0 weight percent."

Subsection 4.3.1.b. states that the spent fuel storage racks are designed and shall be maintained with: "k,< 1.0 if fully flooded with unborated water which includes an allowance for uncertainties as described in WCAP-14416-NP-A, ' Westinghouse Spent Fuel Rack Criticality Analysis Methodology.'"

Subsection 4.3.1.c. states that the spent fuel storage racks are designed and shall be A-2 l J

maintained with: "k,s 0.95 if fully flooded with water borated to 550 ppm, which .

includes an allowance for uncertainties as described in WCAP-14416-NP-A,

' Westinghouse Spent Fuel Rack Criticality Analysis Methodology.'"

Subsection 4.3.1.d. states that the spent fuel storage racks are designed and shall be maintained with: "A nominal 10.32 inch north-south and 10 42 inch east-west center to center distance between fuel assemblies placed in Region 1 racks."

Subsection 4.3.1.e states that the spent fuel storage racks are designed and shall be maintained with: "A nominal 9.03 inch center to center distance between fuel assemblies placed in Region 2 racks."

TS Section 4.3.3, states: "The spent fuel pool is designed and shall be maintained with a storage capacity limited to no more than 2870 fuel assemblies."

1' C. BASES FOR THE CURRENT REQUIREMENTS The Byron and Braidwood spent fuel pools were reracked in the 1980's to avoid a potential loss of full core discharge capability. The current requirements were issued in approved License Amendment No. 25, dated March 17,1989, for Byron Station, and License

, Amendment No. 20, dated July 20,1989, for Braidwood Station.

Backoround The spent fuel pool provides for storage of fuel assemblies of different initial fuel enrichments and exposure histories in two distinct regions. There are 23 separate racks, which provide placement locations for a total of 2870 new or used fuel assemblies, included in this are six specific storage locations in one of the racks for placement of failed fuel assemblies. These locations are identified as the failed fuel storage cells. Of the 23 racks, four are designated " Region 1," with the remaining 19 racks designated as '

" Region 2."

{

Region 1 racks contain 392 cells which are analyzed for storing fuel assemblies in an "All Cells" arrangement (i.e., the criticality analysis assumes that spent fuel assemblies reside in )

all available celllocations, with the exception of the boundary requirements). The stored l fuel assemblies may contain an initial nominal enrichment of s 4.7 weight percent U-235 )

(i.e., without Integral Fuel Burnable Absorbers (IFBAs) installed) or up to an initial nominal j enrichment of s 5.0 weight percent U-235, provided that the requirement for a minimum I number of 16 IFBAs is met.

Region 2 recks contain 2472 cells which are also analyzed for storing fuel assemblies in a combination of storage configurations. These patterns are:

1. "All Cells" Storage;

2. "3-out-of-4 Checkerboard" Storage; and

3. "2-out-of-4 Checkerboard" Storage.

For the "All Cells" storage configuration, the stored fuel assemblies may contain an initial nominal enrichment of s 1.14 weight percent U-235 (i.e., without taking credit for fuel burnup or radioactive decay of fuel constituents) up to an initial nominal enrichment of s 5.0 A-3 1

weight percent U-235, when fuel bumap and radioactive decay of fuel constituents are credited.

For the "3-out-of-4 Checkerboard" storage configuration, the stored fuel assemblies may ;

contain an initial nominal enrichment of s 1.64 weight percent U-235 (i.e., without taking

- credit for fuel bumup~or radioactive decay of fuel constituents) up to an initial nominal enrichment of s 5.0 weight percent U-235, when fuel bumup and radioactive decay of fuel constituents are credited. In this storage pattem, there can be no more than three stored assemblies in any 2X2 matrix of celllattices.

For the "2-out-of-4 Checkerboard" storage configuration, the stored fuel assemblies may contain an initial nominal enrichment of s 4.10 weight percent U-235 (i.e., without taking credit for fuel burnup) up to an initial nominal enrichment of s 5.0 weight percent U-235, when fuel bumup is credited. In this storage pattern, no two fuel assemblies may be stored .

" face adjacent"(i.e., there must be an empty cell opposite each face of the fuel assembly).

The water in the spent fuel pool normally contains soluble boron which results in large suberiticality margins under actual operating conditions.

The criticality analyses, performed using NRC approved methodologies, for the spent fuel assembly storage racks confirm that k., remains < 1.0, including uncertainties and tolerances, at a 95% probability with a 95% confidence level (i.e.,95/95 basis), based on the accident condition of the pool being flooded with unborated water. Thus, the design of both regions assumes the use of unborated water while maintaining stored fuelin a subcritical condition.

l However, the presence of soluble boron has been credited to provide adequate safety margin to maintain spent fuel assembly storage rack k , s 0.95 (i.e., also on a 95/95 basis) for all postulated accident scenarios involving dropped or misloaded fuel assemblies and loss of spent fuel pool temperature control. Crediting the presence of soluble boron for mitigation of these scenarios is acceptable based on applying the " double contingency principle" which states that there is no requirement to assume two unlikely, independent, concurrent events to ensure protection against a criticality accident. j The accident analyses address the following five postulated scenarios:

1. fuel assembly drop on top of rack;

2. fuel assembly drop between rack modules;

3. fuel assembly drop between rack modules and spent fuel pool wall;

4. change in spent fuel pool water temperature; and i

._5. fuel assembly loaded contrary to placement' restrictions.

Of these, only the last two have the capacity to increase reactivity beyond the analyzed condition.

Calculations were performed to determine the reactivity change caused by a change in spent fuel pool water temperature outside the normal range (50 - 160 F). For the change in spent fuel pool water temperature accident, a temperature range of 32 - 240 F is

. considered in all cases, additional reactivity margin is available to the 0.95 k , limit to allow for temperature accidents. The temperature change accident can occur at any time during A-4

l operation of the spent fuel pool.

I For the fuel assembly mistoad accident, calculations were performed to show the largest reactivity increase caused by a fuel assembly misplaced into a storage cell for which the restrictions on location, enrichment, or burnup are not satisfied. The assembly misload l accident can only occur during fuel handling operations in the spent fuel pool.

Spent fuel pool soluble boron has been credited in the criticality safety analysis to offset storage rack and fuel assembly tolerances, calculational uncertainties, uncertainty associated with bumup credit and the reactivity increase caused by postulated accident conditions. l i

Based on the above discussion, should a spent fuel pool water temperature change accident or a fuel assembly misload accident occur in the Region 1, Region 2, or failed fuel storage cells, k., will be maintained s to 0.95 due to the presence of at least 550 ppm (i.e.,

no fuel handling) or 1650 ppm (i.e., during fuel handling) of soluble boron in the spent fuel pool water.

A spent fuel pool dilution analysis has been performed as required. The analysis assumes an initial boron concentration of 2000 ppm. The dilution analysis concludes that an !

unplanned or inadvertent event that would result in the dilution of the spent fuel pool boron I concentration from 2000 ppm to 550 ppm (i.e., minimum non-accident boron concentration) !

is not credible.

I interface requirements have been established to ensure k ,is maintained within the {

appropriate limits. There are interface requirements between Region 1 racks, between !

Region 1 and Region 2 racks, between Region 2 racks, and within racks between different checkerboard configurations. These requirements are necessary to account for unique geometries and configurations which exist at the interfaces, interface requirements exist between adjacent racks to account for the potential reactivity increase in 3-out-of-4 and 2-out-of-4 storage configurations along the interface with non-aligned racks.

Specification 3.7.15. " Spent Fuel Pool Boron Concentration" The presence of soluble boron is assumed in the criticality analyses and is credited for ensuring that spent fuel pool k , will be maintained s 0.95 at a 95% confidence level for all storage configurations. The 2000 ppm minimum boron concentration is also an initial condition in the spent fuel pool dilution analysis. Therefore, the restriction on soluble boron concentration in the spent fuel pool water is maintained at all times when fuel assemblies ,

are stored in the spent fuel pool. j When the concentration of boron in the spent fuel pool is less than required, immediate action must be immediately taken to suspend the movement of fuel assemblies. Immediate actions are required to restore spent fuel pool boron concentration to 2 2000 ppm.

In addition, a surveillance requirement is provided to verify that the concentration of boron in the spent fuel pool is within the required limit once per 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . The 48 nour frequency is based on operating experience and takes into consideration the fact that significant changes in the boron concentration in the spent fuel pool are difficult to produce without detection, considering the large volume of water contained in the spent fuel pool. However, an analysis has concluded that a spent fuel pool boron dilution event of sufficient magnitude to A-5

w I

l reduce boron concentration below the minimum non-accident requirement is not credible.

Specification 3.7.16. "Soent Fuel Assembly Storaae" 1

The restrictions on the placement of fuel assemblies, including interface requirements, l within the spent fuel pool Jn accordance with the requirements in the accompanying LCO l ensure that the k, of the spent fuel pool will always remain < 1.0 assuming the pool is flooded with unborated water and s 0.95 assuming the presence of 550 ppm soluble boron in the pool. These restriction are appl.icable whenever fuel assemblies are stored in the

. .sperit fuel pool.

In Figures 3.7.16-1 and 3.7.16-2, the Acceptable Burnup Domain is provided as a function of bumup,' enrichment, and decay time for the fuel assembly to be stored. In Figure 3.7.16-3, the Acceptable Burnup Domain is provided as a function of burnup and enrichment for the fuel assembly to be stored. The Acceptable Burnup Domain and the Unacceptable Burnup Domain are separated by a single line because decay time is not credited in the 2-out-of-4 Checkerboard storage configuration.

When the configuration of fuel assemblies stored in the spent fuel pool is not in accordance with the requirements of the LCO, immediate action must be taken to make the necessary fuel assembly movement (s) to bring the configuration into compliance.

A surveillance requirement (i.e., SR 3.7.16.1) is provided to verify by administrative means that the initial nominal enrichment of the fuel assembly or a minimum number of 16 IFBAs is met to ensure that the assumptions of the safety analyses are preserved. This surveillance is performed prior to storing the fuel assembly in the intended spent fuel pool storage location.

Another surveillance requirement (i.e., SR 3.7.16.2) is provided to verify by administrative means that the combination of initial enrichment, burnup, and decay time of the fuel assembly are within the Acceptable Burnup Domain of Figure 3.7.16-1,3.7.16-2, or 3.7.16-3 for the intended storage configuration to ensure that the assumptions of the safety analyses are preserved. This surveillance is performed prior to storing the fuel assembly in the intended spent fuel pool storage location.

Verification by administrative means that the interface requirements within and between adjacent racks are met is required by SR 3.7.16.3 to ensure that the assumptions of the safety analyses are preserved. This verification is performed prior to storing the fuel assembly in the intended spent fuel pool storage location.

1 Soent Fuel Pool Bases in contrast to many spent fuel pools in light water reactor installations, the Byron and Braidwood Stations spent fuel pool structures are founded on the base mat and have no sub-base trenches or penetrations which may interrupt the diffusion of the structural weight I into the sub-surface rock. The Byron and Braidwood Stations, designed as exact replicas of ;

each other, feature identical fuel pool structures: the two pool structures are identical l except for their designation of " plant north." The geological terrain at the two sites is also somewhat different. While the geological rock formation at the E!yron Station is very close to the surface, permitting the base mat to be poured directly on the rock bed, an intervening j layer of soil at Braidwood Station serves as a less rigid buffer between the base mat and i

A-6 '

w the geologic rock half-space. In summary, aside from the characten, tic of sub-surface foundation, the Byron Station and Braidwood Station pools are structurally identical.

The pool liner in the Byron Station and Braidwood Station pools is not a safety-related component. The "non-safety related" designation for the pool liner arises from the fact that its failure would not cause a rapid lowering of the water level in the fuel pool. However, in a i pool filled with closed packed array of rack modules, accessing of the liner on the pool slab for inspection and repair is not feasible or practicable. Therefore, it was necessary to ensure that the rack pedestal and bearing pad design do not produce a state of overstress in the liner, leading to failure from tearing or cyclic fatigue during seismic events.

Like all fuel pool liners, the Byron Station and Braidwood Station pool liners are subjected to the weight of water, lateral and vertical forces, plus any seismic additive forces from individual spent fuel rack support pedestals. The liner can be thought of as a stainless steel plate resting on a concrete substrate that acts as an elastic foundation. Because the coefficient of friction at the stainless steel-to-stainless steel interface exceeds that at the stainless steel-to-concrete interface, the portion of the liner not welded to embedments in the concrete may sustain in-plane strains under the action of the friction-induced lateral I forces induced by fuel rack motion during a seismic event. I Calculations were made to establish that the liner would not tear or rupture under all loading conditions in the pool and that the liner could withstand one Safe Shutdown Earthquake (SSE) and 20 Operating Basis Earthquake (OBE) seismic events without fatigue failure. The cumulative damage factor, under one SSE and 20 OBE events, was calculated to be well below the acceptance limit of 1.0. An evaluation of the liner plate section subjected to vertical and horizontal static pedestal loading has also been performed and shown to be satisfactory.

I Existino Fuel Storaae Racks Bases The spent fuel storage racks provide a place in the spent fuel storage pool for storing the spent fuel discharged from the reactor vessel. They are top entry racks, designed to preclude the possibility of criticality under both normal and abnormal conditions. Total j capacity of the racks is 2870 fuel assembly storage cells. j The design of the spent fuel storage rack assembly prevents any possibility of accidental criticality.

i A total of 392 storage cells are provided in four racks in Region 1 which are analyzed for storing Westinghouse OFAs in an "All Cells" arrangement (i.e., the criticality analysis assumes that spent fuel assemblies reside in all available cell locations, with the exception of the boundary requirements). The stored fuel assemblies may contain an initial nominal enrichment of s 4.7 weight percent U-235 without integral Fuel Burnable Absorbers (IFBAs) installed or up to an initis nominal enrichment of s 5.0 weight percent U-235, provided that the requirement for a minimum number of 16 IFBAs is met.

A total of 2472 storage cells plus six failed fuel storage cells are provided in Region 2 in 19 racks. The six failed fuel storage cells are located in Rack J1. The Region 2 cells are i capable of accommodating fuel assemblies with various initial enrichments and minimum )

accumulated burnup requirements. ]

l l

A-7 )

There are 23 discrete rack modules arranged in the fuel pool. The racks are freestanding l and self-supporting. Each rack module is equipped with girdle bars, one-inch thick by three and one-half inches high. The nominal gap between cell walls of adjacent modules is two inches. The modules make surface contact between contiguous sides at the girdle bar locations and thus maintain a specified gap between them.

Region 1 Racks The rack module is fabricated from ASTM A-240-304L austenitic stainless steel sheet and plate material, ASME SA351-CF3 and SA217-CA15 casting material, and SA479-410 steel material. The weld filler material utilized in body welds is ASME SFA-5.9, Types 308L and 308LSI. Boraflex and Boral serve as the neutron absorber material.

A typical module contains storage cells, which have an 8.85-inch nominal square cross-sectional opening. This dimension ensures that fuel assemblies with maximum expected axial bow can be inserted and removed from the storage cells without any damage to the fuel assemblies or the rack modules.

The cells provide a smooth and continuous surface for lateral contact with the fuel l assembly. The rack module components are as follows:

I

a. Intemal Square Box 1 This element provides the lateral bearing surface to the fuel assembly. It is fabricated by joining two formed channels using a controlled seam welding operation. This element has an 8.85-inch square internal cross-section and is 168-7/8 inches long.

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I

b. Neutron Absorber Material (Boraflex) l Boraflex is placed on all four sides of a square tube over a length of 139.5 inches for Byron Station and 144 inches for Braidwood Station. In addition to the Boraflex, Boral sheets were added to the flux traps of the Region 1 racks.

c. Poison Sheathing The poison sheathing (i.e., cover plate) serves to position and retain the poison material in its designated space. This is accomplished by spot welding the cover plate to the square tube along the cover plate's edges at numerous (i.e., at least 20) locations. This manner of attachment ensures that the poison material will not sag or laterally displace during fabrication processes and under any subsequent loading conditions.

d. Gap Element Gap elements position two inner boxes at a predetermined distance to maintain the minimum flux trap gap required between two boxes. The gap element is welded to the inner box by fillet welds. An array of composite box assemblies forms the honeycomb gridwork of cells which harnesses the structural strength of all sheet and plate type members in an efficient manner. The array of composite boxes has overall bending, torsional, and axial rigidities, which are an order of magnitude greater than configurations utilizing a grid bar type of construction.

A-8 1

L

e. Gap Element-The baseplate is a 5/8-inch thick plate type member which has six-inch diameter holes concentrically located with respect to the internal square tube, except at support leg locations, where the hole size is five inches in diameter. These holes provide the primary path for coolant flow. Secondary flow paths are available between adjacent l cells via the lateral flow holes (one inch in diameter) near the root of the honeycomb '

which preclude flow blockages. The honeycomb is welded to the baseplate with 3/32-inch fillet welds.

f. Support Assembly Each module has at least four support legs. Most supports are adjustable in length to enable leveling of the rack. The variable height support assembly consists of a flat-footed spindle, which rides into an internally threaded cylindrical member. The cylindrical member is attached to the underside of the baseplate through fillet and partial penetration welds. The base of the flat-footed spindle sits on the pool floor on shim plates that distribute floor loading. Leveling of the rack modules is accomplished by turning the square sprocket in the spindle using a long arm (i.e., approximately 46 feet long) square head wrench. The supports elevate the module baseplate approximately 7-1/2 inches above the pool floor, thus creating ine water plenum for coolant flow. The lateral holes in the cylindrical member provide the coolant entry path leading into the bottom of the storage locations. Short adjustable legs are used on racks E1, L2, and C2. In addition, one small fixed support leg is used on rack D4.

g. _ Top Lead-in Lead-ins are provided on each cell to facilitate fuel assembly insertion. The lead-ins of contiguous walls of adjacent cells are structurally connected at the lead-in. These lead-in joints aid in reducing the lateral deflection of the inner square tube due to the impact l of fuel assemblies during the ground motion (i.e., postulated seismic motion). This type !

of construction leads to natural venting locations for the inter-cell space where the neutron absorber material is located.

Region 2 Racks I

The rack modules in Region 2 are fabricated from the same material as that used for Region 1 modules with the exception that the Region 2 racks have no Boral flux traps. A typical Region 2 module storage cell also has an 8.85-inch nominal square cross-sectional opening. The rack construction varies from that for Region 1 as the stainless steel cover plates, gap elements, and top lead-ins are eliminated. Hence, the basic components of this design are as follows:

a. . Internal square box,

b. Neutron absorber material, l

c. Side strips,

d. Baseplate, and I 4

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e. Support assemb;y. I in this construction, two channel elements form a box with an 8.85-inch nominal square cross-sectional opening. The poison material (Boraflex,144 inches in length) is placed between two boxes. Stainless steel side strips are inserted in both sides of the poison i materia! to firmly locate it in the lateral direction. The bottom strip positions the poison materialin the vertical direction to envelop the entire active fuellength of a fuel assembly. '

Two adjacent boxes and the side strip between boxes are welded together to form the honeycomb rack module. The baseplate and adjustable support assemblies are i incorporated in exactly the same manner as described for Region 1. j Failed Fuel Storage Cells Fuel assemblies that are damaged or failed such that they may release radioactivity can be stored in special containers in the failed fuel rack. A special storage rack (i.e., rack J1) provides for 35 Region 2 storage cells and six failed fuel storage cells. The failed fuei ,

storage cell consists of 1/4 inch thick stainless steel square tubes spaced at 21 inches center-to-center and is welded to an elevated stainless steel baseplate, which also supports the 35 Region 2 storage cells. The base contains an opening at each storage location to allow natural circulation of pool water up through this opening to remove decay heat from the stored assembly.

Fuel Pool Coolino System Bases The spent fuel pool cooling system consists of two complete cooling trains designed to l remove decay heat from the spent fuel pool. The spent fuel pool cooling system (i.e., l piping, pumps, valves, and heat exchangers) is Safety Category 1, Quality Group C. The three-inch piping from the refueling water storage tanks to the refueling water purification ,

pump, the pump itself, and its associated piping and valves, are Safety Category 1, Quality i Group C. A two-inch Safety Category 1, Quality Group C line from the discharge of the i refueling water purification pump to the spent fuel pool is permanently installed. This is the '

Category I water makeup circuit. ihe backup Safety Category I makeup system consists of -

piping and hoses from the Safe ( .lategory I fire protection system. The primary water makeup system non-Category I takes water from both primary water storage tanks and )

routes the water through the spent fuel pool water filter and then to the return header. In !

summary, there are three sources of makeup water available: a primary unborated non- !

Category I source, a borated Safety Category I source, and an unborated fire protection I Safety Category I water system.

)

i Each cooling train incorporates one heat exchanger and pump, one purification loop with demineralizer and filter, associated piping, valving, and instrumentation. One surface skimmer loop is also provided. Each cooling train is designed to service the spent fuel pool, with design spent fuel assembly loading, and to maintain the bulk fluid temperature of the ,

pool less than 140 F for a normal one-third core discharge.

When either cooling train is in operation, water flows from the spent fuel pool to the spent fuel pool pump suction, is pumped through the tubeside of the heat exchanger, and is 1 returned to the pool. The suction line, which is protected by a strainer, is located at an ,

elevation 4 feet below the normal spent fuel pool water level, while the return line contains '

an antisiphon hole near the surface of the water to prevent gravity drainage of the pool.

1 A - 10 1

While the heat removal operation is in process, a portion of the spent fuel pool water, approximately 80 gpm, may be diverted through a demineralizer and a filter to maintain spent fuel pool water clarity and purity. Transfer canal water may also be circulated through the same demineralizer and filter by having the gate between the canal and the spent fuel pool open. This purification loop is sufficient for removing fission products and other contaminants that may be introduced if a leaking fuel assembly is transferred to the spent fuel pool.

The demineralizer and filter of either cooling train may be isolated from the heat removal portion of the spent fuel pool cooling system. By so doing, the isolated equipment may be used in conjunction with the refueling water purification pump to clean and purify the refueling water while spent fuel pool heat removal operations proceed. Connections are provided such that the refueling water may be pumped from either the refueling water storage tank or the refueling cavity of either unit, through a filter and demineralizer, and discharged to the refueling cavity or refueling water storage tank of either unit.

To assist further in maintaining spent fuel pool water clarity, the water surface is cleaned by a skimmer loop. Water is removed from the surface by the skimmer, pumped through a filter, and returned to the pool surface at two locations remote frcm the skimmers at a rate of approximately 50 gpm.

The spent fuel pool is initially filled for use with water that is at the same boron concentration as that in the refueling water storage tank. Borated water may be supplied from the refueling water storage tank via the refueling water purification pump connection, l or by a line from the boric acid blender, located in the chemical and volume control system.

Demineralized water can be added for makeup purposes (e.g., to replace evaporative j losses) through a connection in each cooling train's purification return loop. !

l i

D, NEED FOR REVISION OF THE REQUIREMENT This request is being submitted to allow replacement of the existing spent fuel storage racks. This replacement is necessary due to the continuing degradation of the rack neutron absorbing material, Boraflex, and a lack of viable economic alternatives for storing spent fuel. The silica releases associated with Boraflex degradation results in reduced spent fuel pool water clarity which requires additional spent fuel pool water filtering and poses fuel handling problems due to potentially reduced visibility during fuel movement. The reduction in the amount of effective Boraflex also adversely affects reactivity control. Boraflex degradation has resulted in a requirement to maintain the spent fuel pool boron concentration at ;t2000 ppm to ensure the spent fuel pool storage racks are maintained with kg 0.95. The existing spent fuel pool storage racks were originally designed to be maintained with ks 0.95 when flooded with unborated water.

Boraflex degradation is an industry wide, irreversible problem that is expected to continue ov'er the life of the spent fuel racks and to worsen as time progresses. Replacement of the existing spent fuel pool storage racks with new racks containing the ' stron absorbing material "Boral*, which has industry wide acceptable performar ce ci aracteristics in the aqueous spent fuel pool environment, is considered to be a prudent, and conservative option.

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E. DESCRIPTION OF THE PROPOSED CHANGES The proposed changes to TS 3.7.15, " Spent Fuel Pool Boron Concentration"; 3.7.16 " Spent Fuel Assembly Storage"; 4.3.1, " Criticality"; and 4.3.3, " Capacity"; support installation of new Boral high-density spent fuel pool storage racks at the Byron and Braidwood Stations.

These proposed changes reflect our plan to remove all 23 of the existing spent fuel storage racks at each station and replace them with 24 new spent fuel storage racks. The existing

- racks utilize Boraflex as the neutron absorber material. Degradation of Boraflex has caused water chemistry and clarity problems and has also resulted in the need to rely on soluble boron in the spent fuel pool oue to reactivity concems. The new spent fuel storage racks utilize Boral as the neutron absorber material.

Boralis a boron carbide aluminum. cermet and has been used successfully at numerous plants in the United States, Korea, Mexico, Brazil, and the United Kingdom. Installation of these new racks will also increase the spent fuel pool storage capacity at Byron and Braidwood Stations from 2670 assemblies to 2984 assemblies.

During the installation of the new Holtec spent fuel pool storage racks, both Holtec and the existing Joseph Oat spent fuel pool storage racks wil! be in the spent fuel pool at the same time. At the completion of installation, only Holtec spent fuel pool storage racks will be in the spent fuel pool. The proposed changes to the TS address the requirements for both the new Holtec racks and the existing Joseph Oat racks. When shuffling fuel during the rack change-out, the fuel assembly burnup, enr;chment, and decay curves requirements, applicable to the Joseph Oat racks, as well as the new bumup and enrichment curve requirements, applicable tc Me Holtec racks, will be met. The requirement of 2 2000 ppm boron in the spent fuel pool will be maintained during the entire rack change out process, thereby ensuring that k,will remain s 0.95. At the completion of installation, only Holtec spent fuel pool storage racks will remain in the spent fuel pool.

The specific proposed TS changes are outlined below.

LCO 3.7.15 currently states: "The spent fuel pool boron concentration shall be 2 2000 ppm." The proposed change adds a requirement for the Holtec racks: "The spent fuel pool boron concentration shall be, as applicable: a. 2 300 ppm for Holtec spent fuel pool storage racks; and b. 2 2000 ppm for Joseph Oat spent fuel pool storcge racks." A higher boron concentration is no longer needed for the Holtec racks due to their improved neutron absorbing characteristics. A boron concentration of 2 300 ppm remains conservative as the enticality analysis calculations of Section 4.7.3 of Attachment E show that under a worst case scenario for a misloaded assembly, a soluble boron concentration of 220 ppm is adequate to assure that the maximum neutron effective multiplication factor (k,) does not exceed 0.945. . The clarification, for Joseph Oat spent fuel pool storage racks, added to the existing 2000 ppm limit, is necessary to ensure the analysis assumptions for the Joseph Oat ,

racks are maintained while the instsilation of the Holtec racks is performed.

SR 3.7.15.1 requires verification that the spent fuel pool boron concentration is within the required limit everv "48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months ." The proposed change revises the frequency for this surveillance to "7 days."

LCO 3.7.16 states: "Each spent fuel assembly stored in the spent fuel pool shall." This statement is changed to: "Each spent fuel assembly stored in the spent fuel pool shall, as applicable." The words "as applicable" are added to address the applicability of LCOs 3.7.16.a A - 12

m i i

l through 3.7.16.e to either the Joseph Oat or Holtec spent fuel pool storage racks, as appropriate.

- LCO 3.7.16.a. places limitations on fuel stored in Region 1 racks. The proposed change clarifies that this requirement is for the Joseph Oat spent fuel poc! storage racks. " Region 1" is changed to " Region 1 of Joseph Oat spent fuel pool storage raAs."

LCO 3.7.16.b. places limitations on fuel stored in Region 2 racks. The proposed change clarifies that this requirement is for the Joseph Oat spent fuel pool storage racks. " Region 2" is changed to " Region 2 of Joseph Oat spent fuel pool storage racks."

LCO 3.7.16.c. piaces res'trictions on assembly placement within and between adjacent recks.

The proposed change clarifies tt.at this requirement is for the Joseph Oat spent fuel pool storage racks: " Interface Requirements" is changed Winterface Requirements for Joseph Oat spent fuel pool storage racks."

LCO 3.7.16.d. is added to address requirements for the Holtec spent fuel pool storage racks. l This requirement states, " Region 1 of Holtec spent fuel pool storage racks have an initial nominal enrichment of s 5.0 weight percent U-235 to permit storage in Rny cell location." This new less restrictive requirement is supported by the criticality analysis presented in Section 4.2.1.1 of Attachment E which states: " Calculations have been performed to qualify the

. Region 1 racks for storage of fresh unburned fuel assemblies with a maximum nominal ;

enrichment of 5.0 weight percent U-235. The criticality analyses for Region 1 of the spent j fuel storage pool are summarized in Table 4.2.1, and demonstrate that for the defined !

acceptance criteria, the maximum k,is less than 0.945."

LCO 3.7.16.e. is added to address requirements for the Holtec spent fuel pool storage racks. j This requirement states, " Region 2 of Holtec spent fuel pool storage racks have a combination I of initial enrichment and bumup within the Acceptable Bumup Domain of Figure 3.7.16-4."

Figure 3.7.16-4 has been added to identify the required assembiv enrichment and burnup for the Holtec racks. The new requirement is supported by the criticality analysis presented in j Section 4.2.1.2 of Attachment E which states: " Calculations have been performed to qualify 1 the Region 2 racks for storage of spent fuel assemblies with a maximum nominal initial enrichment of 5.0 weight percent U-235 which have accumulated a minimum burnup of 40.0

= GWD/MTU or fuel of initial enrichment and bumup combinations within the acceptable domain I depicted in Figure 4.1.1. The criticality analyses for Region 2 of the spent fuel storage pool l

are summarized in Table 4.2.2, and demonstrate that for the defined acceptance criteria,

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the maximum k,is less than 0.940."

SR 3.7.16.1 states to, " Verify by administrative means the initial nominal enrichment of the fuel assembly is s 4.7 weight percent U-235 or a minimum number of IFBAs is met." This surveillance requirement has been divided into two different surveillance items, item a and item b. A note has been added to SR 3.7.16.1 that states: " Item a is only applicable for storage of fuel assemblies in Region 1 Holtec spent fuel pool storage racks. Item b is only applicable for storage of fuel assemblies in Region 1 Joseph Oat spent fuel pool storage racks." The existing SR, renumbered to item b, is revised for clarification and requires verification by administr:Kive means that the initial nominal enrichment of the fuel assembly is s 4.7 weight percent U-235 with less than the minimum number of IFBAs or j

.~s 5.0 weight percent U-235 with the minimum number of IFBAs. The new SR item l (i.e., item a) applicable to the Holtec racks requires verification by administrative means that j the initi9l nominal enrichment of the fuel assembly is s 5.0 weight percent U-235. Region 1 A - 13 '

of tne new racks are designed to accept 5.0 weight percent U-235 fuel assemblies without the use of IFBAs as supported by the criticality analysis presented in Section 4.2.1.1 of Attachment E.

SR 3.'7.16.2 states to, " Verify by administrative means the combination cf initial enrichment,'

bumup,' and decay time of the fuel assembly is within the Acceptable Bumup Domain of Figure 3.7.16-1, 3.7.16-2, cr 3.7.16-3." Figure 3.7.16-4 has been added which is applicable to fuel stored in the Holtec spent fuel pool storage racks. A note has been added to this SR which states: " Figures 3.7.16-1,3.7.16-2, and 3.7.16-3 are only applicaole for storage of fuel assembEes in Region 2 Joseph Oat spent fuel pool storage racks," and " Figure 3.7.16-4 is only applicable for storage of fuel assemblies in Region 2 Holtec spent fuel pool storage racks." ~ The surveillance requirement, SR 3.7.16.2 is subsequently changec' to read, " Verify by administrative means the combination of initial ennchment, bumup arH decay, as applicable, of the fuel assembly is within the Acceptable Bumup Domain of Figure 3.7.16-1, 3.7.16-2,' 3.7.16-?. or 3.7.16-4." Region 2 of the new racks is only subject to the enrichment and bumup limitations presented in Figure 3.7.16-4. This new requirement is supported by the criicality analysis presented in Section 4.2.1.2 of Attachment E.

SR 3.7.16.3 states to " Verify by administrative means the interface requirements within and

- between adjacent racks are met." A Note has been added to this SR wqich states that this requirement is "Only applicable for storage of fuel assemblies in Joseph Oat fuel pool storage racks."

TS Figures 3.7.16-1,3.7.16-2, and 3.7.16-3 have been revised to indicate that they are only applicable to Joseph Oat spent fuel pool storage racks. The Figure titles now include,

"(Joseph Oat Spent Fuel Pool Storage Racks)."

TS Figure 3.7.16-4 has been added which is only applicable to Holtec spent fuel pool storage l racks. Figure 3.7.16-4 provides the Region 2 fuel assembly limitations based on initial U-235 I enrichment and burnup. The Figure is entitled," Region 2 Fuel Assembly Burnup Requirements (Holtec Spent Fuel Pool Storage Racks)."

TS Section 4.3.1 states: "The spent fuel storage racks are designed and shall be maintained with." This statement 's revised to read: "The spent fuel storage racks are designed and shall be maintained, as applicable, with." This change is made for clarification purposes as Design Features Specification 4.3.1.b through e are ~only applicable to Joseph Oat spent fuel pool

, storage racks, and Design Features Specification 4.3.1.f through h are only applicable to

. Holtec spent fuel pool storage racks.

TS Section 4.3.1.b. states that the spent fuel storage racks are designed and shall be maintained with: "k,< 1.0 if fully flooded with unborated water which includes an allowance for uncertainties as described in WCAP-14416-NP-A, ' Westinghouse Spent Fuel Rack Criticality Analysis Methodology.'" This requirement is only applicable to Joseph Oat spent fuel pool storage racks, therefore the statement, "For Joseph Oat spent fuel pool storage racks," has been added to this design feature.

TS Section 4.3.1.c. states that the spent fuel storage racks are designed and shall be ,

maintained with: "ks; 0.95 if fully flooded with water borated to 550 ppm, which includes an !

allowance for uncertainties as described in WCAP-14416-NP-A, ' Westinghouse Spent Fuel l Rack Criticality Analysis Methodology.'" This requirement is only applicable to Joseph Oat

{

spent fuel pool storage racks, therefore the statement, "For Joseph Oat spent fuel pool !

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r storage racks," has been added to this design feature.

TS Section 4.3.1.d. states that the spent fuel storage racks are designed and shall be maintained with: "A nominal 10.32 inch north-south and 10 42 inch east-west center to center distance between fuel assemblies placed in Region 1 rack This requirement is only applicable to Joseph Oat spent fuel pool storage racks, therefore the statement, "For Joseph Oat spent fuel pool storage racks," has been added to this design feature.

TS Section 4.3.1.e states that the spent fuel storage racks are designed and shall be maintained with: "A nominal 9.03 inch center to center distance between fuel assemblies placed in Region 2 racks." This requirement is only applicable to Joseph Oat spent fuel pool storage racks, therefore the statement, "For Jvseph Oat spent fuel pool storage racks," has been added to this design feature.

TS Section 4.3.1.f has been added to address Holtec racks and states that the spent fuel storage racks are designed and shall be maintained with: "For the Holtec spent fuel pool storage racks, kg 0.95 if fully flooded with unborated water which includes an allowance for uncertainties as described in Holtec International Report, HI-982094, " Criticality Analysis for the Byron /Braidwood Rack Installation Project," Project No. 80944,1998."

TS Section 4.3.1.g has been added to address Holtec racks and states that the spent fuel storage racks are designed and shall be maintained with: "For Holtec spent fuel pool storage racks, a nominal 10.574 inen north-south and 10.888 inch east-west center to center distance between fuel assemblies placed in Region 1 racks."

TS Section 4.3.1.h has been added to address Holtec racks and states that the spent fue storage racks are designed and shall be maintained with: "For Holtec spent fuel pool storage racks, a nominal 8.97 inch center to center distance between fuel assemblies placed in Region 2 racks."

TS Section 4.3.3 states: "The spent fuel pool is designed and shal1 be maintained with a storage capacity limited to no more than 2870 fuel assemblies." The number of fuel assemblies is being changec' to 2984 to reflect the number of storage locations in the new racks once installation is complete. The new limit of 2984 cells bounds the previous limit. The current capacity of 2870 cells consist; of 2864 normal fuel cells and six failed fuel cells. The failed fuel cells are special containers found in Rack J1. The failed fuel cells have not been used, nor plan to be used, therefore, failed fuel cells were not included in the new racks.

The removal of the existing Byron and Braidwood Stations spent fuel racks will eliminate the current reliance on soluble boron credit for reactivity control, necessitated by the degradation of Boraflex. Removal of the existing racks from the spent fuel pools has the added benefit of improving water chemistry and clarity which had been adversely affected by the Boraflex degradation.

When installation of the new racks is complete, the Byron and Braidwood Stations spent fuel pools will each contain 24 racks with a total of 2984 fuel stoiage. cells. The storage cells will be divided into two regions based upon rack type. A group of four racks will store the most reactive fuel (i.e., up to 5-weight percent U-235 enrichment) without any burnup limitations. These four racks, which use a flux trap design, will be referred to as Re0 i on 1.

The remaining 20 racks, which do not have flux traps, will have fuel storage limitations based on a function of enrichment versus burnup. These racks will be collectively referred A - 15

to as Region 2.

All storage rack arrays consist of freestanding modules made from Type 304L austenitic stainless steel containing prismatic storage cells interconnected through longitu:'nal welds.

A panel of Boral cermet containing a high arealloading of the "B isotope provides the appropriate neutron attenuation between adjacent storage cells.

The baseplates on all rack modules extend out beyond the rack module wall such that the contiguous edges of the plates establish a geometric separation between the facing cells in the modules.

Each new rack mbdule is supported by legs, which are remotely adjustabie. The racks can i be leveled, and the top of the racks can easily be made co-planar with each other. The rack module support legs are engineered to accommodate undulations in the surface of the pool

{'

floor. Given the maximum support leg upward adjustment, the top of the active fuel will remain greater than 10 feet below the surface of the spent fuel pool water level in the event ,

of an inadvertent draining of the spent fuel pool, as required by the Standard Review Plan.

A bearing pad interposed between the rack pedestals and the pool liner serves to diffuse l the dead weight of the loaded racks into the reinforced concrete structure of the pool slab. I Holtec International is performing the design, fabrication, and safety analysis of the new l high density spent fuel pool storage racks. The rack design and analysis methods l employed in the Byron and Braidwood Stations rerack projects are a direct evolution of previous rerack license applications. The new Holtec racks meet all governing j requirements of the applicable codes and standards, in particular, "OT Position for Review i and Acceptance of Spent Fuel Storage and Handling Applications," issued in a letter from B.

K. Grimes (NRC) to A!! Power Reactor Licensees, dated April 14,1978, and the associated addendum, issued in a letter from B. K. Grimes (NRC) to All Power Reactor Licensees, l dated January 18,1979. The analysis, material procurement, and fabrication of the rack modules conform to 10 CFR 50, Appendix B," Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants," requirements.

The design of the Byron and Braidwood racks is similar to Holtec racks approved by the NRC and presently in service at many other nuciear plants in the United States and overseas. Among these plants are Salem, Watts Bar, and Vogtle. )

l' The NRC has recently approved the installation of new spent fuel pool racks designed and analyzed by Holtec International. Amendment No. 6 for Tennessee Valley Authority's (WA)

Watts Bar Nuclear Plant was issued in a letter from R. E. Martin (NRC) to O. D. Kingsley (TVA), dated July 28,1997. More recently, Amendment No.144 for Entergy's Waterford Steam Electric Station was issued in a letter from C. P. Patel (NRC) to C. M. Dugger (Entergy Operations), dated July 10,1998. In addition, other plants currently requesting ;

license amendments to replace old spent fuel pool racks with new Holtec racks include j Davis Besse, Callaway, Wolf Creek, Vermont Yankee, Hatch, Shearon Harris, and Nine Mile Point.

l l

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F i

F. SAFETY ANALYSIS OF THE PROPOSED CHANGES The replacement of the existing 23 spent fuel storage racks containing Boraflex in each of the spent fuel pools at the Byron and Braidwood Nuclear Stations with 24 new racks containing Boral has been thoroughly analyzed and documented in the Holtec International report, " Licensing Report for Spent Fuel Rack Installation at Byron and Braidwood Stations,"

Hl-982083. This report, which is company proprietary, is contained in Attachment E to this submittal. A non-proprietary version of this report is contained in Attachment F.

The report addresses the areas of:

1. Introduction - Project Overview;

2. High Density Spent Fuel Rack Design;

3. Material and Heavy Load Considerations;

4. Criticality Scfety Evaluation;

5. Thermal-Hydraulic Considerations;

6. Structural Seismic Considerations;

7. Fuel Handling and Construction Accidents;

8. Fuel Pool Structural Integrity Considerations;

9. Radiological Evaluation;

10. Boral Surveillance Program;

11. Rack Installation; and

12. Environmental Cost-Benefit Assessment.

During the installation of the new Holtec spent fuel pool storage racks, both holtec and the existing Joseph Oat spent fuel pool storage racks will be in the spent fuel pool at the same time. The criticality analysis for the Joseph Oat spent fuel pool storage racks states that should a spent fuel pool water temperature change accident or a fuel assembly mistoad accident occur in the Region 1, Region 2, or failed fuel storage cells, ko nwill be maintained s 0.95 due to the presence of at least 550 ppm (i.e., no fuel handling) or 1650 ppm (i.e.,

during fuel handling) of soluble boron in the spent fuel pool water. These assLmptions are more conservative than the requirements stated in the criticality analysis for the Holtec spent fuel pool storage racks which only requires 220 ppm boron to maintain k ns0.95 during the worst case fuel assembly misload accident. The new Holtec racks have a superior neutron attenuation capability due to their improved design. The requirement of 2 2000 ppm boron in the spent fuel pool will be maintained during the entire rack change out process, thereby ensuring that k onwill remain s 0.95. At the completion of installation, only Holtec spent fuel pool storage racks will remain in the spent fuel pool.

SR 3.7.15.1 requires verification that the spent fuel pool boron concentration is within the required limit every "48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months ." The proposed change revises the frequency for this surveillance to "7 days." The 7 day frequency is appropriate based on operating experience and takes into consideration that no major replenishment of pool water is expected to occur over such a short period of time. A review of the past four years of surveillance history has sh'own that the spent fuel pool boron concentration has always been within required limits.

The spent fuel poolis kept at 2 2000 ppm during normal operations. Therefore, significant margin exists to the limiting boron concentration values assumed in the analyses (i.e.,220 ppm for Holtec racks and 1650 ppm for Joseph Oat racks). There is no known phenomenon, such as boron plate out or boron stratification, which has been observed in the spent fuel pool that would significantly reduce the spent fuel pool boron concentration over a short time period such as 7 days. Therefore, the proposed 7 day frequency provides A - 17

adequate assurance that boron concentration will be maintained consistent with the analysis assumptionsc When installation of the new racks is complete, the Byron and Braidwood Stations spent fuel pools will each contain 24 racks with a total of 2984 fuel storage cells. The storage cells will be divided into two regions based upon rack type. A group of four racks will store the most reactive fuel (i.e., up to 5-weight percent U-235 enrichment) without any burnup limitations. These four racks, which use a flux trap design, will be referred to as Region 1.

The remaining 20 racks, which do not have flux traps, will have fuel storage limitations bassd on a function of enrichment versus burnup. These racks will be collectively referred to as Region 2.

The new racks do not require interface restrictions because the minimum water gap between racks, which is 0.875 inches between neighboring Region 2 style racks and 1.75 inches between Region 1 and Region 2 style racks, constitutes an effective neutron flux-trap for the storage cells of neighboring Region 1 to Region 2 racks. Further, the racks are constructed with the base plates extending beyond the edge of the cells which assures that the minimum spacing between storage racks is maintained under all credible conditions.

Region 2 racks do not contain intemal flux traps, and thus, this water gap flux-trap between '

the Region 1 and Region 2 racks will act to maintain reactivity below the required maximum value.

The new Holtec racks, which are free-standing and self-supporting, are designed to the stress limits of, and analyzed in accordance with, Section Ill, Division 1, Subsection NF of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV)

Code,1995 Edition. The principal construction materials for the new racks are SA240-Type 304L stainless steel sheet and plate stock, and SA564-630 (i.e., precipitation hardened stainless steel) for the adjustable support spindles. The only non-stainless material used for construction is the Boral neutron absorber material.

A criticality safety evaluation for the storage of PWR spent nuclear fuel in Holtec Region 1 and 2 high-density spent fuel storage racks at the Byron and Braidwood Stations was performed. The objective of this analysis is to ensure that the effective neutron multiplication factor (k.n) is s 0.95 with the storage racks fully loaded with fuel of the highest anticipated reactivity and the pool flooded with ur:oorated water at a temperature corresponding to the highest reactivity. The maximum calculated reactivity includes a margin for uncertainty in reactivity calculations including manufacturing tolerances and is shown to be s 0.95 with a 95% probability at a 95% confidence level. Reactivity effects of abnormal and accident conditions have also been evaluated to assure that under all credible abnormal and accident conditions, the reactivity will not exceed the regulatory limit of 0.95 kon.

The NRC OT Position Paper, "OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications," dated April 14,1978, Regulatory Guide, RJ 1.13, Revision 1, " Spent Fuel Storage Facility Design Basis," and NRC Memorandum from L.

Kopp to T. Collins, " Guidance on the Regulatory Requirements for Criticality Analysis of I Fuel Storage at Light-Water Reactor Power Plants," dated August 19,1998, specify that the maximum effective multiplication factor, k.,,, including bias, uncertainties, and calculational statistics, shall be s 0.95, with 95% probability at the 95% confidence level. ;

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I The decay heat load calculation is conservatively performed in accordance with the provisions of U.S. NRC Branch Technical Position ASB9-2, " Residual Decay Energy for Light Water Reactors for Long Term Cooling", Rev. 2, July,1981. The new racks increase the number of fuel cells in the spent fuel pool by 114, an increase of approximately 4%.

This results in a smallincrease to the decay heat load of the spent fuel pool which can be removed with the existing spent fuel pool cooling systems without any modifications. The existing spent fuel pool cooling system has been shown to be capable of removing the decay heat generated by the additional spent fuel assemblies utilizing the standard Byron and Braidwood Station operating procedures. It has been shown that the spent fuel pool cooling system will maintain the spent fuel pool water temperature within the existing design basis, as detailed in the Byron and Braidwood Updated Final Safety Analysis Report (UFSAR).

To determine the end-of-life decay heat in the Byron and Braidwood Stations spent fuel pools, the historical and projected discharges are cor sidered. A total of 2864 and 2821 fuel assemblies will be accumulated from previous dischargos in the Byron and Braidwood Stations spent fuel pools, respectively. A fresh full core discharge of 193 assemblies will increase the fuel inventory to 3057 and 3014 fuel assemblies which conservatively bounds the maximum fuel storage capacity for decay heat evaluation purposes.

The bulk pool temperature and decay heat load profile in the spent fuel pool as a function of time has been calculated for the Byron and Braidwood Stations refueling discharge scenarios. The Byron and Braidwood Stations' spent fuel pool cooling systems were evaluated. Calculations performed demonstrate that the resulting maximum bulk spent fuel pool temperatures do not exceed the Byron and Braidwood Stations design basis spent fuel pool maximum bulk temperature acceptance limit of 140 F, for the normal one-third core discharge scenario, which is consistent with the Standard Review Plan. Spent fuel pool temperature is routinely monitored and recorded on operator logs every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . In addition, an alarm in the control room will sound if temperature goes outside an acceptable band.

An evaluation of " time-to-boir if all heat rejection paths from the pool were lost was also performed. Although the probability of a loss of cooling event coinciding with the instant when the pool water has reached its peak value is extremely remote, the calculations performed under this scenario show comparable time-to-boil times with other Pressurized Water Reactor (PWR) spent fuel pools.

A determination of maximum local water temperature in the spent fuel pool racks at the instant when the bulk pool water temperature reaches its maximum value was also made.

Using the maximum decay heat load case scenario of a full core discharge with a 100 hour0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months decay time after shutdown, following a normal discharge occurring 17 days earlier, the local water temperature was shown to be subcooled. Under the same maximum decay heat load scenario, the maximum fuel cladding temperature was also evaluated for the hottest cell location in the pool. The results show that Departure from Nucleate Boiling (DNB), a condition leading to failure of cladding material due to extreme thermal stress, is not credible and cladding integrity is ensured under all discharge scenarios. The core discharge scenario assumption of 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months decay time after reactor shutdown is not a limiting parameter for the thermal-hydraulic analysis of the spent fuel pool.

The new spent fuel storage racks, like the existing racks, are freestanding self-supporting structures. The analysis of the racks must include a large number of parameters to A - 19 ^

e

accurately predict the behavior of a submerged structure under dynamic loading. The analysis includes the effects of impacts and friction, fluid coupling, fuel rattling, lift off, fuel loading scenarios and the racks elastic behavior. Both single rack scenarios, (i.e., where the motion of a rack is calculated without consideration to other racks), and whole pool multi-rack (WPMR) simulations, (i.e., where the motion of all modules is simultaneously calculated), were performed.

Thirty discrete freestanding dynamic simulations of maximum density spent fuel storage racks have been performed to establish the structural margins of safety. Of the thirty parametric analyses, six simulations were performed modeling all 24 fuel racks in the pool in one comprehensive Whole Pool Multi Rack (WPMR) model. The remaining twenty-six runs were carried out with the classical single rack 3-D mode l, The parameters which were varied in the different runs consisted of the rack / pool liner interface coefficient of friction, extent of storage locations occupied by spent nuclear fuel (i.e., ranging from nearly empty to full) and the type of seismic i . 'ut, either Safe Shutdown Earthquake (SSE) or Operating Basis Earthquake (OBE). Mc num (i.e., maximum in time and space) values of pedestal vertica! and shear forces, displacements and stress factors (i.e., the normalized stresses for NF class 3 linear type structures) have been post-processed from the array of runs. The results show that. l (i) All stresses are well below their corresponding "NF" limits, (ii) There is no rack-to-rack or rack-to-wall impact anywhere in the cellulac region of the rack modules; and (iii) The factor of safety against overturning of a rack is in excess of 60.

)

l An evaluation of the fatigue expenditure in the most stressed location in the most heavily i loaded rack module, under combined effect of one SSE and twenty OBE events, shows that the Cumulative Damage Factor, using the classical Miner's rule, is conservatively r alculated io oe 0.95, which satisfies the permissible value of 1.0. Prior to installation of the racks, the vendor recommended actions to be taken in the aftermath of an OBE or SSE, wil be included into an appropriate station procedure.

The Byron and Braidwood Stations spent fuel pool structures are identical, with the exception of the sub-surface foundation material. The Byron Station spent fuel pool base mat, consisting of a reinforced concrete slab, is founded directly on rock outcrop, while Braidwood Station has an intervening layer of soil. The wall and base mat thickness is i similar to other PWR spent fuel pool structures, ranging in thickness from 5 ft. to 6 ft. The l'

concrete compressive strength is 3500 psi. The new loads, calculated by the Holtec WPMR simulations, were used to evaluate the effect on the existing calculations performed by the plant designer, Sargent and Lundy. The methodology for comparing the new loads is consistent with that used by Sargent and Lundy. The results show that all safety factors remain valid. The spent fuel pool liner was evaluated with maximum stresses shown to be less than the u:timate strength of the liner material under SSE conditions.

The fuel drop accidents evaluated for the design change consist of a " shallow drop," where a fuel assembly falls onto the top of a storage rack, and a deep drop, where a fuel assembly l

falls into the bottom of a rack storage cell. The damage due to a perfectly vertical drop, on the top of a . ack, bounds an inclined fuel assembly drop because the impact energy is focused on a single cell wall which results in maximum cell blockage. Both fuel assembly drop accident events were analyzed and found to produce localized damage well within the design limits for the racks. This is allowed by the NRC position paper (Reference 5).

A - 20

1 The Boral surveillance program for Byron and Braidwood consists of coupon placement in a

! designated rack cell surrounded by fuel with periodic removal and testing of the coupon for certain physical and chemical characteristics. The surveillance prograrn elements include procedural controls of schedule, methodology, measurement, acceptance criteria and reporting. Boral is considere>d a satisfactory material for reactivity control in a spent fuel pool environment for the lifetime of the racks. However, should the surveillance process detect indications of Boral degradation, blackness testing can be employed in-situ to I evaluate the Boral material in the installed racks.

Rack removal and installation will ne performed in accordance with Byron and Braidwood Station commitments to NUREG-0612 " Control of Heavy Loads at Nuclear Power Plants,"

July 1980, and ANSI N14.6-1978, " Radioactive Materials - Special Lifting Devices for Shipping Containers Weighing 10,000 Pounds (4500 kg) or More." Rack change-out will require adherence to TS requirements for assembly location based on burnup, enrichnent 4 and decay time when shuffling fuel within existing racks, as well as adherence to the new bumup and enrichment requirements when shuffling fuel within the new Holtec racks. The racks will be lifted by the 125 ton double girder bridge crane, keeping the height of the racks at a minimum wherever possible. A remotely engageable lift rig meeting NUREG-0612 stress criteria will be used to lift the new modules. The lift rig employs independently loaded lift rods with a cam lift configuration. This ensures that failure of one traction rod will not result in uncontn ,lled lowering of the load. Additional measures to ensure installation safety include preventative maintenance inspection of cranes and lifting devices, safe load paths for rack movement, prohibition of racks being lifted over irradiated fuel, rack upending carried out away from the spent fuel pool, int allation training for personnel, and load testing of the lift rig. A leak chase system that is present beneath the spent fuel pool liner is capable of the collection of and isolation of spent fuel poolleakage that could occur in the extremely unlikely event of a rack drop during rack removal or installation. In addition, a potential rack drop would present limited structural damage to the spent fuel pool reinforced concrete slab due to the slab being founded on rock and soil. Local concrete crushing and liner puncture are expected. Failure of the liner would not result in a significant loss of spent fuel pool water and no safety-related equipment would be affected by the leakage. Makeup water is available from three separate sources, Borated water is available from the two, Safety Category 1,450,000 gallon refueling water storage tanks. A reliable backup source of unborated water is available through the fire protection system cross-tie to the essential service water system. Primary backup water is available from two 500,000 gallon primary water storage tanks via a return line. The redundancy and backup water sources' capability is assured via administrative controls in plant procedures.

Accident Analysis Consecuences The previously evaluated Byron and Braidwood Stations accidents, relative to spent fuel storage, are discussed in the UFSAR Section 15.7.4, " Fuel Handling Accidents," and UFSAR Section 15.7.5, " Spent Fuel Cask Drop Accident." These accidents were considered for the new Holtec spent fuel pool racks and are listed below

1. Spent fuel assembly dropped onto the spent fuel pool floor.

2. Spent fuel assembly dropped between racks.

3. Spent fuel assembly dropped between a rack and the spent fuel pool wall.

4. Spent fuel assembly loaded contrary to placement restrictions.

5. Spent fuel assembly dropped onto a rack.

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y.

'6. Spent fuel cask drop.

7. Change in spent fuel pool water temperature.

Spent Fuel Assembly Dropped onto the Spent Fuel Pool Floor

~

The drop of a spent fuel assembly onto the spent fuel pool liner has been evaluated an.

shown to be bounded by the existing design basis as described in the Byron and Braidwood Station UFSAR.~ The maximum drop distance for a fuel assembly will not change as a result of this design change. The initiators of fuel handling accidents are not changed by the installation of new Holtec spent fuel pool storage racks or by the small inc ease (i.e.,

. approximately 4.0 %) in spent fuel storage capacity as the spent fuel handling procedures

' and equipment are unaffected by the change. Also, the number of spent fuel assemblies is not an input to the initial conditions of this accident evaluation.

Spent Fuel Assembly Dropped Between Racks The drop of a fuel assembly between rack modules was previously evaluated under the safety analysis that supports LCO 3.7.15, " Spent Fual Boron Concentration," and was ,

shown to have no effect on reactivity. - This is considered a bounding analysis and is '

applicable to this design change since the new Holtec rack layout still precludes a reactivity increase due to this fuel hand!!ng accident. The initiators of this fuel handling accident are unaffected due to the similarity between the new Holtec spent fuel pool rack layout and the

- existing Joseph Oat spent fuel pool rack layout.

Spent Fuel Assembly Dropped Between a Rack and the Spent Fuel Pool Wall The drop of a spent fuel assembly between a rack module and the spent fuel wall has been l

evaluated for the new Holtec spent fuel pool racks. ' The worst case scenario consists of a l fresh fuel assembly, of the highest allowed enrichment, accidentally placed in a cut out area between a rack and the new fuel elevator or tool bracket. The resulting k, for this event I remains within the design basis criticality limit of s 0.95 k,, assuming a minimum soluble boron concentration of 220 ppm in the spent fuel pool water. The initiators of this event are 1 unaffected due to the similarity between the new Holtec spent fuel pool rack layout and the j existing Joseph Oat spent fuel pool rack layout. This event is bounded by the analysis of misloading an assembly into a Region 2 rack, discussed below.

Spent Fuel Assembly Loaded Contrary to Placement Restrictions Ttie loading of a fuel assembly contrary to placement restrictions has been evaluated for the Holtec racks. A worst case scenario of placing a fuel assembly of the highest enrichment

. (i.e.,5.0 weight percent U-235) into a Region 2 rack ce!! was shown to remain within the design basis criticality limit of 0.95 kg. assuming a minimum soluble boron concentration of

-220 ppm in the spent fuel pool water. The current required soluble boron concentration in the spent fuel poolis 2000 ppm. The minimum soluble borcn concentration, proposed in conjunction with this design change, is 300 ppm for conservatism. The initiators of this event are unaffected by this design change since the existing pool already includes a two region layout, similar to the new Holtec racksi Further, the possibility of a misloaded fuel assembly is minimized by an independent verification of the Nuclear Component Transfer j List that prescribes the exact location of each fuel assembly. After an assembly is placed in a spent fuel pool storage cell, station personnel once again independently verify it.

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1 Spent Fuel Assembly Dropped onto a Rack

. The drop of a spent fuel assembly onto a spent fuel storage rack has been evaluated for the Holtec racks. The results of this evaluation meet all existing design basis requirements as described in the Byron and Braidwood Station UFSAR. Analyses of the spent fuel drop accidents onto the top of a spent fuel pool storage rack (i.e., shallow drop), and a deep drop into the bottom of a cell, ,esulting in impact at the bottom of the rank cell, were performed to demonstrate that the spent fuel rack retains its structural integrity und capability to safely store spent fuelin adjacent cells. The damage due to a perfectly vartical drop, on the top of a rack, bounds an inclined fuel assembly drop because the impact unergy is focused on a single cell wall, which results in maximum cell blockage. i ne radiological consqquences of the drop onto the spent fuel pool liner, shallow drop onto to the top of the rack, and deep drop into the bottom of a rack cell, are bounded by the existing UFSAR assumptions that 314 fuel rods rupture. The UFSAR design basis dose is shown to be much less than the'10 CFR 100 limits of 300 rem to the thyroid and 25 rem to the whole body. The initiation of these fuel handling accidents are unaffected by the installation of new spent fuel storage racks. The spent fuel handling procedures and equipment are unaffected by this change and therefore there is no increase in the probability of these fuel handling accidents.

Spent Fuel Cask Drop The cask drop was evaluated and shown to be unaffected by the replacement of the existing Joseph Oat spent iuel pool storage racks with Holtec racks. There are no changes to any of the systems, structures, components or equipment associated with the movement of a spent fuel cask. The cask is shown by the Byron and Braidwood Station UFSAR to be isolated from the spent fuel pool by the combination of guard walls, which are designed to withstand the impact of a cask drop, and both administrative and physical controls. These controls are designed to preclude the fuel handling building crane from traveling over the spent fuel pool. There are also trolley stops on the crane bridge which physically prevent the main hook of the crane from traveling into the spent fuel pool storage area when ;

handling a spent fuel cask. Spent fuel pool rack installation activities and cask handling will not be performed simultaneously, thus minimizing the possibility of improper movement of the cask. This practice is consistent with the Byron and Braidwood Station UFSAR assumptions relative to new fuel operations.

Change in Spent Fuel Pool Water Temperature The change in the temperature of the spent fuel pool water was evaluated for the potential increase in reactivity. The new Holtec rack analysis was performed assuming a spent fuel pool water temperature of 4*C (39 F), which is well below the lowest normal operating temperature of 50*F. Because the reactivity temperature coefficient in the spent fuel poolis negative, temperatures greater than 4 C will result in a decrease in reactivity. The initiators of this event are unaffected by the spent fuel pool rack replacement because there are no features of this design change affecting the spent fuel pool cooling system or that would prompt a spent fuel pool water temperature decrease.

Radiolooical Evaluation Two analyses of a postulated fuel handling accident are performed in the current accident analysis: (1) a realistic analysis, and (2) an analysis based on Regulatory Guide 1.25,

" Assumptions Used for Evaluating the Potential Radiological Consequences of a Fuel A - 23

m 3

l l

l Handling Accident in the Fuel Handling and Storage Facility of Boiling and Pressurized Water Reactors." The parameters used for each of these analyses are listed in UFSAR Table 15.7-7. The activity release to the atmosphere is given in UFSAR Table 15.7-8 for both the realistic and Regulatory Guide 1.25 analyses. ;

i The short-term (i.e., 0-2 hour) atmospheric dispersion factor at the site boundary and the I dose modeis, are used in the analysis. The thyroid and gamma doses from a postulated fuel handling accident at the site boundary and low population zone are given in UFSAR

- Tables 15.0-11 and 15.0-12 for the realistic and Regulatory Guide 1.25 analyses. These doses are much less than the 10 CFR 100 reference values of 300 rem to the thyroid and ;

25 rem to the whole body. '

increases in the capacity of the fuel storage pools at the Byron and Braidwood Stations are not accompanied by an associated increase in the radiological consequences of fuel-handling accidents. The potential offsite doses are not increased over the values given in the UFSAR. The reasons there are no changes are as follows.

1. The accident scenario presented involves the failure of all the rods in a single fuel assembly, which is not a function of the storage capacity of the fuel pools or the rack design.

2. The basic methodology used to determine the doses, as presented in Regulatory Guide 1.25, does not change. Note that the peak gap activity of I-131 is assumed to be 12%

per NUREG/CR-5009, " Assessment of the Use of Extended Burnup Fuelin Light Water Power Reactors."

3. The short lived fission products are generally independent of fuel assembly burn-up, so the radionuclides that affect the radiation doses do not change, and their release fractions from the fuel do not change.

4, The basic nuclear data that affects radiation dose (e.g., beta and gamma energies from the decay of the radionuclides) does not change.

5. The spent fuel pool water depths do not change, therefore, the effectiveness of iodine retention does not change.

6. The height of the active fuel in relation to the spent fuel pool floor is comparable to the storage height of the current spent fuel rack design.

' 7. The physical plant filter designs do not change, therefore, the capabilities to remove the iodines from the gas streams being released from the buildings do not change.

8. The atmospheric dispersion factors for the plants at their Exclusion Area Boundaries (EAB) and Low Population Zones (LPZ) do not change.

9. The dose conversion factors do not change.

10. The cask drop accident, discussed in UFSAR Section 15.7.4, will not occur because of crane interlocks which prevent the cask from traveling over the spent fuel pool.

l A - 24 l l

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I

11. The accident detection and mitigation equipment features and capabilities are not altered by this design change. In the unlikely event of a dropped fuel assembly and associated release of radioactive noble gas and gaseous iodine to the spent fuel pool water, i

a. the nuclear safety-related radiation detection monitors near the pool begin to detect l the gamma radiation as the gas reaches and emerges from the spent fuel pool surface, and b, the ventilation of the fuel handling building is routed through the emergency exhaust filter train which routs air throu,,,h the HEPA and charcoal filters.

The potential doses from fuel handling accidents at the Byron and Braidwood Nuclear Stations, as presented in the UFSAR, continue to meet the guidelines set forth in NUREG-0800, Section 15.7.4, " Radiological Consequences of Fuel Handling Accidents."

Perso,'..iel Occupational Radiation Exposure During normal operations, personnel working in the fuel storage area are exposed to radiation from the spent fuel pool. The dose rates experienced by personnel are not expected to increase with the increased storage capacity of the Holtec racks because the dose rate from the fuel in stcrage is negligible. The water above the stored fuel is sufficiently deep such that the dose rate from that fuel is orders of magnitude lower than the j dose rate contribution from the radionuclides in the pool water itself. Consequently, though J the dose rate from stored fuel may increase slightly because more spent fuel assemblies j may be stored in the pool, it will not increase to levels comparable to those caused by the radionuclides in the pool water. .

The radionuclide concentrations in the pool water are not expected to increase significantly.

Radionuclide concentration levels are determined principally from the mixing of primary system water with the spent fuel pool water and crud deposits from spent fuel assemblies moved into the storage pool during refueling operations. Although the overall capacities of the pools are being increased, the number of fuel assemblies moved during refueling operations is independent of storage capacity.

Operating experience has shown that there have been negligible concentrations of airborne radioactivity and no increases are expected as a result of the expanded storage capacities.

Area monitors for airborne activities are available in the immediate vicinities of the spent fuel pools.

In summary, no increases in radiation exposure to operating personnel are expected.

Consequently, neither the current health-physics programs nor the area monitoring systems needs to be modified.

Anticioated Exposure Durina Rerackino All of the operations involved in reracking will utilize detailed procedures that implement "As Low as Reasonably Achievable" (ALARA) principles. Similar operations have been performed at a number of facilities in the past, and there high confidence that re-racking can be safely and efficiently accomplished at the Byron and Braidwood Stations with minimum radiation exposure to personnel.

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l l

. Total occupational exposure for the re-racking operation is estimated to be between 6 and 12 person-rem, as indicated in Attachment E, Table 9.1. While individual task efforts and exposures may differ from those in Table 9.1, the total exposure is believed to be a reasonable estimate for planning purposes. Divers will be used only if necessary, however, the estimated dose includes a figure for potential diver exposure.

The existing radiation protection programs at the plants are adequate for re-racking operations. Where there is a potential for airbome activity, continuous air samplers will be in operation. Personnel will wear protective clothing and, if necessary, respiratory protective l equipment will be wom if determined to be ALARA. Radiation Work Permits will govern all radiological work activities and personnel dosimetry will be issued to each individual. As a minimum, this will include thermoluminescent dosimeters and electronic dosimeters.

Additional personnel monitoring equipment (e.g., extremity badges or direct coverage by Radiation Protection Technicians) may be utilized as required. Work activities, personnel traffic, and the movement of equipment, will be monitored and controlled to avoid personnel contamination and assure that exposures are maintained ALARA. j During the re-racking process, the existing spent fuel pool storage racks will be removed, then -

washed down in preparation for packaging and shipment. Estimates of the person-rem exposures associated with washdown and preparing the old racks for shipment are included in Attachment E, Table 9.1. Shipping containers and procedures will conform to Federal )

regulations as specified in 10 CFR 71, " Packaging and Transportation of Radioactive Material," and to the requirements of any state through which the shipment may pass, as set forth by the State Department of Transportation (DOT) office. {

Further details of the proposed changes and the associated analysis are provided in the Holtec international report, " Licensing Report for Spent Fuel Rack Installation at Byron and

. Braidwood Stations," HI-982083. This report can be found in Attachment E.

Solid Radwaste Generation The necessity for resin replacement is determined primarily by the requirement for water clarity. The spent fuel pool demineralizer resin is normally replaced approximately once per year. No significant increase in the volume of solid radioactive wastes is expected with the expanded storage capacity. During reracking operations, small amounts of additional waste resin may be generated by the pool cleanup systems on a one-time basis.

6 No effects are anticipated on non-radiological waste stream generation, specifically for air, wastewater, solid waste or hazardous waste, therefore, no changes to the National Pollutant Discharge Elimination System permit are required.

Gaseous and Liouid Releases Gaseous releases from the Fuel Handling Building are combined with other plant exhausts.

Normally, the contribution from the Fuel Handling Building is negligible compared to the gaseous and liquid releases from other sources. No significant increases are expected as a f

result of the minimal storage capacity expansion (i.e.,114 storage cells or a 4% increase in storage capacity).

i A - 26

The storage of spent fuel assemblies does not directly affect the release of radioactive liquids from the plant, since radioactive liquids are not directly discharged from the spent fuel pool.

Conclusion i The new spent fuel pool storage racks are designed to meet all applicable requirements for safe storage of spent fuel and are functionally compatible with the spent fuel pool in addition, the new spent fuel pool storage racks and fuel configurations have been analyzed considering criticality, thermal hydraulic, seismic, structural, and radiological effects. The results of these analyses show that the proposed changes are adequate for ensuring spent fuel assemblies are maintained in a safe, subcritical, and coolable configuration under all normal and postulated conditions.

G. IMPACT ON PREVIOUS SUBMITTALS We have reviewed the proposed changes for impact on any previous submittals, and have determined that there is an outstanding previous license amendment submittal that has requested a change to page 4.0-2 of Technical Specifications which is also a page affected by this amendment request. _ The previous amendment request, submitted in a letter from R.

M. Krich (Comed) to the NRC, dated October 30,1998, requested a change to Design Feature Specification 4.3.2,." Drainage." The proposed change requested that the pool elevation specified in the statement, "The spent fuel pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 423 ft,2 inches (Byron),423 ft, O inches (Braidwood)," be changed to "410 feet 0 inches" for both Byron and Braidwond. The previous change is unaffected by the changes requested within this submittal. The nominal top of active fuel in the new racks will be at the 398' 7" elevation.

The maximum rack support leg upward adjustment, for leveling purposes, is 3 % inches.

Therefore the maximum possible active fuel elevation would be 398' 10 %", which is greater than 10 feet below the lowest possible spent fuel pool water level (i.e.,410 feet 0 inches) as required by the Standard Review Plan, SRP Section 9.1.3, " Spent Fuel Pocl Cooling and Cleanup System."

H. SCHEDULE REQUIREMENTS We request approval of these proposed changes prior to December 24,1999, to support activities associated with rack installation. Byron Station plans to install the new Holtec spent fuel pool storage racks in January 2000. Braidwood Station plans to install the new Holtec spent fuel pool storage racks in January 2001.

As the proposed TS changes address both the existing Joseph Oat and Holtec spent fuel pool storage racks, the amendment may be made effective upon issuance, to be implemented within 30 days.

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p

]

L REFERENCES (1) Byron and Braidwood Technical Specifications 3.7.15,3.7.16,4.3.1,4.3.3 and associated Bases (2) ' Byron and Braidwood UFSAR, Revision 7, December 1998 i

)

(3) Byron Station Units 1 and 2, License Amendment No. 25 (4) Braidwood Station Units 1 and 2, License Amendment No. 20 (5) NRC Position Paper, "OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications," issued in a letter from B. K. Grimes (NRC) to All Power Reactor Licensees, dated April 14,1978 (6) NUREG-0612, " Control of Heavy Loads at Nuclear Power Plants," July 1980 ,

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SUMMARY

OF REVISIONS I

Revision 0: Inititialissue. l Revision 1: Figures 11.1 through 11.13, which showed a preliminary rack change-out sequence, were removed from Chapter 11 (Installation) and the List of Figures.

Minor corrections were also made to the Table of Contents.

Revision 2: Text was added to Sections 11.5 and 11.6, which describe the fuel shuffling and the new rack installation. Minor changes were also made to the Table of Contents.

Revision 3: Minor changes were made to Paragraph 11.1 h. (ALARA Procedure), Subsection 11.7.2, and Subsection 11.7.3.

Revision 4: Editorial changes were made on pages 1-1,5-2,5-17,8-7, and 8-11.

1 i

1 Holtec Intemational i Report HI-982083 j

m

TABLE OF CONTENTS -

1.0 ' INTRODUCTION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 - 1 1.1 References for Section 1 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-4 2.0 HIGH DENSITY SPENT FUEL RACKS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-1

' 2.1 General Description . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2- 1 2.2 Summary of Principal Design Criteria . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-2 2.3 Applicable Codes and Standards . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-4 2.4 Quality Assurance Program . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-11 2.5 Mech anical Design . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2- 12 2.6 Rack Fabricati on . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 2-12 2.6.1 Fabrication Requirements . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-13 l'

.2.6.2 Byron and Braidwood Rack Modules . . . . . . . . . . . . . . . . . . . . . . . . 2-13 3.0 MATERIAL AND HEAVY LOAD CONSIDERATIONS . . . . . . . . . . . . . . . . . . . . . 3-1 3.1 Introduc ti on . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3- 1 3.2 Structural Materials . . . . . . . ............................. ...... 3-1 3.3 Neutron Absorbing Materials . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-1 3.4 Compatibility with Coolant . . .................................... 3-3 3.5 Heavy Load Considerations for the Proposed Reracking Operation . . . . . . . . 3-3

! 3.6 References for Section 3 ........ ............ .................. 3-7 f

4.0 CRITICALITY S AFETY EVALUATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-1 4.1 De si gn B ase s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4- 1 4.2 Summary of Criticality Analyses . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-4 4.2.1 Normal Operating Conditions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-4 4.2.1.1 Regi on I . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-4 4.2.1.2 Regi on 11 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-4 4.2.2 Abnormal and Accident Conditions . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-5 4.3 Reference Fuel Storage Cells ....................................4-7

! 4.3.1 Reference Fuel Assembly . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-7 L 4.3.2 Region 1 Fuel Storage Cells . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-7 l 4.3.3 Region II Fuel Storage Cells . . . . . . . . . . . . . . . . . . . . ........... 4-8 4.4 Analytical Methodology . . . . . . . . . . ............................ 4-9 4.4.1 Reference Design calculations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-9 4.4.2 Fuel Bumup Calculations and Uncertainties . . . . . . . . . . . . . . . . . . . 4-10 4.4.3 Effect of Axial Bumup Distribution . . . . . . . . . . . . . . . . . . . . . . . . 4-11 j 4.4.4 Long-Term Changes in Reactivity . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-12 l

i I

j - Holtec Intemational- ii Report HI-982083

?

t

l J

TABLE OF CONTENTS i

4.5 Region I Criticality Analyses and Tolerances . . . . . . . . . . . . . . . . . . . . . . . . 4-13 4.5.1 Nominal Design Case . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-13 4.5.2 Uncertainties Due to Burnup . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-13 4.5.3 Uncertainties Due To Tolerances . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-13 J 4.5.4 Eccentric Fuel Positioning . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-13 )

4.5.5 Water-Gap Spacing Between Racks . . . . . . . . . . . . . . . . . . . . . . . . . 4-14 l 4.6 Region H Criticality Analyses and Tolerances . . . . . . . . . . . . . . . . . . . . . . . 4-15 4.6.1 Nominal Design Case . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-15 .

4.6.2 Uncertainties Due to Bumup . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-15 4.6.3 Uncertainties Due to Tolerances . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-15 q 4.6.4 Eccentric Fuel Positioning . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-16 )

4.6.5 Water-Gap Spacing Between Racks . . . . . . . . . . . . . . . . . . . . . . . . . 4-16

, 4.7 Abnormal and Accident Conditions . . . . . . . . . . . . . . . . . ............... 4-17 I 4.7.1 Temperature and Water Density Effects . . . . . . . .......... 4-17 4.7.2 Lateral Rack Movement . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-17 4.7.3 Abnormal Location of a Fuel Assembly . . . . . . . . . . . . . . . . . . . . . . 4-18 j j

. 4.7.4 Dropped Fuel Assembly . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-19 4.8 References for Section 4 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-20 i

! Appendix 4A: Benchmark Calculations 4 i l i

5.0 THERMAL-HYDR AULIC CONSIDER ATIONS . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-1 i 5.1 Introducti on . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ........ 5-1 5.2 Spent Fuel Pool and Cooling System Descriptions . . . . . . . . . . . . . . . . . . . . . 5-2 5.3 Decay Heat Load Calculations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-4 5.4 Discharge Scenarios . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-5 5.5 Bulk Pool Temperatures . . . . . . . . . . . . . ...........................54 ,

local Pool Water Temperature . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-10 l 5.6

.! 5.6.1 B as i s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 - 10 1 ;

5.6.2 Local Temperature Evaluation Model . . . . . . . . . . . . . . . . . . . . . . . . 5-11 5.7 Cladding Temperature . . . . . . . . . . . . . . . . . . . . . . . . . ........ ....... 5-13 i 5.8 Resul ts . . . . . . . . . . . . . . . . . . . . .......................... .. ..., 5-15 l 5.8.1 Bulk Pool Temperature . . . . . . . . 4 ............... ......... 5-15 5.8.2 Ti me-to-B oil . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 - 16 ,

5.8.3 Local Water and Cladding Temperature . . . . . . . . . . . . . . . . . . . . . . 5-16 5.9 References for Section 5 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-18 i i

I Holtec International iii Report HI-982083 L

v TABLE OF CONTENTS, -

6.0 . STRUCTURAL SEISMIC CONSIDERATIONS . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-1 6.1 In troduction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6- 1 6.2 Overview of Rack Structural Analysis Methodology . . . . . . . . . . . . . . . . . . . 6-1 6.2.1. Background of Analysis Methodology . . . . . . . . . . . . . . . . . . . . . . . . 6-2 6.3 Description of Racks and Fuel . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-5

' 6.4 Synthetic Time-Histories . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-5 6.5 3-D Nonlinear Rack Model for Dynamic Analysis . . . . . . . . . . . . . . . . . . . . . 6-7 6.5.1 General Remarks . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-7 !

6.5.2 ' Element Details . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6- 11 11 6.5.3 Fluid Coupling Effect .................................... 6-11 6.5.4 Stiffness Element Details . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-13 1 6.6 Whole Pool Multi-Rack Methodology . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-15 '

6.6.1 General Remarks . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 15 6.6.2 Multi-Body Fluid Coupling Phenomena . . . . . . . . . . . . . . . . . . . . . . 6-15 6.6.3 Coefficients of Friction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-16 6.6.4 Governing Equations of Motion . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-17 6.7 Structural Evaluation of Spent Fuel Rack Design . . . .'. . . . . . . . . . . . . . . . . 6-18 6.7.1 Kinematic and Stress Acceptance Criteria . . . . . . . . . . . . . . . . . . . . . 6-18 i

-l 6.7.2 Stress Limit Evaluations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-19 6.7.3 Dimensionless Stress Factors . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-22 6.7.4 Loads and leading Combinations for Spent Fuel Racks . . . . . . . . . . 6-22 j 6.8 Parametric Simulations. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-24 l 6.9 Time History Simulation Results . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-26 6.9.1 Rack Displacements . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 26

6.9.2 Pedestal Venical Load . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-28 6.9.3 Pedestal Friction Forces . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-30 6.9.4 Rack Impact Loads . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-31 y 6.9.4.1 Fuel to Cell Wall Impact Imads . . . . . . . . . . . . . . . . . . . . . . . 6-32

- 6.9.5 Rack Venical Displacement . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-35 l

6.10 Rack Stmetural Evaluation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-37 l 6.10.1 Rack Stress Factors . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-37 6.10.2 Pedestal Thread Shear Stress . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-38 6.10.3 Local Stresses Due to Impacts . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-40 6.10.4 Assessment of Rack Fatigue Margin . . . . . . . . . . . . . . . . . . . . . . . . 6-41 6.10.5 Weld Stresses . . . . . . . . . , . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-4 2 6.11 le vel A Evaluation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-44 6.12 Hydrodynamic Loads on Pool Walls . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6- 44 6.13 Thermal Stresses From Asymmetric Heat Generation . . . . . . . . . . . . . . . . . 645 I 6.14 ' Overhead S torage . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-46 i

Holtec International iv Report HI-982083

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-TABLE OF CONTENTS

! 6.15 . Conc l usion . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-46 i 6.16 References for Section 6 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-48

- 7.0 FUEL HANDLING AND CONSTRUCTION ACCIDENTS . . . . . . . . . . . . . . . . . . . 7-1 7.1. Introduc tion . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7- 1 7.2 Description of Fuel Handling Accidents . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7-1 7.3 ' Incident Fuel Assembly Velocity . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7-2 7.4 Mathematical Model . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7-4 7.5 Re s ul t s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7-4 7.5.1 Shallow Drop Events . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7-4

! 7.5.2 Deep Drop Events . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7-5 7.6 Conc l usi on . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 -5 7.7 - References for Section 7 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7-6 8.0 FUEL POOL STRUCTURAL INTEGRITY CONSIDERATIONS . . . . . . . . . . . . . 8-1 8.1 ' In troduc ti on . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8- 1 8.2 Description of the Spent Fuel Pool Stmeture . . . . . . . . . . . . . . . . . . . . . . . 8-2

8.3 M aterial Properties . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8-3 8.4 Load Combinations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ............ 8-4 8.5 Analysis Methodology . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8-6

- 8.6 Pool Liner Integrity Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8-9

' 8.7 . Bearing Pad An alysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8- 11 8.8 Results and Conclusions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8-12 8.9 References for Section 8 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8-13 9.0 RADIOLOGICAL EVALU ATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9-1 9.1 . Fuel Handling Accident . . . . . . . . . . . . . . . . . . . . . .................... 9-1

9.2 S olid Radwaste . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9-2 i 9.3 Gaseous and Liquid Releases . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9-2 9.4 Personnel Ex posure s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9-2 9.5 Anticipated Exposures During Reracking . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9-3 10.0 BOR AL SURVEILLANCE PROGRAM . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10-1 10.1 Purpose . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10- 1 10.2- Coupon Surveillance Program . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10-2 10.2.1 Coupon Description . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10-2 10.2.2 Surveillance Coupon Testing Schedule . . . . . . . . . . . . . . . . . . . . . . . 10-3 i

10.2.3 Measurement Program ...................................10-4 I 10.2.4 Surveillance Coupon Acceptance Criteria . . . . . . . . . . . . . . . . . . . . 10-4 Holtec International v Report HI-982083 i

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i TABLE OF CONTENTS l ! !

10.3 In-Service Inspection (Blackness Tests) . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10-5 10.4 References for Section 10 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10-7 11.0 INSTA11ATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 1 - 1

11.1 Introducti on . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 1 - 1 j

11.2 Rack Arrange ment . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11-4 l

. I1.3 Pool Survey and Inspection . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11-5 11.4 Pool Cooling and Purification . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11-5 i 1 1 1.4.1 Pool Cooling . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 1 -5

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11.4.2 Purification . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11 -5 i I 11.5 Fuel S huffling . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11 -5 I 11.6 Installation of New Racks . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11-6 )

11.7 Safety, Radiation Protection, and ALARA Methods . . . . . . . . . . . . . . . . . . . . Il-7 ,

1 1.7.1 S afe ty . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 1 -7 j 11.7.2 Radiation Protection . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11-7 l 1 1.7.3 ALARA . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 1 - 8 l 11.8 Radwaste Material Content . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11-9 12.0 ENVIRONMENTAL COST / BENEFIT ASSESSMENT . . . . . . . . . . . . . . . . . . . . . 12-1 12.1 Introduction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12- 1 12.2 Imperative for Rack Replacement . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12-1 12.3 Appraisal of Altemative Options . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12-1 i 12.4 Cost Es timate . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12-6 l

l 12.5 Resource Commitment . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12-6 i l

l 12.6 Environmental Considerations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12-7

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12.7 References for Section 12 . . . . . . . .............................. 12-7 i

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LIST OF FIGURES i

1-1 New Rack Layout for Byron and Braidwood Nuclear Stations 2-1 Pictorial View of Typical Region I Rack 2-2 Pictorial View of Typical Region II Rack 2-3 Seam Welding Precision Formed Channels 2-4 A Cross Sectional View of an Array of Region I Storage Cells 2-5 A Cross Sectional View of an Array of Region II Storage Cells 2-6 Elevation View of Region I Cells 2-7 Elevation View of Region II Cells f

2-8 Sheathing Shown Installed on the Box 2-9 Adjustable Support Ixg 4.1.1 Minimum Required Fuel Assembly Burnup As A Function of NominalInitial Enrichment 235 to Permit Storage in Region II (Fuel assemblies with enrichments less than 2.0 wt% U 1

will conservatively be required to meet the burnup requirements of 2.0 wt% 235U j assemblies) l 4.3.1 ' A Cross-Sectional View of the Calculational Model Used for the Region I Rack Analysis (NOT TO SCALE) 4.3.2 A Cross-Sectional View of the Calculational Model Used for the Region II Rack Analysis (NOT TO SCALE) 4A.1 Calculated k-eff Values for Various Values of the SpectralIndex ,

4A.2 Calculated k-eff Values for Various Values of the Spectral Index 4A.3 MCNP Calculated k-eff Values at Various U-235 Enrichments 4A 4 KENO Calculate k-eff Values at Various U-235 Enrichments Holtec International vii Report HI-982083

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LIST OF FIGURES

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4A.S . Comparison of MCNP and KEN 05A Calculations for Various Fuel Enrichments 4A.6 Comparison of MCNP and KENO 5a Calculations for Various Boron-10 Areal Densities 5.4.1 Byron and Braidwood Spent Fuel Pools Discharge Scenario Case (i)

-5.4.2 Byron and Braidwood Spent Fuel Pools Discharge Scenario Case (ii) 5.4.3 Byron and Braidwood' Spent Fuel Pools Discharge Scenario Case (iii) 5.5.1 Spent Fuel Pool Cooling Model 5.8.1 Bulk Pool Transient Temperature Plot for Case (i) Normal Discharge Scenario 5.8.2 Bulk Pool Transient Temperature Plot for Case (ii) Full Core Discharge Scenario 5.8.3 Bulk Pool Transient Temperature Plot for Case (iii) Back-to-Back Discharge Scenario 5.8.4 Fuel Pool Decay Heat lead for Case (i) Normal Discharge Scenario 5.8.5 Fuel Pool Decay Heat Load for Cas: (ii) Full Core Discharge Scenario 5.8.6 Fuel Pool Decay Heat Load for Case (iii) Back-to-Back Discharge Scenario

{ . 5.8.7 Post Loss of Forced Cooling Transient Pool Depth Plot

_5.8.8 Byron & Braidwood Pool Local Temperature Plot 5.8.9 Byron & Braidwood Pool Velocity Vectors Plot 6.4.1 Acceleration Time-History SSE x-direction (49 damping) i 6.4.2 Acceleration Time-History SSE y-direction (4% damping) 6.4.3 Acceleration Time-History SSE z-direction (4% damping) 6.4.4 Acceleration Time-History OBE x-direction (2% damping) 6.4.5 Acceleration Time-History OBE y-direction (2% damping)

Holtec International viii Report HI-982083 h

LIST OF FIGURES 1

6.'4.6 Acceleration Time-History OBE z-direction (2% damping) l l

-6.5.1 Schematic of the Dynamic Model of a Single Rack Module Used in DYNARACK

,6.5.2 Fuel-to-Rack Gap / Impact Elements at Level of Rattling Mass 1

6.5.3 Two Dimensional View of the Spring-Mass Simulation !

l 6.5.4 Rack Degrees-of-Freedom for X-Z Plane Bending with Shear and Bending Spnng 6.5.5 Rack Periphery Gap / Impact Elements

-6.8.1 Gap Spring Identification Scheme (At Rack Bottom) j

)

6.8.2 Gap Spring Identification Scheme (At Rack Top) 6.10:1' Rack Fatigue Analysis 6.13.1 Quarter Symmetric Model for " Hot Cell" Thermoelastic Analysis 7.2.1 Shallow Drop on a Penpheral Cell l l

7.2.2 Deep Drop on a Support Leg Locatio, 7.2.3 Deep Drop on a Center Cell Location 7.5.1 Plan View of Shallow Drop Scenario 7.5.2 Maximum Cell Deformation for Shallow Drop 7.5.3 Plan View of Deep Drop Scenarios L7.5.4 ' Maximum Von Mises Stress of the Liner for Deep Drop Scenario 2 l 7.5.5 Maximum Compressive Stress of the Concrete Slab for Deep Drop Scenario SC2 i _7.5.6' Maximum Baseplate Deformation for Deep Drop Scenario SCl 8.2.1 Isometric (Pictorial) View of Byron /Braidwood Nuclear Plant Spent Fuel Storage Pool I

Holtec International ix Report HI-982083 I :

LIST OF FIGURES 8.2.2 Plan View of Byron /Braidwood Nuclear Station Spent Fuel Storage Pool 8.2.3. Section View (A- A) of Byron /Braidwood Nuclear Station Spent Fuel Storage Pool 8.2.4 ' Section View (B-B) of Byron /Braidwood Nuclear Station Spent Fuel Storage Pool I

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Holtec Intemational X Repon HI-982083 I

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1.0 ' INTRODUCTION This report describes the design, fabrication, and safety analysis of high density spent fuel storage rarh manufactured by Holtec Intemational for the Byron and Braidwood Nuclear Stations, which are owned and operated by Commonwealth Edison Company (Comed). The rack design and analysis methods employed in the Byron and Braidwood rerack projects are a direct evolution of previous rerack license applications. The new Holtec racks meet all goveming requirements of the applicable codes and standards,in particular,"OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications," USNRC (1978) and the 1979

' Addendum thereto [1.0.1]. The analysis, material procurement, and fabrication of the rack modules conform to 10CFR50 Appendix B requirements.

Byron is 'a twin unit pressurized water reactor (PWR) site located in rural Illinois, approximately

' four miles south of the town of Byron. The site has been in commercial operation since 1985 (Unit 1) and 1987 (Unit 2). The Braidwood site, which lies close to the town of Braidwood, is a sister plant to Byron. Unit I went into commercial operation in 1988, followed by Unit 2 also in 1988. Each of the reactor cores contains 193 fuel assemblies and is rated for 1175 Mwe. The two units, on each site, share one common spent fuel storage pool, which is currently licensed to store 2,870 fuel assemblies (2,864 cells plus 6 failed fuel locations).

The existing fuel racks n Byron and Braidwood feature the neutron absorber material, Boraflex, which was used almost exclusively for rack installations in the early to mid-1980s. As many plants discovered, Boraflex tends to harden and becomes increasingly more fragile in spent fuel pool environments. This creates criticality and water chemistry hazards. Boraflex degradation is

. the main reason why Comed has decided to install new racks at Byron and Braidwood. The new racks utilize Boral, a boron carbide almninum cermet, as the neutron absorber material. Boral, which was first developed in 1949, has been used successfully at numerous plants in the U.S.,

Korea,-Mexico, Brazil, and the U.K., This license amendment seeks USNRC approval to remove

all of the existing Boraflex racks from the spent fuel pools at Byron and Braidwood and install twenty-four new Boral racks at each site.

SHADED TEXT CONTAINS PROPRIETARY INFORMATION

' Holtec Intemational 1-1 Report HI 982083

7 A plan view of the new fuel rack array is von in Figure 1-1. The proposed rack layouts for

" Byron and Braidwood are identical. The total storage capacity of the new racks is 2,984 fuel assemblies per site. This represents an increase of only 114 fuel assemblies, or 4 percent, from -

the current licensed capacity. As a result, the proposed rerack has minor impact on the existing design basis.

The new Holtec racks, which are free-standing and self-supporting, are designed to the stress ;

limits of, and analyzed in accordance with, Section IH, Division 1, Subsection NF of the ASME Boiler and Pressure Vessel (B&PV) Code. A fatigue assessment of the rack structure is also performed, which follows the guidelines in Su'o cction NB, Paragraph NB-3222.4 to the extent practical. The principal construction materials for the new racks are SA240-Type 304L stainless l steel sheet and plate stock, and SA564-630 (precipitation hardened stainless steel) for the l

adjustable suppon spindles. The only non-stainless material used for const'ruction is the Boral l neutron absorber material. Sections 2 and 3 of this repon provide an abstract of the design and material information for the new racks.

The criticality safety analysis requires tnat the neutron multiplication factor for the stored fuel array be bounded by the kerr limit of 0.95 under assumptions of 95% probability and 95%

confidence. The criticality safety analysis provided in Section 4 sets the requirements on the Boral panel length and the ' B areal density for the new high density racks.

The thermal-hydraulic requirements are that fuel claddmg does not fait due to excessive thermal stress and that the steady state pool bulk temperature remains within the prescribed structural, operational, and regulatory limits. The thermal-hydraulic analyses carried out in support of this storage expansion effort are described in Section 5.

D'emonstrations of seismic and stmetural adequacy are presented in Section 6. The analysis shows that the primary stresses in the rack module structure remain below the ASME B&PV Code (Subsection NF) [1.0.2] allowables. The structural qualification also provides conclusive evidence that the subcriticality of the stored fuel is maintained under all postulated accident SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 1-2 Report HI-982083

scenarios in the Updated Final Safety Analysis Reports (UFSAR) for Byron and Braidwood. The structural consequences of these postulated accidents are presented and evaluated in Section 7 of this report.

The structural evaluation of the reinforced concrete spent fuel pool is presented in Section 8 of this report, which includes a general description of the Byron and Braidwood pool geometries.

~

The radiological considerations are documented in Section 9, and Section 10 presents the In-Service Boral Surveillance Program.

- Section' 11 discusses all aspects related to the on-site field work The many steps and requirements for completing the installation of the new racks are detailed in this section. .

Additionally, one possible rack installation sequence is presented in this section.

l Finally, Section 12 presents a cost / benefit and environmental assessment to establish the prudence of Comed's decision to proceed with the rerack project.

k

' All computer programs utilized to perform the analyses documented in this licensing report are benchmarked and verified.These programs have been utilized by Holtec Intemational in numerous rcrack applications over the past decade.

The analyses presented herein clearly demonstrate that the rack module arrays possess wide margins of safety in respect to all considerations of safety specified in the USNRC OT Position Paper, namely, nuclear subcriticality, thermal-hydraulic safety, seismic and structural adequacy, radiological compliance, and mechanical integrity.

SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 1-3 Report HI-982083 ,

E, 1.1 References for Section 1

'[1.0.1]- USNRC 12tter to All Power Reactor Licensees transmitting the "OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications," April 14,1978, and Addendum dated January 18,1979. ,

1

[1.0.2] ASME Boiler & Pressure Vessel Code,Section III, Subsection NF and Appendices (1995).

SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 1-4 Report HI-982083

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2.0 HIGH DENSITY SPENT FUEL RACKS This section provides general information on the new storage modules proposed for the Byron i and Braidwood spent fuel pools.' It also describes the basis for the detailed criticality, thermal-hydraulic, and seismic /stmetural analyses, which are presented in subsequent sections of this report.

2.1 General Description When the installation is complete, the Byron and Braidwood pools will each contain twenty-four racks with a total cell count of 2,984 fuel storage cells. The storage cells will be divided into two .

regions based upon the rack type. _ A group of four racks will store the most reactive fuel (up to 5 weight % enrichment) without any bum-up limitation. These four racks will use a flux trap design, and they will be referred to as Region I. The remaining twenty racks, which do not have flux traps, will have an enrichment / burn-up limitation for fuel storage. These racks will be collectively referred to as Region II.

All storage rack arrays consist of freestanding modules, made from Type 304L austenitic stainless steel containing prismatic storage cells interconnected through longitudinal welds. A panel of Boral cermet containing a high areal loading of the ' B isotope provides appropriate neutron attenuation between adjacent storage cells. Sketches of typical Region I and Region II rack modules proposed for Byron and Braidwood are provided in Figures 2-1 and 2-2, respectively. Data on the cross sectional dimensions, gross weight and cell count for each rack inodule in the spent fuel pool are presented in Table 2-1.

The baseplates on all rack modules extend out beyond the rack module wall such that the contiguous edges of the plates establish a geometric separation between the facing cells in the modules.

Each new rack module is supported by legs, which are remotely adjustable. Thus, the racks can be leveled, and the top of the racks can easily be made co-planar with each other. The rack SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec Intemational 21 Report HI-982083

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. module support legs are engineered to accommodate undulations in the surface of the pool floor.

LA bearing pad interposed between the rack pedestals and the pool liner serves to diffuse the dead weight of the loaded racks into the reinforced concrete structure of the pool slab.

The current cell indexing system, which places placards in two opposing corner cells of a rack module, will continue to be used with the new racks. The placards will be removed if fuel is inserted in the those cell locations.

The overall design of the Bymn and Braidwood racks is similar to those presently in service at '

. many other nuclear plants. Among these plants are Kuosheng of Taiwan Power Company, James A. Fitzpatrick of New' York Power Authority, Zion of Commonwealth Edison, and Duane Arnold ofIowa Electric and Light. Altogether, there are thousands of storage cells of this design that have been provided by Holtec International to various nuclear plants around the world.

2.2 Summary of Principal Design Criteria The key design criteria for the new Byron and Braidwood spent fuel racks are set fonh in the USNRC memorandum entitled "OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications," April 14,1978 as modified by amendment dated January 18,1979.

The individual sections of this report address the specific design bases that are described in the above-mentioned "OT Position Paper?' A brief summary of the design basis requirements for the i l

Byron and Braidwood racks is provided below. l 1

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a. Disposition: All new rack modules are required to be free-standing.

b. ' Kinematic Stability: All free-standing modules must be kinematically stable (against tipping or overturning)if a seismic event (which is 150% of the

. postulated OBE or 110% of the postulated SSE) is imposed on any module.

c. . Structural Compliance: All primary stresses in the rack modules must satisfy the limits postulated in Section III, Subsection NF of the 1995 ASME Boiler and Pressure Vessel Code. Also, the cumulative damage factor for fatigue life must be less than 1.0 when the rack is subjected to 1 SSE and 20 OBE events. The fatigue SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 2-2 Report HI-982083 i

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assessment of the fuel rack follows the guidelines in Subsection NB, Paragraph NB-3222.4 to'the extent practical.

' d.' Thermal-Hydraulic Comoliance: The spatial average bulk pool ' temperature is required to remain under 140 F following a normal refueling, which is defined as a partial core off-load (i.e.,84 fuel assemblies). For a full core off-load, it is required to demonstrate that bulk pool boiling does not occur.

In addition to the limitations on the bulk pool temperature, the local water

- temperature in the Byron and Braidwood pools must remain below the boiling temperature coincident with local hydraulic pressure conditions.

e. Criticality Compliance: The maximum calculated reactivity of the storage rack is such that the true kerrs hall be less than 0.95 with a 95% probability at a 95%

confidence level for normal and accident conditions.

f. Radiological Compliance: The reracking of Byron and Braidwood must not lead to violation of the off-site dose limits, or adversely affect the area dose environment as set forth in the UFSARs. The radiological implications of the installation of the new racks also need to be ascenained and deemed to be acceptable.

g. Pool Structure: The ability of the reinforced concrete structure to satisfy the load combinations set forth in NUREG-0800, SRP 3.8.4 must be demonstrated.

h. Rack Cyclic Stress Fatigue: In addition to satisfying the primary stress criteria of Subsection NF, the alternating local stresses in the rack structure during a seismic )

event are also required to be sufficiently bounded such that the " cumulative damage factor" due to one SSE and twenty OBE events does not exceed 1.0.

i .' Liner Integrity: The integrity of the liner under cyclic in-plane loading during a j seismic event must be demonstrated.

j. Bearing Pads: The bearing pads must be sufficiently thick such that the pressure on the liner continues to satisfy the ACI limits during and after a design basis seismic event.

k. Accident Events: In the event of a load drop (e.g., uncontrolled lowering of a fuel assembly), it is necessary to demonstrate that the subcriticality of the rack struen::e is not compromised.

1. Construction Events: The field constmetion services required to be canied out for executing the reracking must be demonstrated to be within the " state of proven an."

SHADED TEXT CONTAINS PROPRIETARY INFORMATIdN Hollec International 2-3 Report HI-982083

The foregoing' design" basis requirements are further articulated in Sections 4 through 9 of this

- licensing report.

- 2.3 . Applicable Codes and Standards The fabrication of the rack modulu is performed under a strict quality assurance system suitable

' for 10CFR50 Appendix B manufacturing.

-The following codes, standards and practices are used as applicable for the design, construction, and assembly of the Byron and Braidwood fuel storage racks. Additional specific references related to detailed analyses are given in each section. l

a. Design Codes (1) AISC Manual of Steel Construction,8th Edition,1980.

(2) ANSI N210-1976, " Design Requirements for Light Water Reactor Spent <

Fuel Storage Facilities at Nuclear Power Stations" (contains guidelines for fuel rack design). j (3) American Society of Mechanical Engineers (ASME), Boiler and Pressure -

Vessel Code Section III(Subsections NB and NF),1995 Edition; ASME Section V,1995 edition; ASME Section VIII,1995 Edition: ASME Section IX,1995 Edition; and ASME Section XI,1995 Edition.

(4) ASNT-TC-1 A, American Society for Nondestructive Testing

' (Recommended Practice for Personnel Qualifications), June 1980.

(5) 'American Concrete Institute Building Code Requirements for Reinforced Concrete (ACI318 63) and (ACI318-71,77,83).

(6) Code Requirements for Nuclear Safety Related Concrete Structures, ACI349-85/ACI349R-85 and ACI349.lR-80.

(7) ASME NQA-1-1989, Quality Assurance Program Requirements for Nuclear Facilities.

(8) ASME NQA-2-1989, Quality Assurance Requirements for Nuclear 1 Facility Applications.

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(9) ASME Y14.5M-1994, Dimensioning and Tolerancing (revision and

redesignation of ANSI Y14.5M-1982)

-(10) ACI Detailing Manual,1980.

b. Material Codes - Standards of ASTM (1) E165-95, Liquid Penetrant Examination.

-(2) A240/A240M-97a, Standard Specification for Heat-Resisting Chromium and Chromium-Nickel Stainless Steel Plate, Sheet, and Strip for Pressure Vessels.

l (3) A262-93a, Detecting Susceptibility to Intergranular Attack in Austenitic Stainless Steel.

4 (4) A276-97, Stainless Steel Bars and 'hapes.

(5) A479/A479M-97a, Stainless Steel Bars and Shapes for Use in Boilers and i Other Pressure Vessels.

t (6) A564/A564M-95, Hot-Rolled and Cold-Finished Age-Hardening Stainless Steel Bars and Shapes.

(7). C750-89, Nuclear-Grade Boron Carbide Powder.

(8) A380-96, Cleaning, Descaling, and Passivation of Stainless Steel Parts, Equipment, and Systems. {

l (9) C992-89, Baron-Based Neutron Absorbing Material Systems for Use in I Nuclear Spent Fuel Storage Racks. J (10) E3-95, Preparation of Metallographic Specimens.

(11) E190-92, Guided Bend Test for Ductility of Welds.

(12) American Society of Mechanical Engineers (ASME), Boiler and Pressure )

Vessel Code, Section II-Parts A and C,1995 Edition.

(13) NCA3800 - Metallic Material Manufacturer's and Material Supplier's Quality System Program.

c. Weldine Codesi ASME Boiler and Pressure Vessel Code,Section IX - Welding i and Brazing Qualifications,1995 Edition.

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d. Ouality Assurance. Cleanliness. Packaging. ShiDDing Receiving. Storage, and Handling Reauirements (1) ANSI N45.2.1-1980, Cleaning of Fluid Systems and Associated Components During Construction Phase of Nuclear Power Plants.

(2) ANSI N45.2.2-1972, Packaging, Shipping, Receiving, Storage and

. Handling of Items for Nuclear Power Plants (During the Construction Phase).

(3). ANSI N45.2.6-1978, Qualifications of Inspection, Examination, and Testing Personnel for Nuclear Power Plants (Regulatory Guide 1.58).

(4) ANSI N45.2.8-1975, Supplementary Quality Assurance Requirements for Installation, Inspection and Testing of Mechanical Equipment and Systems for the Construction Phase of Nuclear Plants.

(5) ANSI N45.2.11 1974, Quality Assurance Requirements for the Design of Nuclear Power Plants.

. (6) ANSI N45.2.12-1977, Requirements for Auditing oT Quality Assurance Programs for Nuclear Power Plants.

(7) ANSI N45.2.13-1976,. Quality Assurance Requirements for Control of Procurement of Equipment Materials and Services for Nuclear Power Plants (Regulatory Guide 1.123).

(8) ANS.I N45.2.15-18, Hoisting, Rigging, and Transporting of Items for Nuclear Power Plants.

(9) ANSI N45.2.23-1978, Qualification of Quality Assurance Program Audit Personnel for Nuclear Power Plants (Regulatory Guide 1.146).

- (10) ASME Boiler and Pressure Vessel,Section V, Nondestructive Examination,1995 Edition.

(11) ANSI N16.9-1975, Validation of Calculation Methods for Nuclear Criticality Safety.

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e. Goveming NRC Design Documents (1) "OT Position for Review nd Acceptance of Spent Fuel Storage and Handling Applications," ted April 14,1978, and the modifications to j this document of January 18,1979, 1 I

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Holtec Intemational 2-6 Repon HI-982083 j

(2) NUREG M12, " Control of Heavy Loads et Nuclear Power Plants",

USNRC, Washington, D.C., July 1980.

(3) NUREG-0800, Radiological Consequences of Fuel Handling Accidents.

f. Other ANSI Standards (not listed in the treceding)

(1) ANSI /ANS 8.1 (N16.1), Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors.

(2) ANSI /ANS 8.17-1984 (R1997), Criteria for the Handling, Storage, and Transponation of LWR Fuel Outside Reactors.

(3) ANSI N4fi.2-1971, Quality Assurance Program Requirements for Nuclear Facilities. .

(4) ANSI N45.2.10-1973, Quality Assurance Terms and Definitions.

(5) ANSI /ANS 57.2 (N210), Requirements for Light Water Reactor Spent Fuel Storage Facilities.

(6) ANSI N14.6-1993, Radioactive. Materials - Special Lifting Devices for Shipping Containers Weighing 10,000 pounds (4,500 kg) or More.

(7) ASME N626.3-1993, Qualification and Duties of Specialized Professional Engineers.

(8)' ANSI N18.2-1973, Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants.

(9) ANSI 8.21, Use of Fixed Neutron Absorbers in Nuclear Facilities Outside j Reactor (supplemented by 4.le (2) as applicable).

g. Code of Federal Regulations (1) 10CFR20, Standards for Protection Against Radiatien,1997 Edition.

I (2) 10CFR21, Reponing of Defects and Non-compliance,1997 Edition.

(3) 10CFR50 Appendix A, General Design Criteria for Nuclear Power Plants, 1997 Edition.

1' (4) 10CFR50 Appendix B, Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing' Plants,1997 Edition.

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.(5) 10CFR61, Licensing Requirements for Land Disposal of Radioactive Material,1997 Edition.

1 (6) 10CFR71, Packaging and Transportation of Radioactive Material,1997

. Edition.

10CFR100, Reactor Site Criteria l (7)

I (8) 49CFR 171-178

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h. - Regulatory Guides (1) RG 1.13 - Rev.1, Spent Fuel Storage Facility Design Basis (Revisiod 2 Proposed).

l (2) RG 1.25 - Rev,0, Assumptions Used for Evaluating the Potential l

. Radiological Consequences of a Fuel Handling Accident in the Fuel )

Handling and Storage Facilit} of Boiling and Pressurized Water Reactors.

(3) RG 1.28 - Rev. 3 (ANSI N45.2), Quality Assurance Program Requirements . l (4) RG 1.29, Seismic Design Classification (Rev. 3).

(5) RG 1.31 - Rev. 3, Control of Ferrite Content in Stainless Steel Weld l Material.

(6) RG 1.38 - Rev. 2 (ANSI N45.2.2), Quality Assurance Requirements for Packaging, Shipping, Receiving, Storage and Handling ofItems for Water-Cooled Nuclear Power Plants.

(7) RG 1.44 - Rev. 0, Control of the Use of Sensitized Stainless Steel.

(8) RG 1 id - Rev.1 (ANSI N45.2.6), Qualification of Nuclear Power Plant Inspe tion, Examination, and Testing Personnel.

(9) RG 1.61, Damping Values for Seismic Design of Nuclear Power Plants, Rev. O,1973.

(10) RG 1.64 - Rev. 2 (ANSI N45.2.11), Quality Assurance Requirements fc.- )

the Design of Nuclear Power Plants.

(11) RG 1.71 - Rev. 0, Welder Qualifications for Areas of Limited l Accessibility.

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, (12) RG 1.74 (ANSI N45.2.10), Quality Assurahce Terms and Definitions,

(- February 1974.

(13) RG 1.85, Materials Code Case Acceptability - ASME Section 3, Div.1.

-(14) RG 1.88 - Rev. 2 (ANSI N45.2.9), Collection, Storage and Maintenance

of Nuclear Power Plant Quality Assurance Records.

(15) RG 1.92, Combining Modal Responses and Spatial Components in Seismic Response Analysis.

(16) RG 1.122, Development of Floor Design Response Spectra for Seismic Design of Floor-Supported Equipment or Components.

(17) RG 1.123 - Rev.1 (ANSI N45.2.13), Quality Assurance Requirements for Control of Procurement of Items and Services for Nuclear Power Plants.

(18) RG 1.124 - Rev.1, Service Lunits and Loading Combinations for Class 1 Linear-Type Component Supports, Revision 1,1978.

(19) RG 3.4, Nuclear Criticality Safety in Operations with Fissionable Materials at Fuels and Materials Facilities.

(20) RG 3.41 - Rev.1, Validation of Calculational Methods for Nuclear Criticality Safety, Revision 1,1977.

(21) RG 8.38, Control of Access to High and Very High Radiation Areas in Nuclear Power Plants, June 1993. .

1 (22) RG 8.8, Information Relative to Ensuring that Occupational Radiation Exposure at Nuclear Power Plants will be as Low as Reasonably Achievable (ALARA).

(23) DG.8006," Control of Access to High and Very High Radiation Areas in Nuclear Power Plants."

(24) IE Information Notice 83 Fuel Einding Caused by Fuel Rack Deformation.

i. Branch Technical Position 4

(1) CPB 9.1-1, Criticalityin Fuel Storage Facilities.

l (2) ASB 9-2, Residual Decay Energy for Light-Water Reactors for Long-Term l Cooling.

l i

! SilADED TEXT CONTAINS PROPRIETARY INFORMATION lioltec International 2-9 Report HI-982083

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- j. : Standard Review Plan -

E(1) SRP 3.2.1. Seismic Classification.

_'(2) SRP 3.2.2, System Quality Group Classification.

(3) SRP 3.7.1,-Seismic Design Parameters.

-(4) SRP,3.7.2, Seismic System Analysis.

(5) SRP 3.7.3, Seismic Subsystem Analysis.

(6) SRP 3.8.4, Other Seismic Category I Structures (including Appendix D),'

Technical Position on Spent Fuel Rack..

.(7) ' SRP 3.8.5, Foundations for Seismic Category I Structures, Revision 1, 1981.

(8) SRP 9.1.2 Spent Fuel Storage, Revision 3,1981.

(9) SRP 9.1.3, Spent Fuel Pool Cooling and Cleanup System.

(10) SRP 9.1.4, Light Load Handling System. ,

(11) SRP 9.1.5, Heavy Load Handling System.

(12) SRP 15.7.4, Radiological Consequences of Fuel Handling Accidents.

k. AWS Standards -

(1) AWS DI.1-94, Standard for Steel - Structural Welding Code.

(2) AWS DI.3-98, Structure Welding Code -- Sheet Steel.

-(3) ' AWS D9.1-90, Sheet Metal Welding Code.

-(4)- AWS A2.4-98, Symbols for Welding, Brazing, and Nondestructive Examination.

(5) ~ AWS A3.0-94, Standard Welding Terms and Definitions.

(6)~ AWS A5.12/A5.12M-98, Tungsten and Tungsten Alloy Electrodes for Arc Welding and Cutting.

I SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec Intemational 2-10 Report HI 982083 1

(7) AWS QCl-96, AWS Certification of Welding Inspectors.

1. Other (1). EPRI: "A Methodology for Assessment of Nuclear Power Plant Seismic Margin (Rev.1), EPRI NP-6041-SL (2) Byron and Braidwood UFS ARs, SERs, and plant technical specifications.

(3) NCIG-01, Rev. 2, Visual Weld Acceptance Criteria for Structural Welding at Nuclear Power Plants.

(4) Byron and Braidwood Chemical Control Program.

I (5) " Critical Experiments Supporting Close Proximity Water Storage of Power Reactor Fuel," B AW-1484-7, July 1979.

(6) Comed QABL.

I 2.4 Ouality Assurance Program 1

The governing quality assurance require ~nts for fabrication of the spent fuel racks are enunciated in 10CFR50 Appendix B. Im . ality assurance program for design of the Byron and Braidwood racks are described in Holtec's Nuclear Quality Assurance Manual, which has been reviewed and approved by Commonwealth Edison Co.. This program is designed to provide a flexible but highly controlled system for the design, analysis and licensing of customized l 1

-components in accordance with various codes, specifications, and regulatory requirements.

l The'manufactuting of the racks will be canied out by Holtec's designated manufacturer, U.S.

1 Tool & Die, Inc. (UST&D). The Quality Assurance System enforced on the manufacturer's shop l floor shall provide for all controls necessary to fulfill all quality assurance requirements with sufficient simplicity to make it functional on a day-to-day basis. UST&D.has manufactured high density racks.for over 60 nuclear plants around the world. Holtec and Comed have audited '

UST&D to ensure that their Quality Assurance Program meets the requirements of 10CFR50 Appendix B. UST&D has also been audited by the industry group NUPIC and the QA branch of NMSS with satisfactory results.

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The Quality Assurance System that will be used by Holtec to install the racks is also controlled by the Holtec Nuclear Quality' Assurance Manual and by any site-specific requirements.

l 2.5 - Mechanical Desig i

[ The rack modules are designed as cellular structures'such that each fuel cell has a prismatic square opening with lateral support and a flat horizontal bearing surface. The basic characteristics of the Byron and Braidwood spent fuel racks are summarized in Table 2-2.

A central objective in the design of the new rack modules is to maximize their stmetural rigidity while minimizing their inenial mass. Accordingly, the Byron and Braidwood modules have been designed to simulate multi-flange beam structures. The multiple flanges are formed from the numerous cell walls in the rack cross-sectional array. These cells are connected through intermittent welds. The weld lengths, location, and size were chosen during the original design of this rack style / series to ensure adequate strength and to adjust the natural frequency of the rack modules to avoid resonance. In general, this effon has resulted in excellent detuning ch'aracteristics with respect to the applicable seismic events.

- 2.6 Rack Fabrication I This subsection provides an item-by-item description of the anatomy of the Byron and Braidwood rack modules in the context of the fabrication method. The object is to provide a self-contained description of rack module construction and to enable an independent appraisal of the adequacy of the design. ,

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2.6.1' Fabricetion Reauiremenis

< There are four basic requirements for the Byron and Braidwood high density storage racks. The .

requirements are:

i. _ The rack modules are fabricated in such a manner that there is no weld splatter on

, the storage cell surfaces which could come in contact with the fuel assembly.

ii. - The storage locations are constructed so that redundant flow paths for the co-o lant are available.

iii. The storage cells are connected to each other by austenitic stainless steel corner welds, which lead to a honeycomb lattice construction. The extent of welding is -

selected to "detune" the racks from the seismic input motion (OBE and SSE). -

. iv. The fabrication process involves operational sequences that permit immediate verification by the inspection staff.

4 2.6.2 ' Byron and Braidwood Rack Modules There are five significant components (discussed below) of the storage racks: (a) the composite box subassembly, (b) the baseplate, (c) the neutron absorber material, (d) the sheathing, and (e) l the support legs.

a. Composite box subassembiv: The rack module manufacturing begins with fabrication of the " box." The boxes are fabricated from two precision formed channels by seam welding in a machine equipped with copper chill bars and pneumatic clamps to minimize distortion due to welding heat input. The minimum weld penetration is 80% of the box metal gage. The nominal inside

~ dimensions of the PWR box (or cell) are !C inches for Region I racks and L inches for Region Il racks. This process results in a square box as shown in Figure 2-3. Metal sheathing is then attached to select sides of each box, and the 3 poison material is installed in the sheathing cavities. 1 The square cross section box with Boral panels affixed to its extemal surfaces is referred to as the "c'omposite box assembly." Each composite box has E . inch diameter holes punched near its bottom edge to provide auxiliary flow. Region I storage cells have four holes at the base, and Region II storage cells have two holes aligned at the base.

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. Region I racks are constructed by welding together an array of composite box

(

assemblies, as shown in Figure 2-4. The cell spacing and the Hux traps are controlled by nan ow strips of sheet metal, which are edge welded to the comers of the composite boxes.

' In Region II racks, the composite boxes are arranged in a checkerboard array and i welded edge-to-edge to form an' assemblage of storage cell locations, as shown in !

Figure 2-5. Filler panels and corner angles are welded to the edges of boxes at the outside boundary of the rack to complete the formation of the peripheral cells. 4 The inter-box welding and pitch adjustment are accomplished by small longitudinal connectors. The connectors are sized and placed to ensure that the

- inside cell dimension for developed boxes is maintained after inclusion of any )

. reductions from the sheathing. This assemblage of composite boxes results in'a honeycomb structure with axial, flexural and torsional rigidity which depend on the extent of inter-cell welding. Figures 2-5 shows that all four comers of each interior box are connected to the adjacent boxes, which creates a well-defined I path for " shear now."

. Elevation views of Region I and Region II storage cells are shown in Figures 2-6 and 2-7, respectively.

b. Baseplate: A- inch thick baseplate provides a continuous horizontal surface for supporting the fuel assemblies. The baseplate has W;9; inch diameter holes at each, cell location, except at lift locations. At the four lift locations, a 63 inch byb4 ; inch rectangular cut-out is centered over a k& inch circular flow hole to allow insertion and engagement of the lifting rig. The location of all baseplate holes coincides with the cell centerlines. The baseplate is attached to the base of I the cell assemblage by fillet welds and extends horizontally beyond the periphery of the rack cells. Refer to Figures 2-6 and 2-7. ;

1

c. Neutron absorber material: As mentioned in the preceding section, Boral is used as the neutron absorber material. For Region 11 racks, every interior cell wall and a selec.t number of boundary walls are equipped with one integral Boral sheet (poison material). For Region I racks, each flux trap (or water gap) is sandwiched between two Boral panels.

d. Sheathing: As described earlier, metal sheathing is fastened to select sides of each l box wall. The design objective calls for attaching Boral tightly on the box surface. This is accomplished by die forming the internal and external boral sheathings to provide end Dares with smooth edges, as shown in Figure 2-8. The Danges of the sheathing are welded to the box using skip welds and spot welds.

The sheathings serve to locate and position the poison sheet accurately and to preclude its movement under seismic conditions. The sheathing also provides a vented enclosure for the Boral.

. SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 2-14 Report HI-982083

e. Support lens: All suppon legs are the adjustable type as shown in Figure 2-9. The top (female threaded) ponion is made of austenitic stainless steel. The bottom (male threaded) ponion is made of 17:4 Ph series stainless steel. Each support leg is equipped with a readily accessible socket to enable remote leveling of the rack after its placement in the pool. The suppon legs are located at the centerlines of cells to ensure accessibility of the leveling tool through the [$5Ij inch diameter flow hole in the baseplate. Four radial holes and one vertical hole are drilled in the bottom portion of the support leg to allow for cooling flow.

The assembly of the rack modules is carried out by welding the composite boxes in a venical fixture with the precision fabricated baseplate serving as the bottom positioner.

SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 2-15 Report HI-982083

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@E in (Region II)

Cell pitch (nominal) EdiShin (Region I, N-S dir.)

M'II)in (Region I, E-W dir.)

de t in (Region II) l

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Storage cell height (above the baseplate) =in Baseplate thickness t iin Baseplate hole size 4.5 in Support leg height (nominal) h .vii in ,

Support leg type - Remotely adjustable legs Number of support pedestals _ per rack - 4 (Racks W, X - 5)

Remote lifting and handling provisions -Yes Poison material- -

Boral Poison length - ! Sin Poison width - rY in

' SHADED TEXT CONTAINS PROPRIETARY INFOR.MATION

.Holtec International ~ 218 Report HI-982083

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f3.0 MATERIAL AND HEAVY LOAD CONSIDERATIONS 3.1 Introduction Safe storage of nuclear fuel in the Byron and Braidwood pools requires that the materials utilized in the rack fabrication be of proven durability and be compatible with the pool water environment. This section provides a synopsis of the considerations with regard to long-term service life and short-teun construction safety.

3.2 Structural Materials The following structural materials are utilized in the fabrication of the new spent fuel racks:

a. ASME SA240-304L for all sheet metal stock

b. Baseplate: ASME SA240-304L c Interne!!v threaded support legs:- ASME SA240-304L

d. Externally threaded support spindle: ASME SA564-630 precipitation hardened stainless steel (heat treated to 1100 F)

e. Bearing pads: ASME SA240-304

f. Weld material - per the following ASME specification: SFA 5.9 Ell 308L 3.3 Neutron Absorber Niaterials In addition to the structural and non-stmetural stainless material, the racks employ Boral, a

, patented product of AAR Manufacturing, as the neutron absorber material. A brief description of Boral, and its pool experience list follows.

Bors' is a thennal neutron poison material composed of boron carbide and 1100 alloy aluminum.

Boron carbide is a compound having a high baron content in a physically stable and chemically inert form. The 1100 alloy aluminum is a lightweight metal with high tensile strength, which is protected from corrosion by a highly resistant oxide film. The two materials, boron carbide and SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 3-1 Report HI-982083

y 1 aluminum, are chemically compatible and ideally suited for long-term use in the radiative,-

thermal and chemical environment of a nuclear reactor or a spent fuel pool.

Boral's use in spent fuel pools as the neutron absorbing material can be attributed to its proven performance (over 150 pool years of experience) and the following unique characteristics:

a. The content and placement of boron carbide provides a very high removal cross-section for thermal neutrons.

b. Boron carbide, in the form of fine particles, is homogeneously dispersed throughout the central layer of the Boral panels,

c. The boron carbida and aluminum materials in Boral do not degrade as a result of long-term exposure to radiation. .

d. The neutron absorbing central layer of Boral is clad with permanently bonded surfaces of aluminum.

e. Boral is stable, strong, durable, and corrosion resistant.

l Holtec International's Q.A. program ensures that Boral is manufactured by AAR Manufacturing under the control and surveillance of a Quality Assurance / Quality Control Program that conforms to the requirements of 10CFR50 Appendix B," Quality Assurance Criteria for Nuclear Power Plants."

As indicated in Tables 3-1 and 3-2, Boral has been licensed by the USNRC for use in numerous PWR and BWR spent fuel storage racks and has been extensively used in nuclear installations worldwide.

l Boral Material Characteristics

' Aluminum: Aluminum is a silvery-white, ductile metallic element that is the most abundant in the earth's crust. The 1100 alloy aluminum is used extensively in heat exchangers, pressure and storage tanks, chemical equipment, reflectors and sheet metal SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 3-2 Report HI-982083

work.

It has high resistance to corrosion in industrial and marine atmospheres. Aluminum has an atomic number of 13, an atomic weight of 26.98, a specific gravity of 2.69 and a valence of 3.' The physical, mechanical and chemical properties of the 1100 alloy aluminum are listed in Tables 3-3 and 3-4.

The excellent corrosion resistance of the 1100 alloy aluminum is provided by the protective oxide film that develops on its surface from exposure to the atmosphere or water. This film prevents the loss of metal from general corrosion or pitting corrosion.

Boron Carbide: The boron carbide contained in Boral is a fine granulated powder that confonns to ASTM C-750-80 nuclear grade Type III. The material conforms to the chemical composition and properties listed in Table 3-5.

3.4 Compatibility with Coolant All materials used in the consuuction of the Holtec racks have an established history of in-pool i usage. Their physical, chemical and radiological compatibility with the pool environment is an established fact. As noted in Tables 3-1 and 3-2, Boral has been successfully used in fuel pools.

i Austenitic stainless steel (304L) is perhaps the most widely used stainless alloy in nuclear power plants.

3.5 _ Heavy Load Considerations for the Proposed Reracking Operation !

l

A 125-ton crane will be utilized for handling all heavy loads in the reracking operation. A temotely engageable lift rig, which meets NUREG-0612 stress criteria, will be used to lift the new modules. It consists of independently loaded lift rods with a " cam type" lift configuration. .

I This ensures that failure of one traction rod will not result in uncontrolled lowering of the load; {

compliant with Section 5.1.6(1) of NUREG-0612. The remotely engageable lift rig also has the I

following attributes:

I SHADED TEXT CONTAINS PROPRIETARY INFORMATION Hollec Intemational - 3-3 Report HI-982083 l

a. ' The stresses in-the lift rods are self limiting inasmuch as an mcrease m the j magnitude of the load reduces the eccentricity between the upward force and downward reaction (moment arm).

b. It is impossible for a traction rod to lose engagement with the lifted rack because the downward load secures each rod in its locking slot. Moreover, the locked configuration can be directly verified from above the pool water without the aid of an underwater camera by the orientation of position locator flags atop each tractica rod.

c. 'A stress analysis of the rig has been performed, and the stress limits postulated in

]

ANSI 14.6 (1978) are shown to be met. l

)

d. The rig is load tested with 300% of the maximum weight to be lifted. The test weight is maintained in the air for one hour. All critical weld joints are liquid- .

penetrant examined, after the load test, to establish the soundness of all critical joints. j i

l Pursuant to the defense-in-depth approach of NUREG-0612, the following additional measures of safety will be undertaken for the reracking operation.

a. The cranes and lifting,ddvices used in the project will be given a preventive maintenance checkup and inspection per Byron and Braidwood plant procedures.

b. Safe load paths will be developed for moving the new racks in the Fuel Building.

The new racks will not be carried over any region of the pool containing fuel.

c. The rack upending will be carried out in an area that is not poolside. Additionally, this area will not overlap any safety related component.

d. All crew members involved in the rack installation will be given training in the use of the lifting, upending equipment, and all other aspects of rack installation.

e. Crane stop blocks will be temporarily installed to prevent movement over fuel. l The fuel shuffle scheme developed for the spent fuel pool is predicated on the following criteria:

a. No heavy load (rack c rig) with a potential to drop on a rack shall be carried over or near active fuel. 'lius shall be accomplished by shuffling fuel into racks that are not in the area of the safe load path.

SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 3-4 Report HI-982083 1

l

u

b. All heavy loads are lifted in such a manner that the C.G. of the lift point is aligned with the C.G. of the load being lifted.

c. Tumbuckles are utilized to " fine tune" the verticality of the rack being lifted.

4 All phases of the rack installation will be conducted in accordance with written procedures, which will be reviewed s. d approved by Comed.

The guidelines contained in NUREG-0612, Section 5 will be followed throughout the. rack i l

installation. The guidelines of NUREG-0612 call for measures to " provide an adequate defense- i I

in-depth for handling of heavy loads near spent fuel..." and cite four major causes of load '

' handling accidents, namely:

\

i. Operator enors ii. Rigging failure lii. Lack of adequate inspection iv. Inadequate procedures

. 1 The Byron and Braidwood rack expansion program mitigates the potential for load drop 1 1

accidents by implementing measures that increase overall quality and safety in the four aforementioned areas. A summary of the measures specifically planned to deal with the major causes is provided below. I Operator errors: As mentioned above, Comed plans to provide comprehensive training -

to the installa' tion crew.

Riggingfailure: The lifting device designed for handling and installation of the racks in

'the fuel pool has redundancies in the lift legs, and lift eyes such that there are four independent load members. Failure of any one load bearing member would not leaa to uncontrolled lowering ~of the load. The dg complies with all provisions of ANSI 14.6 -

1978, including compliance with the primary stress cdteria, load testing at 300% of maximum lift load, and dye examination of critical welds.

The Bjron and Braidwood rig design is similar to the rigs used in the rerack of numerous other plants, such as Sequoyah, J. A. Fitzpatrick, Duane Amold, Three Mile Island Unit

, SHADED TEXT CONTAINS PROPRETARY INFORMATION

- Holtec Intemational 3-5 Report HI-982083 i

1, D.C. Cook, and Connecticut Yankee.

Lack of adequate inspection: The designer of the racks will develop a set of inspection points which have proven to eliminate any incidence of re-work or erroneous installation in numerous prior rerack projects. Inspection of lifting equipment will be performed per NUREG-0612.

Inadequate procedures: Comed plans to implement a multitude of procedures to cover

- ihe entire rack installation, such as mobihzation, rack handling, upending, lifting, installation, verticality, alignment, dummy gage testing, site safety, and ALARA compliance.

The operating procedures planned for the Byron and Braidwood rack installations are the successors of procedures that were implemented successfully for other projects.

In addition to the above, a complete inspection and preventive maintenance program of all the cranes and lifting equipment used in the project are planned prior to the start of rack installation.

Safe load paths will be developed as required by NUREG-0612.

Table 3-6 provides a synopsis of the requirements delineated in NUREG-0612 and our intended

.cmpliance.

in sum. mary, the measures implemented in the Byron and Braidwood rack installations are similar to the those utilizw m all recun reracks in the U.S..

i

- i SHADED TEXT CONTAINS PROPRIETARY INFORMATION l Holtec Intemational 3-6 Report HI-982083 !

l

3.6 References for Section 3

[3.2.1] " Nuclear Engineering International," July 1997 issue, pp 20-23.

[3.3.1] " Spent Fuel Storage Module Corrosion Report," Brooks & Perkins Report 554, June 1,1977.

[3.3.2] "Snitability of Brooks & Perkins Spent Fuel Storage Module for Use in PWR Sturage Pools," Brooks & Perkins Report 578, July 7,1978.

[3.3.3] "Boral Neutron Absorbing / Shielding Material - Product Performance Report,"

Brooks & Perkins Report 624, July 20,1982.

i l

l l

h l

SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 3-7 Report HI-982083

Table 3-1 BORAL EXPERIENCE LIST- PWR PLANTS Plant Utility Docket No. Mfg. Year Maine Yankee Maine Yankee Atomic Power 50-309 1977 Donald C. Cook Indiana & Michigan Electric 50-315/316 1979 Sequoyah1,2 Tennessee Valley Authority 50-327/328 1979 Salem 1,2 Public Service Electric & Gas 50-272/311 1980 Zion 1,2 Cammonwealth Edison Co. 50-295/304 1980 .

Bellefonte 1. 2 Tennessee Valley Authority 50-438/439 1981 Yankee Rowe Yankee Atomic Power 50-29 1964/1983 Indian Point 3 NY Power Authority 50-286 1987 Byron 1,2 Commonwealth Edison Co. 50-454/455 1988 Braidwood 1,2 Commonwealth Ediscn Co. 50-456/457 1988 Yankee Rowe Yankee Atomic Power 50-29 1988 Three Mile Island I' GPU Nuclear 50-289 1990 Sequoyah (rerack) Tennessee Valley Authority 50-327 1992 Donald C. Cook (rerack) American Electric Power 50-315/316 1992 Beaver Valley Unit i Duquesne Light Company 50-334 1993 Fort Calhoun Omaha Public Power District 50-285 1993 Zion ! & 2 (rerack) Commonwealth Edison Co. 50-295/304 1993 Salem Units 1 & 2 Public Gas and Electric Company 50-272/311 1995 (rerack)

Haddim Neck Connecticut Yankee Atomic Power 50-213 1996 Company Wate: ford Unit 3 Entergy Operations,Inc. 50-382 1997 Callaway Union Electric Company 50-483 1997 Gosgen Kernkraftwerk Gosgen-Daniken AG -- 1984 (Switzerland)

SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 3-8 Report HI-982083 j l

t i

i

Table 3-1 BORAL EXPERIENCE LIST- PWR PLANTS Plant Utility Docket No. Mfg. Year Koeberg 1,2 ESCOM (South Africa) -- 1985 Beznau 1,2 Nordostschweizerische Kraftwerke AG -- 1985 (Switzerland) 12 Various Plants Electricite de France (France) - 1986 Ulchin Unit i Korea Electric Power Company (Korea) -- 1995 Ulchin Unit 2 Korea Electric Power Company (Korea) -- 1996 Kori-4 Korea Electric Power Company (Korea) -- 1996 Yonggwang 1,2 - Korea Electric Power Company (Korea) -- 1996 Sizewell B '

Nuclear Electric, plc (United Kingdom) -- 1997 Angra 1 Furnas Centrais-Electricas S A (Brazil) -- 1997 l

1 1

l SHADED TEXT CONTAINS PROPRIETARY INFORMATION l Holtec International 3-9 Report HI-982083 I

l

. l

1 Table 3-2 BORAL EXPERIENCE LIST - BWR PLANTS Plant Utility Docket No, Mfg. Year Cooper Nebraska Public Power 50-298 1979 l J. A. FitzPatrick NY Power Authority 50-333 1978 Duane Arnold Iowa Electric Light & Power 50-331 1979 Browns Ferry 1,2,3 Tennessee Valley Authority 50-259/260/296 '1980 Brunswick 1,2 Carolina Power & Light 50-324/325 1981 Clinton Illinois Power 50-461/462 1981 Dresden 2,3 Commonwealth Edison Company 50-237/249 1981 E.1. Hatch 1,2 Georgia Power 50-321/366 1981 Hope Creek Public Service Electric & Gas 50-354/355 1985 Humboldt Bay Pacific Gas & Electric Company 50-133 1985 Lacrosse Dairyland Power 50-409 1976 Limerick 1,2 Philadelphia Electric Company 50-352/353 1980 Monticello Northern States Power 50-263 1978 Peachbottom 2.3 Philadelphia Electric 50-277/278 1980 Perry 1,2 Cleveland Electric Illuminating 50-440/441 1979 Pilgrim Boston Edison Company 50-293 1978 50-387,388 Susquehanna 1,2 Pennsylvania Power & Light 1979 f Vermont Yankee Vermont Yankee Atomic Power 50-271 1978/1986 Hope Creek Public Service Electric & Gas 50-354/355 1989 ;

Sheaton Harris Pool B Carolina P'. ver & Light 50-401 1991 I Duane Arnold Iowa Electtw Light & Power 50-331 1993 Pilgrim Boston Edison Company 50-293 1993 LaSalle 1 Commonwealth Edison Company 50-373 1992 l l

Millstone Unit i Northeast Utilities 50-245 1989 l l

SHADED TEXT CONTAINS PROPRIETARY INFORMATION

.Holtec International 3-10 Report HI-982083 l 1

1 1

I

. Table 3-2

BORAL EXPERIENCE LIST- BWR PLANTS Plant Utility Docket No. Mfg. Year James A.FitzPatrick - NY Power Authority 50-333 1990 Hope Creek ' Public Service Electric & Gas Company 50-354 1991 Duane Arnold Energy Iowa Electric Power Company 50-331 1994,.

Center Limerick Units 1,2 PECO Energy - 50-352/50-353 1994 Shearon Harris Pool'B' Carolina Power & Light Company - 50-401 1996 Nine Mile Point Unit i Niagara Mohawk Power Corporation 50-220 1997 J.A. FitzPatrick NY Power Authority 50-333' 1997 (racks added)

Chinshan 1,2 - Taiwan Power Company (Taiwan) -. 1986

! Kuosheng 1,2 Taiwan Power Company (Taiwan) -- 1991 Lag:una Verde 1.2 ' Comision Federal de Electricidad -- 1991 f <

(Mexico)

J ll l

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, .1 1

I SHADED TEXT CONTAINS PROPRIETARY INFOR$ATION Holtec International 3 Report HI-982083

r

- Table 3-3 1100 ALLdY ALUMINUM PHYSICAL CHARACTERISTICS Density 0.098 lb/in' !

2.713 g/cm'

. Melting Range 1190 F- 1215 F 643* - 657 C

~

2 Theimal Conductivity (77 F) 128 BTU /hr/ft /F/ft 2

0.53 cal /sec/cm / C/cm 4

Coefficient of Thermal Expansion 13.1 x 10 in/in- F 4

(68 F - 212 F) 23.6 x 10 cm/cm *C Specific Heat (221 F) 0.22 BTU /lb/ F 0.23 cal /g/ C Modulus of Elasticity 10 x 10' psi Tensile Strength (75*F) 13,000 psi (annealed) 18,000 psi (as rolled)

Yield Strength (75*F) 5,000 psi (annealed) 17,000 psi (as rolled)

Elongation (75 F) 35-45% (annealed) 9-20% (as rolled)

Hardness (Brinell) 23 (annealed) 32 (as rolled)

Annealing Temperature 650 F 343 C SHADED TEXT CONTAINS PROPRIETARY INT C dMATION Holtec International 3-12 Report HI-982083

Table 3-4 CHEMICALCOMPOSITION- ALUMINUM (l100 ALLOY)

Aluminum 99.00% min.

Silicone and Iron 1.00% max.

Copper 0.05-0.20% max.

Manganese 0.05% max.

Zinc 0.10% max.

Other 0.15% max. !

l i

i l

SHADED TEXT CONTAINS PROPRIETARY INFORMATION lloltec International 3-13 Report HI-982083

r.

i i

Table 3-5 CHEMICAL COMPOSITION AND PHYSICAL PROPERTIES OF BORON CARBIDE 1

CHEMICAL COMPOSITION (WEIGHT PERCENT) i Total boron 70.0 min. i B' isotopic cs . rnt in natural boron 18.0 Boric oxide t. 3.0 max. .

Iron t 2.0 max.

l Total boron plus total carbon 94.0 min. j PHYSICAL PROPERTIES Chemical formula BC 4

Boron content (weight percent) 78.28 %

Carbon content (weig.ht percent) 21.72 % l l

Crystal structure rhombohedral

' Density 0.0907 lb/in' 3

2.51 g/cm Melting Point 4442 F 2450 C Boiling Point 6332*F 3500 C i

l Impurity that forms due to the chemical manufacturing process.

SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 3-14 Report HI-982083 u 1

Table 3-6 HEAVY LOAD HANDLING COMPLIANCE MATRIX (NLREG-0612)

Criterion . Compliance

1. Are safe load paths defined for the movement of heavy loads to Yes minimize the potential of impact if dropped on irradiated fuel and safe shutdown equipment?

2. Will procedures be developed to cover; identification of required Yes equipment, inspection, and acceptance criteria required before movement of load, steps and proper sequence for handling the load, defining the safe load paths, and special precautions?

3. Will crane operators be trained and qualified? Yes

4. Will special lifting devices meet the guidelines of ANSI 14.6-1978? Yes

3. Will non-customer lifting devices be installed and used in Yes accordance with ANSI D30.9-1971?

6. Will the cranes be inspected ar.J tested prior to use in rerack? Yes

7. Does the crane meet the intent of ANSI B30.2-1976 and CMMA- Yes 70?

I I

SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 3-15 Report HI-982083 I

h 4.0 - CRITICALITY SAFETY EVALUATION 4.1 - Design Bases This report documents the criticality safety evaluation for the storage of PWR spent nuclear fuel in Holtec Region I & II style high-density spent fuel storage racks at the Byron and Braidwood nuclear power plants operated by Commonwealth Edison (Comed). The ' objective of this analysis is to ensure that the effective neutron multiplication factor (kerr) is less than or equal to 0.95 with the storage racks fully loaded with fuel of the highest anticipated reactivity and the pool flooded with unborated water at a temperature corresponding to the highest reactivity. The maximum calculated reactivity includes a margin for uncertainty in reactivity calculctions including manufacturing tolerances and is shown to be less than 0.95 with a 95% probability at a 95%, confidence level [4.1.1]. Reactivity effects of abnormal and accident conditions have also been evaluated to assure that under all credible abnormal and accident conditions, the reactivity will not exceed the regulatory limit of 0.95.

4 Applicable codes, standards, and regulations or pertinent sections thereof, include the following:

Code of Federal Regtdations, Title 10, Part 50, Appendix A, General Design Criterion 62, " Prevention of Criticality in Fuel Storage and Handling."

i e USNRC Standard Review Plan, NUREG-0800, Section 9.1.2, Spent Fuel Storage, Rev. 3

]

- July 1981. l

.- USNRC letter of April 14,1978, to all Power Reactor Licensees - OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications, including modification letter dated January 18,1979.

L. Kopp," Guidance on the Regulatory Requirements for Criticality Analysis of Fuel ,

Storage at Light-Water Reactor Power Plants," NRC Memorandum from L. Kopp to T.

Collins, August 19,1998.

1

. . . USNRC Regulatory Guide 1,13, Spent Fuel Storage Facility Design Basis, Rev. 2 )

(proposed), December 1981. l l

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e ANSI'ANS-8.17-1984, Criticality Safety Criteria for the Handling, Storage and Transportation of LWR Fuel Outside Reactors.

USNRC guidelines [4.1.2] and the applicable ANSI standaros specify that the maximum effective multiplication factor, ke , including bias, uncertainties, and calculational statistics, shall be less than or equal to 0.95, with 95% probability at the 95% confidence level.

To assure the true reactivity will always be less than the calculated reactivity, the following conservative design criteria and assumptions were employed:

. Moderator is unborated water at a temperature that results in the highest reactivity (4'C, corresponding to the maximum possible moderator density,1.000 g/cc).

. The racks were assumed to be fully lo~aded with the most reactive fuel authorized to be stored in the racks without any control rods or bumable poison, such as Integral Fuel Burnable Absorber (IFBA) rods.

No' soluble poison (boron) is assumed to be present in the pool water under normal operating conditions.

e Neutron absorption in minor structural members is neglected, i.e., spacer grids are replaced by water, e The effective multiplication factor of an infinite radial array of fuel assemblies was used in the analyses, except for the assessment of peripheral effects and certain abnormal / accident h 1

conditions where neutron leakage is inherent.

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In-core depletion calculations assume conservative operating conditions, highest fuel and moderator temperature, and an allowance for the soluble boron concentrations during in-core 1* operations.

The spent fuel storage racks are designed to accommodate the fuel assembly types listed in Table 235 4.1.1 with a maximum nominal initial enrichment of 5.0 wt% U. 1 Two separate storage regions are provided in the Byron and Braidwood spent fuel pools. The independent acceptance criteria for storage in each of the regions are as follows:

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I s

=> Region Iis designed to accommodate fresh unburned fuel assemblies with a maximum nominal enrichment of 5.0 wt% 235U.'

=6 Region 11 is designed to accommodate fuel assemblies with a maximum nominalinitial enrichment of 5.0.wt% 2nU which have accumulated a minimum burnup of 40.0 GWD/MTU )

or fuel ofinitial enrichment and burnup combinations within the acceptable domain depicted in Figure 4.1.1'. j The water in the spent fuel storage pool normally contains soluble boron, which would resulf in a large sub<riticality margin under actual operating conditions. However, the NRC guidelines, based upon the accident condition in which all soluble poison is assumed to have been lost, specify that the limiting ken of 0.95 for normal storage be evaluated for the accident condition that assumes the loss of soluble boron. The double contingency principle of ANSI N-16.1-1975 and of the April 1978 NRC letter allows credit for soluble boron under other abnormal or accident conditions, since only a single independent accident need be considered at one time.

Consequences of abnormal and accident conditions have been evaluated, where " abnormal" refers ,

to conditions which may reasonably be expected to occur during the lifetime of the plant and

" accident" refers ta conditions which are not expected to occur but nevertheless must be protected against.

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m 14.2 Summarv of Criticality Analyses 4.2.1 Normal Operating Conditions The criticality analyses for each of the two separate regions of the spent fuel storage pool are summarized in Tables 4.2.1 and 4.2.2, for the design basis storage conditions. For the fuel acceptance enteria defined in the previous section, the maximum kerr values are shown to be less than 0.945 (95% probability at the 95% confidence level)in each of the regions.

4.2.1.1 Region I

, Calculations have been performed to qualify the Region I racks for storage of fresh unbumed fuel assemblies with a maximum nominal enrichment of 5.0 wt% 235 U. The criticality analyses for Region I of the spent fuel storage pool are summarized in Table 4.2.1, and demonstrate that for the defined acceptance criteria, the maximum ken is less than 0.945.

4.2.1.2 Region II Calculations have been performed to qualify the Region II racks for storage of spent fuel 235 assemblies with'a maximum nominal initial enrichment of 5.0 wt% U which have accumulated a minimum bumup of 40.0 GWD/MTU or fuel of initial enrichment and burnup combinations within the acceptable domain depicted in Figure 4.1.1. The criticality analyses for Region II of

-~ the spent fueI storage pool are summarized in Table 4.2.2, and demonstrate that for the defined acceptance criteria, the maximum ken is less than 0.940.

4

' The calculated maximum reactivity in Region II includes the reactivity effect of the axial distdbution in bumup and provides an additional margin of uncertainty for the depletion calculations. The data points shown in Figure 4.1.1 are tabulated in Table 4.2.3. For

. convenience, the minimum (limiting) burnup data may be described as a function of the nominal 2

initial enrichment, E, in wt% "U by a bounding polynomial expression as follows:

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y B = 0.1861xE'- 2.3287xE* + 21.377xE - 31.248, where B is the minimum burnup in GWD/MTU and E is the enrichment in wt% 2"U (for initial enrichments from 2.0 to 5.0 wt% 2"U). Fuel assemblies with enrichments less than 2.0 wt% 2"U will conservatively.be required to meet the burnup requirements of 2.0 wt%2 "U assemblies.as shown in Fig 4.1.1. Alternatively, since the data are nearly linear, linear interpolation between the points listed in Table 4.2.3 is acceptable.

The criteria identified above for acceptable storage in each of the regions will be implemented by.

appropriate administrative procedures.

4.2.2 Abnormal and Accident Conditions Although credit for the soluble poison normally present in the spent fuel pool water is permitted under abnormal or accident conditions, most abnormal or accident conditions will not result in exceeding the limiting reactivity even in the absence of soluble poison. The effects on reactivity ;

1

- of credible abnormal and accident conditions are discussed in Section 4.7 and summarized in i Table 4.2.4. Strict administrative procedures to assure the presence of soluble poison will preclude the possibility of the simultaneous occurrence of the two independent accident conditions.-

The abnormal location of a fresh fuel assembly has the potential for exceeding the limiting reactivity, should there be a concurrent and independent accident condition resulting in the loss of

' all soluble poison. Assuring the presence of soluble poison during fuel handling operations will

~

preclude the possibility of the simultaneous occurrence of the two independent accident conditions. The largest reactivity increase would occur if a fresh fuel assembly of the highest permissible enrichment (5.0 wt% 2"U) were to be inadvertently loaded into a Region II storage cell with the remainder of the rack fully loaded with fuel of the highest permissible reactivity.

Under this accident condition, credit for the presence of soluble poison is permitted by the NRC SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 4-5 Report HI-982083

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t guidelines , Calculations indicate that a minimum soluble boron concentration of 220 ppm, which is less than the concentration required by the Technical Specifications during fuel handling operations, is more than adequate to assure that keff remains below 0.945. l 1

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I t Double contingency principle of ANSI N16.1-1975, as specified in the April 14,1978 NRC letter (Section 1.2) and implied in the proposed revision to Reg._ Guide 1.13 (Section 1.4, Appendix A).

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4.3 Reference Fuel Storage Cells 4.3.1 -Reference Fuel Assembly.

The spent fuel storage racks are designed to accommodate Westinghouse 17x17 OFA,17x17 Vantage 5, and 17x17 Vantage + fuel as.semblies. The design specifications for these fuel assemblies,'which were used for this analysis, are given in Table 4.1.1. In terms of dimensions that are important to reactivity, all of the assembly types listed in Table 4.1.1 are identical.

Therefore, calculations to determine the most reactive assembly type are not necessary. Herein, the reference fuel assembly is referred to as a Westinghouse 17x17 assembly, with dimensions listed in Table 4.1.1.

~4.3.2 Region I Fuel Storage Cells Figure 4.3.i shows the calculational model of the nominal Region I spent fuel storage cell containing the Westinghouse 17x17 fuel assembly. The Region I storage cells are composed of stainless steel boxes separated by a gap'with fixed neutron absorber panels, Boral, centered on each side in a 1 2 inch channel.The ! : . ' thick steel walls define the storage cells, which have a lii M E , inch nominalinside dimension. A ; E inch stainless steel sheath supports the Boral panel and defines the boundaryof the flux-trap water-gap used to augment reactivity control. The cells are located on a lattice spacing of dii . m. ,~. inch in one l

direction and M inch in the other direction. Stainless steel channele connect the l storage cells in a rigid structure and define the flux-trap between the Boral ptnels, which are 1 1

IG$WdINE inch in one direction and 3 M ' inch in the other direction. The Boral absorber has a thickness of ' M.R 3 inch and a nominal B-10 areal density of M#

E93bM The Boral absorber panels are FPF9 inches in width and

$6M?Nd inches in length. Boral panels are installed on all exterior walls facing other racks, as well as non-fueled regions, i.e., the pool walls. The minimum gap between neighboring Region I style racks and between Region I and Region II style racks is 1.75 inches.

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4.3.3. Region II Fuel Storage Cells Figure 4.3.2'shows the calculational model of the nominal Region II spent fuel storage cell containing the Westinghouse 17x17 fuel assembly. The Region II storage cells are composed of stainless steel walls with a single fixed neutron absorber panel, Boral, (attached by a 1 -. M i 1

stainless steel sheathing) centered on'each side in a Ench channel. Stainless steel boxes are i arranged in an alternating pattern such that the connection of the box corners form storage cells between those of the stainless steel boxes.' These cells are located on a lattice spacing of l M444..i inch. The .

thick steel walls define a storage cell, which has a ,

4:.. A.E, . - inch nominal inside dimension. The Boral absorber has a thickness of l E.11.M ' ? inch and a nominal B-10 areal density of 1 - .

a,- .. I The Boral absorber panels are ... . : inches in width and

. . inches in length.

. Boral panels are installed on one side of neighboring Region II racks. Boral panels are not installed on exterior walls facing non-fueled regions, i.e., the pool walls. The minimum gap between neighboring Region II style racks is 0.875 inches, while the minimum gap between

' Region I and Region Il style racks is 1.75 inches.

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h 4.4 Analytical Methodolony 4.4.1 ' Reference Design Calculations The principal methods for the criticality analyses of the high density storage racks include the following codesi (1) MCNP4a [4.4.1], (2) KENO 5a [4.4.2], and (3) CASMO-4 [4.4.5-4.4.7].

MCNP4a is a~ continuous energy three-dimensional Monte Carlo code developed at the Los -

Alamos National Laboratory. KENO 5a is a three-dimensional multigroup Monte Carlo code

< . developed at the Oak Ridge National Laboratory as part of the SCALE 4.3 package [4.4.3]. The

, KENO 5a calculations used the 238-group SCALE cross-section library and NITAWL [4.4.4] for

U resonance shielding effects (Nordheim integral treatment). - Benchmark calculations, presented in Appendix 4A, indicate a bias of 0.0009 with an uncertainty ofi 0.0011 for MCNP4a and 0.0030 i 0.0012 for KENO 5a, both evaluated with the 95 % probability at the 95%

confidence level [4.1.1].

Fuel depletion analyses during core operation were performed with CASMO-4, a two-

' dimensional multigroup transport theory code based on capture probabilities [4.4.5 4.4.7]. )

Restarting the CASMO-4 calculations in the storage rack geometry yields the two-dimensional

! infinite multiplication factor (k ) for the storage rack. Parallel calculations with CASMO-4 for l the storage rack at various enrichments enable a reactivity equivalent enrichment (fresh fuel) to be t'etermined that provides the same reactivity in the rack as the depleted fuel. CASMO-4 was also used to determine the reactivity uncertainties (differential calculations) of manufacturing talerances and the reactivity effects of variations in the water temperature and density.

In the geometric models used for the calculations, each fuel rod and its cladding were described explicitly and reflecting boundary conditions were ut d in the radial direction, which has the effect of creating an infinite radial array of storage cells. Monte Carlo calculations inherently include a statistical uncertainty due to the random nature of neutron tracking. To minimize the statistical uncertain:y of the MCNP4a and KEN 05a calculated reactivities and to assure j convergence, a minimum of 1 million neutron histories were accumulated in each calculation.

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'4.4.2 Fuel Bt;mup Calculations and Uncertainties J

l CASMO-4 was used for burnup calculations in the hot operating condition. To the e: 'ent .

i possible, CASMO-4 has been benchmarked [4.4.6, 4.4.7] against cold, clean, critical- !

experiments (including plutonium-bearing fuel) and Monte Carlo calculations. l In the CASMO-4 geometric models, epch fuel rod and its cladding were described explicitly and -

reflective boundary conditions were used between storage cells. These boundary conditions have the effect of creating an infinite array of storage cells. q

= Conservative assumptions of moderator and fuel temperatures and the average operating soluble l boron concentrations were used to assure the highest plutonium production and hence -

conservatively high values of reactivity as a function of burnup. Since critical experiment data with spent fuel is not available for determning i the uncertainty in depletion calculations, an t

allowance for uncertainty in reactivity was assigned based upon other considerations [4.1.2].

Assuming the uncertainty in depletion calculations is less than 5% of the total reactivity decrement, a bumup dependent uncertainty in reactivity for bumup calculations was assigned.

Thus, the burnup uncertainty varies (increases) with burnup. This allowance for bumup -

uncertainty was included in determination of the acceptable bumup versus enrichment combinations.

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t The majority of the uncertaintv in depletion calculations derives from uncertainties in fuel and moderator

. temperatures and the effect of reactivity control methods (e.g., soluble boron). For depletion calculations, I bounding values of these operating parameters were assumed to assure conservative results in the analyses.

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e 4.4.31 Effect of Axial Burnup Distribution' Initially, fuel loaded into the reactor will burn with a slightly skewed cosine power distribution.

As burnup progresses, the burnup distribution will tend to' flatten, becoming more highly burned in the central regions than in the upper and lower regions. At high bu'rnup, the more reactive

- fuel near the ends of the fuel assembly (less than average burnup) occurs in regions of high neutron leakage. Consequently, it is expected that over most of the burnup history, fuel assemblies with distributed burnups will exhibit a slightly lower reactivity than that calculated for the uniform average burnup. As burnup progresses, the distribution, to some extent, tends to be self-regulating as controlled by the axial power distribution, precluding the existence of large regions of significantly reduced burnup.

' Among others, Turner [4.4.8] has provided generic analytic results of the axial burnup effect based upon calculated and measured axial burnup distributions. These analyses confhm the minor and generally negative reactivity effect of the axially distributed burnups at values less than about 30 GWD/MTU with small positive reactivity effects at higher burnup values. For the present criticality analyses, the reference calculations utilized actual axial burnup distributions !

a from Byron and Braidwood plant data. Burnup-equivalent enrichments were determined with CASMO-4 for each of 24 axial zones and used in three-dimensional Monte Carlo calculations.

Results of t!'ese calculations, therefore, inhere.aly include the effect of the axial distribution in ;

burnup. Comparison of these results to results of calculations with uniform axial burnup allows ]

the reactivity effect of the axial burnup distribution to be quantified. This reactivity effect is

' included, where applicable, in the calculation of the maximum km values.

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4.4.4 Lonn-Term Channes in Reactivity i

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l At reactor shutdown, the reactivity of the fuel initially decreases due to the growth of Xe-135.

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Subsequently, the Xenon decays and the reactivity increases to a maximum at several hundred {

hours when the Xenon is gone. Therefore, for conservatism, the Xe is set to zero in the calculations to assure maximum reactivity. During the next 50 years, the reactivity continuously l l

decreases due primarily to Pu-241 decay and Am-241 growth. No credit is taken for this long- I 1

, term decrease in reactivity other than to indicate additional and increasing suberiticahty margm. i l

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4.5 Region I Criticality Analyses an'd Tolerances d5.1; Nominal Design Case f !

For the nominal storage cell design in Regicu I, the criticality safety analyses are summarized in Table 4.2.1_. - These data confirm that the maximum reactivity in Region I remains conservatively -

-less than the regulatory limit (k,, s 0.95). An jridependent calculation with the KEN 05a code provides confirmation of the validity of the reference MCNP4a calculations.

4.5.2 - ]Jncertainties Due to Burnuo For storage in Region I, consideration of fuel burnup is not necessary, and thus, burnup related

- uncertainties are not applicable.

4.5.3 Uncertainties Due to Tolerances The reactivity effects of manufacturing tolerarices are tabulated, along with the tolerances, in Table 4.5.1.' The mdividual tolerances were calculated for the design basis fresh unburned fuel assembly.

4.5.4 Eccentric Fuel Positionine The fuel assembly is assumed to be normally located in the center of the storage rack cell.

However, calculations were also made with the fuel assemblies assumed to be in the corner of the l

I storage rack cell (four-assembly cluster at closest approach). These calculations indicated that the reactivity effect is small and negative. Therefore, the reference case in which the fuel assemblies are centered is bounding and no uncertainty for eccentricity is necessary.

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4.5.5 ' Water-Gao Snacine Between Racks 1The Boral panels installed on the external surfaces ofinterfacing Region I and Region II racks and the minimum water-gap spacing between racks, which is 1.75 inches between neighboring Region I style racks and also 1.75 inches between Region I and Region II style racks, constitutes a neutron flux-trap between the storage cells of facing racks. The racks are constructed with the

~ base plates extending beyond the edge of the cells which assures that the minimum spacing between storage racks is maintained under all credible conditions. This water-gap flux-trap is larger than those between Region I cells (1.337 inches in one direction and 1.651 inches in the other direction), and thus, will act to reduce the reactivity below the cited maximum. Therefore, -

boundary constraints between rack moc ales are not necessary or required.

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.Ifoltec International 4-14' Report HI-982083

E j 4.6 Region II Criticality Analyses and Tolerances -

4.6.1 Nominal Design Case For the nominal storage cell design in Region II, the criticality safety analyses are summarized in Table 4.2.2. These data confirm that the maximum reactivit y in Region II remains conservatively less than the regulatory limit (bn s 0.95). An independent calculation with the KEN 05a code ~

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provides confirmation of the validity of the reference MCNP4a calculations.

4.6.2 Uncenainties Due to Bumuo

' CASMO-4 was used for the depletion analysis and the restart option was used to analytically _

transfer the spent fuel into the storage rack configuration at a reference temperature of 4 C (corresponding to the highest reactivity, see Section 4.7.1). Calculations were also made for fuel of several different initial enrichments and interpolated to define the burnup-dependent equivalent enrichmentst , at each burnup. MCNP4a calculations were then made for the equivalent enrichment to establish the limiting bn value, which includes all applicable uncertainties and the effect of the axial burnup distribution. These calculations were used to define the boundary of the acceptable domain shown in Figure 4.1.1.

4.6.3 Ulncertainties Due to Tolerances The reactivity effects of manufacturing tolerances are tabulated, along with the tolerances, in Table 4 ', . All of the individual reactivity allowances were calculated for th reference fuel assembly at zero burnup and at burnups enveloping the criteria for storage (i.e.,35,40, and 45 GWD/MTU). The largest statistical combination of uncertainties from either the fresh or bumed condition was conservatively used.

t The (reactivity) equivalent enrichment is the fresh un-burned fuel enrichment that yields the same reactivity as the depleted fuel. both evaluated in the storage rack configuration. The equivalent enrichment may then be used in three; dimensional MCNP4a or KENO 5a calculations.

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' 4.6.4 Eccentric Fuel Positioning The fuel assembly is assumed to be normally located in the center of the storage rack cell.

However, calculatiens were also made with the fuel assemblies assumed to be in the corner of the storage. rack cell (four-assembly cluster at closest approach). These calculations indicated that the reactivity effect is small and negative. Therefore, the reference case in which the fuel assemblies are centered is bounding and no uncertainty for eccentricity is necessary.

'4 6 5 Water-Gap Spacing Between Racks

. The minimum water-gap between racks, which is 0.875 inches between neighboring Region II style racks and 1.75 inches between Region I and Region II style racks, constitutes a neutron flux-trap for the storage cells of facing racks. The racks are constructed with the base plates extending beyond the edge of the cells which assures that the minimum spacing between storage racks is maintained under all credible conditions. Region II style racks do not contain intcmal water gaps, and thus, this water-gap flux-trap will act to reduce the reactivity below the cited maximum.

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r L 4.7 - ' Abnormal and Accident Conditions.

4.7.1 Temocrature and Water Density Effects -

1 The moderator temperatum coefficient of reactivity in both Region I and Region II is negative.

- Therefore, a moderator temperature of 4'C (39 F) was assumed for the reference calculations, 3 which assures that the true reactivity will always be lower over the expected range of water temperatures. Temperature effects on reactivity have been evaluated (CASMO-4) and the results are shown in Table 4.7.1. In addition, the introduction of voids in the water intemal to the storage cell (to simulate boiling) decreased reactivity, as shown in Table 4.7.1.

' .With soluble boron present, the temperature coefficients of reactivity would differ from those listed in Table 4.7.1. However, the reactivities would also be substantially lower at all temperatures with soluble boron present. Tne data in Table 4.7.1 are pertinent to the higher-reactivity unborated case.

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.4.7.2 Lateral Rack Movement Lateral motion of the storage racks under seismic conditions could potentially alter the spacing between racks. In Region I, the minimum water gap between racks (1.75 inches, as limited by the base plate extensions)is larger than the corresponding design water-gap spacing (1.337 i

inches in one direction and 1.651 inches in the other direction) internal to the racks. Similarly, the. minimum water gap in the interface between Region I and Region II style racks (1.75 inches,

.as limited by the base plate extensions) is also larger than the water gap internal to the Region I racks and Boral panels are installed on the external surfaces of interfacing Region I and Region II racks. Consequently, there will be no positive effect or: reactivity as a result of lateral rack movement.

~ Region II storage cells do not use a flux-trap, and thus, the calculated maximum reactivity does not rely on spacing between racks. Nevertheless, the minimum water gap between Region II SHADED TEXT CONTAIN3 PROPRIETARY INFORMATION

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racks (0.875 inches, as limited by the base plate extensions) and the Boral panels, which are installed on one side of neighboring Region H racks, assure that the reactivity is always less than the design limitation. Furthermore, soluble poison would assure that a reactivity less than the design limitation is maintained under seismic conditions. Consequently, there will be no positive effect on reactivity as a result of lateral rack movement.

4.73 Abnormailocation of a Fuel Assembly i

l The misplacement of a fresh unbumed fuel assembly could, in the absence of soluble poison, result j l

- in exceeding the regulatory limit (ken 5 0.95). This could occur if a fresh fuel assembly of the l 2

highest permissible enrichment (5.0 wt% nU) were to be inadvertently loaded into a Region D storage cell. Calculations confirmed that the highest reactivity, including uncertainties, for the worst case postulated accident condition (fresh fuel assembly in Region H) would exceed the limit on reactivity in the absence of soluble boron. Soluble boron in the spent fuel pool water, for which credit is permitted under these accident conditions, would assure that the reactivity is maintained substantially less than the design limitation. Calculations indicate d.at a soluble boron concentration of 220 ppm is adequate to assure that the maximum ken does not exceed 0.945.

In addition, the mislocation of a fresh unburr ed fuel assembly could,in the absence of soluble i 1

poison, result in exceeding the regulatory limit (keno f 0.95). This could possibly occur if a fresh fuel assembly of the highest permissible enrichment (5.0 wt%) were to be accidentally mislocated I

outside of a Region H storage rack adjacent to other fuel assemblies. The worst case would be an assembly mislocated in a comer of the cut-out areas for the fuel elevator and tool bracket. The rack exterior walls in these cut-out areas have Boral panel's attached. Therefore, this condition is bounded by the abnormal location of fuel assembly in a Region H storage cell, which is controlled by a minimum soluble boron concentration of 220 ppm.

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t 4.7,4' Dropped Fuel ssembly

. For the case in which a fuel assembly is assumed to be dropped on top of a rack, the fuel assembly '

L will come to rest horizontally on top of the rack with a minimum separation distance from the active fuel region of mom than 12 inches, which is sufficient to pmclude neutron coupling (i.e., an effectively infinite separation). ' Maximum expected deformation under seismic or accident conditions will not reduce the minimum spacing to less than 12 inches. Consequently, the horizontal fuel assembly drop accident will not result in a significant increase in reactivity.

Furthermore, the soluble boron in the spent fuel pool water assures that the tme reactivity is always less than the limiting value for this dropped fuel accident.

It is also possible to vertically drop an assembly into a location occupied by another assetably.

Such a vertical impact, would, at most cause a small compression of the stored assembly,

. redu'cing the water-to-fuel ratio and thereby reducing reactivity. In addition the distance between the active fuel regions of both assemblies will be more than sufficient to ensure no neutron interaction between the two assemblies.

Strt:ctural analysis has shown that dropping an assembly into an unoccupied cell could result in a

. localized deformation of the baseplate of the rack. The resultant effect would be the lowering of a single fuel assembly by the amount of the deformation. This could potentially result in the active fuel height of that assembly no longer being completely covered by the Boral. The immediate eight surrounding fuel cells could also be affected. However, the amount of deformation for these cells would be considerably less. Stmetural analysis has shown that the amount of localized

. deformation will not exceed two inches. The reactivity consequence of this situation was calculated and found to be statistically insignificant. For simplicity in modeling, the calculation

' conservatively assumed an infinite array of assemblies in this damaged condition, and demonstrated the reactivity effect to be negligible. Since this is a localized event (nine storage cells at most) the actual reactivity effect will be even less than the calculated value. Furthermore, the soluble boron in the pool water assums that the true' reactivity is always less than the limiting value for this dropped fuel accident.

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'4.8 ' References for Section 4

[4.1.1] ' M. G. N' atrella, Experimental Statistics, National' Bureau of Standards Handbook 91 August 1%3.-

1

_ [4.1.2] _' L.I. Kopp," Guidance on the Regulatory Requirements for Criticality Analysis of Fuel Storage at Light-Water Reactor Power Plants," NRC Memorandum from L

' Kopp to T, Collins, August 19,1998.'

[4.4.1] J.F. Briesmeister, Editor, "MCNP - A General Monte Carlo N-Particle Transport Code, Version 4A," LA-12625, Los Alamos National Laboratory (1993).

[4.4.2] _ LM. Petrie and N.F. Landers, " KENO Va - An Improved Monte Carlo Criticality Program with Supergrouping," Volume 2, Section F11 from " SCALE: A Modular System for Performing Standardized Computer analysis for Licensing Evaluation" i NUREG/CR-0200, Rev. 4, January 1990.

[4.4.3] " SCALE 4.3: A Modube System for Performing Standardized Computer Analysis for Licensing Evaluation For Workstations and Personal Computers, Volume 0," CCC-545, ORNL-RSICC, Oak Ridge National Laboratory (1995).

[4.4.4] 'N.M. Greene, L.M. Petrie and R.M. Westfall, "NITAWL-II: Scale System Module for Performing Shielding and Working Library Production," Volume 1, Section F1 from

" SCALE: A Modular System for Performing Standardized Computer Analysis for i

Licensing Evaluation" NUREG/CR-0200, Rev. 4, January 1990. ]

i l

[4.4.5] M. Edenius, K. Ekberg, B.H. Forss6n, and D. Knott, "CASMO-4 A Fuel Assembly Burnup Program User's Manual," Studsvik/SOA-95/1, Studsvik of 1

6' erica,Inc. and Studsvik Core Analysis AB (proprietary).

[4.4.6] D. Knott,"CASMO-4 Benchmark Against Critical Experiments", SOA-94/13, Studsvik of America,Inc.,(proprietary).

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j (4.4.7] D. Knott,"CASMO-4 Benchmark Agamst MCNP," SOA-94/12, Studsvik of America,Inc., (proprietary). 4

[4.4.8] S.E. Turner, " Uncertainty Analysis - Burnup Distributions, presented at the DOE /SANDIA Technical Meeting on Fuel Burnup Credit, Special Session, ANS/ ENS Conference, Washington, D.C., November 2,1988.

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I Table 4.1.1 i

Fuel Assembly Specifications I i

1 Fuel Rod Data l l

Westinghouse Assembly type - OFA / Vantage 5 / Vantage +

Fuel pellet outside diameter, in. 0.3088 Cladding thickness, in. 0.0225 Cladding outside diameter, in. 0.360

(

Cladding material Zr Maximum stack density, g/cc 10.522 i i

Maximum enrichment, wt% *U 5.0010.05 i Fuel Assembly Data !

l Fuel rod a:Tay 17 x 17 l Number of fuel rods 264 Fuel rod pitch,in. 0.496 Number of control rod guide and 25 instrument thimbles Thimble outside diameter, in. 0.474 Thimble thickness, in. 0.016 Active fuel length,in. 144 l

l k

i 1

SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec Intemational 4-22 Report HI-982083

s 7,- l Table 4.2.1 Summary of the Criticality Safety Analyses for Region I Storage Arrangement- Unrestricted Design Basis Burnup at 5.0 wt% "U 0 Uncenainties Bias Uncertainty (95%/95%) i0.0011 Calculational Statistics * (95%/95%,2.0xo) i0.0016 Depletion Uncertainty N/A Fuel Eccerrricity _ negative j Manufacturing Tolerances . t 0.0093 Statistical Combination of Uncertaintiest i0.0095 i i

. Reference ken (MCNP4a) 0.9318 Total Uncertainty (above) 0.0095 Axial Bumup Distribution N/A Calculational Bias (see Appendix A). 0.0009 Maximum k,g 0.9422" Regulatory Limiting k,n 0.9500 t The value used for the MCNP4a (or KEN 05a) statistical uncertainty is 2.0 times the estimated standard deviation. ,

Each final k value calculated by MCNP4a (or KENO 5a)is the result of averaging a minimum of 200 cycle k values, and thus, is based on a minimum sample size of 200. The K multiplier, for a one-sided statistical tolerance with 95%

probability at the 95% confidence level, corresponding to a sample size of 200,is 1.84. However, for this analysis a value of 2.0 was assumed for the K multiplier, which is larger (more conservative) than the value corresponding to a sample size of 200.

l t Square root of the sum of the squares, j SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 4-23 Report HI-982083 l

1

)

Table 4.2.2 Summary of the Criticality Safety Analyses for Region II Design Basis Burnup at 5.0 wt% "U 40.0 GWD/MTU l I

l i

Uncenainties ]

Bias Uncertainty (95%/95%) i0.0011 f Calculational Statistics * (95%/95%,2.0xo) 1 0.0011 l

Depletion Uncenainty i0.0142 Fuel Eccentricity negative J Manufacturing Tolerances t 0.0054 l Statistical Combination of Uncertaintiest i 0.0153 ,

I 1

i 4

Reference ken (MCNP4a) 0.9129 Total Uncertainty (above) 0.0153 Axial Burnup Distribution -

0.0086 l Calculational Bias (see Appendix A) 0.0009 Maximum k,n 0.9377" l Regulatory Limiting k,n 0.9500 t The value used for the MCNP4a (or KENO 5a) statistical uncertainty is 2.0 times the estimated standard deviation.

Each final k value calculated by MCNP4a (or KENO 5a) is the result of averaging a minimum of 200 cycle k values, and thus, is based on a minimum sample size of 200 The K multiplier, for a one-sided statistical tolerance with 95%

probability c: the 95% confidence level, correspondi g to a sample size of 200, is 1.84. However, for this analysis a value of 2.0 was assumed for the K multiplier, which is larger (more conservative) than the value corresponding to a sample size of 200.

t Square root of the sum of the squares.

SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 4-24 Report HI-982083

Table 4.2.3 Bumup-Enrichment Limits in Region U

- Nominal Initial Fuel Enrichment Minimum Fuel Bumup 235 (wt% U) . (GWD/MTU) 2 3.085

'2.5 10.215 3 16.506 3.5 22.410 4 28.048 4.5 34.751 5 40.000 k

l 1

SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 4-25 Report HI-982083

Table 4.2.4 Reactivity Effects of Abnormal and Accident Conditions in Regicns I and II Abnormal / Accident Conditions Reactivity Effect Temperature Increase (above 4 C) Negative (Table 4.7.1)

Void (boiling) Negative (Table 4.7.1)

Assembly Drop Negligible Lateral Rack Movement Negligible Misplacement of a Fresh Fuel Assembly in Positive - controlled by a minimum of 220 ppm ,

Region II - soluble boron Mislocatian of a Fresh Fuel Assemblyin a Positive - controlled by 220 ppm soluble baron RegionIICut-Out Area 1

l 3

SHADED TEXT CONTAINS PTOPRIETARY INFORMATION Hollec Intemational 4-26' Report HI-982083

Table 4.5.1 Reac'.ivity Effects of Manufacturing Tolerances in Region I Tolerance Reactivity Effect, Ak Minimum Boralloading 6 nominal) 10.0020 Minimum Boral width l _

i nominal) i0.0007 Maximum box I.D. (M nominal) 10.0086 Maximum box wall thickness (ETJXGEW] nominal) 10.0006 Density tolerance M,10.522 g/cm nominal) 3 i0.0021 235 Enrichment ( ' i,5.00 wt% U nominal) 10.0016 Total (statistical sum)t i0.0093 i i

l I

l 1

1 l

I l

t Square root of the sum of the squares.

SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 4-27 Report HI-982083

Table 4.6.1 Reactivity Effects of Manufacturing Tolerances in Region II Tolerance Reactivity Effect, Ak Minimum Boralloading ( .' - nominal) 10.0028 Minimum Boral width l :.

nominal) i0.0007 Minimum box I.D. (M nominal) i0.0006 Maximum box wall thickness (o/S4 Ark 11 nominal) 10.0002 Density ( 11'..';

. ,10.522 g/cm 3nominal) 10.0029 Enrichment ( . ,5.00 wt% 2"U nominal) 10.0034 Total ( atistical sum)t 10.0054 I

t Square root of the sum of the squares.

SHADED TEXT CONTAINS PROPRIETARY INFORMATION lloltec International 4-28 Report HI-982083

~ '

i Table 4.7.1 Reactivity Effects of Temperature and Void in' Regions I and II ,

Reactivity Effect, Ak 1

- Case Region I Region II

{

4 C (39 F) reference - reference

-0.00155 -0.0024 20 C (68 F)

-0.00874 -0.0103 60 C (140 F)- l

-0.01356 -0.0149 j 80 C (176 F)

-0.02532 -0.0258 120 C(248 F) 120 C w/10% void -0.05580 -0.0499 ,

l 120 C w/ 20% void -0.08849' -0.0786 120 C w/ 30% void -0.12342 -0.1129 I

l l

i l

l SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 4-29 Report HI-982083

g 45 , . , , ,

bounding curve ,

actual data points e 40 (5.00,40.00 [0 35 (4.50,34.751) -

ACCEPTABLE DOMAIN 30 -

S e (4.00,28.048) 2 D 25 -

3:

5c.

e (3.50,22.410).

$ 20 -

j (3.00,16.506)

15 UNACCEPTABLE DOMAIN 10 (2.50,10.215) -

5 o (2.00,3.085)

,0

2. 2.5 3 .

3.5 4 4.5 5 Initial Fuel Enrichment (wt% U-235)

- Figure 4.1.1: Minimum Required Fuel Assembly Burnup as a Function of Nominal Initial Enrichment to Peimit Storage in Region II (Fuel assemblies with 235 enrichments less than.2.0 wt% U will conservatively be required to meet 235 the burnup requirements of 2.0 wt% U assemblies).

SIIADED TEXT CONTAINS PROPRIETARY INFORMATION

. Holtec International 4-30 Report HI-982083

n .

1 l

1

=

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i .

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l I M

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. , , , , , , , , , , , ,d,' O.I10' CHANNEL 0 0235 SS SHEATH]NG DETAll "A" Figure 4.3.1: A Cross-Sectional View of the Calculational Model Used for the Region I Rack Analysis (NOT TO SCALE).

SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 4-31 Report 111-982083

=

8.97* S0. =

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~

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( BORAL THICKNESS )/2

( 0 110 - 0 10I )/2 = 0.0045'

( 0.1011/2 = 0 0505 DETAll "A" Figure 4.3.2: A Cross-Sectional View of the Calculational Model Used for the Region II Rack Analysis (NOT TO SCALE). ,

I SilADED TEXT CONTAINS PROPRIETARY INFORMATION lloltec International 4-32 Report 111-982083

e l

APPENDIX 4A: BENCHMARK CALCULATIONS

- 4A.1 '- INTRODUCTION AND

SUMMARY

f 1

Benchmark calculations have been made on selected critical ex9eriments, chosen, in so far as possible, to bound the range of variables in the rack designs. Two independent methods of analysis were used, differing in cross section libraries and in the treatment of the crc., )

sections. MCNP4a [4A.1] is a continuous energy Monte Carlo code and KENO 5a [4A.2]

uses group-dependent cross sections. For the KENO 5a analyses reported here, the 238- 1 group library was che:=, processed through the NITAWL-II [4A.2] program to create a working library ar'a to account fer resonance self-shielding in uranium-238 (Nordheim t

integral treatment). The 238 group library was chosen to avoid or minimize the errors (trends) that have been reported (e.g., [4A.3 through 4A.5]) for calculations with collapsed cross section sets.

in rack designs, the three most significant parameters affecting criticality are (1) the fuel enrichment, (2) the ' B loading in the neutron absorber, and (3) the lattice spacing (or l' water-gap thickness if a flux-trap design is used). Other parameters, within the normal range of rack and fuel designs, have a smaller effect, but are also included in the analyses.

~

Table 4A.1 summarizes results of the benchmark calculations for all cases selected and analyzed, as referenced in the table. The effect of the major variables are discussed in l l

subsequent sections below. It is important to note that there is obviously considerable overlap in parameters since it is not possible to vary a single parameter and maintain criticality; some other parameter or parameters must be concurrently varied to maintain criticality.

I One possible way of representing the data is through a spectrum index that incorporates all of the variations in parameters. KENO 5a computes and prints the " energy of the average lethargy causing fission" (EALF). In MCNP4a, by utilizing the tally option with the identical 23'd-group energy structure as in KENO 5a, the number of fissions in each group may be collected and the EALF determined (post-processing).

' Small but observable trends (errors) have been reported for calculations with the -

27-group and 44-group collapsed libraries. These errors are probably due to the use of a single collapsing spectrum when the spectrum should be different for the various cases analyzed, as evidenced by the spectrum indices.

Holtec International Proprietary Information Appendix 4A, Page !

Figures 4A.1 and 4A.2 show the calculated kerrfor the benchmark critical experiments as a function of the EALF for MCNP4a and KEN 05a, respectively (UO2 fuel only). The scatter in the data (even for comparatively minor variation in critical parameters) represents experimental errort in performing the critical experiments within each laboratory, as well as between the various testing laboratories. The B&W critical !

experiments show a larger experimental error than the PNL criticals. This would be expected since the B&W criticals encompass a greater range of critical parameters than the PNL criticals.

Linear regression analysis of the data in Figures 4A.1 and 4A.2 show that there are no trends, as evidenced by very low values of the correlation coefficient (0.13 for MCNP4a i

and 0.21 for KENO 5a). The total bias (systematic error, or mean of the deviation from a kerrof exactly 1.000) for the two methods of analysis are shown in the table below.

Calculational Bias of MCNP4a and KENO 5a '

MCNP4a 0.0009 t0.0011 KENO 5a 0.0030t0.0012 l

The bias and standard error of the bias were derived directly from the calculated k,r, values in Table 4A.1 using the following equationt", with the standard error multiplied by the one-sided K-factor for 95% probability at tl e 95% confidence level from NBS Handbook 91 [4A.18] (for the number of cases analyzed, the K-factoc is -2.05 or slightly more than 2).

k=I k, (4A.1)

' A classical example of experimental error is the corrected enrichment in the PNL experiments, first as an addendum to the initial report and, secondly, by revised values in subsequent reports for the same fuel rods.

" These equations n,J be found in any sin.dard text on statistics, for example, reference

[4 A.6] (or the MU;P4a manual) and is the same methodology used in MCNP4a and in KENO 5a.

Holtec International Proprietary Information Appendix 4A, Page 2

l l

kf - ( k,)2 f, (4A.2) 2 ,.i ..i

=

. o* n (n-1)

Blas = (1 - k )

K og (4A.3) where k, are the calculated reactivities of n critical experiments; o, is the unbiased

- estimator of the standard deviation of the mean (also called the standard error of the bias (mean)); K is the one-sided multiplier for 95% probability at the 95% confidence level (NBS Handbook 91 [4A.18]).

Formula 4.A.3 is based on the methodology of the National Bureau of Standards (now NIST) and is used to calculate the values presented on page 4.A-2. The first portion of the equation, ( 1- E ), is the actual bias which is added to the MCNP4a and KENO 5a results.

The second term, Kog, is the uncertainty or standard error associated with the bias. The K values used were obtained from the National Bureau of Standards Handbook 91 and are for one-sided statistical tolerance limits for 95% probability at the 95% confidence level. The actual K values for the 56 critical experiments evaluated with MCNP4a and the 53 critical 1

experiments evaluated with KENO 5a are 2.04 and 2.05, respectively.

The bias values are used to evaluate' the maximum ke rr values for the rack designs.

KENO 5a has a slightly larger systematic error than MCNP4a, but both result in greater i

precision than published data [4A.3 through 4A.5] would indicate for collapsed cross section s'ets in KENO 5a (SCALE) calculations.

l 4A.2 Effect of Enrichment The benchmark critical experiments include those with enrichments ranging from 2.46 w/o to 5.74 w/o and therefore span the enrichment range for rack designs. Figures 4A.3 and 4A.4 show the calculatede k rr values (Table 4A.1) as a function of the fuel enrichment reported for the critical experiments. Linear regression analyses for these data cotfirms that there are no trends, as indicated by low values of the correlation coefficients (0.03 for MCNP4a and 0.38 for KENO 5a). Thus, there are no corrections to the bias for the various enrichments.

Holtec International Proprietary Information Appendix 4A, Page 3 l

I l

1 1

c - ,

As further confirmation of the absence of any trends with enrichment, a typical configuration was calculated with both MCNP4a and KEN 05a for various enrichments.

The cross-comparison of calculations with codes of comparable sophistication is suggested in Reg. Guide 3.41. Results of this comparison, shown in Table 4A.2 and Figure 4A.5, confirm no significant difference in the calculated values of k,n for the two independent codes as evidenced by the 45 slope of the curve. Since it is very unlikely that two independent methods of analysis would be subject to the same error, this comparison is considered confirmation of the absence of an enrichment effect (trend) in the bias.

4A.3 Effect of 'B Loading Several laboratories have performed critical experiments with a variety of thin absorber ,

panels similar to the Boral panels in the rack designs. Of these critical experiments, those ,

performed by B&W are the most representative of the rack designs. PNL has also made ;

j some measurements with absorber plates, but, with one exception (a flux-trap experiment),

j the reactivity worth of the absorbers in the PNL tests is very low and any significant errors that might exist in the treatment of strong thin absorbers could not be revealed.

Table 4A.3 lists the subset of experiments using thin neutron absorbers (from Table 4A.1) and shows the reactivity worth (Ak) of the absorber.'

No trends with reactivity worth of the absorber are evident, although based on the calculations shown in Table 4A.3, some of the B&W critical experiments seem to have unusually large experimental errors. B&W made an effort to report some of their l

experimental errors. Other laboratories did not evaluate their experimental errors.

l To further confirm the absence of a significant trend with B concentration in the absorber, a cross-comparison was made with MCNP4a and KENO 5a (as suggested in Reg.

Guide 3.41). Results are shown in Figure 4A.6 and Table 4A.4 for a typical geometry.

These data substantiate the absence of any error (trend) in either of the two codes for the conditions analyzed (data points fall on a 45 line, within an expected 95% probability limit).

The reactivity worth of the absorber panels was determined by repeating the calculation with the absorber analytically removed and calculating the incremental (Ak) change in reactivity due to the absorber.

Holtec International Proprietary Information Appendix 4A, Page 4

j l

4A.4 Miscellaneous and Minor Paramme 4A.4.1 Reflector Material and Spacings PNL has performed a number of critical experiments with thick steel and lead reflectors.t Analysis of these critical experiments are listed in Table 4A.5 (subset of data in Table 4A.1). There appears to be a small tendency toward overprediction of k,, at the lower

_ spacing, although there are an insufficient number of data points in each series to cllow a

, quantitative determination of any trends. The tendency toward overprediction at close spacing means that the rack calculations may be slightly more conservative than otherwise.

4A.4.2 Fuel Pellet Diameter and I attice Pitch The critical experiments selected for analysis cover a range of fuel pellet diameters from 0.311 to 0.444 inches, and lattice spacings from 0.476 to 1.00 inches. In the rack designs, {

the fuel pellet diameters range from 0.303 to 0.3805 inches O.D. (0.496 to 0.580 inch lattice spacing) for PWR fuel and from 0.3224 to 0.494 inches O.D. (0.488 to 0.740 inch

)

{

lattice spacing) for BWR fuel. Thus, the critical experiments analyzed provide a reasonable I rep'resentation of power reactor fuel. Based on the data in Table 4A.1, there does not appear to be any observable trend with either fuel pellet diameter or lattice pitch, at least over the range of the critical experiments applicable to rack designs.

l 4A.4.3 Soluble Boron Conceniration Effects l I

Various soluble boron concentrations were used in the B&W series of critical experiments and in one PNL experiment, with boron concentrations ranging up to 2550 ppm. Results of MCNP4a (and one KEN 05a) calculations are shown in Table 4A.6. Analyses of the very high boron concentration experiments (> 1300 ppm) show a tendency to slightly overpredict reactivity for the three experiments exceeding 1300 ppm. In turn, this would suggest that the evaluation of the racks with higher soluble boron concentrations could be I

slightly conservative.

1 l

l t Parallel experiments with a depleted uranium reflector were sis ~ performed but not (

included in the present analysis since they ere not pertinent to the Holtec rack design. l l

Holtec International Proprietary Information Appendix 4A, Page 5 i

l

l l

1 4A.5 MOX Fuel The number of critical experiments with PuO2 bea-ing fuel (MOX) is more limited than for !

UO2fuel. However, a number of MOX critical experiments have been analyzed and the I results are shown in Table 4A.7. Results of these analyses are generally above a k,y of 1.00, indicating that when Pu is present, both MCNP4a and KENO 5a overpredict the reactivity. This may indicate that calculation for MOX fuel will be expected to be !

conservative, especially with MCNP4a. It may be noted that for the larger lattice spacings, the KENO 5a calculated reactivities are below 1.00, suggesting that a small trend may exist with KENO 5a. It is also possible that the overprediction in k,y for both codes may be due to a small inadequacy in the determination of the Pu-241 decay and Am-241 growth. This I possibility is supported by the consistency in calculated k,y over a wide range of the '

spectral index (energy of the average lethargy causing fission). .

~

i Holtec Intemational Proprietary Information Appendix 4A, Page 6

, i

I 4A.6 References

[4A.1] J.F. Briesmeister, Ed., "lACNP4a - A General Monte Carlo N-Particle Transport Code, Version 4A; Los Alamos National Laboratory, LA-12625-M (1995).

[4A.2] SCALE 4.3, "A Modular Code System for Performing Standardized -

Computer Analyses for Licensing Evaluation", NUREG-0200 (ORNL-NUREG-CSD-2/U2/R5, Revision 5, Oak Ridge National Laboratory, September 1995.

[4A.3] M.D. DeHart and S.M. Bowman, " Validation of the SCALE Broad Structure 44-G Group ENDF/B-Y Cross-Section Library for Use in Criticality Safety Analyses", NUREG/CR-6102 (ORNL/TM-12460)

Oak Ridge National Laboratory, September 1994.

[4A.4] W.C. Jordan et al., " Validation of KENOV.a", CSD/TM-238, Martin Marietta Energy Systems, Inc., Oak Ridge National Laboratory, December 1986.

[4A.51 0.W. Hermann et al., " Validation of the Scale System for PWR Spent Fuel Isotopic Composition Analysis", ORNL-TM-12667, Oak Ridge National Laboratory, undated.

[4A.6] R.J. Larsen and M.L. Marx, An Introduction to Matheh11ical Statistics and its Applications, Prentice-Hall,1986.

-[4A 7) M.N. Baldwin et al., Critical Experiments Supporting Close Proximity Water Storage of Power Reactor Fuel, BAW-1484-7, Babcock and Wilcox Company, July 1979.

[4A.8] G.S. Hoovier et al., Critical Experiments Supporting Underwater Stocage of Tightly Packed Configurations of Spent Fuel Pins, BAW-1645-4. Thbcock & Wilcox Company, November 1991.

[4A.9] L.W. Newman et al., Urania Gadolinia: Nuclear Model Development and Critical Experiment Benchmark, BAW-1810, Babcock and Wilcox Company, April 1984.

Holtec International Proprietary Information Appendix 4A, Page 7

[4A.10] J.C. Manaranche et al., " Dissolution and Storage Experimental Program with 4.75 w/o Enriched Uranium-Oxide Rods," Trans.

Am. Nucl. Soc. 33: 362-364 (1979).

. [4A.11] S.R. Bierman and E.D. Clayton, Criticality Experimcats with 2

Suberitical Clusters of 2.35 w/o and 4.31 w/o "U Enriched UO 2 Rods in Water with Steel Reflecting Walls, PNL-3602, Battelle Pacific Northwest Laboratory, April 1981.

[4A.12] S.R. Bierman et al., Criticality Experiments with Subcritical Clusters of 2.35 w/o and 4.31 w/o2 "U Enriched UO2 Rods in Water with Uranium or Lead Reflecting Walls, PNL-3926, Battelle Pacific Northwest Laboratory, December,1981.

[4A.13] S.R. Bietman et al., Critical Separation Between Suberitical Clusters of 4.31 w/o2 "U Enriched UO2 Rods in Water with Fixed Neutron Poisons, PNL-2615, Battelle Pacific Northwest Laboratory, October 1977.

[4A.14] S.R. Bierman, Criticality Experiments with Neutron Flux Traps Containing Voids, PNL-7167, Battelle Pacific Northwest Laboratory, April 1990.

[4A.15] B.M. Durst et al., Critical Experiments with 4.31 wt %2 "U Enriched UO2 Rods in Highly Borated Water Lattices, PNL-4267, Battelle Pacific Northwest Laboratory, August 1982.

[4A.16] S.R. Bierman, Criticality Experiments with Fast Test Reactor Fuel Pins in Organic Moderator, PNL-5803, Battelle Pacific Northwest Laboratory, December 1981.

[4A.17] E.G. Taylor et al., Saxton Plutonium Program Critical Experiments for the Saxton Partial Plutonium Core, WCAP-3385-54, l Westinghouse Electric Corp., Atomic Power Division, December 1%5.

i

[4A.18] M.G. Natrella, Egerimental Statistics, National Bureau of Standards, Handbook 91, August 1963.

i i

Holtec International Proprietary Information Appendix 4A, Page 8

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Table 4A.2 COMPARISON OF MCNF4a AND KEN 05a CALCULATED REACTIVITIESt FOR VARIOUS ENRICHMENTS Calculated Fw 10 MCNP4a KENO 5a Enrichment 3.0 0.8465. 0.0011 0.8478 0.0004 3.5 0.8820 i 0.0011 0.8841 i 0.0004 ;

3.75 0.9019 i 0.0011 0.8987 i 0.0004 4.0 0.9132 0.0010 0.9140 i 0.0004 4.2 0.9276 i 0.0011 0.9237 i 0.0004.

4.5 0.9400 t 0.0011 ,

0.9388 0.0004 I

l 1

' Based on the GE 8x8R fuel assembly, i

Holtec International Proprietary Information Appendix 4A, Page 14

Table 4A.3 MCNP4a CALCULATED REACTIVITIES FOR CRITICAL EXPERIMENTS WITH NEUTRON ABSORBERS 1

Ak MCNP4a Worth of Calculated EALF' !

Ref. Experiment Absorber k,, (eV) 4A.13 PNL-2615 Boral Sheet 0.0139 0.9994 i 0.0012 0.1165 4 A.7 B&W-1484 Core XX 0.0165 1.0008 i 0.0011 0.1724 4A.13 1.62% Boron-steel 0.0165 0.9996 i0.0012 0.1161 1 PNL-2615 4A.7 B&W-1484 Core XIX 0.0202 0.9961 i 0.0012 0.2103 4A.7 B&W-1484 Core XXI 0.0243 0.9994 i 0.0010 0.1544 4 A.7 B&W-1484 Core XV11 0.0519 0.9962 i 0.0012 0.2083 4A.11 PNL-3602 Boral Sheet 0.0708 0.9941 i 0.0011 0.3135 4A.7 B&W-1484 Core XV 0.0786 0.9910 i 0.0011 0.2092 4A.7 B&W-1484 Core XVI 0.0845 0.9935 i o.0010 0.1757 4A.7 B&W-1484 Core XIV 0.1575 0.9953 i 0.0011 0.2022 4A.7 B&W-1484 Core X111 0.1738 1.0020 i0.0011 0.1988 !

l 0.1931 0.9991 i 0.0011 0.3722 j 4A.14 PNL-7167 Expt 214R flux trap l

l l

l tEALF is the energy of the average lethargy causing fission.

Holtec International Proprietary Information Appendix 4A, Page 15

j

(

l l

1 i

i-Table 4A.4 ;

COMPARISON OF MCNP4a AND KENO 5a '

CALCULATED REACTIVITIESt FOR VARIOUS ' B LOADINGS Calculated k,n lo

' B, g/cm2 MCNP4a KENO 5a 0.005 1.0381 i 0.0012 1.0340 0.0004 0.010 0.9960 i 0.0010 0.9941 0.0004 0.015 .0.9727 i 0.0009 0.9713 i 0.0004 l 0.020 0.9541 0.0012 0.9560 i 0.0004 0.025 0.9433 i 0.0011 d.9428 0.0001 0.03 0.9325 i 0.0011 0.9338 i 0.0004 0.035 0.9234 0.0011 0.9251 0.0004 0.04 0.9173 i 0.0011 0.9179 i 0.0004 l

I l

l t Based on a 4.5% enriched GE 8x8R fuel assembly.

Holtec Intemational Proprietary Information Appendix 4A, Page 16 l 1' )

I l

Table 4A.5 CALCULATIONS FOR CRITICAL EXPERIMENTS WITH

' t ITIICK LEAD AND STEEL REFLECTORS Separation, Ref. Case E, wt% cm- MCNP4a k,, KENO 5a k,n d

4A.11 Steel 2.35 1.321 0.9980 i 0.0009 0.9992 i 0.0006 Reflector 2.35 2.616 0.9968 i 0.0009 0.9964 i 0.0006 2.35 3.912 0.9974 i 0.0010 0.9980 i 0.0006 2.35 = 0.9962 i 0.0008 0.9939 0.0006 4A.ll Steel 4.306 1.321 0.9997 0.0010 1.0012 0.0007 Reflector 4.306 2.616 0.9994 i 0.0012 0.9974 i 0.0007 4.306 3.405 0.9969i0.00ll 0.9951 i 0.0007 4.306- = 0.9910 i 0.0020 0.9947 i 0.0007 4A.12 Lead ~ 4.306 0.55 1.0025 i 0.0011 0.9997 i 0.0007 Reflector 4.306 1.956 1.0000 i 0.0012 0.9985 i 0.0007 4.306 5.405 0.9971 i 0.0012 0.9946 i 0.0007 t Armnged in order ofincreasing reflector-fuel spacing.

Holtec International Proprietary Information Appendix 4A, Page 17

Table 4A.6 CALCULATIONS FOR CRITICAL EXPERIMENTS WITH VARIOUS SOLUBLE BORON CONCENTRATIONS

- Calculated k, Boron Concentration, Reference Experiment ppm MCNP4a KENO 5a 4A.15 PNL-4267 0 0.9974 0.0012 -

4A.8 B&W-1645 886 0.9970 i 0.0010 0.9924 t 0.0006 4A.9 B&W-1810 1337 1.0023 t 0.0010 -

4A.9 B&W-1810 1899 1.0060 0.0009 -

4A.15 PNI 4267 2550 1.0057 0.0010 -

Holtec International Proprietary Information Appendix 4A, Page 18

U!!

1

. Table 4A.7 CALCULATIONS FOR CRITICAL EXPERIMENTS WITH MOX FUEL MCNP4a KENO 5a k,, EALF" k,, EALF" Reference Case' 0.9171 1.0046 i 0.0006 0.8868 PNL-5803 MOX Fuel - Exp. No. 21 1.0041 10.0011

[4A.16) 0.2944 MOX Fuel - Exp. No. 43 1.0058 i 0.0012 0.2 % 8 1.0036 1 0.0006 0.1665 0.9989 1 0.0006 0.1706 MOX Fuel - Exp. No.13 1.0083 i 0.0011 0.1139 0.9966 1 0.0006 0.1165 MOX Fuel - Exp. No. 32 1.0079 i 0.0011 0.8665 1.0005 i 0.0006 0.8417 WCAP- Saxton @ 0.52* pitch 0.9996 1 0.0011 l

3385 0.5197 l Saxton @ 0.56' pitch 1.0036 10.0011 0.5289 1.0047 i 0.0006 (4A.17) 1.000'io.0010 0.6389 NC NC Saxton @ 0.56' pitch borated 0.1520 1.0133 1 0.0006 0.1555 l Saxton @ 0.79" pitch 1.0063 1 0.0011 l

Note: NC stands for not calculated -

t Arranged in order of increasing lattice spacing.

" EALF is the energy of the average lethargy causing fission.

Holtec International Proprietary Information Appendix 4A, Page 19

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e

't 5.0 : THERMAL-HYDR AULIC CONSIDER ATIONS 5.1 _ introduction .

~

This section.provides a summary of the methods, models, analyses and numerical results to

" demonstrate the compliance of the reracked Byron and Braidwood spent fuel pools within the provisions of Section III of the USNRC "OT Position Paper for Review and Acceptance of Spent

. Fuel Storage and Handling Applications", (April 14,1978).

Similar methods of thermal-hydraulic analysis have been used in other rerack licensing projects (see Table 5.1.1).

The thermal-hydraulic qualification analyses' for the rack array may be broken down into the following categories:

i. Evaluation of the long-term decay heat loali, which is the cumulative spent fuel decay heat generation from all fuel assemblies stored in the pool from previous discharges to the final discharge, ii. Evaluation of the decay heat load and pool bulk temperature as a function of time during the postulated final discharge scenarios.

Jiii. Evaluation of the " time-to-boil" if all forced heat rejection paths from the pool are lost.

iv. Determination of maximum local water temperature in the fuel rack cells at the instant when the bulk temperature reaches its maximum value.

v. Evaluation of peak fuel cladding temperature in the hottest fuel cell.

The following sections present the plant system description, analysis assumptions, a synopsis of the analysis methods employed, and final results.

SHADED TEXT CONTAINS PROPRIETARY INFORMATION Hohec Internationaj 5-1 Report HI-982083

5.2 Spent Fuel Pool and Cooling System Descriptions The spen: fuel pool cooling system consists of two complete cooling trains. The spent fuel pool '

cooling system (piping, pumps, valves, and heat exchangers) is Safety Category I, Quality Group C. 1

< The 3-inch piping from the refueling water storage tanks to the afueling water purification pump, the pump, and its associated piping and valves are Safety Category I, Quality Group C. A 2-inch I i

Safety Category I, Quality Group C line from the discharge of the refueling water purification pump j to the spent fuel pool is permanently installed. This is the Category I water makeup circuit. The backup Safety Category I makeup system consists of piping and hoses from the Safety Category I fire protection system. The non-Category I primary water makeup system takes water from both primary water storage tanks and routes the water through the spent fuel pool water filter and then to ,

the return header. In summary, there are three sources of makeup water available: a primary unborated non-Category I source, a borated Safety Category I source, and an unborated fire protection Safety Category I water system.

Each cooling train incorporates one heat exchanger and pump, one purification loop with demineralizer and filter, associated piping, valving, and instrumentation. One surface skimmer loop is also provided. Each cooling train is designed to service the spent fuel pool, with design spent fuel assembly loading,.and to maintain the bulk fluid temperature of the pool below 140*F for normal discharge. The fuel pool cooling system design data is presented in Table 5.2.1.

The spent fuel pool cooling system removes decay heat from fuel stored in the spent fuel pool. Spent fuel is placed in the pool during refueling sequences at each outage and the heat is transferred from the sper fuel pool cooling system through the heat exchanger to the component cooling (CC) system.' e essential service water system (SX), in :um, removes heat from the component cooling water system.The minorexpansion of the SFP storage capacity does not affect the performance of either the CC or SX System.

4

, SHADED TEXT CONTAINS PROPRETARY INFORMATION Holtec Intemational 5-2 Report HI-982083

f I

I When either cooling train is in operation, water flows from the spent fuel pool to the spent fuel pool pump suction, is pumped through the tubeside of the heat exchanger, and is returned to the pool. The suction line, which is protected by.a strainer, is located at an elevation 4 feet below the normal spent fuel pool water level, while the return line contains an antisiphon hole near the surface of the water to prevent gravity drainage of the pool.

While the heat removal operation is in process, a portion of the spent fuel pool water, approximately 80 gpm, may be divened through a demineralizer and a filter to maintain sper,st fuel pool water clarity and purity. Transfer canal water may also be circulated through the same demineralizer and filter by having the gate between the canal and the spent fuel pool open. This purification loop is sufficient for removing fission products and other contaminants that may be introduced if a leaking fuel assembly is transferred to the spent fuel pool.

1-1 l

The demineralizer and filter of either cooling train may be isolated from the heat removal portion of the spent fuel pool cooling system. By so doing, the isolated equipment may be used in conjunction with the refueling water purification pump to clean and purify the refueling water while spent fuel pool heat removal operations proceed. Connections are provided such that the refueling water may be pumped frcm either the refueling water storage tank or the refueling cavity of either unit, through a filter and demineralizer, and discharged to the refueling cavity or refueling water storage tank of either unit.

To assist further in maintaining spent fuel pool water clarity, the water surface is cleaned by a skimmer loop. Water is removed from the surface by the skimmer unaer, pumped through a filter, and returned to the pool surface at two locations remote from the skimmers at a rate of approximately 50 gpm.

i i

SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 5-3 Report HI-982083 i

)

The spent fuel pool is initially filled for use with water that is at the same boron concentration as that in the refueling water storage tank. Borated water may be supplied from the refueling water storage tank via the refueling water purification pump connection, or by a line from the boric acid blender, 1

located in the chemical and volume control system. Demineralized water can be added for makeup purposes (i.e., to replace evaporative losses) through a connection in each cooling train's purification return loop.

5.3 Decay Heat Load Calculations 1

The decay heat load calculation has been performed in accordance with the provisions of USNRC l Branch Technical Position ASB9-2, " Residual Decay Energy for Light Water Reactors for Long Term Cooling", Rev. 2, July,1981. l To determine the end-of-life decay heat in the Byron and Braidwood spent fuel pools, the historical and projected discharges are considered as shown in Tables 5.3.1 to 5.3.6. All future fuel cycles for Byron and Braidwood are anticipated to be 18 months. A total of 2,864 and 2,821 assemblies will be accumulated from previous discharges in the Byron and Braidwood pools, respectively. A fresh )

1 full core discharge of 193 assemblies will increase the fuel inventory to 3,057 and 3,014 fuel assemblies which conservatively bounds the maximum fuel storage capacity for decay heat evaluation purpose.

A bounding long-term decay heat calculated from the discharge schedule will be considered for the analysis. Since the decay heat load from the old assemblies varies very slowly as a function of time, the long-term decay heat is assumed to be constant.

SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 5-4 Report HI-982083

p q l

5.4 Discharge Scenarios -

1 The following discharge scenarios are considered for bulk pool thermal-hydraulic analysis:

A normal discharge scenario consisting of an equilibrium reload batch size l

Case (i):

of 84 fuel assemblies to be transferred to the spent fuel after a 100-hour in-J reactor hold time. The duration of this batch discharge is 10.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months . The heat )

from this fresh batch and background heat from old discharges is removed by 1 I

one cooling train.

Case (ii): A full-core discharge scenario in which 193 fuel assemblies are to be )

transferred to the spent fuel pool after a 100-hour in-reactor hold time. The i time duration of this batch discharge is 24.1 hours1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . The heat from this fresh )

batch and background heat from old discharges is removed by one cooling l train. .'

Case (iii): An abnormal back-to-back discharge scenario in which a normal discharge is followed by a full-core discharge 17 days later. The old discharge that occuned prior to the normal discharge is 6 months old. The heat from the background, normal, and full core discharges is removed by one cooling train. This case is also analyzed with two cooling trains in operation, as stipulated in UFS \R Subsection 9.1.3.1.

In all cases considered the design maximum CCW temperature at 105 F is assumed in the analysis. j

' Decay heat input into the Byron and Braidwood poo:s from old and fresh fuel discharges is based on Branch Technical Position ASB 9-2 methodology, Thermal inenia of the spent fuel pool, credited

]

in the dynamic analysis, is limited to the quantity of water in the pool. Th!'s represents a significant j conservatism since the thermal capacity of the installed racks, spent fuel, liner, and concrete is not credited in the analysis. The influence of thermal inenia in pool transient analysis is discussed in the i

. results Section 5.8.1, j Heat rejection from the spent fuel pool occurs as a consequence of forced cooling via the SFP l 1'

cooling system, heat loss from SFP cooling system piping, natural convection and evaporative cooling from the fuel pool surface and conduction through the pool concrete walls. Evaporative heat j

~

loss is evaluated at maximum fuel handimg building room temperature. Heat loss from th'e pool concrete walls and SFP system piping is neglected. This is a conservative assumption. ;

SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 55 Report H1982083 4

Power input to the SFP cooling system circulation pumps is converted to heat as a result of frictional

' losses in the piping, valves, fittings, and other equipment. This represents an additional means of heat input into the overall fuel pool cooling system. Based on brake hersepower curves specified by the pump manufacturer, the pump heat input is included in the background decay heat input calculations in all scenarios.

The fuel discharge scenarios described above as Case (i), Case (ii), and Case (iii) are schematically depicted in Figures 5.4.1, 5.4.2, and 5.4.3, respectively. Principal thermal-hydraulic data is summarized in Tables 5.4.1 and 5.4.2.

5.5 Bulk Pool Temperatures In this section, we present the methodology fo'r calculating the bulk pool temperature as a function of the time coordinate. The method used to calculate the rate of pool water temperature rise and the time-to-boil when all forced cooling paths are unavailable is also presented.

In order to perform a conservative analysis, the heat exchangers are assumed to be fouled to their design basis level. Thus, the temperature effectiveness, p, for the heat exchangers utilized in the analysis is the lowest postulated value based on the exchanger specification sheets.

The following assumptions also apply for the bulk pool temperature evaluation.

All old spent fuel assemblies discharged into the pool are assumed to have 1,642 Effective Full Power Days (EFPD) of operation for all cycles. This assumption will provide a conservative decay heat calculation.

. . In calculating the spent fuel pool evaporation heat losses, the building housing the spent fuel pool is assumed to have the maximum ambient air temperature of 104 F and 100% relative humidity.

This assumption will result in a conservative credit for the evaporation heat losses.

l

Heat conduction _ losses through the pool walls and slab are neglected.

SHADED TEXT CONTAINS PROPRIETARY INFORMATION

' Holtec International '5-6 Report HI-982083

1 The mathematical formulation can be explained with reference to the simplified heat exchanger alignment of Figure 5.5.1. Referring to the spent fuel pool cooling system, the governing differential equation can be written by utilizing conservation of energy: l C dT = Ot -Q, (5.5.1) dr Qt = P, + Q(r)-4 (T,t,) (5.5.2) where:

C: Themial capacity of the pool (net water volume times water density and times heat capacity), Btu / F , ,

1 Qt.: Net heat load in the SFP, Btu /hr.

Q(t): Heat generation rate from recently discharged fuel, which is a specified function of time, T, Btu /hr.

Peons: Heat generation rate from "old" fuel, Blu/hr. It is also stated as long-term decay heat load. The SFP pump heat is included in Peon, for maximum bulk pool temperature calculation.

Qux: Heat removal rate by the heat exchanger, Btu /hr.

Qev (T,t.): Heat loss to the suiroundings, which is a function of pool temperature T and ambient temperature t., Blu/hr.

i i

Qux is a nonlinear function of time and can be written in terms of effectiveness p as follows: l Qax = W Ci p (Ti- ti) (5.5.3) to - t' (5.5.4) p = T - ti where:

W:i Coolant flow rate, lb./hr.

C:i Coolant specific heat, Btu /lb. F.

SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 5-7 Report HI-982083

rm

,3 9

p: 1 Temperature effectiveness of heat exchanger I, Pool water temperature, F t:

i Coolant inlet temperature, *F to: Coolant outlet temperature, F The temperature effectiveness p is calculated from data on the exchanger specification sheets. Q(T)-

,is determined according to the provisions of USNRC Branch Technical Position ASB9-2, " Residual Decay Energy for Light Water Reactors for long Term Cooling", Rev. 2, July,1981 Q(t) is a function of decay time, number of assemblies, and in-core exposure time. During the fuel transfer, the heat load in the pool will increase with respect to the rate of fuel transfer and equals Q(T) after ;

the fuel transfer.

Qsv.is a nonlinear function of pool temperature and ambient temperature. Qsv contains the heat evaporation loss through the pool surface, natural convection from the pool surface and heat

- conduction through the pool walls and slab. Experiments show that'the heat conduction takes only about 4% of the total heat loss [5.5.1], and can therefon: be neglected. The evaporation heat loss and naturai connction heat loss can be expressed as:

Qg = m V A, + hc A,0 (5.55) l where:

m: Mass evaporation rate, Ib/hr. ft.2 j V: Latent heat of pool water, Btu /lb.

' A.: Pool surface awa, ft.2 he: Convection heat transfer coefficient at pool surface, Bm/ft.2 hr. F 0 = T-t.: The tempera:ure difference between pool water and ambient air, 'F

. SHADED TEXT CONTAINS PROPRIETARY INFORMATION Hoitec International 5-8 Report 111-982083 a-

n The mass evaporation rate m can be obtained as a nonlinear function of 0. We, therefore, have m = hol0 )(Wp,- Wa ) (5.5.6)

W ps: Humidity ratio of saturated moist air at pool water temperature T.

Wa: Humidity ratio of saturated moist air at ambient temperature t i ho(0) Diffusiou coefficient at pool water surface. ho is a nonlinear function of 0, i lb/hr. ft.2 F .

The nonlinear single order differential equation (5.5.1) is solved using Holtec's QA-validated numerical integration code "ONEPOOL".

The next step in the analysis is to determine the time-to-boil if all forced cooling paths become unavailable. Clearly, the most critical instant of loss-of-cooling is when pool water temperature has reached its maximum value. It is assumed that makeup water is added at the rate of G lb/hr. The makeup water is at temperature tcoa. The governing enthalpy balance equation for this condition can be written as (C + C'] dT = Pc,, + Q ( r + r,, ) + G ( tcoa - T)- Qg, (5.5.7) l dr 1 7 + 1,gy3 wisere C'= G(r - r,) L Qu.. ,

L y In the foregoing, water is assumed is have specific heat of unity and latent heat L, and the time com enate t is measured from the instant maximum pool water temperature is reached. To is the time coordinate when the makeup watcr application is begun. %, is the time coordinate measured from the instant of reactor shutdown to when maximum pool wa:er temperature is reached. T is the dependent variable (pool water temperature). For conservati Qsv is assumed to remain constant after pool water terr.perature reaches and rises above 170 F. The time-to-boil calculations are

' performed assuming no makeup water is available.

SHA DED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 5-9 Report HI 982083

A QA-validated numerical quadrature code is used to integrate the foregoing equation. The pool water heat up rate, time-to-boil, and subsequent water pool depth time profile are generated and reported in this chapter.

5.6 Local Pool Water Temperature In this section, a summary of the methodology for evaluating the local pool water temperature is presented.

5.6.1 Basis The local Water temperature analysis uses bounding bulk pool conditions coincident with peak bulk temperature response. In order to determine an upper bound on the maximum local water temperature, a series of conservative assumptions are made. The principal assumptions are listed j i

below:

i e A 2-inch downcomer gap along the racks outline is conservatively used in the local analysis.

. No downcomer flow is assumed to exist between the rack modules.

. With a full core discharged into the fuel pool, the peak bulk pool temperature is evaluated as a steady sfate condition for local temperature analysis, e The water inlet-outlet lines are modeled to be in close proximity to each other so as to increase the propensity for flow bypass in the model. Th: inlet / outlet lines in the pool are separated by a large distance to preclude such a possibility.

e The hydraulic characaristics of the fuel assembly types stored in the Byron /Braidwood pools are conservatively determined. j l

1 I

l I

SHADED TEXT CONTAINS PROPRIETARY INFORMATION I Holtec International 5 10 Report 111-982083

5.6.2 Local Temperature Evaluation Method l

Local temperature evaluation of the Byron and Braidwood pools is performed with the sparger line truncated. Adequate cooling of hot fuel in the fuel pool is demonstrated by performing a rigorous evaluation of the velocity and temperature fields in the pool created by the interaction of buoyancy-driven flows and water injection / egress. A Computational Fluid Dynamic:; (CFD) analysis for this demonstration is implemented. The objective of this study is to demonstrate that the principal thermal-hydraulic criteria of ensuring local subcooled conditions in the pool is met for the maximum fuel pool decay heat scenario. The local thermal-hydraulic analysis also considers the effect of a hot fuel assembly dropped above the racks and laying horizontally on top of the hot fuel rack cells coincident with a 50% fuel cell exit blockage ur$ der this postulated mechanical accident scenario.

l An outline of the CFD approach is de ,cribed as follor

There are several significant geometric and thermal-hydraulic features of the Byron and Braidwood spent fuel pools that need to be considered for a nigorous CFD analysis. From a fluid flow modeling perspective, there are two regions to be considered. One region is the bulk pool region where the !

l' classical Navier-Stokes equations are so'Ived with turbulence effects included. The other region is the heat generating fuel assemblies located in the spent fuel racks located near the bottom of the ,

spent fuel pool. In this region, water flow is directed vertically upwards due to buoyancy forces ,

I through relatively small flow channels formed by the fuel assembly rod an ays in each rack cell. {

n l

l l

w

~

(5.6.1) n . .

The permeability and inertial resistance parameters for the rack cells loaded with Byron and Braidwood fuel is determined based on friction factor correlations for laminar flow conditions typically encountered due to low buoyancy induced velocities and small size of the flow channels.

SHADED TEXT CONTAINS prs "GETARY LhFORMATIO.;

Holtec International 5 11 Regrt III-982083 I

t I

l

r The Byror, and Braidwood pools geometry requires an adequate portrayal of large scale and small scale features, spatially distributed heat sources in the spent fuel racks and water inlet / outlet configuration. Relatively cooler bulk pool wate normally flows down through the narrow fuel rack

. outline to pool wall liner clearance known as the downcomer. Near the bottom of the racks, the flow

- turns from a vertical to horizontal direction into the bottom plenum supplying cooling water to the

~

rack cells. Heated water issuing out of the top of the racks mixes with the bulk pool water. An adequate modelit.g of these features on the CFD program involves meshing the large scale bulk pool region and small scale downcomer and bottom plenum regions with sufficient num%r of computational cells to capture the bulk and local features of the flow field.

The distributed heat sources in the spent fuel pool racks are modeled by identifying distinct heat generation zones considering full-core discharge, peaking effects, and presence of background decay

. heat from old discharges. Three heat generating zones were modeled. The first consists of background' fuel from previous discharges, the remaining two zones consist of fuel from a full-core-discharge scenario. The two full core discharge zones are differentiated by one zone with higher than average decay heat generation and the other with less than average decay heat generation. This is a conservative model, since all of the fuel with higher than average decay heat is postulated to be placed in a contiguous area. The analysis has been performed for a limiting full-com offload scenario selected from the offload scenarios listed earlier.

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5.7 Claddine Temperature In this section, the method to calculate the temperature of the fuel cladding is presented.

The maximum spacific power of a fuel anay g3 can be given by:

q, = q F,, (5.7.1) where:

F n- =- radial peaking factor q = average fuel assembly specific power SHADED TEXT CONTAINS PROPRIETARY II" FORMATION Holtec International 5-13 Report HI-982083

I The pe'aking factors are given in Table 5.4.1. The maximum temperature rise of pool water in the

? most dir.af /antageously placed fuel assembly, defined as one which is subject to the highest local pool water temperature, is computed for the maximum decay heat load full core discharge scenario.

Having determined the maximum local water temperatun: in the pool, it is now possible to detennine the maximum fuel cladding temperature. A fuel rod can produce F, times the average heat emission .

rate over a small length, where F is the axial rod peaking' factor. The axial heat distribution in a rod I is generally a maximum in the central region, and tapers off at its two extremities. Thus, peak cladding heat Oux is given by the equation:

9 "' '

q, = (5.7.2)

.A ,

where Ae is the total cladding external heat transfer area in the active fuel length region.

Within each fuel assembly sub-channel, water is continuously heated by the cladding as it moves axially upwards from bottom to top under laminar now conditions. Rohsenow and Hartnett [5.7.1]

repon Nusselt-number based heat transfer correlatian for laminar flow in a heated channel. The film temperature driving force (ATr) at the peak cladding flux location is calculated as follows: .

hf D,, / K, = Nu (5.7.3)

ATf = 1 (5.7.4) h, I where, hr is the waterside film heat transfer coefficient, Da is sub-channel hydraulic diameter, K is l 1

water thermal conductivity and Nu is Nusselt number from heat transfer correlation.

I SHADED TEXT CONTAINS PROPRIETARY INFORMATION Hol:ec International 5-14 Repon HI-982083

I 1

Ir. order to introduce some additional conservatism in the analysis, we assume that the fuel cladding 2

has a crud deposit resistance Re (equal to 0.0005 ft -lir- F/ Btu), which covers the entire surface.

Thus, including the temperature drop across the crud resistance, the cladding to water local temperature difference (ATc)is given by:

ATc = ATr + Re qc (5.7.5) 5.8 Results This section contains results culled from the analyses performed for each of the three postulated discharge scenarios.

f.8.1 Bu'k Pool Temperature The bulk pool temperature and the decay heat load profile in the spent fuel pool as a function of time (after reactor shutdown) are shown in Figures 5.8.1 through 5.8.6 for the three discharge scenarios I postulated in Section 5.4. The maximum values of the pool bulk temperature and the coincident heat load to the spent fuel pool cooler are summarized in Table 5.8.1. As would be expected from physical considerations, the thermal inertia of the pool water causes the bulk pool temperature to reach its maximum value within a short time after the occurrence of the peak decay heat load, the lag is a direct result of the system thermal capacitance. The fact that the coincident heat load is not the maximum decay heat load (due to aforementionad system inertia) warrants emphasis; since this distinction is often neglected in evaluation of the system performance data. The coincident time to the maximum temperature is the summation of in-core hold time, fuel transfer time, and the lag time due to thermal inertia. It is shown from the results that the pool water temperatures is kept below 140 F during normal refueling offload (Case i). For the Case (ii) and (iii) scenarios, the full core discharge peak bulk pool temperature is below boiling by a substantial margin.

The USNRC criteria stipulated in the Standard Review Plan require a demonstration that bulk pool temperature be maintained below 140 F with a single active failure for normal discharges and that the pool should not boil under abnormal (full-core) discharges. The maximum bulk temperatures SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 5-15 Report Hi-982083

1 l

l tabulated in Table 5.8.1 demonstrate the compliance of the Byron and Braidwood cooling system with the USNRC acceptance criteria. Furthermore, the maximum calculated bulk pool temperature is less than the minimum design temperature for the spent fuel pool cooling and purification system.

Therefore, no physical modification of the spent fuel pool cooling and purification system is necessary.

5.8.2 Time-to-Boil If all heat exchanger assisted forced pool cooling becomes unavailable, then the pool water will begin to rise in temperature and eventually will reach the normal bulk boiling tem.o, erature at 212 FJ The time to reach the boiling point will be the shortest when the loss of forced cooling occurs at the point in time when the pool bulk temperature is at its maximum calculated value. Although the probability of the loss-of-cooling event coinciding with the instant when the pool water has reached its peak value is extremely remote, the calculations are performed under this extremely unlikely scenario. Table 5.8.2 contains the results with the additional proviso that no makeup water was added to the pool Figure 5.8.7 shov. ; tne variations of SFP water elevation as a function of time after loss of cooling for all the analyzed scenarios. The time-to-boil results for the Byron and Braidwood pools are comparable to other PWR pools with densified fuel storage.

5.8.3 Imcal Water arn Cladding Temperature Consistent with our approach to make the most pessimistic assessments of temperature, the local water temperature calculations are performed when the pool is at its peak bulk temperature. Thus, the local water temperature evaluation is a calculation of the temperature increment over the theoretical spatially uniform value due to local hot spots (due to the presence of a highly heat emissive fuel bundle). The CFD study has analyzed the maximum decay heat input Case (iii) scenario. In this scenario, a full-core discharge with a 100-hour hold time is considered in which 193 assemblies are transferred to the pool following a previous normal discharge which occurred 17 days earlier.

SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 5-16 Report HI-982083

1 Converged temperature contour and velocity vector plots obtained from the FLUENT model 'are i

presented in Figures 5.8.8 and 5.8.9. The peak local temperature is conservatively estimated to be j t

l- less than 208.8 F. This result ensures local subcooled conditions in all fuel rack cells with an {

adequate margin of safety.

~

The peak cladding temperature is determined for the hottest cell location in the pool as obtained from the CFD model for the Byron and Braidwood pools, which includes the postulated dropped assembly scenario and 50% cell exit blockage. The maximum cladding-tclocal water temperature difference (ATc) is calculated to be 38.3*F. Applying this calculated cladding ATc to the peak local water temperature at the top of the active fuel length, a 245.l*F peak cladding temperature is determined.

This is a few degrees hotter than 239 F local saturation temperature on top of the racks.

DNB Evaluation Under a condition in which fuel cladding temperature exceeds the local saturation temperature, a concem related to departure from nucleate boiling (DNB) must be addressed. During DNB, the surfaces of the fuel rods will experience a sudden, large increase in surface temperature due to vapor blanketing at high heat fluxes. This sudden increase can lead to failure of the cladding material due 1 to extreme thermal stresses, a phenomenon commonly known as burnout.

In order for DNB to occur, the surface heat flux must be greater than a critical flux which is I dependent on local conditions. At oie atmosphere absolute pressure, the critical flux is 5

approximately 11.6x10 97,2 [5.8.1] for water. The critical flux is higher at higher pressures, so using this value for comparison with the surface heat flux is conservative. In English units, the 2

critical flux is approximately 367,718 Btu /ft -hr. This critical flux is over 200 times greater than the peak cladding heat flux in the fuel pool storage condition. Consequently,in view of this creedingly

- large safety margin, DNB is concluded to be non-credible. The fuel cladding will not be subjected to extreme thennal stresses, and the cladding integrity is maintained.

Results of DNB evaluation confirm that adequate cooling at local hot spots on the cladding surfaces is guaranteed by a large margin of sifety. In conclusion, the local water temperature remains subcooled and cladding integrity is ensured under all discharge scenarios.

SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 5-17 Report HI-982083

5.9 ' ' References for Section 5 l

l [5.5.1] " Heat Loss to the Ambient from Spent Fuel Pools: Corrdlation of Theory with l Experiment", Holtec Report HI-90477, Rev. O, April 3,1990.

[5.6.1] Batchelor, G.K., "An Introduction to Fluid Dynamics", Cambridge University Press, 1967, l

[5.6.2] Hinze, J.O., " Turbulence", McGraw Hill Publishing Co., New York, NY,1975.

[5.6.3] Launder, B.E., and Spalding, D.B., " Lectures in Mathematical Models of' l Turbulence", Academic Press, London,1972.

[5.6.4] "QA Documentation and Validation of the FLUENT Version 4.3 CFD Analysis Program", Holtec Report HI-951444.

[5.7.1] Rohsenow, N.M.,'and Hartnett, J.P., " Handbook of Heat Transfer", McGraw Hill Book Company, New York,1973.

~

[5.8.'1] John H. Lienhard, "A heat transfer textbook", Prentice-Hall Inc., (1981). !

l

)

4 SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 5 18 Report H1-982083

p i

Table 5.1.1 PARTIAL LISTING OF RERACK APPLICATION USING <

SIMLAR METHODS OFTHERMAL-HYDRAULIC ANALYSIS l:

PLANT DOCKET NO.

Enrico Fermi Unit 2 USNRC 50-341 Quad Cities 1 and 2 USNRC 50-254,50-265 Rancho Seco USNRC 50-312 t Grand Gulf Unit 1 USNRC 50-416 'l Oyster Creek USNRC 50-219 Pilgrim USNRC 50-293 V.C. Summer USNRC 50-395 i I

Diablo Canyon Units 1 and 2 USNRC 50-275,50-455 Byron Units 1 and 2 USNRC 50-454,50-455

]

Braidwood Units 1 and 2 USNRC 50-456,50-457 Vogtle Unit 2 USNRC 50-425 l'

St. Lucie Unit 1 USNRC 50-425 Millstone Point Unit 1 USNRC 50-245 D.C. Cook Units 1 and 2 USNRC 50-315,50-316

'e Indian Point Unit 2 USNRC 50-247 j Three Mile Island Unit 1 USNRC 50-289 J.A. FitzPatrick USNRC 50-333 f Shearon Harris Unit 2 USNRC 50-401 Hope Creek USNRC 50-354 ;

Kuosheng Units I and 2 Taiwan Power Company ,

Chinshan Units I and 2 Taiwan Power Company i

S I

f SH ADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 5-19 Report 111982083 l

l L '

i

F:

4 Table 5.1.1 (continued) ;

PART.AL LISTING OF FUEL RERACK APPLICATIONS USING SIMILAR METHODS OF THERMAL-HYDRAULIC ANALYSIS .

PLANT DOCKET NO.

Ulchin Unit 2 Korea Electric Power Corporation Laguna Verde Units 1 and 2 Comision Federal de E!cetricidad

' Zion Station Units I and 2 USNRC 50-295,50-304 ;

Sequoyah USNRC 50-327,50-328 La Salle Unit One USNRC 50-373 Duane Arnold USNRC 50-331 Fort Calhoun USNRC 50-285 Nine Mile Point Unit One USNRC 50-220 i

Beaver Valley Ur.it One USNRC 50-334

~

Limerick Unit 2 USNRC 50-353 Ulchin Unit 1 Korea Electric Power Corporation W aterford 3 USNRC 50-382 i

SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec Internationa 5-20 Report HI-982083

,l

Table 5.2.1.

BYRCN AND BRAIDWOOD SPENT FUEL POOLS HEAT EXCHANGER DESIGN DATA Type- Shell and Tube Quantity 2 6

Heat Duty 15.S33 x 10 Btu /hr Tubeside

Fluid Flow 2.23 x 106 lb/hr

Pool Water Inlet Temperature 120 F e Outlet Temperature 112.9 F ,

2 e Fouling Factor 0.0005 ft -hr- F/ Btu Shellside

Fluid Flow 2.72 x 106 lb/hr e Coolant Inlet Temperature 105 F !

. Outlet Temperature 110.82 2

Fouling Factor 0.0005 ft -hr- F/ Btu l

SHADED TEXT CONTAINS PROPRIETARY INFORMATION I Holtec 'nternational 5-21 Report 111-982083

hk 4 I

i l

I

, 'l Table 5.3.1-Byron Unit 1 Past Discharges History NO. OF ASSEMBLIES CYCLE-START DATE END DATE DISCHARGED BYlC1 2/2/85 2/14/87 '88 BYIC2 ' 6/1/87 9/3/88 76 BY1C3 11/9/88 1/5/90 88 BYIC4 3/4/90 ... 9/6/91 88 ,

BYlCd llD/91 2/5/M 88 1'YaC6 4/10/93 9/8/94 92 BY1C7- 11/2/94 4/5/96 76 BYlC8 7/3/96 1In/97 76 TOTAL 672 SHADED TEXT CONTAINS PROPPIETARY INFORMATION Holtec International 5-22 Report HI-982083 I

m ,

s J

I Table 5.3.2 Byron Unit 2 Past Disci arges liistory

, NO. OF. .j i

ASSEMBLIES c CYCLE START DATE END DATE DISCHARGED BY2Cl 2/6/87 1/7/89 88- i l

BY2C2 3/8/89 9/1/90 84 j i

BY2C3 11/21/90 2/28/92 . 84 BY2C4 4/30/92 9/3/93 88 BY2C5, 10/25/93 2/10/95 88 BY2C6 3/25/95 8/8/96 84

.BY2C7 10/5/96 4/10/98 84 l TOTAL 600 SHADED TEXT CONTAINS PROPRIETARY INFORMATION

. Holtec larbrnational 5-23 Report HI-982083 i

i

/ t j

-4

l l

c Table 5.3.3 Braidwood Unit 1 Past Discharges History NO. OF ASSEMBLIES CYCLE START DATE END DATE DISCHARGED R1C1 5/29/87 9/2/89 88 I RIC2 12/15/89 .12/30/90 64 RIC3 5/18/91 9/5/92 84 R1C4 11/3/92 3/4/94 92 RIC5 5/13/94 9/30/95 -80 RIC6 12/14/95 3/29/97 56 RIC7 5/27/97 9/5/98 84 TOTAL 548 r

i 4

SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec Intemational 5-24 ' Report H'-982083 s_

a :

Table 5.3N Braidwood Unit 2 Past Discharges History NO.OF ASSEMBLIES CYCLE- START DATE END DATE DISCHARGED R2Cl 3/8/88 3/16/90 88 R2C2 5/28/90 9/13/91 84 R2C3 11/26/91 3/5/93 '84 R2C4- 5/2/93 10/f194 88 R2C5' 11/17/94 3/16/96 84 R2C6 5/14/96 9/27/97 81 TOTAL 509 I

SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 5 25 Report HI-982083 l

Table 5.3.5 BYRON UNITS 1 AND 2 PROJECTED DISCHARGES CUMULATIVE ENRDATE NO. OF FUEL (MONTH AND ASSEMBLIES INVENTORY IN CYCLE YEAR) DISCHARGED POOL BYlC9 March 99 80 1,352 BY2C8 Octobe:r 99 84 1,436 BYIC10 October 00 84 1,520 BY2C9 April 01 84 1,604 BYICI1 March 02 84 1,688 BY2C10 November 02 84 1,772 BYlCl2 September 03 84 1,856 BY2Cl1 May 04 84 1,940 BYlCl3 March 05 84 2,024 BY2Cl2 November 05 84 2,108 BYlCl4 September 06 84 2,192 BY2Cl3 May 07 84 2,276 BYlC15 March 08 84 2,360 BY2Cl4 November 08 84 2,444 BYIC16 September 09 84 2,516 BY2C15 May 10 84 2,612 BYlCl? March 11 84 2,696 BY2C16 November i1 84 2,780

-BYlC18 September 12 84 2,864 BY2C17 May 13' 84 2,948 Last discharge after which there is a loss of normal discharge capability.

SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 5-26 Report HI-982083

r i

Table 5.3.6 1 BRAIDWO D UNITS 1 AND 2 PROJECTL '; CHARGES END DATE NO. OF CUMULATIVE FUEL (MONTH AND ASSEMBLIES INVENTORY IN l CYCLE YEAR) DISCHARGED POOL R2C7 May 99 84 1,141 l

RIC8 February 00 84 1,225 l l

R2C8 November 00 84 1,309 RIC9 September 01' 84 1,393 )

l R2C9 April 02 84 1,477 I RIC10 February 03 84 1,561 R2C10 October 03 84 1,645 RICll September 04 84 1,729

)

R2Cl1 April 05 84 1,313 l RICl2 February 06 84 1,897 l R2Cl2 October 06 84 1,981 R1Cl3 August 07 84 2,065 R2C13 April 08 84 2,149 R1C14 February 09 84 2,233 R2C14 October 09 84 2,317 RIC15 August 10 84 2,041 R2C15 April i1 84 2,485 RIC16 January 12 84 2,569 R2C16 October 12 84 2,653 RIC17 July 13 84 2,737 R2C17 April 14 84 2,821 RIC18 January 15' 84 2,905 Last discharge after which there is a loss of normal discharge capability.

SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 5-27 Report HI-982083 L.

p I

l Table 5.4.1

SUMMARY

OF TIIERMAL-HYDRAULIC INPUTS FOR BYRON AND BRAIDWOOD SPENT FUEL POOLS PARAMETER VALUE Building Temperature 104 F Reactor Power 3,425 MW Core Size . 193 assemblies Fuel Irradiation Time, EFPD 4.5 years Fuel Transfer Time Normal batch: 10.5 hrs.

Full core: 24.1 hrs.

Radial Peaking Factor 1.70 Total Peaking Factor 2.60 Holtec Racks:

Baseplate flow hole diameter :"

Cell wall flow hole diameter Pedestal flow holes 4 Pedestal flow hole diameter i Overall rack height (including bearing pad) ;. . . _ ,

PN" Cell length Bottom plenum height (min.)

Cell opening (Region I)

(Region II)

Cell pitch ,' ? (Region I, E-b)

- ( - (Region I, N-S)

' (Region II)

Pool Dimensions:

Length 62.0 ft Width 33.08 ft Normal Water depth 39.5 ft Cask Pit Region 13.5 ft. x 13.5 ft.

SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 5-28 Repon 111-982083 l

l L

i Table 5.4.2 BYRON AND BRAIDWOOD FUEL DESIGN DATA WESTINGHOUSE 17x17, FUEL TYPES OFA,17x17 VANTAGE 5 and 17x17 VANTAGE +

Fuel Rod O.D. 0.360" Rod Pitch 0.496" Number of Fuel Rods 264 Active Fuel 12ngth 144" l 1

Number of Guide Tubes 24 Guide Tubes O.D. 0.474" Number ofInstrument Tubes 1 Instrument Tubes O.D. 0.474" ,

1 Number of Grid Spacers ' Structural (8)

Intermediate Flow Mixing (3)

Protection Grid (1)

Grid Spacer Thickness Inner Strap (0.018")

Outer Strap (0.026")

SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 5-29 Report H1-982083

t l-

--Table 5.8.1 !

. RESULTS OF BULK POOL TRANSIENT EVALUATION l-Coincident Coincident Coincident Peak Bulk . Time (Hrs- Exchanger Evaporation Heat Discharge Scenario Temperature after Reactor Heat Removal Removal (Bru/hr)

(F) Shut 6own)

(Btdr) 1 6

Case (i) : 138.32 122 35.15x10 0.84x10 6 (Normal Core ^

Discharge,100 Hour Hold) 1 6 6 Case (ii) . 157.13 133 54.97x10 2.18x10 (Full Core Discharge -

100-Hour Hold) 6 6 Case (iii) 166.49t 132- 64.9x10 3.38x10 (Back-to-Back Discharge,100-Hour Hold) l

-t The peak bulk temperature is liniited to 137.2'F when both cooling trains are in operation.

SHADED TEXT CONTAINS PROPRIETARY INFORMATION

- Holtec Intemational- 5-30 Report HI-982083 L

IL[

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Table 5.8.2 RESULTS OF LOSS-OF-COOLING Case Number Time-to-Boil (Hours)

(Without Make-up. Water)

Case (i) 8.43 Case (ii) 3.82 Case (iii) 2.63t .

t The time-to-boil for this case with both cooling unins in operation is 4.26 hours3.009259e-4 days 0.00722 hours 4.298942e-5 weeks 9.893e-6 months .

SHADED TEXT CONTAINS PROPRIETARY INFORMATION Hollec International - 5-31 Report HI-982083 fal V -

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? 6.0 STRUCTURAL SEISMIC CONSIDERATIONS 6.1- Introduction This chapter considers the stmetural adequacy of the new maximum density spent fuel racks under all loads postulated f" armal, seismic, and accident conditions at Byron and Braidwood Nuclear stations.

1

- As discussed in Chapter 1, the reracking of the Byron and Braidwood pools involves replacement l of existing high-density storage racks with new racks with a slight increase in the total capacity.

, The reracking is being undena':en to remove the Boraflex neutron material from the two pools because of its ongoing degradation and loss in neutron attenuation ability. The new racks, like the j existing racks, will be installed in a free-standing configuration. At the time of the previous I rerack, however, the state-of-the-art limited the seismic evaluation to single rack 3-D simulations. As we discuss in this chapter, it is now possible to model the entire assemblage of rack modules in one comprehensive simulation known as the 3-D Whole Pool Multi-Rack ,

(WPMR) analysis. In order to maintain continuity with the previous analysis methods, both i single rack and WPMR analyses have been performed to establish the structural margins of safety in the Byron /Braidwood racks.

The analyses undertaken to confirm the structural integrity of the racks are in full compliance I

with the USNRC Standard Review Plan (6.1.1) and the OT Position Paper [6.1.2]. For each of the analyses, an abstract of the methodology, modeling assurnptions, key results, and summary of parametric evaluations are presented. Delineation of the relevant criteria are discussed in the text associated with each analysis. !

6.2 Overview of Rack Structural Analysis Methodology The response of a free-standing rack module to seismic inputs is highly nonlinear and involves a complex combination of motions (sliding, rocking, twisting, and turning), resulting in impacts and friction effects. Some of the unique attributes of the rack dynamic behaviorinclude a large fraction of the total structural mass in a confined rattling motion, friction suppon of rack pedestals against lateral motion, and large fluid coupling effects due to deep submergence and moti m of closely spaced prismatic structures.

SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 61 Report HI-982083

- Linear methods, such as modal analysis and response spectrum techniques, cannot accurately simulate the structural response of such a highly nonlinear structure to seismic excitation. An -

accurate simulation is obtained only by direct integration of the nonlinear equations of motion i with the three pool slab acceleration time-histories applied as the forcing functions acting l simultaneously.

1 i Both Whole Pool Multi-Rack (WPMR) and Single Rack (SR) analysis are used in this project to simulate the dynamic behavior of the high density rack structres described in Chapter 2 of this report. The following sections provide the basis for the se';cGon of the appropriate methodology and discussion on its development. !

i ,

6.2.1 Background of Analysis Methodology i l

I Reliable assessment of the stress field and kinematic behavior of the rack modules calls for a l

l conservative dynamic model incorporating all key attributes of the actual structure. This means l that the model must feature the ability to execute the concurrent motion forms compatible with l

l the free-standing configuration of the modules, i-l The model must possess the capability to effect momentum transfers which occur due to rattling i of fuel assemblies inside storage cells and the capability to simulate lift-off and subsequent .

impact of support pedestals with the pool liner (or bearing pad). The contribution of the water mass in the interstitial spaces around the rack modules and within the storage cells must be modeled in 'an accurate manner since erring in quantification of fluid coupling on either side of the actual value is no guarantee of conservatism.

The Coulomb friction coefficient at the pedestal-to-pool liner (or bearing pad) mterface may lie in a rather wide range and a conservative value of friction cannot be prescribed a priori. In fact, a perusal of results of rack dynamic analyses in numerous dockets (Table 6.2.1) indicates that an j upper bound value of the coefficient of friction often maximizes the computed rack l displacements as well as the equivalent clastostatic stresses.

In short, there are a large number of parameters with potential influence on the rack kinematics. ,

The comprehensive structural evaluation must deal with all of these without sacrificing conservatism.

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)

The three-dimensional single rack dynamic model introduced by Holtec International personnel in the Enrico Femti Unit 2 rack project (ca.1980) and used in some 50 rerack projects since that 1

time, including Byron and Braidwood in the late 80s (Table 6.2.1) addresses most of the above j mentioned array of parameters. The details of this methodology are also published in the I permanent literature [6.2.1]. Despite the versatility of the 3-D seismic model, the accuracy of the single rack simulations has been suspect due to one key element; namely, hydrodynamic participation of water around the racks. During dynamic rack motion, hydraulic energy is either l drawn from or added to the moving rack, modifying its submerged motion in a significant l manner. Therefore, the dynamics of one rack affects the motion of all others in the pool.

However, Single Rack analysis is still a va!uable tool to examine the behavior of a rack under i different load conditions. It is used here as a first step in evaluating the racks. WPMR analysis l l

builds upon the Single Rack model. The worst-case loads and stresses that result from these two models are used to determine the structural adequacy of the racks.

4 The 3-D rack model dynamic simulation, involving one or more spent fuel racks, handles the array of variables as follows:

Interface Coefficient of Friction Param'etric runs are made with upper bound and lower bound values of the coefficient of friction (COF). The limiting values are based on experimental data that have been found to be bounded by the values 0.2 and 0.8. Simulations are also performed with the array of pedestals having randomly chosen coefficients of friction in a Gaussian distribution with a mean of 0.5 and lower and upper limits of 0.2 and 0.8, respectively. In the fuel rack simulations, tb, Coulomb friction interface between rack support pedestal and liner is simulated by piecewise linear (friction) elements. These elements function only when the pedestal is physically in contact with the pool liner.

Rack Elastic Behaviol Rack elasticity, relative to the rack base, is included in the model by introducing linear springs to represent the elastic bending action, twisting, and extensions.

Impact Phenomena Compression-only gap elements are used to provide for opening and closing of interfaces such as the pedestal-to-bearing pad interface, and the fuel assembly-to-cell wall interface. These interface gaps are modeled using nonlinear spring elements. The term

" nonlinear spring" is a genene term used to denote the mathematical representation of the condition where a restoring force is not linearly proportional to displacement.

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l Fuel Loadine Scenarios The fuel assemblies are conservatively assumed to rattle in unison, which obviously exaggerates the effect of fuelimpacts against the cell walls. Panial fuel l loadings (e.g., a rack that has fuel assemblies in only half of its cells) are simulated by offsetting l

the center of gravity of the stored fuel mass with respect to the rack center of gravity, as appropriate.

Fluid Couplina Holtec International extended Fritz's classical two-body fluid coupling model to multiple bodies and utilized it to perform the first two-dimensional multi-rack analysis (Diablo Canyon, ca.1987). Subsequently, laboratory experiments were conducted to validate the multi-rack fluid coupling theory. This technology was incorporated in the computer code DYNARACK [6.2.4] which handles simultaneous simulation of all racks in the pool as a Whole Pocl Multi-Rack 3-D analysis. This development was first utilized in Chinshan, Oyster Creek, and Shearon Harris plants in the 80's [6.2.1,6.2.3] and, subsequently, in numerous other rerack projects. The WPMR analyses have corroborated the accuracy of the single rack 3-D solutions in predicting the maximum structural stresses, and also serve to improve predictions of rack kinematics..

The Whole Pool Multi-Rack (WPMR) mo, del used to predict the dynamic behavior of the storage racks contains elements specifically designed to represent the attributes necessary to simulate rack motions during earthquakes. These elements include non-linear springs to develop the interaction between racks, between racks and walls, and between fuel assemblies and rack internal cell walls. Hydrodynamic effects within these interstitial spaces are accounted for using Fritz's classical method which relates the fluid kinetic energy in the annulus due to relative motion to an equivalent hydrodynamic mass.

The modeling technique used was chosen based on the applicable Codes, Standards and Specifications given in Section IV (2) of the NRC guidance on spent fuel pool modifications entitled," Review and Acceptance of Spent Fuel Storage and Handling Applications," dated April 14,1978, which states that " Design...may be performed based upon the AISC specification or Subsection NF requirements of Section III of the ASNE B&PV Code for Class 3 component supports." The rack modeling technique is consistent with the linear suppon beam-element type members covered by these codes.

SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 6-4 Report HI-982083 l

r 1

( Although it is acknowledged that finite element models could be developed using plate and fluid elements which may also provide satisfactory simulated behavior for a single rack, there is no known commercial finite element code which can treat multi-body fluid interaction correctly and sufficiently account for near and far field fluid effects involving many bodies (racks) in a closed pool. It is for this reason that the global dynamic analysis uses the formulation specifically developed and contained within DYNARACK.

1 The computer software validation of the DYNARACK program is documented in validation manual HI-91700. The validation manual demonstrates that the DYNARACK code verification is adequate for engineering applications without further experimental verification.

Foi closely spaced racks, demonstration of kinematic compliance is verified by including all .

modules in one comprehensive simulation using a WPMR model. In WPMR analysis, all rack modules are modeled simultaneously and the coupli,ng effect due to this multi-body motion is included in the analysis. Due to the superiority of this technique in predicting the dynamic behavior of closely spaced submerged storage racks, the Whole Pool Multi-Rack analysis l methodology is used as the principal vehicle for seismic qualification in the Byron /Braidwood )

project.

6.3 Description of Racks and Fuel As discussed in Chapter 3, the proposed rack layouts for Byron and Braidwood are identical. A total of twenty-four racks is proposed to be installed in each pool. Four new racks use a flux-trap design and are referred to as Region I racks. The remaining racks do not utilize flux-traps and are referred to as Region II racks. The dynamic rack models include all twenty-four racks. For dynamic simulations, the dry fuel weight is conservatively taken to be 1600 lbs.

6.4 Synthetic Time-Histories Synthetic time-histories in three orthogonal directions (N-S, E-W, and vertical) are generated in accordance with the provisions of SRP 3.7.1 [6.4.1]. In order to prepare an acceptable set of acceleration time-histories, Holtec Intemational's proprietary code GENEQ [6.4.2) is utilized.

A preferred criterion for the synti etic time-histories in SRP 3.7.1 calls for both the response spectrum and the power spectral density (PSD) corresponding to the generated acceleration time-SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 6-5 Report HI-982083

I l j history to envelope their target (design basis) counterparts with only finite enveloping infractions.

The time-histories for the pools have been generated to satisfy this preferred (and more rigorous)

! criterion.

i The target PSD is generated using a program called GENEQ. GENEQ is a Q.A. validated synthetic time-history generator and has been used by Holtec International to generate statistically independent artificial acceleration time histories in over 40 reracking projects.

GENEQ accepts an initial digitized response spectra as input and generates a new bounding response spectra along with a PSD corresponding to both the target and generated spectra along with an acceleration time history.

To prepare the PSD, the digitized design basis response spectra for the elevation of interest (at the floor of the pool) was initially input to obtain the target response spectra and target PSD and a generated spectra and PSD. If the generated spectra and PSD does not bound the initially input target, then an iterative process begins which involves revising the input spectra (by broadening, increasing peak values, etc.) to establish a bounding spectra and PSD. At the completion of each iterative step the newly generated spectra and PSD are compared against the target spectra and PSD developed from the initially input digitzed design basis response spectra. If necessary, the 3 response spectra data is smoothed to prepare comparable results.

The synthetic time-histories in the three directions also satisfy the requirements of statistical independence mandated by SRP 3.7.1.

1 Figures 6.4.1 through 6.4.3 and 6.4.4 through 6.4.6 provide plots of the time-history I accelerograms that were generated over a 20-second duration for SSE and OBE events, respectively. These artificial time-histories are used in all non-linear dynamic simulations of the racks. !

Results of the correlation function of the three time-histories are given in Table 6.4.1. It is noted that the absolute values of all correlation coefficients are well below 0.15, indicating that the desired statistical independence of the three data sets has been met. l 1

l !

I SHADED TEXT CONTAINS PROPRIETARY INFORMATION 0 Holtec International 6-6 Report HI-982083 ;

I 1 i i !

p i

6.5 ' 3-D Nonlinear Rack Model for Dynamic Analysis 6.5.1 Geno.d kemarks v

The single rack 3-D model of the Byron /Braidwood racks has been prepared with due consideration of the following characteristics, which are typical of high-density modules designed by Holtec Intemational.

i. As a continuous stmeture, the rack possesses an infinite number of degrees-of-freedom (DOF), of which the cantilever beam type modes r.re most pronounced under seismic excitation if the rack is of the honeycomb construction genre. (The Byron /Braidwood racks, like all prior Holtec designs, are of the honeycomb type.)

ii. The fuel assemblies are " nimble" structures with a relatively low beam mode fundamental frequency.

iii. The interstitial gap between the storage cells and the stored fuel assemblies leads to a rattling condition in the storage cells during a seismic event.

2 iv. The lateral motion of the rack due to seismic input is resisted by the pedestal-to-pool slab interfacial friction and is abetted or retarded by the fluid coupling forces I produced by the proximity of the rack to other structures. (The fluid coupling forces are distinct from the nonconservative forces such as fluid " drag" which are, by NRC regulations, excluded from the analysis). The construction of a 3-D single rack dynamic model consists of modeling the rack as a multi-degree-of-freedom I

)

stmeture in such a manner that the selected DOFs capture all macro-motion !

I' modes of the rack, such as twisting, overtuming, lift-off, sliding, flexing, and combinations thereof. Particular attention must be paid to incorporating the potential for the friction-resisted sliding of the rack on the liner, lift-off and i subsequent impact of the pedestals on the slab, collision of the rack with adjacent !

structures, and most important, rattling of the fuel in the storage cells. The dynamic model must also provide for the ability to simulate the scenarios of l I

panially loaded racks with arbitrary loading pattems.

i I

SHADED TEXT CONTAINS PROPRIETARY INFORM ATION Holtec International 6-7 Report HI-982083 i

i l

1 As the name implies, the Single Rack (SR) dynamic model is a 3-D structural model for one rack in the pool. The rack selected for the SR analysis in this project is the one with the most mass, or most non-square cross section (i.e., pool aspect ratio). The dynamic model of this rack,i.e., its structural stiffness characteristics, rattling effect of the stored fuel, etc., can be prepared with extreme diligence in the manner described in the following, resulting in an excellent articulation 1 of the rack structure. Even the fluid coupling effects between the fuel assemblies and the storage cell can be modeled with acceptable accuracy [6.5.2]. If the rack is adjacent to a wall, the fluid )

coupling effects between the reck and the wall can also be set down deterministically because the l wall is a fixed structure. Such a definitive situation does not exist, however, when the !

neighboring structure to the subject rack is another free-standing rack. During a seismic event, the subject rack and the neighboring rack will both undergo dynamic motions which will be governed by the interaction among the inenia, fluid, friction, and rattling forces for each rack.

The fluid coupling forces between two racks, however, depend on their relative motions. j

) Because the motion of the neighboring rack is undefined, it is not possible to characterize the )

hydrodynamic forces arising from the fluid coupling between the neighboring rack and the subject rack. This inability to accurately model the inter-rack fluid coupling effects is a central limitation in the single rack analysis.

1 To overcome this limitation intrinsic to the single rack solutions, an anificial boundary condition, )

1 referred to as the "oM-of-phase" assumption, has been historically made to bound the problem. l l

i In the out-o'f-phase assumption, it is assumed that all racks adjacent to the subject rack are j vibrating 180 out-of-phase, r; sulting in a plane of symmetry between the subject rack and the neighboring rack, across which water will not flow. Thus, the subject rack is essentially surrounded by a fictitious box with walls that are midway to the adjacent racks. Impact with the adjacent rack is assumed to have occurred if the subject rack contacts the " box wall",

in summary,in the out-of-phase motion analysis the analyst makes the election that the adjacent racks are moving at 180 out-of-phase from the subject rack at all times during the seismic event.

This is an artificial technical construct, albeit one that is known to predict rack-to-rack impact conservatively.

However, this assumption also increases the relative contribution of fluid coupling, which depends on fluid gaps and relative movements of bodies, making overall conservatism a less SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 6-8 Report HI-982083

certain assertion. As is well known, the fluid forces between adjacent rack modules can reach

rather large values in closely spaced rack geometries. It is, therefore, essential that the contribution of the fluid forces be included in a comprehensive manner. This is possible only if all racks in the pool are allowed to execute 3-D motion in the mathematical model. For this reason single rack, or even multi-rack models involving only a portion of the racks in the

. pool, are inherently inaccurate. The Whole Pool Multi-Rack model removes this intrinsic limitation of the rack dynamic models by simulating the 3-D motion of all modules-

- simultaneously. The fluid coupling effect, therefore, encompasses interaction between every set of racks in the pool, i.e., the motion of one rack produces fluid forces on all other racks and on the pool walls. Stated more formally, both near-field and far-field fluid coupling effects are included in the WPMR analysis.

Therefore, to maintain consistency with past anrjyses, an array of single rack 3-D simulations were carried out, principally to compare the re.sults (viz., rack-to-rack impact, maximum primary stress levels, pedestal loads, etc.) whh the more definitive WPMR analysis. The description below provides the essentials of the 22 DOF model for a single rack. This model is used in both 3-D single rack simulatiens and as the building block for the more complicated WPMR analyses, which are described later in this chapter.

The dynamic modeling of the rack stmetufe is prepared with special consideration of all ;

nonlinearities and parametric variations. Paniculars of modeling details and assumptions for the 4 rack analysis are given in the following:

a. The fuel rack structure motion is captured by modeling the rack as a 12 degree-of-freedom structure. Movement of the rack cross-section at any height is uescribed by six degrees-of-freedom of the rack base and six degrees-of-freedom at the rack top. In this manner, the response of the module, relative to the baseplate, is captured in the dynamic analyses once suitable springs are introduced to couple the rack degrees-of-freedom and simulate rack stiffness.

b. Rattling fuel assemblies within the rack are modeled by five lumped masses located at H, .75H, .5H, .25H, and at the rack base (H is the rack height measured above the baseplate). Each lumped fuel mass has two horizontal displacement degrees-of-freedom. Vertical motion of the fuel assembly mass is assumed equal to rack vertical motion at the baseplate level. The centroid of each fuel assembly mass can be located off-center, relative to the rack structure centroid at that level, to simulate a partially loaded rack.

SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 6-9 Report HI-982083

c. Seismic motion of a fuel rack is characterized by random rattling of fuel assemblies in their individual storage locations. An upper bound on'the effective cumulative fuel assembly mass is established using the previously described artificial time histories,

d. Fluid coupling between rack and fuel assemblies, and between rack and wall, is simulated by appropriate inertial coupling in the system kinetic energy. Inclusion of these effects uses the methods of (6.5.2,6.5.3] for rack / assembly coupling and for rack-to-rack coupling.

e. Fluid damping and form drag are conservatively neglected.

f. Sloshing is found to be negligible at the top of the rack and is, therefore, neglected in the : analysis of the rack,

g. Potential impacts between the cell walls of the new racks and the contained fuel assemblies are accounted for by appropriate compression-only gap elements between masses involved. The possible incidence of rack-to-wall or rack-to-rack impact is simulated by gap elements at the top and bottom of the rack in two

~

horizontal directions. Bottom gap elements are located at the baseplate elevation.

The initial gaps reflect the presence of baseplate extensions, and the rack stiffnesses are chosen to simulate local structural detail.

h. Pedestals are modeled by gap elements in the vertical direction and as " rigid links" for transferring horizontal stress. Each pedestal support is mathematically linked to the pool liner (or bearing pad) by two friction springs. The spring rate for the friction springs includes any lateral elasticity of the stub pedestals. Local pedestal vertical spring stiffness accounts for floor elasticity and for local rack elasticityjust above the pedestal.

i. Rattling of fuel assemblies inside the storage locations causes the gap between fuel assemblies and cell wall to change from a maximum of twice the nominal gap l to a theoretical zero gap. Fluid coupling coefficients are based on the nominal gap l in order to provide a conservative measure of fluid resistance to gap closure. j l

j. The model for the rack is considered supported, at the base level, on four )

pedestals modeled as non-linear compression only gap spring elements and eight !

piecewise linear friction spring elements; these elements are properly located with !

respect to the centerline of the rack beam, and allow for arbitrary rocking and !

sliding motions. 1 SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 6-10 Report HI-982083

i

)

6.5.21 Element Details ~ ,

1 l

. Figure 6.5.1 showr 3 schematic of the dynamic model of a single rack. The schematic depicts I many of the characteristics of the model including all of the degrees-of-freedom and some of the spring restraint elements.

~

Table 6.5.1 provides a complete listing of each of the 22 degrees-of-freedom for a rack model. )

' Six transitional ' and six rotational degrees-of-freedom (three of each type on each end) describe l the motion of the rack structure. Rattling fuel mass motions (sh'own at nodes 1*,2*,3*,4*, and 5*

~

in Figure 6.5.1) are described by ten horizontal transitional degrees-of-freedom (two at'each of the five fuel masses).' The vertical fuel mass motion is assumed (and modeled) to be the same as

-I that of the rack baseplate. -

1 Figure 6.5.2 depicts the fuel to rack impact springs (used to develop potential impact loads-between the fuel assembly mass and rack cell inner walls)in a schematic isometdc. Only one of the five fuel masses is shown in this figure. Four compression only springs, acting in the j horizontal direction, are provided at each fuel mass. Figure 6.5.3 provides a 2-D schematic elevation of the storage rack model, discussed in more detail in Section 6.5.3. This view shows l the vertical location of the five storage masses and some of the support pedestal spring members. j l

Figure 6.5.4'shows the modeling technique and degrees-of-freedom associated with rack elasticity. In each bending plane a shear and bending spring simulate elastic effects (6.5.4]. .

L Linear elastic springs coupling rack vertical and torsional degrees-of-freedom are also illustrated in this figure.

l Figure 6.5.5 depicts the inter-rack impact springs (used to develop potential impact loads between racks or between rack and wall). The approximate spring contact location and numbering of each impact spring used in the model are shown in Figure 6.8.1 and Figure 6.8.2.

6.5.3 Fluid Coupling Effect In its simplest form, the so-called " fluid coupling effect" [6.5.2,6.5.3] can be explained by considering the proximate motion of two bodies under water. If one body (mass mi) vibrates adjacent to a second body (mass m2), and both bodies are submerged in frictionless fluid, then Newton's equations of motion for the two bodies are:

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1 4

2 (mi + M ) X i+ Mi2 X 2= applied forces on mass mi + O(Xi )

ii M 2i X i+ (m2 + M 22 ) X =2 applied forces on mass m2 + O(X22 ) I X iand X denote 2 absolute accelerations of masses mi and m2, respectively, and the notation 2

j -- O(X ) denotes nonlinear terms. The hydrodynamic coupling effect is shown to be comr,osed of i an added nonlinear term which varies with geometry and a component which varies with the square of velocity. This is easily shown by considering a typical example where fluid coupling l

! plays a significant role. Consider two long beams of length "1" width "h" and a distance "s" apart:

l

/ /y '// [t

/

~~/ h~~

~~/ '~~

/

~~B j /A~~

~~/ // I // V L - It is assumed that s hul which is always the case for spent fuel racks. It is shown by levy~~

~~[6.6.1] that the force exerted by the fluid on " beam" A is given by j~~

\

~~l l~~

~~lh i F,,,,,,, = P ' _ , _ s~~

~~g g j; 12s 2s~~
= Plh' _ , _ s*
F-- ,2, 2, g g The above solution is valid at each instant in time so that as the beams (racks) approach each otherlarger forces result which tend to reduce rack motion and preclude rack-to-rack impact. For conservative results, the rack analyses are based only on the nominal gap that exists prior to any seismic event. ,Therefore the forces exerted have the form.
SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 6-12 Report HI-982083 x
n
~
l F = C(-s ias*) -
' The non-linear term for the case of a prismatic fuel assembly in a square cell has been derived by
' Soler & Singh (1982). " Dynamic Coupling in a Closely Spaced Two-Body System Vibrating in Uquid Medium: The Case of Fuel Racks". ,
i In the actual spent fuel rack analyses, the geometry is more complex but the resulting non-linearities have the same character of an added mass multiplier by the acceleration of the rack plus a velocity squared fluid damping term. As the interstitial gap changes, the resulting fluid mass changes also result in non-linear terms. The non-linear terms, O (Xi 2) and O (X2 2), been consistently neglected in order to maintain the requirement that no credit be taken for fluid damping in the seismic analysis.-
' M ii, M , i2M , and 2i M22 are fluid coupling coefficients which depend on body shape, relative disposition, etc. Fritz [6.5.3] gives data for My for various body shapes and arrangements. The fluid adds mass to the body (M ii ot mass mi), and an inertial force proportional to acceleration of
. the adjacent body (mass m2). Thus, acceleration of one body affects the force field on another.
. This force field is a function of inter-body gap, reaching large values for small gaps. Lateral motion of a fuel assembly inside a storage location encounters this effect. For example, fluid coupling behavic .i will be experienced between nodes 2 and 2* in Figure 6.5.1. The rack analysis also contains inertial fluid coupling terms that model the effect of fluid in the gaps between
. adjacent racks. ,
1 These terms are usually computed assuming that all racks adjacent to the rack being analyzed are vibrating in-phase or 180 degrees out of phase. The WPMR analyses do not require any assumptions with regard to phase.
Rack-to-rack gap elements have initial gaps set to 100% of the physical gap between the racks or between outermost racks and the adjacent pool walls. )
6.5.4 ' Stiffness Element Details 1 Table 6.5.2 lists all spring elements used in the SD 22-DOF single rack model. It helps to explain the stiffness details. Byron and Braidwood are mirror images about the E-W direction.
SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International . 6-13 Report HI-982083
m 1
The analysis for Braidwood serves for Byron as well. In the table, the following coordinate system applies:
x= ' Horizontal axis along Braidwood plant North (Byron, South) y .= Horizontal axis along Braidwood plant West z.=. Vertical axis upward from the rack base l If the simulation model is restricted to two dimensions (one horizontal motion plus one venical motion, for example), for the purposes of model clarification only, then Figure 6.5.3 describes the configuration. This simpler model is used to elaborate on the various stiffness modeling elements.
' Type 3 gap elements modeling impacts between fuel assemblies and racks have local mifuess Ki in Figure 6.5.3. In Table 6.5.2. for example, type 3 gap elements 5 through 8 act on the rattling fuel mass at the rack top. Support pedestal spring rates Ks are modeled by type 3 gap elements 1 through 4, as listed in Table 6.5.2. Local compliance of the concrete floor is included in Ks. The type 2 friction elements listed in Table 6.5.2 are shown in Figure 6.5.3 as Kr. The spring elements depicted in Figure 6.5.4 represent type 1 elements.
Friction at support / liner interface is modeled by the piecewise linear friction springs with suitably large stiffness Kr up to the limiting lateral load :N, where N is the current compression load at the interface between support and liner. At every time-step during transient analysis, the current -
value of N (either zero if the pedestal has lifted off the lmer, or a compressive finite value) is p.nputed. j The' gap element Ks, modeling the effective compression stiffness of the structure in the vicinity of the support, includes stiffness of the pedestal, local stiffness of the underlying pool slab, and local stiffness of the rack cellular structure above the pedestal.
The previous discussion is limited to a 2-D model solely for simplicity. Actual analyses incorporate 3-D motions and include all stiffness elements listed in Table 6.5.2.
SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 6-14. ,
Report HI-982083
~~5 .~~
!6.6 Whole Pool Multi-Rack Methodology L6.6.1E General Remarks
' The single rack 3-D (22-DOF) models for the new racks outlined in the preceding subaction are used as a first step to evaluate the structural integrity and physical stability of the rack modules.
However, prescribing the motion of the racks adjacent to the module being analyzed is an assumption in the single rack simulations that cannot be defended on the grounds of conservatism. For closely spaced racks, demonstration of the kinematic compliance is further verified by including all modules in one comprehensive simulation using a Whole Pool Multi-Rack (WPMR) model.' The WPMR analysis builds on the Single Rack model by simultaneously
' modeling all racks with full consideration of the multi-body fluid coupling effects (discussed in .,
the next subsection).
Recognizing that the analysis work effort must deal with both stress and displacement criteria, the sequence of model development and analysis steps that are undertaken are summarized in the following:
9 Prepare 3-D dynamic models suitable for a time-history analysis of the new maximum density racks. These models include the assemblage of all rack modules in the pool. Include all fluid coupling interactions and mechanical coupling appropriate to performing an accurate non-linear simulation. This 3-D simulation is referred to as a Whole Pool Multi-Rack model.
b. Perform 3-D dynamic analyses on various physical conditions (such as coefficient of friction and extent of cells containing fuel assemblies).- Archive appropriate displacement and load outputs from the dynamic model for post-processing.
c. Perform stress analysis of high stress areas for the limiting case of all the rack dynamic analyses. Demonstrate compliance with ASME Code Section III, Subsection NF limits on stress and displacement.
6.6.2 Multi-Body Fluid Coupling Phenomena DurinF the' seismic event, all racks in the pool are subject to the input excitation simultaneously.
The motion cf each free-standing module would be autonomous and independent of others as ,
long as they do not impact each other and no water is present in the pool. While the scenario of b SHADED TEXT CONTAINS PROPRIETARY INFORMATION
- Holtec International - 6-15 Report HI-982083
inter-rack impact is not a common occurrence and depends on rack spacing, the effect of water )
the so-called fluid coupling effect ) is a universal factor. As noted in Ref. [6.5.2,6.5.3], the fluid forces can reach rather large values in closely spaced rack geometries. It is, therefore,
- essential that the contribution of the fluid forces be included in a comprehensive manner. This is possible only if all racks in the pool are allowed to execute 3-D motion in the mathematical model. For this reason, single rack or even multi-rack models involving only a portion of the racks in the pool, are inherently inaccurate. The Whole Pool Multi-Rack model removes this intrinsic limitation of the rack dynamic models by simulating the 3-D motion of all modules simultaneously. The fluid coupling effect, therefore, encompasses interaction between every set
, of racks in the pool, i.e., the motion of one rack produces fluid forces on all other racks and on the pocl walls. Stated more formally, both near-field and far-field fluid coupling effects are included in the analysis.
The derivation of the fluid coupling matrix (6.6.2] celies on the classical inviscid fluid mechanics principles, namely the principle of continuity and Kelvin's recirculation theorem. While the derivation of the fluid coupling matrix is based on no artificial construct, it has been nevertheless verified by an extensive set of shake table experiments [6.6.2].
6.6.3 Coefficients of Friction To eliminate the last significant element of uncertainty in rack dynamic analyses, multiple simulations are performed to adjust the friction coefficient ascribed to the support pedestal / pool bearing pad interface. These friction' coefficients are chosen consistent with the two bounding extremes from Rabinowicz's data (6.5.1]. Simulations are also performed by imposing intermediate value friction coefficients developed by a random number generator with Gaussian normal distribution characteristics. The assigned values are then held constant during the entire simulation in order to obtain reproducible results. ' Thus, in this manner, the WPMR analysis
. results are brought closer to the realistic structural conditions.
The coefficient of friction ( ) between the pedestel supports and the pool floor is indeterminate.
-According to Rabinowicz [6.5.1], results of 199 tests performed on austenitic stainless steel plates submerged in water show a mean value of to be 0.503 with standard deviation of 0.125.
' It is noted tha' DYNARACK has the capabihty to change the coefficient of friction at any pedestal at each instant of contact based on a random reading of the computer clock cycle. However. exerciLng this option would yield results that could not be reproduced. ~Iherefore. the random choice of coefficients is made only once per run.
SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 6-16 Report HI-982083
-1 . ' . .- : . . - . . .
1 i
Upper and lower bounds (based on twice standard deviation) are 0.753 and 0.253, respectively.
Analyses are therefore performed for coefficient of friction values of 0.2 (low limit),0.8 (upper limit), and for random friction values clustered about a mean of 0.5. The bount 1g values of =
0.2 and 0.8 have been found to envelope the upper limit of rnodule response in previous rerack projects.
The bearing pads, which are inserted between the support pedestals and the pool liner, may require additional shim plates in order to span liner weld seams. These shims will be welded to the bearing pads prior to final installation in the spent fuel pool. The presence of these shims does not affect the range of friction coefficients that is used in the dynamic rack simulations. If sliding does occur, the bearing pad is expected to remain stationary as the support pedestal moves on its surface. This is because the interface between the bearing pad and the support pedestal is the only friction surface that involves two different materials. The liner plate, the bearing pads, and the shim plates are all fabricated with SA240-304 stainless steel, whereas the support pedestal is fabricated with SA564-630 stainless steel.
i 6.6.4 Governing Eauntions of Motion i
i Using the structural model discussed in the foregoing, equations of motion corresponding to each degree-of-freedom are obtained using Lagrange's Formulation [6.6.1]. The system kinetic energy includes contributions from solid structures and from trapped and surrounding fluid. The final system of equations obtained have the matrix form:
~ g2q
[M] . , , , . = @ + [W where:
[M] -
total mass matrix (including structural and fluid mass contributions). The size of this matrix will be 22n x22n for a WPMR analysis (n = .. amber of racks in the model).
q -
the nodal displacement vector relative to the pool slab displacement (the term with q indicates the second derivative with respect to time, i.e., acceleration)
I
[G] -
a vector dependent on the given ground acceleration SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 6-17 Report HI-982083 l I
1
l
[Q] -
a vector dependent on the spring forces (linear and nonlinear) and l the coupling between degrees-of-freedom
~ The above column vectors have length 22n. The equations can be rewritten as follows:
= [M J' [Q] + [M l' [G]
_dt.
This equation set is mass uncoupled, displacement coupled at each instant in time. The numerical solution uses a central difference scheme built into the proprietary computer program DYNARACK [6.2.4]. 1 No corivergence problems were experienced during any of the simulations. As demonstrated during an intensive NRC review of DYNARACK in the V.C. Summer station rerack license (ca.
1983) the central difference iteration scheme used in DYNARACK ensures that the solution will be unconditionally convergent. DYNARACK has been used to perform over 2000 seismic simulations of fuel racks in more than 40 dockets over the past 20 years. Stability is achieved after a cenain minimum time step (interval between solutions) is established. The time step is chosen based on previous experience with the solver and is adjusted during the initial runs if solutions are not obtained.
6.7 Structural Evaluation of Spent Fuel Rack Design 6.7.1 Kinematic and Stress Acceptance Criteria There are two sets of criteria to be satisfied by the rack modules:
~~a. - Kinematic Criteria ,~~
1 An isolated fuel rack situated in the middle of the storage cavity is most !
vulnerable to overturning because such a rack would be hydrodynamically
' uncoupled from any adjacent structures. Therefore, to assess the margin against overturning, a single rack module is evaluated. According to the O.T. Position .
paper (USNRC, ca 1978) the minimum required safety margins under the OBE and SSE events is 1.5 and 1.1, respectively. The maximum rotation of the rack (about its two principal axes) is obtained from a post processing of the rack time history response output. The ratio of the rotation required to produce incipient tipping in either principal plane to the actual maximum rotation in that plane from SHADED TEXT CONTAINS PROPRIETARY INFORMATION l
' 'Holtec International 6-18 Report HI-982083
p the time history solution is the margin of safety. All ratios available for the OBE and SSE events should be greater than 1.5 and 1.1, respectively to satisfy the agulatory acceptance criteria.
~~b. Stress Limit Criteria Stress limits must not be exceeded under the postulated load combinations provided herein.~~
~~L 6.7.2 Stress Limit Evt!uations -~~
L i l \
The stress limits pmsented below apply to the rack structure and are derived from the ASME )
Code,Section III, Subsection NF [6.7.1]. Parameters and terminology are in accordance with the ASME Code. Material properties are obtained from the ASME Code Appendices [6.7.2], and are listed in Table 6.3.1. 1 (i) Normal and Upset Conditions (I2 vel A or Level B) i
~~a. Allowable stress in tension on a net section is:~~
i Fi = 0.6 S y Where, Sy = yield stress at temperature, and Fi is equivalent to primary membrane stress.
b, Allowable stress ir shear on a net section is:
F, = .4 S y L
1
~~c. Allowable stress in compression on a net section F = S, .47 -~~
kA'
< 444 r>
kA/r for the main rack body is based on the full height and cross section of the
{- honeycomb region and does not exceed 120 for all sections.
SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 6-19 Report HI-982083 l'
X= unsupported length of component k= length coefficient which gives influence of boundary conditions. The following values are appropriate for the described end conditions:
= 1 (simple support both ends)
= 2 (cantilever beam)
= % (clamped at both ends) r= radius of gyration of component
~~d. Maximum allowable bending stress at the outermost fiber of a net section, due to flexure about one plane of symmetry is:~~
~~F3= 0.60 S y (equivalent to primary bending)~~
~~c. Combined bending and compression on a net section satisfies:~~
f, C f s, + C,'ir /s, <1
-+
F, D, Fs, D, Fe, I
l where: 1 1
l f, = Direct compressive stress in the section i
fex = Maximum bending stress along x-axis fey = Maximum bending stress along y-axis Cmx = 0.85 l
l C,y = 0.85 j Dx = 1 - (f,/F,x)
Dy = 1 - (f,/F,y) 2 F,x.,y = (B E)/(2.15 (kl/r)2x y) l E= Young's Modulus Where subscripts x,y reflect the particular bending plane.
~~- f. Combined flexure and compression (or tension) on a net section~~

fa + f s, + f av <1.0 0.6S, Fe, Fe, SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 6-20 Report HI-982083
The above requirements are to be met for both direct tension or wmpression.
~~g. Welds Allowable maximum shear stress on the net section of a weld is given by:~~
F = 0.3 S ,
where So is the weld material ultimate strength at temperature. For fillet weld legs in contact with base metal, the shear stress on the gross section is limited to 0.4Sy, where Syis the base material yield strength at temperature.
(ii) level D Service Limits SNtion F-1334 (ASME Section III, Appendix F) [6.7.2], states that the limits for the I2 vel D condition are the minimum of 1.2 (S y/FJ or 0.7 (S,/FJ times the corresponding limits for the level A condition. So is ultimate tensile stress st the specified rack design temperature. Examination of material propenies for 304L stainless steel demonstrates that 1.2 times the yield strength is less than the 0.7 times the ultimate strength.
Exceptions to the above general multiplier are the following:
a) Stresses in shear shall not exceed the lesser of 0.72S yor 0.42So. In the case of the
( austenitic stainless steel material used here,0.72Sy governs.
b) ' Axial Compression Loads shall be limited to 2/3 of the calculated buckling load. ,
1 c) Combined Axial Compression and Bending - The equations for level A conditions j shall apply except that: )
F, = 0.667 x Buckling Load / Gross Section Area, ,
l and the terms F,x and F',y may be increased by the factor 1.65.
d) For welds, the level D allowable maximum weld stress is not specified in Appendix F of the ASME Code. An appropriate limit for weld throat stre ss is conservatively set here as:
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Holtec Intemational 6-21 Report HI-982083
1 2
I F. = (0.3 S.) x factor ;
where:
\
factor = (Level D shear stress limit)/(Level A shear stress limit) 6.7.3 ' Dimensionless Stress Factors l
I For convenience, the stress results are presented in dimensionless form. Dimensionless stress factors are defined as the ratio of the actual developed stress to the specified limiting value. The limiting valuc of each stress factor is 1.0, based on the allowable strengths for each level, for Levels A, B, and D (where 1.2S 3. < .7S ). Stress factors reported are:
1 l
R=i Ratio of direct tensile or compressive stress on a net section to its allowable value !
(note pedestals only resist compression)
R=2 Ratio of gross shear on a net section in the x-direction to its allowable value l
R=3 Ratio of maximum x-axis bending stress to its allowable value for the section R=4 Re'io of maximum y-axis bending stress to its allowable value for the section
'R 5= Combined flexure and compressive factor (as defined in the foregoing) i R=6 Combined flexure and tension (or compression) factor (as defined in the i foregoing)
R'=
7 Ratio of gross shear on a net section in the y-direction to its allowable value 6.7.4 leads and Loading Combinations for Spent Fuel Racks The applicable loads and their combinations which must be considered in the seismic analysis of rack modules is excerpted from Refs. [6.1.2] and [6.7.3]. The load combinations considered are identified below:
l SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 6-22 Report HI-982083
Loading Combination Service level D+L Ie vel A D+L+To D + L + To + E D + L.+ T + E Ievel B D + L + To + Pr -
D + L + T. + E' level D D + L + To + Fa The functional capability of the fuel racks must be demonstrated.
where:
D= Dead weight-induced loads (including fuel assembly weight) .
L= Live lead (not applicable for the fuel rack, since there are no moving objects in the rack load path)
Pr = Upward force on the racks caused by postulated stuck fuel assembly Fa = Impact force from accidental drop of the heaviest load from the maximum possible height.
E= Operating Basis Eanhquake (OBE)
E' = Safe Shutdown Earthquake (SSE)' ,
To =- Differential temperature induced loads (normal operating or shutdown condition based on the most critical transient or steady state condition)
T. = Differential temperature induced loads (the highest temperature associated with the postulated abnormal design conditions)
T, and To produce local thermal stresses. The worst thermal stress field in a fuel rack is obtained when an isolated storage location has a fuel assembly generating heat at maximum postulated rate and surrounding storage locations contain no fuel. Heated water makes unobstructed contact with the inside of the storage walls, thereby producing muimum possible temperature difference between adjacent cells. Secondary stresses produced aro limited to the body of the rack; that is, support pedestals do not experience secondary (thermal) stresses.
SHADED TEXT CONTAINS PROPRETARY INFORMATION Holtec Intemational- 6-23 Report HI-982083
6.8 Parametric Simulations Consideration of the parameters described above results in a number of scenarios for both the WPMR and the Single Rack analyses. The single rack analysis considers only one rack in the analysis model whereas the WPMR analysis considers all racks in the model. Since the proposed pool layout and rack modules for both plants are exactly alike, the set of WPMR and Single Rack
- models developed fer one plant is equally valid for the other plant. The pool layout is shown la Figure 2.1. The rack numbering scheme used in the dynarack model for WPMR simulation i.,
introduced in Figure 6.8.1 along with the gap spring numbering scheme.
The Single Rack analysis considers the heaviest rack module and the rack module with highest aspect ratio (i.e. the ratio of length to width of a rack). Rack K (one of the heaviest racks) and Rack J (featuring the highest aspect ratio) are selected for single rack analysis. The single rack model with the heaviest rack is most likely to produce the largest pedestal loads while the model l
with the highest aspect ratio rack is highly susceptible to large displacements using the Single Rack method. In addition to these Single Rack simulations, a Single Rack mn that exhibits the greatest displacement is re-run solely for the purpose of checking the potential for overtuming. l The table below presents a complete listing of the simulations discussed herein.
, J LIST OF RACK SIMULATIONS Run Model Load Case COF Event 4
1 WPMR Full Pool 0.2 SSE 2 WPMR Full Pool 0.8 SSE 3 WPMR Full Pool Random SSE 4 WPMR Full Pool 0.2 OBE 5 WPMR Full Pool 0.8 OBE 6 WPMR Full Pool Random OBE 7 Single Rack (J) Full Rack 0.8 SSE 8 Single Rack (J) Full Rack 0.2 SSE 9 Single Rack (J) Half Full Rack (symmetric 0.8 SSE about short axis) 10 Single Rack (J) Half-Full Rack (symmetric 0.2 SSE about short axis) . 1 11 Single Rack (J) Nearly Empty 0.8 SSE SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec Intemational 6-24 Report HI-982083
LIST OF RACK SIMULATIONS Run Model Load Case COF Event 12 Single Rack (J) Nearly Empty 0.2 SSE 13 Single Rack (K) Full Rack 0.8 SSE 14 Single Rack (K) Full Rack 0.2 SSE 15 Single Rack (K) Half-Full Rack (symmetric 0.8 SSE about short axis) 16 Single Rack (K) Half-Full Rack (symmetric 0.2 SSE about short axis) 17 Single Rack (K) Nearly Empty 0.8 SSE 18 Single Rack (K) Nearly Empty 0.2 SSE 19 Single Rack (J) Full Rack 0.8 OBE 20 Single Rack (J) Full Rack 0.2 OBE 21 Single Rack (J) Half-Full Rack (symmetric 0.8 OBE about short axis) 22 Single Rack (J)- Half-Full Rack (symmetric 0.2 OBE about short axis) 23 Single Rack (J) Nearly Empty 0.8 OBE 24 Single Rack (J) Nearly Empty 0.2 OBE 25 Single Rack (K) Full Rack 0.8 OBE 26 Single Rack (K) Full Rack 0.2 OBE 27 Single Rack (K) Half-Full Rack (symmetric 0.8 OBE about short axis) i 28 Single Rack (K) Half-Full Rack (symmetric 0.2 OBE about short axis) 29 Single Rack (K) Nearly Empty 0.8 OBE 30 Single Rack (K) Nearly Empty 0.2 OBE 31 Single Rack Full Rack 0.8 SSE Overturning Check (K) ,
I where:
Random = Gaussian distribution with a mean coefficient of friction (COF) of 0.5 and a standard deviation equal to 0.15.
SHADED TEXT CONTAINS PROPRIETA RY INFORMATION Holtec International 6-25 Report HI-982083 1
Note that Run No. 31 is a re-run of Run No.13 except that the rack module in this run is simulated as an isolated rack in the pool as described earlier in subsection 6.7.1.
6.9 - Time History Simulation Results The results from the DYNARACK runs may be seen in the raw data output files. However, due to the huge quantity of output data, a post-processor is used to scan for worst case conditions and develop the stress factors. Further reduction in this bulk of information is provided in this section by extracting the worst case values from the parameters of interest; namely displacements, support pedestal forces, impact loads, and stmss factors. This section also- <
l summarizes other analyses performed to develop and evaluate structural member stmsses which l are not determined by the post processor. For each table, the Pool Condition /COF column refers to whether the pool is full, half full or nearly empty (a few cells loaded). COF is the interface j coefficient of friction discussed in subsection 6.2.1. The " Rack" column denotes racks by number (applicable to the DYNARACK model) and by letter (applicable to the pool layout drawing).
l 6.9.1 Rack Displacements l
l A tabulated summary of the maximum displacement for each simulation is provided below. It is noted that all of the maximum displacements occurred at the tops of the storage racks, as expected, from swaying, bending and tipping. The location / direction terms are defined as follows:
uxt, uyt = - displacement of top corner of rack, relative to the slab, in the North-South and East-West directions, respectively, in the Braidwood pool. The maximum ;
displacements for every simulation, including the single rack tipover mn, occurred l l at the top of the racks shown in the last table column.
l
~
SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International '6-26 Report HI-982083
l RACK DISPLACEMENT RESULTS Run Model Pool Event Max. Direction Rack Condition / Disp. (in) l COF l
i 1 WFMR Full /0.2 SSE 0.688 uyt 14(P) 2 WPMR Full /0.8 SSE 0.865 uxt 12(M) .
3 WPMR Full / Rand. SSE 0.823 uxt 12(M) 4 WPMR Full /0.2 OBE 0.468 uyt 12(M) 5 WPMR Full /0.8 OBE 0.467 uyt 12(M) 6 WPMR Full / Rand. OBE 0.467 uyt 12(M) 7 Single Rack Full /0.8 SSE 0.234 uyt 9(J) 8 Single Rack Full /0.2 SSE 0.207 uyt 9(J) 9 Single Rack Half /0.8 SSE 0.124 uyt 9(J)
Single Rack Half /0.2 SSE 0.116 10 uyt 9(J) )
11 Single Rack Empty /0.8 SSE 0.0256 uyt 9(J) 12 Single Rack Empty /0.2 SSE 0.0232 uyt 9(J) l Single Rack Full /0.8 SSE 0.162 uxt 13 10(K) l l 14 Single Rack Full /0.2 SSE 0.143 uxt 10(K) 15 Single Rack Half /0.8 SSE 0.0766 uxt 10(K) 16 Single Rack Half /0.2 SSE 0.0682 uyt 10(K) 17 Single Rack Empty /0.8 SSE 0.0262 uyt 10(K) 18 Single Rack Empty /0.2 SSE 0.0265 uyt 10(K) 19 Single Rack Full /0.8 .OBE 0.09 uxt 9(J) 20 Single Rack Full /0.2 OBE 0.0897 uxt 9(J) 21 Single Rack Half /0.8 OBE 0.0665 uxt 9(J) 22 Single Rack Half /0.2 OBE 0.0634 uxt 9(J) 23 Single Rack Empty /0.8 OBE 0.0118 uxt 9(J) 24 Single Rack Empty /0.2 OBE 0.0121 uxt '
9(J) 25 Single Rack Full /0.8 OBE 0.0915 uyt 10(K) 26 Single Rack Full /0.2 OBE 0.072 uyt 10(K) 27 Single Rack Half /0.8 OBE 0.058 uyt 10(K)
SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 6-27 Report HI-982083
RACK DISPLACEMENT RESULTS Run Model Pool Event Max. Directin Rack
, Condition / Disp. On)
.C9f 28 Single Rack Half /0.2 - OBE 0.050 uyt 10(K) 29 Single Rack Empty /0.8 OBE 0.0148 uyt 10(K) 30 Single Rack Empty /0.2 OBE 0.0145 uyt 10(K) q 31 Single Rack Single Rack SSE 1.500 uyt 10(K) 1 Ovenurning Check The table shows that the maximum rack displacement is 1.50 inches (Ruri 31). With this given value, an evaluation of rack overtuming is performed. The factor of safety obtained from this evaluation is 61, which is much higher than th~e prescribed limit of 1.5 for OBE conditions. This indicates that tipover is not a concem. Table 6.9.1 shows the maximum calculated and maximum allowed rotation of the rack.
The maximum rack displacements at the baseplate elevation are 0.5975 inches (Run No.1) and 0.2358 inches (Run No. 4) for the SSE event and the OBE event, respectively.
I The displacement shape of each rack, from the baseplate elevation to the top of the rack, is nearly l
linear. This indicates tha the primary displacement mode of the fuel racks is rigid body motion (i.e., sliding and tilting). In other words, the clastic deformation of the cell stmeture due to bending is negligible compared to the rigid body displacements.
6.9.2 Pedestal Venical Load Pedestal number 1 for each rack is located in the northeast and the southwest corner of racks of the Braidwood and Byron station, espectively. Numbering increases counterclockwise around the 'petiphery of each rack. The following bounding vertical pedestal foires are obtained for each run:
l 1
SHADED TEXT CONTAINS PROPRIETARY INFORMATION !
' Holtec International 6-28 Report HI-982083 j i
~
i l
MAXIMUM VERTICAL LOADS Run Model Pool Event Max. Vertical Rack Condition / Load (Ib)
COF 1 WPMR Full /0.2 SSE 198000 18(T) 2 WPMR Full /0.8 SSE 229000 18(T) 3 WPMR Full / Rand. SSE 238000 2(B) l 4 WPMR Full /0.2 OBE 131000 23(Y) j 5 WPMR Full /0.8 OBE 144000 11(L) 6 WPMR Full / Rand. OBE 144000 11(L) 7 Single Rack Full /0.8 SSE 169000 9(y) 8 Single Rack Full /0.2 SSE 131000 9(y) j 9 Single Rack Half /0.8 SSE 78200 9(y) 10 Single Rack Half /0.2 SSE 68700 o(y) ;
11 Single Rack Empty /0.8 SSE 15000 9(y) 12 Single Rack Empty /0.2 SSE 13900 9(y)
Single Rack Full /0.8 SSE 109000 13 10(K) 14 Single Rack Full /0.2 SSE 101000 10(K) i 15 Single Rack Half /0.8 SSE 52900 10(K)
Single Rack Half /0.2 SSE 56500
( 16 10(K) 17 Single Rack Empty /0.8 SSE 206M 10(K) ;
Single Rack Empty /0.2 SSE 19600
! 18 10(K) j 19 Single Rack Full /0.8 OBE 92300 9(y) j 20 Single Rack Full /0.2 OBE 92500 9(y) 21 Single Rack Half /0.8 OBE 51400 9(y) 22 Single Rack Half /0.2 OBE 51300 9(y) 23 Single Rack Empty /0.8 OBE 10800 9(3) 24 Single Rack Empty /0.2 OBE 10900 9(y) 25 Single Rack Full /0.8 OBE 81400 10(K) 26 Single Rack Full /0.2 OBE 79700 10(K) 27 Single Rack Half /0.8 OBE 42800 10(K)
SHADED TEXT CONTAINS PROPRIETARY INFORMATION HeLc International 6-29 Report HI-982083
MAXIMUM VERTICAL LOADS Run Model Pool Event Max. Vertical Itack Condition / Load Oh)
COF 28 Single Rack Half /0.2 OBE 42800 10(K) 14500 29 Single Rack Empty /0.8 OBE J(K) 30 Single Rack Empty /0.2 OBE 14300 10(K)
The highest maximum vertical pedestal loads from all simulations for the SSE and OBE conditions are 238,000 lbs and 144,000 lbs, respectively. The effect of these loads is evaluate'd in bearing pad and rack fatigue analyses.
6.9.3 Pedestal Friction Forces The maximum (x or y direction) shear load bounding all pedestals in the simulation are reponed below and are obtained by compilation of the complete tabular data produced by time history solution of whole pool multirack and single rack simulations.
MAXIMUM NORIZONTAL LOADS Run Model Pool Event Max. Shear Rack Condition / Load Ob)
COF 1 WPMR Full /0.2 SSE 33200 20(V) 2 WPMR Full /0.8 SSE 101000 11(L) 3 WPMR Full / Rand. SSE 83300 3(C) 4 WPMR Full /0.2 OBE 24200 16(R) 5 WPMR Full /0.8 OBE 50100 3(C) 6 WPMR Full / Rand. OBE 48500 4(D) 7 Single Rack Full /0.8 SSE 55200 9(y) 8 Single Rack Full /0.2 SSE 24600 9(y) 9 Single Rack Half /0.8 SSE 33800 9(y) 10 Single Rack Half /0.2 SSE 13200 9(y) 11 Single Rack Empty /0.8 SSE 4990 9(y)
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L MAXIMUM HORIZONTAL LOADS Run Model Pool Event Max. Shear Rack Condition / Load Ob) l COF 12 Single Rac.k Empty /0.2 SSE 2740 9(y) l 13 Single Rack Full /0.8 SSE 50000 10(K) 14 Single Rack Full /0.2 SSE 20000 10(K)
! Single Rack Half /0.8 SSE 19200 15 10(K) 16 Single Rack Half /0.2 SSE 9720 10(K) _
17 Single Rack Empty /0.8 SSE 6270 10(K) 18 Single Rack Empty /0.2 SSE 3600 l 10(K) 19 Single Rack Full /0.8 OBE 10000 9(y) 20 Single Rack Full /0.2 OBE 16000 9(y) 21 Single Rack Fw]f/0.8 OBE 7540 9(y) 22 Single Rack Half /0.2 OBE 10100 9(y)
' 23 Single Rack Empty /0.8 OBE 1380 9(3) 24 Single Rack Empty /0.2 OBE 1940 9(y) 25 Single Rack Full /0.8 OBE 8660 10(K) 26 Single Rack Full /0.2 OBE 11800 go(g) 27 Single Rack Half /0.8 OBE 14600 10(K) 28 Single Rack Half /0.2 OBE 8170 10(K) 29 Single Rack Empty /0.8 OBE 2130 10(K) 30 Single Rack Empty /0.2 OBE 2390 10(K)
The largest horizontal pedestal load of 101,000 lbs occurs in run 2. The effect of this load is evaluated in the liner fatigue analysis.
6.9.4 Rack Impact Loads I
A freestanding rack, by definition, is a stnicture subject to potential impacts during a seismic l
event. Impacts arise from rattling of the fuel assemblies in the storage rack locations and,in !
some instances, from localized impacts between the racks, or between a peripheral rack and the l
. pool wall. j I SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 6-31 Report HI-982083 i
Results of simulations predict no impact (in the rack cellular region) between racks, or with walls under any simulation. The time history solution does indicate the occurrence of localized impacts
' at rack baseplate location. The maximum local rack-bottom impact force from each set are reported as follows:
Run hnpact Force,Ibf Analysis 2 80,070 WPMR 8 13,980 SINGLE RACK The rack baseplates, which am manufactured from a continuous steel plate, which is 0.75 inch thick, can sustain impacts at the baseplate level that are greater than the forces listed above. The
'* compressive stress in the baseplate due to the maximum impact load of 80,070 lbs is 8,897 psi, which is less than the yield stmss of the baseplate material (21,300 psi). Therefom, the calculated rack-to-rack impact loads at the baseplate are acceptable.
6.9.4.1 Fuel to Cell Wall Impact Loads As discussed in subsection 6.5.1 the fuel assemblies are modelled using five lumped masses.
Each of these masses interacts with the rack via four nonlinear compression only spring elements. These elements are termed nonlinea , since they have the capability of being inactive
~ (non-load bearing) until the fuel assembly mass comes in contact with the cell wall. The loads developed by these elements are conservative, since the actual assembly to cell wall impacts will be experien'c ed at the assembly spacer grids. Since the number of spacer grids are greater than the number of lumped masses used in the model, the impact loadings will actually occur at more locations, resulting in lower loads at each point of contact.
The DYNARACK program produces a complete time history of the loadings within these non-linear compression only gap / spring elements and archives the results for later review or post-processing. Post-processors enable scanning of the large number of time steps for instants where loads are ac+ually present and allows for easy retrieval of the bounding values.
Fuel assembly integrity is assured by comparing the calculated impact load agdast
- manufacturers test data for assembly grid side loadings. Additionally, it should be noted that the impact loads mperienced by the fuel assemblies from postulated seismic event during storage in l
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the racks is expected to be exceeded by the loadings experienced during service within the reactor under similar events as previously analyzed and accepted during original plant licensing.
The cell wall integrity is determined by comparison of the calculated load with an allowable impact load developed using plastic analysis of the local cell wall impact zone.
The fluid coupling between the fuel assembly and the cell wall is treated by inenial coupling in the system kinetic energy. The methodology is taken from classical mechanics as described by Fritz in, 'The Effects of Liquids on the Dynamic Motions of Immersed Solids" [6.5.3]and by Singh and Soler in " Dynamic Coupling in a Closely Spaced Two-Body System Vibrating in Liquid Medium: The Case of Fuel Racks" (6.5.2].
A review of all simulations performed allows determination of the maximum instantaneous ,
impact load between fuel assembly and fuel cell wall at any modeled impact site. The maximum fuel to cell wall impact load values are reported in the following table.
FUEL-TO-CELL WALL IMPACT Run Model Pool Event Impact Load Rack Condit!rg/ Ub)
COF 1 WPMR Full /0.2 SSE 822 4(D) 2 WPMR Full /0.8 SSE 830 4(D) 3 WPMR Full / Rand. SSE 828 4(D) 4 WPMR Full /0.2 OBE 428 10(K) 5 WPMR Full /0.8 OBE 458 4(D) 6 WPMR Full / Rand. OBE 458 4(D) 7 Single Rack Full /0.8 SSE 559 9(y) 8 Single Rack Full /0.2 SSE 570 9(y) 9 Single Rack Half /0.8 SSE 529 9(y) 10 Single Rack Half /0.2 SSE 514 9(y) 11 Single Rack Empty /0.8 SSE 485 9(y) 12 Single Rack Empty /0.2 SSE 565 9(y)
Single Rack Full /0.8 SSE 829 13 10(K)
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FUEL-TO-CELL WALL IMPACT Run Model Pool Event IMDact Load Rack 1
Condition / db)
COF 14 Single Rack Full /0.2 SSE 822 16(K) i 15 Single Rack Half /0.8 SSE 780 to(g) 16 Single Rack Half /0.2 SSE 559 10(K)
' Single Rack Empty /0.8 SSE 808 17 10(K) 18 Single Rack Empty /0.2 SSE 448 10(K) 19 Single Rack Full /0.8 OBE 292 9(y) 20 Single Rack Full /0.2 OBE 291 9(y) 21 Single Rack Half /0.8 OBE 329 9(y) 22 Single Rack Half /0.2 OBE 341 9(3) 23 Single Rack Empty /0.8 OBE 1148 9(y)
24 Single Rack Empty /0.2 OBE 1113 9(y) 25 Single Rack Full /0.8 OBE 387 10(K) 26 Single Rack Full /0.2 OBE 389 10(K) 27 Single Rack Half /0.8 - OBE 893 10(K) 28 Single Rack Half /0.2 OBE 893 10(K)
Single Rack Empty /0,8 OBE 1118 29 10(K) 30 Single Rack Empty /0.2 OBE 1120 10(K)
The maximum fuel-to< ell wall impact load is recorded to be 1,148 lbs during mn no. 23. The structural integrity of the cell wall under the impact of this load is evaluated. The discussion of this evaluation is provided in section 6.10.3 of this report.
The permissible lateral load on an irradiated spent fuel assembly has been studied by the Lawrence Livermore National Laboratory. The LLNL report [6.10.1] states that "...for the most vulnerable fuel assembly, axial buckling varies from 82g's at initial storage to 95g's after 20 years' storage. In a side drop, no yielding is expected below 63g's at initial storage to 74g's after 20 years' [ dry] storage". The most significant load on the fuel assembly arises from rattling during the seismic event. For the fiv'e lumped mass model, the limiting lateral load, therefore, is equal to F,, where SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 6-34 Report HI-982083
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F, = (w x a)/5 where:
w= weight of one fuel assembly (upper bound value = 1600 lbs) a= permissible lateral acceleration in g's (a = 63)
Therefore F, = 20,160 lbs.
The maximum fuel-to-cell wall impact force from the array of parametric runs listed in the above table is 1,148 lbs. Therefore, the nominal factor of safety against Spent Nuclear Fuel (SNF) failure is computed to be 17. ,
-6.9.5 Rack Vertical Displacement A tabulated summary of the maximum vertical displacement for each simulation is provided below. Note that these displacements represent the rack lift-off during the seismic event, as a result of rocking, sliding, swaying, bending, and tipping behavior of the rack module.
RACK VERTICAL DISPLACEMENT RESULTS Run Model Pool Condition / Event Max. direction Rack COF Vertical Disp. (in)
WPMR Full /0.2 SSE 0.0998 upward 23/Y 1 )
2' WPMR Full /0.8 SSE 0.1334 upward 2/B 3 WPMR Full / Rand. SSE 0.1301 upward 11/L 4 WPMR Full /0.2 OBE 0.0754 upward 21/W 5' WPMR Full /0.8 OBE 0.0766 upward 21/W 6 WPMR Full / Rand. OBE 0.0761 upward 21/W 7 Single Rack Full /0.8 SSE 0.087 upward 9(y) ]
8 Single Rack Full /0.2 SSE 0.087 upward 9(y) l 9 Single Rack Half /0.8 SSE 0.043 upward 9(y) r,iA26 upward 10 Single Rack Half /0.2 SSE 9(y)
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RACK VERTICAL DISPLACEMENT RESULTS Run Model Pool Condition / Event Max. direction Rack COF Vertical Disp. (in) 11 Single Rack Empty /0.8 SSE 0.0084 upward 9(y) 12 Single Rack Empty /0.2 SSE 0.0084 upward 9(y) 13 Single Rack FulUO.8 SSE 0.0443 upward 10(K) 14 Single Rack FulUO.2 SSE 0.0431 upward 10(K) 15 Single Rack Half /0.8 SSE 0.020 upward 10(K) 16 Single Rack Half /0.2 SSE 0.0219 upward 10(K) 17 Single Rack Empty /0.8 SSE 0.0070 upward 10(K) 18 Single Rack Empty /0.2 SSE 0.0069 upward 10(K) 19 Single Rack Full /0.8 OBE 0.0675 upward 9(y) 20 Single Rack Full /0.2 OBE 0.0675 upward 9(y) 21 Single Rack Half /0.8 OBE 0.0343 upward 9(y) 22 Single Rack Half /0.2 OBE 0.0343 upward 9(y) 23 Single Rack Empty /0.8 OBE 0.0069 upward 9(y) 24 Single Rack Empty /d.2 OBE 0.0069 upward 9(y) 25 Single Rack Full /0.8 OBE 0.0344 upward 10(K) 26 Single Rack Full /0.2 OBE 0.0344 upward 10(K) 27 Single Rack Half /0.8 OBE 0.0176 upward 10(K) 28 Single Rack Half /0.2 OBE 0.0176 upward 10(K) 29 Single Rack Empty /0.8 OBE 0.0058 upward 10(K) 30 Single Rack Empty /0.2 OBE 0.0058 upward 10(K)
-31 Single Rack Single Rack SSE 0.3199 upward 10(K.)
Overtuming Check -
The maximum vertical displacement is 0.3199 inches which occurs in run No. 31.
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6.10 Rack Stmetural Evaluation
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6.10.1 Rack Stress Factors I
With time history results available for pedestal normal and lateral interface forces, the maximum values for the previously defined stress factors can be determined for every pedestal in the array l of racks. With this information available, the stmetural integrity of the pedestal can be assessed i and reported. The nel section maximum (in time) bending moments and shear forces can also be I i
determined at the bottom casting-rack cellular structure interface for each spent fuel rack in the pool. With this information in hand, the maximum stress in the most stressed rack cell (box) can be evaluated.
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i An evaluation of the stress factors for all of the simulations performed leads to the conclusion j that all stress factors are less than the mandated limit of 1.0 for the load cases examined. From
' all of the simulations, the bounding stress factors for each run, in either the cellular or the 1
pedestal region, are summarized below :
MAXIMUM STRESS FACTORS Run Model Pool Event Stress Factor Rack / Factor l
Condition / Type COF 1 WPMR Full /0.2 SSE 0.252 18(T)/R5 2 WPMR Full /0.8 SSE 0.443 3(C)/R6 l 3 WPMR Full / Rand. SSE 0.407 ll(L)/R6 4 WPMR Full /0.2 OBE 0.428 12(M)/R6 5 WPMR Full /0.8 OBE 0.494 12(M)/R6 i 6 WPMR Full / Rand. OBE 0.459 12(M)/R6 7 Single Rack Full /0.8 SSE 0.302 9(J)/R6 8 Single Rack Full /0.2 SSE 0.173 9(J)/R6 9 Single Rack Half /0.8 SSE 0.140 9(J)/R6 10 Single Rack Half /0.2 SSE 0.091 9(J)/R5 11 Single Rack Empty /0.8 SSE 0.025 S(J)/R6 12 Single Rack Empty /0.2 SSE 0.018 9(J)/R5 SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec Intemational 6-37 Report HI-982083 1
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MAXIMUM STRESS FACTORS j Run Model Pool Event Stress Factor Rack / Factor !
Condition / Type COF 13 Single Rack Full /0.8 SSE 0.248 10(K)/R6 14 Single Rack Full /0.2 SSE 0.125 10(K)/R5 0.087 10(K)/R6 l 15 Single Rack Half /0.8 SSE 16 Single Rack Half /0.2 SSE 0.067 10(K)/R5,R6 17 Single Rack Empty /0.8 SSE 0.028 10(K)/R5 18 Single Rack Empty /0.2 SSE 0.023 10(K)/R5 19 Single Rack Full /0.8 OBE 0.228 9(J)/R5 3 20 Single Rack Full /0.2 OBE 0.229 9(J)/R5 0.131 1 21 Single Rack Half /0.8 OBE 9(J)/R5 22 Single Rack Half /0.2 OBE 0.130 9(J)/R5 23 Single Rack Empty /0.8 OBE 0.028 9(J)/R5 24 Single Rack Empty /0.2 ' OBE 0.029 9(J)/R5 j 25 Single Rack Full /0.8 OBE 0.170 10(K)/R5 26 Single Rack Full /0.2 OBE 0.169 10(K)/R5 27 Single Rack Half /0.8 OBE 0.137 10(K)/R6 28 Single Rack Half /0.2 OBE 0.103 10(K)/R5 29 Single Rack Empty /0.8 OBE 0.031 10(K)/R5,R6 30 Single Rack Empty /0.2 OBE 0.032 10(K)/R5 The maximum stress factor scanned from above table is 0.443 for SSE and 0.494 for OBE, which is below the prescribed Code limit of 1.0. Therefore, the stress allowables are indeed satisfied for the load levels considend for every limiting location in every rack in the array.
6.10.2 Pedestal Thread Shear Stress The complete post-processor results give thread stresses under faulted conditions for every pedestal for every rack in the pool. The average shear stress in the engagement region is given below for the limiting pedestal in each simulation.
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! THREAD SHEAR STRESS l
l Run Model Pool Event Stress (psi) Rack
! Condition / Numher l- COF l 1 WPMR Full /0.2 SSE 8404 18(T) 2 WPMR Full /0.8 SSE 9719 18(T) l 3 WPMR Full / Rand. SSE 10102 2(B) 4 WPMR Full /0.2 OBE 5560 23(Y 5 WPMR Full /0.8 OBE 6112 11(L) 6 WPMR Full / Rand. OBE 6112 ll(L) 7 Single Rack Full /0.8 SSE 7173 9(y) 8 Single Rack Full /0.2 SSE 5560 9(y) l 9 Single Rack Half /0.8 SSE 3319 9(y) 10 Single Rack Half /0.2 SSE 2916 9(y) 11 Single Rack Empty /0.8 SSE 637 9(y) l 12 Single Rack Empty /0.2 SSE 590 9(y) 13 Single Rack Full /0.8 SSE 4626 10(K) 14 Single Rack Full /0.2 SSE 4287 10(K) 15 Single Rack Half /0.8 SSE 2245 10(K) 16 Single Rack Half /0.2 SSE 2398 10(K) 17 Single Rack Empty /0.8 SSE 874 10(K) 18 Single Rack Empty /0.2 SSE 832 10(K) 19 Single Rack Full /0.8 OBE 3918 9(y) 20 Single Rack Full /0.2 OBE 3926 9(y) .
21 Single Rack Half /0.8 OBE 2181 9(y) 22 Single Rack Half /0.2 OBE 2177 9(y) 23 Single Rack Empty /0.8 OBE 458 9(y) 24 Single Rack Empty /0.2 OBE 462 9(y) 25 Single Rack Full /0.8 OBE 3455 10(K) 26 Single Rack Full /0.2 OBE 3383 10(K) 27 Single Rack Half /0.8 OBE 1817 10(K)
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1 THREAD SHEAR STRESS I Run Model Pool Event Stress (psi) Rack Condition / Number i COF 28 Single Rack Half /0.2 OBE 1817 10(K) l 29 Single Rack Empty /0.8 OBE- 615 10(K) 30 Single Rack Empty /0.2 OBE 607 10(K) l l
t The ultimate strength of the internally threaded part of the pedestal is 66,200 psi. The yield stress for this material is 21,300 psi. The allowable shear stress for Level B (OBE) conditions is 0.4 times the yield stress which gives 8,520 psi and the allowable shear stress for level D is 0.72 times the yield stress which gives 15,336 psi. The maximum calculated shear stress value for the SSE is 10,102 psi and 6112 psi for the OBE which are less than their respective allowables.
Therefore, thread shear stresses are acceptable under all conditions.
6.10.3 Local Stresses Due to Impe;ts Impact loads at the pedestal base (discussed in subsection 6.9.2) produce stresses in the pedestal l for which explicit stress limits are prescribed in'the Code. However, impact loads on the cellular l
region of the racks, as discussed in subsection 6.9.4.1 above, prod"ce stresses which attenuate rapidly away from the loaded region. This behavior is characteristic of secondary stresses.
Even though limits on secondary stresses are not prescribed in the Code for Class 3 NF structures, evaluations must be made to ensure that the localized impacts do not lead to plastic deformations in the storage cells which affect the subcriticality of the stored fuel array.
Local cell wall integrity is conservatively estimatG from peak impact loads. Plastic analysis is
.used to obtain the limiting impact load which wotid lead to gross permanent deformation. Table 6.9.1 indicates that the limiting impact load (3,698 lbf, including a safety factor of 2.0) is much .
greater than the highest calculated impact load s :lue (1,148 lbf, see subsection 6.9.4.1) obtained from any of the rack analyses. Therefore, tuci impacts do not represent a concern with respect to fuel rack cell deformation.
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- 6.10.4 Assessment of Rack Fatinue Marnin Deeply'sub' merged high density spent fuel storage racks arrayed in close proximity to each other L - in a free-standing configuration behave primarily as a nonlinear cantilevered structure when subjected to 3-D seismic excitations. In addition to the pulsations in the vertical load at each
- pedestal, lateral friction forces at the pedestal / bearing pad-liner interface, which help prevent or mitigate lateral sliding of the rack, also exert a time-varying moment in the baseplate region of the rack. The friction-induced lateral forces act simultaneously in x and y directions with the
~~requirement that their vectorial sum does not exceed N where is the limiting interface~~
- coefficient of friction and N is the concomitant vertical thrust on the liner (at the given time
. instant).: As the vertical thrust at a pedestal location changes, so does the maximum friction- .
l force, F, that the interface can exen. In other words, the lateral force at the pedestal / liner interface, F,is given by _
- F $; N (r) where N (vertical thrust)is the time-varying function of T. F does not always equal N; rather, pN is the maximum value it can attain at any time. The actual value, of course, is determined by the dynamic equilibrium of the rack structure. In summary, the horizontal friction force at the l pedestal / liner interface is a function of time. Its magnitude and direction of action varies during the canhquake event.
The time-varying lateral (horizontal) and vertical forces on the extremities of the support pedestals produce stresses at the root of the pedestals in the manner of'an end-loaded cantilever. j The stress field in the cellular region of the rack is quite complex, with its maximum values )
located in the region closest to the pedestal. The maximum magnitude of the stresses depends on the severity of the pedestal end loads and on the geometry of the pedestal / rack baseplate region. ;
Alternating stresses in metals produce metal fatigue if the amplitude of the stress cycles is L
sufficiently large. In high density racks designed for sites with moderate to high postulated seismic action, the stress intensity amplitudes frequently reach values above the material-endurance limit, leading to expenditure of the fatigue " usage" reserve in the material.
Because the locations of maximum stress (viz., the pedestal / rack baseplate junction) and the close placement of racks, a post-eanhquake inspection of the high stressed regions in the racks is SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 6-41 Report HI-982083 l
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not feasible. Therefore, the racks must be engineered to withstand rnultiple earthquakes without L reliance on nondestructive inspections for post-earthquake integrity assessment. The fatigue life
~~j. evaluation of racks is an integral aspect of a sound design.~~
The time history method of analysis employed in this report provides the means to obtain a l complete cycle history of the stress intensities in the highly stressed regions of the rack. Having determined the amplitude of the stress intensity cycles and their number, the cumulative damage factor, U, can be determined using the classical Miner's rule U=I A Ni where ni is the number of stress intensity cycles of amplitude 4, 4 and Ni is the permissible number of cycles corresponding to @if rom the ASME fatigue curve for the material of construction. U must be less than or equal to 1.0.
To evaluate the cumulative damage factor, a conservative model of a portion of the spent fuel rack in the vicinity of a support pedestal is constructed. Using the archived results of the spent fuel rack dynamic analyses (pedestal load histories versus time) enables a time-history of stress intensity to be established at the most limiting location. This permits establishing a set of alternating stress intensity ranges versus cycles for an SSE and an OBE event. Based on ASME Code Subsection NF guidelines, the cumulative damage factor (U) is conservatively calculated to l . be 0.950 due to the combined effect of one SSE and twenty OBE events. This value is below the l ASME Code limit of 1.0.
6.10.5 Weld Stresses l Weld locations subjected to significant seismic loading are at the bottom of the rack at the baseplate-to-cell connection, at the pedestal-to-baseplate connection, and at cell-to-cell l
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~ connections. Bounding values of resultant loads are used to qualify these connections. The l
i paragraphs below summarize each of the weld evaluations. The numerical results are also summarized in Table 6.9.1.
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a. BaseDlate-to-Cell Welds The highest predicted weld stress for SSE is calculated from the highest R6 value (provided in 6.10.1 above). The ratio of 2.15 is developed from the differences in material thickness and length versus weld throat dimension and length:
~~RATIO = (D[D$in~~
~~b155 in) / (K6Iis3 in~~
~~0.7071~~
~~ilin) = 2.15 R6 (obe) * [(0.6) Fy]~~
~~RATIO = 0.494 * [0.6~~
~~21300 psi)~~
~~2.15 = 13,574 psi R6 (sse) *. [(1.2) Fy]~~
~~RATIO' = 0.443 * [1.2~~
~~21300 psi)~~
2.15 = 24,345 psi The above calculated values are less than the OBE allowable weld stress value of 19,860 psi and weld stress allowable value of 35,748 psi. Therefore, weld stresses between the baseplate and cell wall base are acceptable.
b. Baseolate-to-Pedestal Welds The maximum weld stress between the baseplate and the support pedestal,15,380 psi under an SSE event and 13,600 psi under an OBE event,is yerified to be less than the allowable values of 35,748 psi and 19,860 psi, respectively.
~~c. Cell-to-Cell Welds Cell-to-cell connections are formed by a series of connecting welds along the cell height.~~
Stresses in storage cell to cell welds develop due to fuel assembly impacts with the cell wall. These weld stresses are conservatively calculated by assuming that fuel assemblies in adjacent cells are moving out of phase with one another so that impact loads in two adjacent cells are in opposite directions; this tends to separate the two cells from each other at the weld.
I Table 6.9.1 gives results for the maximum allowable load that can be transferred by these welds based on the available weld area. An upper bound of the transferred load is also given in Table 6.9.1, and it is much lower than the allowable load. This upper bound value is conservatively c,btained by applying the maximum rack-to-fuel impact load from any simulation in two orthogonal directions simultaneously and multiplying the result by SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec Intemational 6-43 Report HI-982083
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2 to account for the simultaneous impact of two assemblies. An equilibrium analysis at the connection then yields an upper bound of the transferred load. It is seen from the result in Table 6.9.1 that the calculated load is well below the allowable.
6.11 level A Evaluation The Level A condition is not a governing condition for spent fuel racks since the general level of loading is far less than 12 vel B loading. To illustrate this, the heaviest spent fuel rack is considered under the dead weight load. It is shown below that the maximum pedestal load is low and that further stress evaluations are unnecessary.
LEVEL A MAXIMUM PEDESTAL LOAD Dry Weight of a 14X11 Rack = 23,679 lbf Dry Weight of 154 Fuel Assemblies = 246,400 lbf Total Dry Weight = 270,079 lbf Total Buoyant Weight (0.87 X Total Dry Weight) = 234,969 lbf Load per Pedestal = 58,742 lbf The stress allowables for the normal condition is the same as for the upset condition, which resulted in a maximum pedestal load of 144,000 lbs. Since this load (and the corresponding stress throughout the rack members) is greater than the 58,742 lb load calculated above, the
! seismic condition controls over normal (Gravity) condition.
6.12 Hydrodynamic Loads on Pool Walls The maximum hydrodynamic pressure (in psi) that develop between the fuel racks and the spent
~~fuel pool walls correspond to the case in which the racks exhibit the largest displacements. The~~
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' maximum pressure 's computed for both the SSE and OBE cases. The results for these worst case conditions are shown in the table below.
Case Maximum Pressure (psi)
SSE 10.5 i
OBE 7.5 SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 6-44 Report HI-982083
4 These hydrodynamic pressures must be considend in the evaluation of the Spent Fuel Pool structure.
6.13 Thermr! Stresses From Asymmetric Heat Generation
+
. Inter-cell welded joints are examined under the loading conditions arising from thennal effects due to an isolated hot cell. A thermal gradient between cells will develop when an isolated storage location contains a fuel assembly emitting maximum postulated heat, while the .
a surrounding locations are empty. The temperature difference between these cells can be obtained from section 5.8.3 of this repon. The maximum temperature diffemnce between the local water -
temperature and the bulk SFP temperature is 38.3 F.
We can obtain a conservative estirnate of weld stresses along the length of an isolated heated cell ,l l
by using a finite element model (illustrated in figure 6.13.1) which is based on the following assumptions: l (a) The cell walls surrounding the " hot" cell are assumed to be at the exit temperature of the coolant. In actuality, the water temperature rises monotonically from the bulk temperature value at the base plate elevation to its maximum value at exit. By assuming the cell wall tc be at exit temperature of water, the state of computed thermal stress will be maximized.
(b) The cell walls contiguous to the hot cell are assumed to be at the pool bulk temperature.
(c) The connectivity between adjacent cells is through six discrete li9eal welds which is <
explicitly modeled in the finite element model.
(d) !The bottom edges of all cell walls are assumed to be fixed.
- (e) The top edges of all cell walls are assumed to be free. ;
The finite element solution exploits the symmetry of the problem about the two venical planes to
= permit a quaner symmetric model (figure 6.13.1). It is clear from the physics of the problem and ;
~~from the above finite element model that the locations of sharp temperature change, namely the~~
- longitudinal welds, are locations of maximurr. stress which arise from restraint of thermal expansion between adjoining cells.
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The fm' ite element solution confirms this expected result. The maximum weld shear stress, however, is limited to 8,384 psi. Thermal stresses, which are " secondary stresses" in the stress hierarchy of the ASME Code, have no prescribed limit for NF Class 3 structures. Since the maximum shear stress in weld material is less than the code allowable limit of 19,860 (Table -
6.9.1), it is concluded that the " isolated cell" scenario will not lead to any primary yielding in the cell connectivity.
6.14 Overhead Storage The spent fuel racks for Byron and Braidwood are also qualified for two additional storage functions.
The Region II racks are designed to accommodate an Overhead Platform, which has a capacity of 3 tons (dry). The platform is movable, and it can be installed on top of any Region II rack by insening its four support legs mto empty storage cells. The surface of the platform is elevated 13 inches above the top of the rack and measures 52 inches square, which covers a six by six area of cells. Multiple items can be stored on the platform as long as (i) the total dry weight is less than 6,000 lb and (ii) each item completely rests on the storage surface (i.e.,52 inch square area). The stored objects are protected from falling off of the platform by 14 inch high side walls.
Both the Region I and Region II racks, when they are completely empty, are also capable of supponing miscellaneous eqsipment (e.g., tools) directly on top of the storage cells. The object must weigh less than 2,000 lb (dry), and it must span a minimum of three storage cells.
I 6.15 Conclusion l
)
l Thiny discrete freestanding dynamic simulations of maximum density spent fuel storage racks have been performed to establish the structural margins of safety. Of the thiny parametnc analyses, six simulations consisted of modeling all 24 fuel racks in the pool in one comprehensive Whole Pool Multi Rack (WPMR) model. The remaining twenty-six mns were carried out with the classical single rack 3-D model. The parameters varied in the different mns consisted of the rack / pool liner interface coefficient of friction, extent of storage locations occupied by spent nuclear fuel (ranging from nearly empty to full) and the type of seismic input (SSE or OBE). Maximum (maximum in time and space) values of pedestal vertical, shear forces, SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec Intemational 6-46 Report HI-982083 s
s
. displacements and stress factors (normalized stresses.for NF Class 3 linear type structures) have been post-pacessed from the array of runs and summarized in tables in this chapter. The results show that:
(i) All stresses am well below their cormsponding "NF" limits.
-(ii) There is no rack-to-rack or rack-to-wall impact anywhere in the cellular region of the rack modules (iii) The factor of safety against ovenurning of a rack is in excess of 63. i An evaluation of the fatigue expenditure in the most stressed location in the most heavily loaded .
rack module under the combined effect of one SSE and twenty OBE events shows that the Cumulative Damage Factor (using Miner's rule) is 0.950, which is less than the permissible value of 1.0.
An evaluation of the thermal (secondary) stress produced by the condition of maximum thermal gradient (obtained when a maximum heat emitting fuel assembly is stored in a cell surrounded by empty storage locations wherein no heat is generated) was performed. The thermal stresses for which no statutory limit in the code (Section III, Subsection NF, Class 3 Structures) exists, is foun'd to be limited to 8,384 psi, which is well below the allowable limit of 19,860 psi. !
j In conclusion, all evaluations of structural safety, mandated by the OT Position Paper [6.1.2] and the contemporary fuel rack structural analysis practice have been carried out. They demonstrate consistently large margins of safety in all new storage modules.
As a final note, the continited compliance of the installed rack arrays with the licensing basis is an essential part of a plant's safety considerations. Since the fuel racks are free-standing structures, the inter-body spacings in the Byron and Braidwood pools, after a site seismic event, may be different from the as-installed values. A plant's procedures would require a comprehensive survey of the inter-module and module-to-wall gaps absequent to a seismic event. If the gaps are found to have changed, then a re-evaluation of the acceptability of the module layout configuration (using the WPMR model) will be performed to complete a no-significant-hazards evaluation pursuant to 10CFR50.59. The rack modules will be restored to their original installed locations (pre-seismic)if the conclusion of the 650.59 evaluation is non-SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 6-47 Report HI-982083
w
' affimiative, or if the plant elects to skip the analytical evaluation (s50.59) step and move directly to reposition the modules.
' 6.16 ' References for Section 6 '
f
[6.1.1] USNRC NUREG-0800, Standard Review Plan, June 1987. l 1
. - 4'
[6.1.2] (USNRC Office of Technology) "OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications", dated. April 14,1978,and January 18,1979 amendment thereto.
1
^
[6.2.1] Soler, A.I. and Singh, K.P., " Seismic Responses of Free Standing Fuel Rack I
Constructions to 3-D Motions", Nuclear Engineering and Design, Vol. 80, pp.
315-329 (1984).
-l
[6.2.2] . Soler, A.I. and Singh, K.P.,"Some Results from Simultaneous Seismic j Simulations of All Racks in a Fuel Pool", INNM Spent Fuel wtanagement j Seminar X, January,1993.
[6.2.3] Singh, K.P. and Soler, A.I., " Seismic Qualification of Free Standing Nuclear Fuel Storage Racks - the Chin Shan Experience, Nuclear Engineering International, UK (March 1991).
[6.2.4] Holtec Proprietary Repon HI-961465 - WPMR Analysis User Manual for Pre & Post Processors & Solver, August,1997.
-[6.2.5] Holtec Proprietary Report HI-91700 - Validation Manual for DYNARACK.
~
[6.4.1] 'USNRC Standard Review Plan, NUREG-0800 (Section 3.7.1, Rev. 2,1989). j Holtec Proprietary Report HI-89364 - Verification and User's Manual for !
[6.4.2]
Computer Code GENEQ, January,1990. l
[6.5.1]' Rabinowicz, E., " Friction Coefficients of Water Lubricated Stainless Steels for a Spent Fuel Rack Facility," MIT, a repon for Boston Edison Company,1976.
[6.5.2] Singh, K.P. and Soler, A.I., " Dynamic Coupling in a Closely Spaced Tv.o-Body System Vibrating in Liquid Medium: The Case of Fuel Racks," 3rd International Conference on Nuclea- Power Safety, Keswick, England, May 1982.
[6.5.3] ' Fritz, R.J., "The Effects of Liquids on the Dynamic Motions of Immersed Solids," ;
Journal of Engineering for Industry, Trans. of the ASME, February 1972, pp 167- ;
172. ;
' SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 6-48. Report HI-982083
I 1
l 1
[6.6.1] Ixvy, S. and Wilkinson, J.P.D., "The Component Element Method in Dynamics j with Application to Earthquake and Vehicle Engineering," McGraw Hill,1976. l
[6.6.2) Paul, B., " Fluid Coupling in Fuel Racks: Correlation of Theory and Experiment",
(Proprietary). NUSCO/Holtec Repon HI-88243.
[6.7.1] ASME Boiler & Pressure Vessel Code,Section III, Subsection NF,1989 Edition.
[6.7.2]- ASME Boiler & Pressure Vessel Code,Section III, Appendices,1989 Edition.
[6.7.3] USNRC Standard Review Plan, NUREG-0800 (Section 3.8.4, Rev. 2,1989).
l
[6.10.1] Chun, R., Witte, M. and Schwanz, M., " Dynamic Impact Effects on Spent Fuel i Assemblies," UCID-21246, Lawrence Livermore National Laboratory, October 1987. 3 l
l l
t l
l I
)
I SHADED TEXT CONTAINS PROPRIETARY INFORMATION
' Holtec International 6-49 Report HI-982083
1 Table 6.2.1 PARTIAL LISTING OF FUEL RACK APPLICATIONS USING DYNARACK' PLANT DOCKET NUMBER (s) YEAR Enrico Fermi Unit 2 USNRC 50-341 1980 Quad Cities 1 & 2 USNRC 50-254, 50-265 1981 Rancho Seco USNRC 50-312 1982 Grand Gulf Unit 1 USNRC 50-416 1984
, Oyster Creek USNRC 50-219 1984 Pilgrim USNRC 50-293 1985 V.C. Summer USNRC 50-395 1984 Diablo Canyon Units 1 & 2 USNRC 50-275,50-323 1986 Byron Units 1 & 2 USNRC 50-454,50-455 1987 Braidwood Units 1 & 2 USNRC 50-456,50-457 1987 Vogtle Unit 2 USNRC 50-425 1988 St. Lucie Unit 1 USNRC 50-335 1987 Millstone Point Unit 1 USNRC 50-245 1989 l
l Chinshan Taiwan Power 1988 D.C. Cook Units 1 & 2 USNRC 50-315,50-316 1992 )
Indian Point Unit 2 USNRC 50-247 1990 Three Mile Island Unit 1 USNRC 50-289 1991 l James A.FitzPatrick USNRC 50-333 1990 1
Shearon Harris Unit 2 USNRC 50-401 1991 Hope Creek USNRC 50-354 1990 Kuosheng Units 1 & 2 Taiwan Power Company 1990 Ulchin Unit 2 Korea Electric Power Co. 1990 Laguna Verde Units 1 & 2 Comision Federal de Electricidad 1991 SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 6-50 Report HI-982083
rs j Zion Station Units 1 & 2 USNRC 50-295,50-304 1992 i Sequoyah USNRC 50-327,50-328 1992 LaSalle Unit 1 USNRC 50-373 1992 l Duane Arnold Energy Center USNRC 50-331 1992 Fort Calhoun USNRC 50-285 1992 Nine Mile Point Unit 1 USNRC 50-220 1993 Beaver Valley Unit 1 USNRC 50-334 1992 Salem Units 1 & 2 USNRC 50-272,50-311 1993 I l
Limerick USNRC 50-352,50-353 1994 Ulchin Unit 1 KINS 1995 Yonggwang Units 1 & 2 KINS 1996 Kori-4 KINS 1996 )
Connecticut Yankee USNRC 50-213 1996 Angra Unit 1 Brazil 1996 Sizewell B United Kingdom 1996 Watts Bar Nuclear Plant, Unit 1 USNRC 50-390 1997 Waterford Unit 3 USNRC 50-382 1997
)
Vogtle Unit 1 USNRC 50-424 1998 Chirsham Taiwan 1998 i i
Callaway Plant, Unit 1 USNRC 50-483 1993 J. A. Fitzpatrick Nuclear Power Plant USNRC 50-333 1998 Certain plants have undergone multiple rerack campaigns leading to their repeat listing in this table.
SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 6-51 Report HI-982083
m Table 6.3.1 RACK MATERIAL DATA (200 F)
(ASME - Section II, Part D)
Material Young's Modulus Yield Strength Ultimate Strength E S, Su (psi) (psi) (psi) 6 SA240-304L 27.6 x 10 21,300 66,200 SUPPORT MATERIAL DATA (200'F) 6 SA240-304L (upper part of 27.6 x 10 21,300 66,200 support feet) 6 SA564-630 (lower part of 28.5 x 10 106,300 140,000 support feet; age hardened at
-i 1100 F)
Note: Spent Fuel Pool bulk temperature is less than 200 F.
l l
l I
4 SHADED TEXT CONTAINS PROPRIETARY INFORMATION IIoltec International 6-52 Report HI-982083
Table 6.4.1 TIME-HISTORY STATISTICAL CORRELATION RESULTS OBE Datal to Data 2 -0.10 Datal to Data 3 -0.02 Data 2 to Data 3 -0.02 SSE Datal to Data 2 0.01 l I
Datal to Data 3 0.02 Data 2 to Data 3 -0.02 Datal corresponds to the time-history acceleration values along the X e xis (wes.;
Data 2 corresponds to the tirne-history acceleration values along the Y axis (North)
Data 3 corresponds to the time-history acceleration values along the Z axis (Vertical) l SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 6-53 Report HI-982083 i
Table 6.5.1 Degrees-of-freedom plSPLACEMENT ROTATION LOCATION (Node)
{
U, Uy U, 0x Oy 0, 1
pi P2 pa 94 45 48 p7 pa ps 4o 41 qi2 2
Nodo 1 is assumed to be attached to the rack at the bottom most point.
Node 2 is assumed to be attached to the rack at the top most point.
Refer to Figure 6.5.1 for node identification.
1 2 pis pt4 S' pts pie 4' pi7 . pis 5' P19 P20 1' P21 p22 where the. relative displacement variables q, are defined as:
p3 = q,(t) + U,(t) i = 1,7,13,15,17,19,21 j
= q,(t) + Uy (t) i :: 2,8,14,16,18,20,22 l
= q,(t) + U,(t) i = 3,9
= q,(t) i = 4,5,6,10,11,12 i pi denotes absolute displacement (or rotation) with respect to inertial space l q, denotes relative displacement (or rotation) with respect to the floor slab
~~denotes fuel mass nodes l~~
U(t) are the three known earthquake displacements 4
l 1
SHADED TEXT CONTAINS PROPRIETARY INFC31ATION i Holtec International 6-54 Report HI-9820G3 l i
_ _ . ~ _ _
l I
Table 6.5.2 (DYNARACK) NUMBERING SYSTEM FOR GAP ELEMENTS AND FRICTION ;
ELEMENTS
~~1. Nonlhear Springs (Type 3 Gap Elements - 824 Total)~~
Node Loc.
Number Description 1 Support S1 Z compression-only element 2 Support S2 Z compression-only element 3 Support S3 Z compression-only element 4 Support S4 Z compression-only element 5 2,2* X rack / fuel assembly impact element between nodes 2
. 1 and 2 .
6 2,2* X rack / fuel assembly impact element between nodes 2 1
and 2' 7 2,2* Y rack / fuel assembly impact element between nodes 2 l l
and 2' 8 2,2* Y rack / fuel assembly impact element between nodes 2 l and 2' 9-576 Impact elements corresponding to the rattling masses at nodes 1',3',4' and 5' (similar to elements 5 thru 8) j l
, 577-824 Bottom and inter-rack impact elements !
l' Top Cross l section of Rack (around edge)
I SHADED TEXT CONTAINS PROPRIETARY INFORMATION i Holtec International 6-55 Report HI-982083 L l
i
~~11. Linear Springs (Type 1 Elements - 144 Total)~~
Rack No. j Number Description ]
1 1 Rack beam bending element (x-z plane) 2 1 Rack shear deformation element (x-z plane) 3 1 Rack beam bending element (y-z plane) I 4 1 Rack shear deformation element (y-z plane) 5 1 Rack be'am axial deformation element 6 's Rack 'aeam torsional deformation element 7-12 2 Sirallar to elements 1 thru 6 13-18 3 Eimilar to olements 1 thru 6, continue to Rack 24 Ill. Piece-wise Ligr.r Friction Springs (Type 2 Elements - 192 Total)
Rack Nc.
Number ,_
Description 1 1 Pedestal 1, X direction 2 1 Pedestal 1. Y direction 3- 1 Pedestal 2, X direction 4 1 Pedestal 2, Y direction 5 1 Pedestal 3, X direction 6 1 Pedestal 3, Y direction 7 1 Pedestal 4, X direction 8 1 Pedestal 4, Y direction 9-16 2 Similar to elements 1 thru 8 17-24 3 Similar to elements 1 thru 8, continue to Rack 24 SHADED TEXT CONTAINS PROPRIETARY INFORM ATION Holtec International 6-56 Report HI-982083
1 Table 6.9.1 J COMPARISON OF BOUNDING CALCULATED LOADS / STRESSES VS CODE ALLOWABLES AT IMPACT LOCATIONS AND WELDS I Item / Location Calculated Allowable Fuel assembly / cell wall impact,Ibf 1,148 3,698* .
l l
Cell- to - baseplate weld stress, psi 13,574 (OBE) 19,860 (OBE) 24,345 (SSE) 35,748 (SSE) ,
Pedestal - to - baseplate weld stress, 15,380 (SSE) 35,748 (SSE) psi 13,600 (OBE) 19,860 (OBE)
Cell- to - cell weld load, Ibf 15,680** (SSE) 35,748 (SSE) 1 l
Maximum rack rotation, degrees 0.478 29.21 1
i 1
~~Based an the limit load for a cell wall. The allowable load on the fuel assembly itself may be !~~
~~less than tqis value but is greater than 1,148 lbs.~~
~~Based on the fuel assembly to cell wall impact load simultaneously applied in two orthogonal ,~~
directions.
l l
SHADED TEXT CONTAhS PROPRIETARY INFORMATION Holtec Intemational 6-57 Report HI-982083
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~~REPORT # HI-982083l~~
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. ' 7.0 ' ' FUEL' HANDLING AND CONSTRUCTION ACCIDENTS
-7.1 IIntroduction
~~The USNRC _OT position paper [7.1] specifies that' the design of the rack must ensure the functional integrity of the spent fuel racks'under all credible fuel assembly drop events.~~
~
. This chapter contains synopses of the analyses carried out to demonstrate the regulatory compliance of the proposed racks under postulated fuel assembly drop scenarios germane to'the fuel pools.
. I 7.2 Descriotion of F -I Handline Accidents i I
Two categories of fuel assembly accidental drop events are considered. In the so-called " shallow drop" event,-a fuel assembly, along with the portion of handling tool, which is severable in the case of a single element failure, is assumed to drop vertically and hit the top of the rack. The damage due to a perfectly vertical drop, on the top of a rack, bounds an inclined fuel assembly drop because the impact energy is focused on a single cell wall, which results in maximum cell blockage. For thermal hydraulic consicferations, the maximum cell blockage is conservatively I applied to every rack cell. Inasmuch as the new racks are of honeycomb construction, the
~
deformation produced by the impact is. expected to be confined to the region of collision.
However, the " depth" of damage to the affected cell walls must be demonstrated to remain limited to the portion of the cell above the top of the " active fuel region", which is essentially the elevation of the' top of the Boral neutron absorber. Stated in qualitative terms, this criterion implies that the plastic deformation of the rack cell walls should not extend more than 17 inches i (downwards) from the top. In order to utilize an upper bound of kinetic energy at impact, the impactor is assumed to weigh 2,300 lbs and the free-fall height is assumed to be 36 inches. ,
i It is readily apparent from the description of the rack mod 'es in Section 3 that the impact resistance of a rack at its periphery is considerably less than u mterior. Accordingly, the limiting shallow drop scenario, which would produce maximum cell wall deformation, consists of the case where the fuel' assembly impacts the peripheral cell wall.
SIIADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec Intemational 71 Report HI-982083 9
l i
The recond class of" fuel drop event" postulates that the impactor falls through an empty storage l cell impacting the rack baseplate. This so-called " deep drop" scenario threatens the structural integrity of the baseplate. If the baseplate is pierced, then the fuel assembly ,vight damage the pool liner (which at 3/16" is considerably thinner) and create an abnormal condition of the enriched zone of fuel assembly outside the " poisoned" space of the fuel rack. To preclude damage to the pool liner and to avoid the potential of an abnormal fuel storage configuration in the aftermath of a deep drop event, it is required that the baseplate remain unpierced and that the maximum lowering of the fuel assembly support surface does not violate subcriticality.
The deep drop event can be classified into two scenarios, namely, drop through a cell located above a support leg (Figure 7.2.2), and drop in an interior cell away from the support pedestal (Figure 7.2.3).
In the first deep drop scenario (Figure 7.2.2), the baseplate is buttressed by the support pedestal and presents a hardened impact surface, resulting in a high load. The principal design objective is to ensure that the support pedestal does not tear the liner that overlays the reinforced concrete l pool slab.
In the second deep drop scenario (Figure 7.2.3), the fuel assembly impacts the baseplate away ;
from the support pedestals, where it is more flexible. Severing and large deflection of the baseplate leading to a potential secondary impact with the pool liner are unacceptable results.
7.3 Incident Fuel Assembly Velocity Fuel assembly drop events fall into two broad categories of underwater motion, which may be denoted as " geometry unconstrained" (GU) and " geometry constrained" (GC). In the shallow drop scenario, the fuel assembly, which falls from 36" above the rack and is accelerated by gravity, is solely opposed by the force arising from the form drag effect of the unconfined body of water. In this case the Newtonian equation of motion has the form:
SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec intemational 7-2 Report HI-982083
CoAjn' ~
(m + mg)x.. = mg -
2 ]
~~l. where:~~
m: ~ mass of the impactor . j l .
j mn: hydrodynamic (virtual) mass (due to submergence) l x: displacement variable -
~~g: acceleration due to gravity I~~
co: form drag coefficient
.A: area subject to drag forces
~
~~p. mass density of water~~
- (dot): . derivative with time r: time coordinate 1
For a drop from 36" height, the initial conditions are:
r = 0,i = 0, x = 0 -
The above nonlinear se:-and order differential equation is readily solved to obtain the incident impact velocity (i.e., i at x = 36 inches).
4 The geometry constrained (GC) drop scenario corresponds to the downward movement of the fuel assembly through a storage cell. In this case, a portion of the linear momentum of the fuu assembly is absorbed by the water, which is displaced by the movement of the fuel and forced to exit through the opening in the baseplate. Additional dissipation of the fuel assembly's kinetic j energy occurs through the forced flow of water into the wake of the advancing fuel assembly.
The equation of motion for this case has the form:
l SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 7-3 Report HI-982083
' (m + m y )# = mg -f(x,i,di,d 2 )
where the termfis the fluid resistive force at time r, which is a function of the coincident axial location of the fuel assembly, x, its velocity, i, size of the bottom opening in the baseplate, d i, and the size of the top opening in the storage cell,' d2 . The constrained geor.xtry equation of i
motion is also a nonlinear second order differential equation which can be numerically integrated
. to detennine the incident velocity of the impactor at the instant ofits collision with the baseplate. l l
1
-7.4 Mathematical Model I l
In the 'first step of the solution process, the velocity of the dropped object (impactor) is computed -i 1'
for the condition of underwater free fall in the manner of the formulation presented in the above section. Table 7.1 contains the resuus of the three drop events. l In the second step of the solution, an elasto-plastic finite element model of the impacted region on Holtec's computer Code PLA9TIPAGT (Los Alamos Laboratory's LS-DYNA implemented
)
on Holtec's QA system) is prepared. PLASTIPACT simulates the transient collision event with full consideration of plastic, large deformation, wave propagation, and elastic / plastic buckling modes. The impactor (spent fuel assembly) is simulated by an equivalent clasto-plastic lineal element with lumped masses at its two extremities. The physical properties of material types undergoing deformation in the postulated impact events are summarized in Table 7.2 7.5 - Results 7.5.1 Shallow Droo Events i 1
Of the Region I and Region 11 racks, module M (10x8) is structurally weakest, and it is used in all impact simulations. The location of the target region in the peripheral scenario is shown in a
. plan view (Figure 7.5.1).
SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 7-4 Report HI 982083 4
I
{
I Dynamic analyses show that the top of the impacted region undergoes severe localized deformation. Figure 7.5.2 shows an isometric view of the post-impact geometry of the rack. The 1
maximum depth of plastic deformation is limited to 16 inches, which is below the de sign limit of 17 inches.
7.5.2 Deen Dron Events The deep drop events wherein the impact region is located above the support pedestal (Figure )
1 7.5.3) is found to produce limited stresses in the liner and the concrete slab. The maximum j stresses are well below the failure limits, as shown in Figures 7.5.4 and 7.5.5. Therefore, we I l
conclude that the pool liner and the concrete slab will not be damaged. *
)
l
{
The deep drop condition through an interior cell does produce some deformation of the baseplate and localized severing of the baseplate / cell wall welds (Figure 7.5.6). However, the fuel l assembly support surface is lowered by a maximum of 1.25 inches, which is much less than the distance of 14 inches from the baseplate to the liner. Therefore, the pool liner will not be damaged. !
I 7.6 Conclusion l
l The fuel assembly drop accident events postulated for the Byron /Braidwood fuel pools were l analyzed and found to produce localized damage well within the design limits for the racks. The shallow drop event is found to produce certain plastic defonnation in the top of the storage cc ,
but the region of permanent strain is limited to the portion of the rack structure situated above the l
top of the active fuel region. The deep drop case analyzed for the scenario to produce maximum pedestal force, indicates that the pedestal axial load is small enough to preclude liner damage.
The analysis of the deep drop event at cell locations selected to maximize baseplate deformation indicates that the downward displacement of the baseplate is limited to 1.25 inches, which ensures that a secondary impact of the fuel assembly with the pool liner would not occur. We therefore concluded that the new high-density spent fuel racks for the Byron /Braidwood pools SFIADED TEXT CONTAINS PROPRIETARY INFORMATION lloltec International 7-5 Report 111-982083
1 possess acceptable margins of safety under the mechanical fuel handling accidents postulated in the OT Position Paper [7.1].
l l
1 '
7.7 References for Section 7
[7.1] "OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications," dated April 14,1978.
- [7.2] - " Analysis of the Mechanical Accidents for Byron /Braidwood Nuclear Station,"
I Holtec Report No. Hi-982086, Rev.1.
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l SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 7-6 Report HI-982083 9
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Table 7.1 IMPACT EVENT DATA Impactor impactor Drop Impact Case - Weight Type Height Velocity (Ib) _
(inches) (inch /sec) -
~~1. Shallow 2,300 Fuel assembly 36 151.1 drop event -~~
~~2. Deep drop 2,300 Fuel assembly 203 243 event (away -~~
~~from pedestal)~~
~~3. Deep drop event 2,300 Fuel assembly 203 104.6 (above pedestal) i i~~
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0 1 0- 0 0- 0 1 0- 0 1 0- 0 1 0 -0 0 0 0 +0 0+ + + + + + + EE E E E E E E E E E E E 2 2 9 5 2 9 8 7 5 3 2 0 0 3 9 3 3 314 5 5 5 2 9 6 3 0 0 3 5 7 6 7 71 8 9 0 0 7 7 6 4 2 0 9 3 5 7 0 0 1 1 1 1 1 1 8 7 5 3 1 0 0 R . O _. F _ N _ O I T A M . R OI m FC _ E S . D O m EI TR AA . LN , PE EC S S A P - BO - MR UD MP I E XE AD M 6 5 7 2 0 E 0- R E 9 U 7 8 G I 9T F 9 4 N 0. 1 E M x = E C EA MLP I T S 1 I D ZA 2 L PA . ET TO 3 ST 8 0 2 8 9 i H m ' 8.0 FUEL' POOL STRUCTURAL INTEGRITY CONSIDERATIONS E _ 8.1 Introduction Section IV_of the USNRC OT position paper [8.1.1) specifies that the ability of the reinforced . concrete structure to withstand the loadings arising from the storage of the rpent fuel in the pool (_ must be established. In contrast to many spent fuel pools in light water reactor installations, the Byron and Braidwood spent fuel pool structures are founded on the base mat and have no sub-base trenches or penetrations which may interrupt the ditTusion of the structural weight into the sub-surface rock. The Byron and Braidwood plants, designed as exact replicas of each other, feature identical fuel pool structures: the two pool structures are identical in every respect except l their designation of " plant north" (which are exactly opposite of each other, m'aking the two

~~plants mirror images of each other). The geological terrain at the two sites are also somewhat~~

'different: while the geological rock formation at the Byron site is very close to the surface, permitting the base mat to be poured directly on the rock bed, an intervening layer of soil at Braidwood serves as a less rigid buffer between the base mat and the geologic rock half-space. In sunimary, aside from the characteristic of sub-surface foundation, the Byron and Braidwood pools are structurally identical. The plants' architect-engineer (A/E), Sargent and Lundy Engineers, designated ACI 318-71 as . the goveming Code for the~ Byron /Braidwood plants. Subsequent to their original design in the , 'I' .1970's (both Byron and Braidwood received their construction permits on December 31,1975), the Byron'and Braidwood pool structures were subjected to several re-evaluations by the plant ' A/E, the most recent of which occurred during the 1986-88 reracking of the two sites. The last rerack campaigns for Byron and Braidwood, in the late 1980's, were carried out by the plant owner and operator (Comed) after securing amendments to the plants' operating licenses [8.1.2, '8.1.3] which included the USNRC's review of the pool qualification per the provisions of ACI ~ 318-71. The plant A/E used Braidwood's structure for load capacity evaluation purposes because of its relatively less rigid sub-surface soil layer -(in comparison to Byron). The plant A/E also . considered bounding inertia loads in the pool structural integrity assessment in support of the last terack licensing effort, and demonstrated large margins of safety (which is typical of pool - SH ADED TEXT CONTAINS PROPRJETARY INFORMATION Holtec International 8-1 Report HI 982083 t i structures founded on grade). Inasmuch as the present rerack campaigns for Byron and Braidwood are not intended to realize a substantial increase in the storage capacity (2,984 cells in ca. 2000 vs. 2,870 cells in the rerack in the late 19%'s), the inertia loads on the pool structure (arising from the dead weight of the stored spent nuclear fuel, contained mass of borated water, seismic events, etc.) are expected to undergo minor changes. The structural evaluation effort in the present rerack project, therefore, consisted of comparing the factored loads from each load combination (germane to the goveming ACI code [8.1.4]), from the existing plant Design Basis analysis and those applicable to the present rerack. In view of the robust status of the existing pool structural qualification [8.1.5], and minor changes in the applied loadings, a completely new finite element model and a complete re-analysis was not deemed necessary. Rather, a supplemental calculation package [8.1.6] to document load comparisons was prepared by Holtec International. The information presented in this chapter draws from the analyses and evaluations documented in references [8.1.5-6]. 8.2 Description of the Soent Fuel Pool Structure i The Byron /Braidwood spent fuel pool structures, both built from monolithic reinforced concrete, are essentially rectangular containers consisting of four reinforced concrete walls and a thick reinforced concrete mat founded on grade. Table 8.2.1 provides essential data on the two spent fuel pools (which, as stated before, are identical). A pictorial view of the Byron /Braidwood pools is presented in Figure 8.2.1, where the locations of an a.itonomous cask pit and the gate to the fuel transfer canal are also indicated. Figures 8.2.2-8.2.4 provide additional plan and sectional views to complete the structural 1 illustration of the Byron /Braidwood spent fuel pools. These illustrations are derived from the plant structural drawings [8.1.7]. I The present reracks of Byron and Braidwood have been necessitated by the ongoing ! degradation of the Borafiex neutron absorber and associated release of soluble silica and ! emulsive precipitates in the pool water. l SHADED TEXT CONTAINS PROPRIETARY INFORMATION Ho%: Intemational 8-2 Report HI-982083 i p ! Table 8.2.1: BYRON /BRAIDWOOD POOL STRUCTURE DATA l l Item Thickness j' Base Mat 6'-0" North and South Walls 5 '-0" East Wall (Byron) 5 '-6" West Wall (Braidwood) West Wall (Byron) 6'-0" West Wall (Braidwood) Height (four walls) 41 '-0" 8.3 Material Pronerties The design basis evaluation [8.1.5] utilized conservative material properties so as to minimize the computed ultimate strength of the pool structure. Table 8.3.1 provides a summary of the key material properties. Table 8.3.1: MATERIAL PROPERTIES l Parameter Value Concrete Compressive Strength (psi) 3.500E+03 Un-Cracked Concrete Elastic Modulus (psi) 3.372E+06 i Concrete Poisson's Ratio 0.167 3 Concrete Weight Density (Ib/ft ) 150.0 ! -Concrete Thermal Expansion Coefficient 5.500E-06 Reinforcement Yield Strength (psi) 6.000E+04 Reinforcement Elastic Modulus (psi) 2.900E+07 l l SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 8-3 Report HI-982083 t. - 8.4 Load Combinations The load combinations. applicable' to the Byron. and Braidwood reinforced concrete pool structures are derived from reference [8.1.4]. Table 8.4.1, extracted from the design basis calculation package [8.1.5], summarizes the governing load combinations. I The loading components denoted as D, L, Eoss, etc., in Table 8.4.1 themselves are comprised of eleven " basic loads". These are: - BLI: Liner expansion; l BL2: Horizontal hydrostatic pressure, i I BL3: . Thermal gradient BL4: Axial expansion; BL5: Dead load (racks + pool + vertical hydrostatic pressure); BL6: Dead loads for combination with accident cask drop; BL7: Accident cask drop; ) BL8: OBE scenario Hydro-Dynamic Effects on E-W Direction ; BL9: SSE scenario Hydro-Dynamic Effects on E-W Direction; BLIO: OBE scenario Hydro-Dynamic Effects on N-S Direction; 4 BL11: SSE scenario Hydro-Dynamic Effects on N-S Direction. The load combinations of Table 8.4.1 can be expressed in terms of the " basic loads"(BL1, BL2, .) etc.) leading to Table 8.4.2 which presents fifteen discrete factor load combinations as a function of the aforementioned " basic loads". j i SHADED TEXT CONTAINS PROPRIETARY INFORMATION r Holtec International 8-4 Report HI-982083 i I l L L. l Table 8.4.1: FACTORED LOAD COMBINATIONS !~ Normal Conditions (N): 1.4*D + 1.7*L + 1.3*T Abnormal Conditions (A): 1.0*D + 1.0*L + 1.0*T + 1.0*Y Severe Conditions (S): 1.4*D + 1.7*L+ 1.3*T + 1.9'Eoss; 1.2*D + 1.3*T + 1.9* Eons Extreme Conditions (E):' 1.0*D + 1.0*L+ 1.0*T+ 1.0*Esss where : .D. is the dead load; L is the live load; ~T is the thermal load; - Y is the load induced by the cask drop; Eoss represents the OBE seismic scenario; Esse represents the SSE seismic scenario. 1 Tab.le 8.4.2: COEFFICIENTS OF BASE LOADS FOR FACTORED LOAD COMBINATIONS Load Basic Load Combination No. Type BL1 BL2 BL4 BL5 . BL6 BL7 BL8 BL9 BL10 BL11 1 N 1.3 1.4 - 1.3 1.4 2 N 1.3 1.4 1.3 1.4 3 A 1.0 1.0 1.0 1.0 1.0 4 S 1.3 1.4 1.3 1.72 1.9 1.9 5' S 1.3 1.4 1.3 1.72 1.9 -1.9 6- S 1.3 1.4 1.3 1.72 -1.9 1.9 7 S 1.3 1.4 1.3 1.72 -1.9 -1.9 8 S 1.3 1.2 1.3 1.52 1.9 1.9 9 S 1.3 1.2 1.3 1.52 1.9 -1.9 j 10 S 1.3 1.2 1.3 1.52 -1.9 1.9 l 11- S 1.3' l.2 1.3 1.52 -1.9 -1.9 ! 12 E. 1.0 1.0 1.0 1.32 1.0 1.0 ; E 1,0 1.0 1.0 1.32 1.0 -1.0 13 l

~~14. E 1.0 1.0 1.0 1.32 -1.0 1.0 l~~

E' l.0 1.0 ' 1.0 1.32 -1.0 -1.0 15 l l SilADED TEXT CONTAINS PROPRIETARY INFORMATION Hohec International 85 Report 111-982083 l i l' Note: The input seismic loads (BLE - BLil) do not include acceleration of the dead loads in BLS. This acceleration is included by deriving load factors as follows: [ LC4-LC7: 1.4*D + 1.9*Eoss = 1.4*(BL5) +.1.9*(gvoss)*(BL5) = 1.72*(BL5) . . LC8-LCl1: 1.2*D + 1.9* Eoss = 1.2*(BLS) + 1.9*(gvoss)*(BLS) = 1.52*(BL5) LC12-LC15: 1.0*D + 1.0* Esse = 1.0*(BL5) + 1.0*(gvsse)*(BL5) = 1.32*(BL5) where: gvoss = 0.17; . gvsss = 0.32. 8.5 Analysis Methodolouv ' - As mentioned earlier, the original investigation of the monolithic reinforced concrete structure of the Spent Fuel Pool was conducted by Sargent and Lundy, the designer of the plant,'in 1977 [8.1.5]. The area ofinterest, namely the four walls and a portion of the on-grade mat associated with the Spent Fuel Pool, were isolated from the Fuel Handling Building and analyzed solely for - its pertinent loads, without any. credit for support from the adjoining structures. This was acceptable because the magnitude. of the pool mass was small enough in comparison to participating building mass so that a change to this mass was not significant to.the overall . auxiliary-fuel handling-turbine building seismic response. Therefore, overall dynamic interaction between the pool constituents and the building was not significant. This allowed a three-dimensional finite-element model representing only half of the Spent Fuel Pool walls and mat, with a symmetry line along the east-west centerline of the Spent Fuel Pool, to be constructed. The Spent Fuel Pool structure was modeled as a symmetric structure under the assumption that the Cask Pit walls induce a negligible perturbation in the displacement field of . the intersecting walls. This assumption conservatively neglects the two 2'-6" thick interior cask _ pit walls. The asymmetric distribution of the loading is solved by considenng two bounding , symmetric loadings. The finite-element analysis of the pool walb and slab was conducted as a three-step process. First, the stress and displacement fields pertinent to all plate type finite-elements, .which represent the walls and the mat, are obtained using the SLSAP computer , i program [8.5.1] for- each of the eleven (11) " basic load" cases. The SLSAP computer code j SHADED TEXT CONTAINS PROPRIETARY INFORMATION ' Holtec Intemational 8-6 Report HI 982083 4 ll 0 , q r 1 l l !~ (which is Sargent and Lundy's QA'd version of the public domain Code SAP) solution is a ' classical linear clastic static analysis. Second, the in-plane stresses and global moments l l 1 calculated for each one of the eleven " basic load" cases for all finite-elements in the three-l l dimensional model wen: comb;ned using the SLSAP post processor. Fifteen (15) load ) combinations, derived from the five (5) " generic Icad combinations" shown in Table 8.4.1, are presented in Table 8.4.2. Third, eight critical elements were selected and analyzed using the TEMCO computer program [8.5.2]. The combined element forces and moments, and thermal gradient were input into TEMCO. The program superimposed the previously combined fields _ and the impact of the thermal gradient on the concrete. The thermal impact on the concrete and the evaluation of cross-section concrete capacity are estimated under the assumption of cracked cross-sections. The most critical finite-element was found to be Element #7. The :e-analysis process follows the three-step process previously explained above. First, the new magnitude of the loads caused by the replacement of the high density racks are computed I and appropriate ratios to the previously evaluated loads are calculated. These ratios are applied to ] the " basic" loads affected by the changes that are shown in Table 8.5.1. l Table 8.5.1: BASIC LOAD RATIOS Basic Load Case Load Name Load Type Load Changed Multiplication Coefficient BL1 Linear To,Ta Ta 1.054 Expansion BL2 Hydrostatic D No 1.00 Pressure BL3 Thermal - To,Ta Ta 1.054 Gradient BL4 Axial Thermal To,Ta Ta 1.054 Expansion BL5 Mat Dead Loads D Rack Weight 0.916 BL6- Cask Drop Mat D No 1.00 l Dead Loads l , BL7 Cask Drop ' Y No 1.00 , SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International - 8-7 Report HI-982083 FL : { i g Basic Load Case Load Name I,oad Type Load Changed Multiplication j Coefficient i L BL8 E-W OBE - E.E' No 1.00 BL9 Hydrodynamic E.E' No 1.00 j l E-W SSE . ) BL10 Hydrodynamic E,E' No 1.00 4 ! N-S OBE l: BL11 Hydrodynamic E,E' No 1.00 l N-S SSE Due to the change in loads, the matrix shown in Table 8.4.2'is modified and depicted in Table 8.5.2. Table 8.5.2i RECALCULATED COEFFICIENTS OF BASE LOADS FOR FACTORED LOAD COMBINATIONS Load Basic Load Combination No. Type BL1 BL2 BL4 BL5 BL6 BL7 BL8 BL9 BL10 BL11 1 N 1.3 1.4 1.3 1.282 2 N 1.4 1.4 1.3 1.4 3 A , 1.0 1.0 1.0 1.0 1.0 4 S 1.3 1.4 1.3 1.576 1.9 1.9 5 S 1.3 1.4 1.3 1.576 1.9 -1.9 6 S 1.3 1.4 1.3 1.576 -1.9 1.9 7 S- 1.3 1.4 1.3 .1.576 -1.9 -1.9 i 8 S 1.3 1.2 1.3 1.392 1.9 1.9 9- S 1.3 l'.2 1.3 1.392 1.9 -1.9 10 S 1.3 1.2 1.3 1.392 1.9 -1.9 11 S 1.3 1.2 1.3 1.392 -1.9 -1.9 12 E 1.054 1.0 2.054 1.209 1.0 1.0 13 E 1.054 1.0 1.054 1.209 1.0 -1.0 i 14 ' E 1.054 1.0 1.054 1.209 -1.0 1.0 E 1.054 1.0 1.054 1.209 -1.0 -1.0 (' 15 SHADED TEXT CONTAINS PROPRIETARY INFORMATION ! Holtec International ' 8-8 Report H1982083 L Using the recalculated load coefficients, the global forces and moments pertinent to critical finite-element #7 are re-evaluated for all load combinations. By comparison with the previous l results from the Sargent and Lundy calculation [8.1.6],' the difference in magnitude is negligible (i.e., less than 1%). Therefore, safety factors previously calculated by Sargent and Lundy remain valid for the present rack installation. The safety factors are reproduced in Table l L 8.5.3. n Table 8.5.3: MINIMUM SAFETY FACTORS Reinforcing Steel Concrete l l Horizontal Direction Vertical Direction Horizontal Vertical Direction Direction Inside Outside Inside Outside TEMPCO 4.4 21 3 51 1850 626 l l Stress (ksi) Allowable 54 54 54 54 2975 2975 Stress (ksi) (*) Safety 12.27 2.57 18.00 1.06 1,61 4.75 Factors-i l Note: (*) Allowable stresses are calculated as: , Reinforcing steel.. 0.9 fy = 0.9 (60 ksi) = 54 ksi Concrete.. 0.85 fe' = 0.85 (3500 psi) = 2975 psi ! 8.6 Pool Liner Integrity Analysis L . ' .The pool liner in the Byron /Braidwood pools is not a safety-related component. The "non-safety related" designation for the pool liner arises from the fact that its failure would not cause a rapid lowering of the water level in the fuel pool. The spent fuel pool liner contains a leak chase system designed to collect leakage. Five 1.5 inch diameter drains are embedded in the concrete i slab beneath the stainless steel liner. This drain piping is located up to column row "N" where it ' SHADED TEXT CONTAINS PROPRIETARY INFORhiATION Holtec International 8-9 Report HI-982083 k.'_ 1 1 joins the auxiliary building floor drain piping system. The system also contains sight glasses and h five normally closed valves. Any leakage occurring past the spent SJ pool liner would be collected and stopped by the piping and valves. However, in a pool filled with closed packed array of rack - ales, accessing of the liner on the pool slab for inspection and repair is not feasible or practicable. Therefore, it is necessary to ensure that the rack pedestal and bearing pad design do not produce a state of overstress in the liner, leading to failure from tearing or cyclic fatigue during seismic events. A description ,f the liner plate design and construction details follows. Embedded wide flanges are arranged in the concrete pool floor in an approximate 6'-0" grid pattern. Individual embedded plates are located at 2'-0" on center within the floor grid system. The liner pa:e is attached to the embedments in 6 foot panels with a continuous groove weld at the seams to the continuous grid embedments and with plug welds at 2'-0" on center to the individual embedments. This anchorage sys:em results in unsupported liner panels of 2'-0" by 2'-0" Lik'e all fuel pool liners, the Byron /Braidwood pool liners are subjected to the weight of water (plus any seismic adder) and lateral plus vertical forces from individual spent fuel rack support pedestals. The liner can be thought of as a stainless steel plate resting on a concrete substrate which acts as an elastic foundation. Because the coefficient of friction at the stainless-steel-to-stainless-steel interface exceeds that at the stainless / concrete interface, the portion of the liner not welded to embedments in the concrete may sustain in-plane strains under the action cf the friction-induced lateral forces induced by fuel rack motion during a seismic event. In the analysis model, the embedments are considered as fixed anchor points. Calculations have been made to establish th:16e liner will not tear or mpture under all loading conditions in the pool and that the liner can withstand one SSE and twenty OBE seismic events $ without fatigue failure. The cumulative damage factor under one SSE and twenty OBE events is 5.2 x 10", which is well below the acceptance limit of 1.0. An evaluation of the liner plate section subjected to vertical and horizontal static pedestal loading has also been performed. The maximum stress in the liner and liner welding during an SSE seismic event has been computed SHADED TEXT CONTAINS PROPRIETARY INFORMATION Iloltec Intemational 8-10 Report HI-982083 l 1 e

~~and found to be less than the ultimate strength of the liner material, i.e., tear failure will not~~

~ occur. , The liner plate and anchorage system also have been designed for the forces resulting from long-term shrinkage of concrete and a temperatum rise tc 158'F from the 70*F ambient temperature with nominal cooling. The maximum compression force in the liner is calculated using the total strain of the long-term shnnkage of the concrete and the temperature rise. This compressive , stress in the liner is limited to 0.90 Fy. The liner and its anchorage were evaluated for the therr.2 loads, the strain induced load from the deformation of the floor, and the horizontal maximum friction load induced by the rack due to seismic effects. The liner was evaluated for no tearing or rupture near the rack pedestals. The pool liner fatigue analysis due to cyclic loading was performed using ASME Code Section III ~ methodology. The liner anchors were evaluated for the unbalanced liner in-place force due to temperature and strain induced loads, as well as the horizontal friction force. Therefore, we conclude that the liner in the Byron /Braidwood pools will maintain its integrity during and after the postulated seismic events. 8.7 Bearing Pad Anai/ sis ' To protect the pool slab from high-localized dynamic loads, beating pads are placed between the pedestal base and the slab. Fuel rack pedestals impact these bearing pads during a seismic event, and the pedestalload is transferred to the slab. Bearing pad dimensions are chosen to ensure that the average pressure on the slab surface, due to a st*: load plus a dynamic impact load, does not exceed the~ American Concrete Institute, ACI-349 [8.7.1] limit on bearing pressures. Section .10.15 of [8.1.4] gives the design bearing strength as f b= $(0.85 fe')e MADED TEXTDONTAINS PROPRIETARY INFORMATION ~ Holtec International . 8-11 Report HI-982083 n .. .. . . I where 4 = 0.7 and fe' is the specified concrete strength for the spent fuel pool. c = 1, except when the supporting surface is wider on all sides than the loaded area. In that case, c = , (A 2/Ai)", but not more than 2. Ai is the actual loaded area, and A2 is an area greater than A i, which is defined in [8.1.4].' Using a value of e > 1 includes credit for the confining effect of.the l surrounding concrete. It is noted that these criteria are in conformance with the ultimate strength l primary design methodology of the American Concrete Institute in use since 1971. For the spent fuel pools at Byron and Braidwood, the concrete compressive strength is fe' = 3,500 psi.. The allowable bearing pressure is conservatively computed by taking c = 1 to account for lack of total concrete confinement in the leak chase region and a stress reduction factor of $ = 0.7. Thus, the maximum allowable concrete bearing pressure is 2,082 psi. The maximum vertical pedestal load is 238,000 lb. The bearing pad is fabricated from 3-% inch thick, austenitic stainless steel plate stock. The average pressure at the pad to liner interface is l computed and compared against the above-mentioned limit. Calculations show that the average pres'sure at the slab interface is 1,452 psi, which is below the allowable value of 2,082 psi. This yields a safety factor of 1.43. l Therefore, the bearing pad design devised for the Byron and Braidwood spent fuel pools is ! l deemed appropriate for the prescribed loadings. l 1 8.8 Results and Conclusions An evaluation of the factored load combinations applicable to the Byron /Braidwood pool structures led to the following results: (

The net section moment and shear in the most heavily stressed 'ocations in the pool structure is lower _ in the reracked pool compared to the values compated in the plants' Design Basis L document [8.1.5] for all applicable load combinations.

l

j.
In one respect, after r: racking, the structural margin in the Byron /Braidwood pools will actually be improved over the existing configuration: the new fuel racks have been designed with a larger platferm area and optimized mass / rigidity ratio leading to the precluding of any SHADED TEXT CONTAINS PROPRIETARY INFORMATION

' Holtec Intemational 8-12 Report HI 982083 l H i rack-to-wall impacts (the existing racks are incapable of preventing rack-to-wall impacts during SSE events because of their non-optimized layout). e' The pool liner can withstand one SSE and twenty OBE events without sustaining failure from .. tearing or fatigue. Therefore, we conclude that the margins of' safety documented in the existing pool structural design basis provide a lowe- bound on the margins that will exist after the planned reracking of the Byron /Braidwood pools. ! 8.9 References for Section 8 - . [8.1.I'] OT Position for Review and Acceptance of Spent Fuel Handling Applications, by i B.K. Grimes, USNRC, Washington, D.C., April 14,1978. [8.1.2] ~ OL Amendment 25, Byron Station, Docket Nos. 50-454 and 50-455, March 17, 1989. [8.1.3] . OL Amendment 20, Braidwood ' Station, Docket Nos. 50-456 and 50-457, July 20, 1989. [8.1.4] American Concrete Institute, ACI 318-71. [8.1.5] Sargent and Lundy Calculation Book No. 8.1.12," Fuel Handling Building Spent Fuel Pool Analysis and Design",1986. [8.1.6] Sargent and Lundy Calculation No. 8.1.12-BRW-96-754, " Evaluation of Spent Fuel Pool for Elevated Temperature," Rev. 3,1996. l [8.1.7] " Structural Re-evalua. ion'of the Spent Fuel Pool for Byron and Braidwood Stations, Holtec Report No. HI-982090 (1998). i [8.1.8] Drawings S-811, S-816, S-820, and S-821 for Byron /Braidwood Stations, Sargent and Lundy Engineers. l l [8.5.1] "SI. SAP and SLSAP Post-Processor", S&L Program Number 09.7.130-4.00. i [8.5.2] "TEMCO Computer Program", S&L Program No. 09.7,072-5.20. [8.7.1] ACI 349_85, Code Requirements for Nuclear Safety Related Concrete Structures, American Concrete Institute, Detroit, Michigan,1985. i SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 8-13 Report HI-982083 ( 'n ~hp $ '% .e -d .g 1 / , g: g~ a. a l . g ) g-k - e 1 / .E l 5 . / = l m E: ._s / ~A, ' l m. E. y . ~~ / EE ~e u4' - . ~& v f. f. G w. i E 25 / 22 Os / / 4 Cm3 l / ,/ - r , / - ~/ 1 l a E - W h-El2 ma-a o 6 . . o - 4 0 - 4 O

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= - 6 O " - 4 5' > O p O - >> O ? > O a O ,g . * = , i < - ' ' 8 0 3 8 0 2 8 9 I H 9.0 -- RADIOLOGICAL EVALUATION ' i l 9.1 Fuel Handline Accident Increases in the capacity of the fuel storage pools at the Byron and Braidwood Nuclear Plants are . 1 not accompanied by ; concomitant increases in the potential severity of fuel-handling accidents, ! nor are the potential offsite doses increased over the values given in the Updated Final Safety / Analysis Report (s). The reasons there are no changes include the following: eo the accident scenario presented involves the failure of all the rods in' a single fuel assembly, which is not a function of the storage capacity of the fuel pools or the rack design; j e ' the basic methodology used to determine the doses, as presented in Regulatory Guide 1.25, does not change '(Note: The peak gap ~ activity of I-131 is , assumed to be 12% per .. NUREG\CR-5009); . .1the short life fission products are generally independent of fuel assembly burn-up, so the radionuclides that affect the radiation doses do not change, and their release fractions from the fuel do not change; e the basic nuclear data that affect the doses - beta and gamma energies from the decay of the radionuclides, for example - do not change; e the pool water depths do not change, so the retention effectiveness for iodine does not - i change; e the height of the active fuel in relation to the SFP floor is comparable to the storage height of the original spent fuel rack design; eL the physical plant filter designs do not change, so'the capabilities to remove the iodines from the gas streams being released from the buildings do not change; e the atmospheric diffusion factors for the plants at th. Exclusion Area Boundaries (EAB) and Low Population Zones (LPZ) do not change; I e the dose conversion factors do not change; and l 1 .. the cask drop accident, which is discussed in UFSAR Section 15.7.4, is not analyzed because i of crane interlocks, which prevent the cask from traveling over the Spent Fuel Pool. SHADED TEXT CONTAINS PROPRIETARY INFORMATION p l Holtec International 91 Report HI 982083 Thus,' the potential doses from fuel handling accidents at the Byron and Braidwood Nuclear Statians, as presented in the UFSAR, continue to meet the guidelines set forth in NUREG-0800, the Standard Review Plan. 9.2 Solid Radwaste The necessity for resin replacement is detennined primarily by the requirement for water clarity, and the resin is normally changed about once a year. No significant increase in the volume of solid radioactive wastes is expected with the expanded storage capacity. During reracking operations, small amounts of additional resins may be generated by the pool cleanup systems on one-time bases. Incidentally, no effects are anticipated relative to non-radiological waste stream generation or alteration, 'specifically for air, wastewater, solid waste or hazardous waste. Therefore, no ; changes to the Natior.a! Pollutant Discharge Elimination System permit are required. 9.3 Gaseous and Liouid Releases Gaseous releases from the fuel storage building are combined with other plant exhausts. Nonnally, the contribution from the fuel storage building is negligible compared to the other releases, and no significant increases are expected as a result of the minimal storage capacity 1 l expansion (i.e.,114 storage cells or a 4% increase in storage capacity). l. The storage of spen't fuel assemblies does not directly affect the release of radioactive liquids from the plant, since radioactive liquids are not directly discharged from the Spent Fuel Pool. 9.4 . Personnel Exoosures During normal operations, personnel working in the fuel storage area are exposed to radiation ! l from the spent fuel pool. The dose rates experienced by personnel are not expected to increase I SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 9-2 Report HI-982083 with the increased storage capacity of the Holtec racks because the dose rate from the fuel in storage is negligible. The water above the stored fuel is sufficiently deep hat the dose rate from that fuel is orders of magnitude lower than the dose rate contribution from the radionuclides in the pool water itself. Consequently, though the dose rate from stored fuel will increase because more spent fuel assemblies are stored, it will not increase to levels comparable to those experienced from the radionuclides in the pool water. The radionuclide concentrations in the pool water are not expected to increase significantly, for the levels are determined principally from the mixing of primary system water with the pool water and the spalling of crud deposits from the spent fuel assemblies as they are moved in the storage pool during refueling operations. Although the overall capacities of the pools are being increased, the movement of fuel during given refuelings is independent of storage capacity. Operating experience has shown that there have been negligible concentrations of airborne radioactivity, and no increases are expected as a result of the expanded storage capacities. Area monitors for airborne activities are available in the immediate vicinities of the spent fuel pools. In summary, no increases in radiation exposure to operating personnel are expected. Consequentl*. neither the current health-physics programs nor the area monitoring systems needs to be modified. I 9.5 Anticioated Exoosures Durine Rerackine All of the operations involved in reracking will utilize detailed procedures, which implement ALARA principles. Similar operations have been performed in a number of facilities in the past, and there is every reason to believe that re-racking can be safely and efficiently accomplished at the I Byron and Braidwood Nuclear Stations with minimum radiation exposure to personnel. . Total occupational exposure for the re-racking operation is estimated to be between 6 and 12 person-rem, as indicated in Table 9.1. While individual task efforts and exposures may differ from SHADED TEXT CONTAINS PROPRIETARY INFORMATION Report HI-982083 ) Holtec International 9-3 l I i p those in Table 9.1, the total is believed to be a reasonable estimate for plannmg purposes. D'ivers will be used only if necessary, but the estimated person-rem burden includes a figure for their 1

~~possible exposure.~~

The existing radiation protection programs at the plants are adequate for the re-racking operations. Where there is a potential for airbome activity, continuous air samplers will be in operation.- Personnel will. wear protective clothing and, if necessary, respiratory protective equipment if determined to be TEDE ALARA. Activities will be governed by Radiation Work Permits, and personnel monitoring equipment will be issued to each individual. As a minimum, this will include thermoluminescent dosimeters and electronic dosimeters. Additional personnel monitoring equipment (i.e., extremity badges or alarming dosimeters) may be utilized as required. Work, personnel traffic, and the movement of equipment will be monitored and controlled to minimize contamination and to assure that exposures are maintained ALARA In re-racking, the existing storage racks will be removed, then washed down in preparation for packaging and' shipment.' Estimates of the person-rem exposures associated with washdown and ~ readying the old racks for shipment are included in Table 9.1. Shipping containers and procedures will confonn to Federal DOT regulations and to the requirements of any state through which the shipment may pass, as set forth by the State DOT office. SH ADED TEXT CONTAINS PROPRIETARY INFORM ATION Holtec International 9-4' Report 111-982083 (. . L I Table 9.1 PRELIMINARY ESTIMATE OF PERSON-REM EXPOSURES DURING RERACKING l L Estimated , Number of Dose Rate Person-Rem l S. ley Personnel Hours in mrem /hrt Exoosure Remove empty racks 5 40 2.5 to 5.0 0.5 to 1.0 Wash and decon racks 3 10 2.5 to 5.0 0.075 to 0.15 j Clean' and vacuum pool 3 25 4.0 to 8.0 0.3 to 0.6 Remove underwater appurtenances" 4 5 20 to 40 0.4 to 0.8 i Install some new rack modules 5 20 2.5 to 5.0 0.25 to 0.5 Move fuel to new racks 2 150 2.5 to 5.0 0.75 to 1.5 Remove remaining old racks 5 120 2.5 to 5.0 1.5 to 3.0 Wash and decon racks 3 30 2.5 to 5.0 0.225 to 0.45 l Install remaining new rack modules 5 40 2.5 to 5.0 0.5 to 1.0 l Prepare old racks for shipment 4 80 -- 1.0 to 2.0nt i TOTAL EXPOSURE, PERSON-REM 5.5 to 11 1 l ' Assumes a dose rate of 2.5 mrem /hr (minimum) to 5 mrem /hr (maximum), except for pool vacuuming operations, which assume 4 to 8 mrem /hr, and diving operations, which assume 20 to 40 mrem /hr. " Although this activity assumes that divers will be used, the rack installation is planned such that divers may not be needed. '" Maximum expected exposure, although details of preparation and packaging of old racks for shipment have not yet been determined. SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec Intemational 9-5 Report HI-982083 f i . 7, , .10.0 -- BORAL SURVEILLANCE PROGRAM 10.1 _P_ge Although the U.S. Nuclear Regulatory Commission (USNRC) has no current requirement for in-service Boral surveillance, as stated in the USNRC letter fmm Dr. L. Kopp to Dr. K. Singh '(Holtec International) [10.1.5], Comed has requested e Boral surveillance pmgram for use as needed. Boral, the neutron absorbing material incorporated in the spent fuel storage rack design to assist in controlling system reactivity, consists of finely divided particles of boron carbide (B4C) - uniformly distributed in type 1100 aluminum powder, clad in type 1100 aluminum and pressed and sintered in a hot-rolling process. Tests simulating the radiation, thermal and chemical environment of the spent fuel pool have demonstrated the stability and chemical inenness of Boral (References [10.1.1] through [10.1.3]). The accumulated dose to the Boral over the expected rack lifetime is estimated to be about 3 x 10' to 1 x 10" rads depending upon how the racks are used and the number of full com off-loads that are necessary. Based upon the accelerated test programs, Boral is considered a satisfactory material for reac-tivity control in spent fuel storage racks and is fully expected to fulfill its design function over the lifetime of the racks. Nevenheless, it is prudent to establish a surveillance program to monitor the integrity and performance of Boral on a continuing basis and to assure that slow, long-term synergistic effects,if any, do not become significant. Furthermore, the April 14,1978 ~ . USNRC letter to all power reactor licensees (Reference [10.1.4]), specities that " Methods for verification of long-term material stability and mechanical integrity of special poison materials utilized for neutron absorption should include actual tests." The purpose of the surveillance program is to characterize cenain properties of the rlo.al with the objective of providing data necessary to assess the capability of the Boral panels in the racks to continue to perform their intended function. The surveillance program is also capable of detecting the onset of any significant degradation with ample time to take such corrective action SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec Intemational 10-1 Repcn HI-982083 s 4 1 ' as may be necessary.

In response to t he need for a comprehensive Boral surveillance program to assure that the subcriticality requirements of the stored fuel array are safely maintained, a surveillance program has been developed incorporating cenain basic tests and acceptance criteria. The Boral surveill-ance program depends primarily on representative coupon samples to monitor performance of the absorber material without disrupting the integrity of the storage system. The principal parameters to be measured are the thickness (to monitor for swelling) and boron content.

10.2 Coupon Surveillance Program; 10.2.1 Couoon Descriotion . The coupon measurement program includes coupons suspended on a mounting (called a " tree"),- . placed in a designated cell, and surrounded by spent fuel. Coupons will be removed from the . array on a prescribed schedule and cenain physical and chemical propenies measured from 'l which the stability and integrity of the Boral in the storage cells may be inferred. Each surveillance coupon will be approximately 4 inches wide and 8 inches long. The coupon surveillance program will use a total of eight test coupons with each coupon mounted in a - stainless steeljacket, simulating as nearly as possible, the actual in-service geometry, physical mounting, materials, and flow conditions of the Boral in the storage racks. The jacket (of the same alloy used in the manufacture of the racks) will be closed by screws or clamps to allow easy opening with minimum possibility of mechanical damage to the Boral specimen inside. In mounting the coupons on the tree, the coupons will be positioned axially within the central 8 feet l ) of the fuel zone where the gamma flux is expected to be reasonably uniform. ! i Each coupon will be carefully pre-characterized prior to insenion in the pool to provide initial ; reference values for comparison with measurements made after irradiation. The surveillance l L coupons will be pre-characterized for weight, length, width and thickness. In addition, two coupons (which need not be jacketed) will be preserved as archive samples for comparison with subsequent test coupon measurements. Wet chemical analyses of samples from the same lot of SHADED TEXT CONTMNS PROPRIETARY INFORMATION Holtec Intemational 10-2 Report HI-982083 q l ~ Boral w'i ll be available from the vendor for comparison. 10.2.2. Surveillance Coupon Testinn Schedule j -l . ~ Since the Spent Fuel Pools at Byron and Braidwood are each shared by two Units, the coupon j schedule at eaah plant should be based on the Unit with the shorter cycle length. For example,if l Unit I has 18 month refueling cycles and _ Unit 2 has 24 month refueling cycles, then the coupon testing schedule should be based on Uriit 1 operation. Then, at each of the first five refuelings for the selected Unit, the coupon tree, which is placed in a Region Il storage cell, is surrounded ) L by freshly discharged fuel assemblies to assure that the coupons experience a slightly higher radiation dose than the Boral in the racks. Beginning with the fifth load of spent fuel, the fuel assembl'i es stay in place for the remaining lifetime of the racks. The coupon management schedule is shown in Table 10-1. At the time of the first fuel off-load following installation of the coupon tree, the (8) storage cells sunounding the tree shall be loaded with freshly-discharged fuel assemblies that had been among t' he higher specific power assemblies in the core. Shortly before the second reload, the coupon . tree is removed and a coupon removed for evaluation. The coupon tree is then re-installed and,' at reload, again surrounded by freshly discharged fuel assemblies. This procedure is continued for the third, founh, and fifth off-loading of spent fuel (except that a coupon is not pulled at the j i founh refueling).. From the fifth cycle on, the fuel assemblies in the (8) surrounding cells remain j in place. Evaluation of the coupons removed will provide information of the effects of the radiation, thermal and chemical environment of the pool, and, by inference, comparable information on the - Boral panels in the racks. Over the duration of the coupon testing program, the coupons will- ; have accumulated more radiation dose than the expected lifetime dose for normal storage cells. . Coupons which have not been destructively analyzed by wet-chemical processes, may optionally ; o be retumed to the' storage pool and re-mounted on the tree. They will then be available for ; subsequent investigation of defects, should any be found. - SHADED TEXT CONTAINS PROPRIETARY INFORMATION , Holtic International 10-3 Report HI-982083 k = i J10.2.3 Measurement Program The coupon measurement program is intended to monitor changes in physical properties of the Boral absorber material by performing the following measurements on a scheduled basis: . a. Visual observation and photography, ~ b. . Neutron attenuation,

~~c. Dimensional measurements (length, width and thickness),~~
~~d. ' - Weight and specific gravity, and~~
~~e. Wet-chemical analysis (a process wherein the aluminum in Boral is chemically disso!.ved in an acid solution leaving Boron carbide precipitate.~~

which can be dried and weighed to determine the B4C content in the coupon). The most significant measurements are thickness (to monitor for swelling) and neutron t attenuation (to confirm the concentration of Boron-10 in the absorber material). In the event that loss of boron is observed or suspected, the data may be augmented by wet-chemical analysis (a destructive gravimetric technique for total boron only). .10.2.4 Surveillance Coupon Acceptance Criteria Of the measurements to be performed on the Boral surveillance coupons, the most important are: !- (1) the neutron attenuation measurements (to verify the continued presence of the boron) and (2) the thickness measurement (as a monitor of potential swelling). Acceptance criteria for these measurements are as follows: ) ' ' ' Netaron attenuation measurements are a precise instrumental meth'od of chemical analysis for , Boron-10 content using a non-destructive technique in which the percentage of thermal neutrons trans- l 'mitted through the panel is measured and compared with pre-determined calibration data. Boron-10 is the nuclide of principal interest since it is the isotope responsible for neutron absorption in the Boral . panel. SHADED TEXT CONTAINS PROPRIETARY INFORMATION ! 'Holtec International 10-4 Report HI-982083 I

a. ~ A decrease of no more than 5% in Boron-10 content, as determined by neutron attenuation, is acceptable. (This is tantamount to a requirement for no loss in boron within the accuracy of the measurement.)
~~b. ' An increase in thickness at any point should not exceed 10% of the initial thickness at that point.~~

Changes in excess of either of these two criteria requires investigation and engineering evaluation which may include early retrieval and measurement of one or more of the remaining I . coupons to provide corroborative evidence that the indicated change (s) is real. If the deviation is determined to be real, an engineering evaluation shall be performed to identify further testing or any corrective action that may be necessary. The remaining measurement parameters serve a supponing role and should be examined for early indications of the potent ial onset of Boral degradation that would suggest a need for funher attention and possibly a change in measurement schedule. These include (1) visual or photographic evidence of unusual surface pitting, corrosior: or edge deterioration, or (2) unac-countable weight loss in excess of the meisurement accuracy. 10.3 In-Service Inspection (Blackness Tests) In-service inspection involves directly testing the Boral panels in the storage racks by neutron logging' (sometimes called " blackness testing"). This technique is able to detect areas of significant boron loss or the existence.of gaps in the Boral, but cannot determine other physical propenies such as those measured in the coupon program. In the event that the surveillance coupon program shows a confirmed indication of degradation, blackness testing may be one of the techniques employed to investigate the extent of degradation, if any,in the racks. Blackness testing is a method of comparing the data found in the pool with a / Neutron logging, is a derivative cf well-logging methods successfully used in the oil industry for many years. SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 10-5 Repon HI-982083 known sample. A comparison of readings taken from the Boral in the racks with those taken i from a known sample allows one to detennine where there may be deficiencies in the Boral - panels in the spent fuct pool storage racks. 1 ( l I J \ l I 1 1 i 1 l ) l l i i I SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 10-6 Report 10-982083 . l '10.4 References for Section 10 i ( [10.1.1) " Spent Fuel Storage Module Corrosion Repon," Brooks & Perkins Repon 554, -June 1,1977. ,, I ~ [10.1.2] " Suitability of Brooks & Perkins Spent Fuel Storage Module for Use in PWR Storage Pools," Brooks & Perkins Repon 578, July 7,1978. , [10.1.3) "Boral Neutron Absorbing / Shielding Material - Product Performance Report," Brooks & Perkins Report 624, July 20,1982. 1 l (10.1.4) ~ USNRC Letter to All Power Reactor Licensees transmitting the "OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications," April 14,1978. i [10.1.5] USNRC Ietter to Dr. K Singh (Holtec International) from Dr. L. Kopp, February l 15,1995. I 1 . l I l l ) l I SHADED TEXT CONTAINS PROPRIETARY IhTORMATION Holtec International 10-7 Repon HI-982083 m- !: \ Ts>le 10-1: COUPON MEASUREMENT SCHEDULE i Coupon ' Refuelingt After i Rerack I l ~- 1st" - 2 2nd" i 3 ~' 3rd" , 4 - 5th"- ) . 5 8th 6 lith 7 14th 8 20th . 9 25th 10' 30th l Remove coupons for evaluation within 1 or 2 months before the next refueling. l "' Place freshly discharged fuelin the 8 surrounding cells at the beginning of the 1st,2nd,3rd,4th, and E 5th refueling cycles after completion of reracking. The refueling cycles are counted based on the Unit with the shorter cycle length. ' SHADED TEXT CONTAINS PROPRIETARY INFORMATION - Holtec International . 10-8 Report HI-982083 L. ) 11.0-- INSTALL.ATION L 11.1 ~ IIntroduction The construction phases of the Byron and Braidwood rack installations will be executed by Holtec International personnel. All construction work at Byron and Braidwood will be performed in compliance with NUREG-0612 and site-specific procedures. Crane and fuel bridge operators are to be adequately trained in the operation ofload handling ' machines'per the requirements of ANSI /ASME B30.2-19% and the plant's specific training program. -l l I l ' The lifting device designed for handling and installation of the new racks at Byron and i l . Braidwood is in compliance n ah the provisions of ANSI N14.6-1978 and NUREG-0612, 1 i including compliance with the primary stress criteria, load testing at a multiplier of maximum -! i l : working load, and nondestructive examination of critical welds. f e An intensive surveillance and inspection program shall be maintained throughout the rack Einitallation phase of the project. A set ofinspection points which have been proven,in numerous previous rack installatien campaigns, to eliminate any incidence of rework or erroneous - installation will be implemented. ! l l ~ 1 Holtec International anh Commonwealth Edison are developing a complete set of operating ! i procedures which cover the entire gamut of operations penaining to the rack installation effort. l 1 Similar procedures have been utilized and successfully implemented by Holtec International on l o previous' rack installation projects. These procedures assure that ALARA practices are followed ; and provide detailed requirements to assure equipment, personnel, and plant safety. The ' following is a list of procedures which will be used to implement the rack installation phase of ' l l the project.' i SHADED TEXT CONTAINS PROPRIETARY INFORMATION j Repon HI-982083 . Holtec International . 11 1 w

~~a. Receipt Inspection Procedure:~~

This procedure delineates the steps necessary to perform a. thorough receipt inspection of the new racks after their arrival on site. The receipt inspection includes dimensional measurements, . cleanliness _ in'spection, and visual ' weld examination. -- b. Cleaning Procedure: This procedure provides for the cleaning of the new racks, if it is required, in order to meet the requirements of ANSI N45.2.1, Level B. Permissible cleaning agents, methods and limitations on materials to be employed are provided.

~~c. Job Site Storage Procedure:~~

This procedure establishes the requirements for safely storing the new racks on site, in the event that long term site storage is necessary. This procedure provides environmental restrictions, i temperature limits, cleanliness requirements, and packaging requirements. I

~~d. Pre-Installation Drag Test Procedure:~~

This procedure stipulates the requirements for performing a functional test on the new racks prior to installation into the spent fuel pool. The procedure provides direction for inserting and withdrawing a " dummy" fuel assembly into designated cell locations, and establishes an

~~acceptance criterion in terms of maximum kinetic drag force during withdrawa'~~
~~e. Installation / Handling / Removal Procedure:~~

1 This procedure provides direction ~ for the handling / installation of the new high density racks and the removal of the existing racks. The procedure delineates the steps necessary to receive new high density racks on site, to unload and upright the racks, to stage the racks prior to installation, SHADED TEXT CONTAINS PROPRIETARY INFORMATION . Holtec International . Il-2 Report HI-982083 1 LL r and to install the racks. The ' procedure also provides' for the installation of new rack bearing pads, adjustment of the new rack pedestals and performance of the as-built field survey. Any y pool modifications that may be necessary, such as protrusion tmneation, are also described in the procedure; Finally, this procedure also outlines the rack disposal sequence.

~~f. ' h)ost-Installation Drag Test Procedure:~~

This procedure stipulates the requirements for performing a functional test on new racks following installation into the spent fuel pool. The procedure will provide direction for inserting d and withdrawing a " dummy" fuel assembly into designated cell locations, and establishes an . acceptance criterion in terms of maximum kinetic drag force. t L g. Underwater Diving Procedure: It is Comed's intention to perform the rack installation without using divers. However, should underwater diving operations become necessary to support the new rack installation, diving activities will be strictly controlled by Comed and Holtec procedures. These procedures describe j the method for introducing a diver into the spent fuel pool or cask pit, provide for radiological l monitoring during the operation, and define the egress of the diver from the fuel pool following j work completion. Funhermore, these procedures require strict compliance with OSHA Standard 29CFR-1910, Subpart T, and establish contingencies in the event of an emergency. l In addition to the procedural controls, underwater cameras will also be used to monitor the - I movements of the diver.

~~h. ALARA Procedure:~~

1 1 l Consistent with both Holtec International's and the Byron and Braidwood plants' ALARA l Programs, this procedure provides details to minimize the total person-rem received during the rack installation project, by. accounting for time, distance, and shielding. Additionally, pre-job - briefings are performed in order to mitigate the potential for overexposure. SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 11-3 Report HI-982083 i 1 l I

~~i. Liner Inspection Procedure:~~

a i In the event that a visual inspection of any submerged portion of the spent fuel pool liner is deemed necessary, this procedure describes the method to perform such an inspection, which ) ' includes the use of an underwater camera, and describes the requirements for documenting any observations. .]

~~j. Leak Detection Procedure:~~

~~1 This procedure describes the method to test the spent fuel pool liner for potential leakage using a vacuum box. This procedure may be applied to any suspect area of the pool liner. l J~~

~~k. Underwater Welding Procedure:~~

I t -In the event of a positive leak test result, an underwater welding procedure may be implemented which provides for the placement of a stainless steel repair patch over the area in question. The procedure contains appropriate qualification records documenting relevant variables, parameters, and limiting conditions. The weld procedure is qualified in accordance with AWS D3.6-93, Specification for Underwater Welding or may be qualified to an alternate code accepted by Commonwealth Edison and Holtec International. 11.2 Rack Arrangement The existing rack arrangement at Byron and Braidwood consists of 23 racks, which represent 2,870 storage locations (2,864 cells plus 6 failed fuel locations). The new rack arrangement consists of 24 free standing Holtee racks, which provide a total of 2,984 storage locations per site. Of these 2,984 storage cells, four racks, which account for 3% cells, are designated as Region I, and the remaining 20 racks totaling 2.588 cells are designated as Region II storage. A i schematic depicting the spent fuel pool after completion of the rerack can be seen in Figure 1-1. SHADED TEXT CONTAINS PROPRIETARY INFORMATION' . Holtec International 11-4 Report HI-982083 L l i t_ p i 11.3 Pool Survey and Inspection 'A pool inspection shall be performed to determine if any items attached to the liner wall or floor will interfere with the placement of the new racks or prevent usage of any cell locations subsequent to installation. In the event that protrusions are found which would pose any interference to the installation process, appropriate actions will be taken to ensure safe rack installation. 'T.4 Pool Cooling and Purification 11.4.1 Pool Cooling The pool cooling system shall be operated in order to maintain the pool water temperature at an acceptable level. It is anticipated that specific activities, such as bearing pad elevation ! measurements, may require the temporary shutdown of the Spent Fuel Pool cooling systera. At no time, however, will pool cooling be terminated in a manner or for a duration which would . create a violation of the Technical Specification. l 11,4.2 Purification . The existing spent fuel pool filtration system shall be operational in order to maintain pool clarity. Additionally, an underwater vacuum system will be used, as racks are removed, to . supplement fuel pool purification. The vacuum system may be employed to remove extraneous' debris, reduce general contamination levels prior to diving operations, and to assist in the i restoration of pool clarity following any pressure washing operations. I 1.5 Fuel ShufDing As new high density racks are installed into the pool, it is anticipated that fuel shufDes will be performed in independent phases in order to transfer irradiated assemblies from existing racks ' into Holtec racks. During the transition phase when both the existing racks and Holtec racks are ^ SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 11-5 Report HI-982083 u in the spent fuel pool, fuel shall be shuffled in accordance with the Technical Specification : . requirements specific for fuel stored in the existing racks and for fuel stored in the Holtec racks. Additionally, the parameters in the spent fuel pool will be maintained to ensure that ken s 0.95 at ! all times with both existing and Holtec racks present in the spent fuel pool. Fuel shuffle operations shall be conducted in accordance with station procedures and in a manner consistent with the rack installation sequence. Final shuffle operations shall not be conducted - during the new rack instaliation. 1 1 11.6 Installation of New Racks ' The new high density racks, which are supplied by Holtec International, will be delivered in the horizontal position. Each new rack will be removed from the shipping trailer using a suitably l rated crane, while maintaining the horizontal configuration, and placed upon an upender and 1 secured. Using two independent overhead hooks, or a single overhead hook and a spreader ' beam, the module will be uprighted into the vertical position. l. l The new rack lifting device will be installed into the rack and each lift rod successively engaged. Thereafter, the rack will be transporter'o a pre-leveled surface where the appropriate quality control receipt inspection will be performed. l As indicated in Section 11.5 above, a procedure shall be developed to shuffle spent fuel in accordance with Technical Specification requirements. Several existing racks will be emptied of spent fuel and removed from the spent fuel pool. Several Holtec racks are then subsequently placed m the pool. The Holtee racks will be loaded with spent fuel fuel in accordance with applicable Technical Specification requirements. The sequence is then repeated until all existing racks are removed from the pool and all Holtec racks are installed. In preparing the spent fuel pool for the rack installation, the pool floor will be inspected and any debris which may inhibit the installation of bearing pads will be removed. After the bearing pads SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International - 11-6 Repon HI-982083 1 ' l i l ) are positioned on the pool floor, elevation measurements will then be taken at each pad location. ] I

~~The rack pedestals will be adjusted in accordance with the bearing pad elevation measurements~~
~~in order to achieve levelness when the racks are installed.~~

l The new racks will be lifted using the 125-ton double girder bridge crane with a 15-ton auxiliary . ; hook. A temporary hoist with a suitably rated capacity may be attached to the overhead crane for installation activities in order to avoid submerging the main hook and causing contamination. For rack movements along the pool floor, the height of the rack above the liner shall not exceed i 1 six ~ inches, except where floor projections obstruct the path. Once the rack has reached its final ) position it will be carefully lowered onto its bearing pad. ) Elevation readings will be taken to confirm that the module is level and as-built rack-to-rack and rack-to-wall offsets will be recorded. The lifting device will be disengaged and removed from the fuel pool under Radiation Protection direction. A post-installation drag test may be performed using an inspection gage to ensure that no cell location poses excessive resistance to the insertion or withdrawal of a bundle. 11.7 Safety. Radiation Protection, and ALARA Methods 1 11.7.1 Safety During the rack installation phase of the project, personnel safety is of paramount importance, outweighing all other concerns. All work shall be carried out in strict compliance with approved procedures. 11.7.2 Radiation Prote . ion Radiation Protection personnel shall provide necessary coverage in order to provide radiological protection and monitor dose rates. The Radiation Protection department shall prepare Radiation . Work Pert .ts (RWPs) that will instruct the project personnel in the areas of protective clothing, general dose rates, contamination levels, and dosimetry requirements. SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 11-7 Report HI 982083 p i't-In addition, no activity within the radiologically posted area shall be carried out without the . knowledge and approval of Radiation Protection. Radiation Protection shall also monitor items removed from the pool, provide for the use of alarming dosimetry and supply direction for the L proper storage of radioactive material. l 1 ' 1-11.7.3 ALARA The key factors in maintaining project dose As Low As Reasonably Achievable (ALARA) are-time, distance, and shi:lding. These factors are addressed by utilizing many mechanisms with respect to project planni.,g and execution. Time l Each member of the project team will be properly trained and will be provided appropriate education and understanding of critical evolutions. Additionally, daily pre-job briefings will be employed to acquaint each team member with the scope of work to be performed and the proper l means of executing such tasks. Such pre-planning devices reduce worker time within the l radiologically posted area and, therefore, decrease project dose. Distance Remote tooling such as lift fixtures, pneumatic grippers, a support leveling device and a lift rod disengagement device have been developed to execute numerous activities from the pool surface, where dose rates are relatively low. By maximizing the distance between radioactive sources and project personnel, pre *t dose is reduced. Shielding During the course of the rack installation, the water in the spent fuel pool provides primary shielding.' The amount of water between an individual at the surface and an irradiated fuel assembly is an essential shie nt reduce: dose. Additionally, other shielding, may be employed to mitigate dose when work is performed around high dose rate sources. SHADED TF" CONTAINS PROPRIETARY INFORMATION Holtec International Il-8 Report HI-982083 e t $ 11.8 ' Radwaste Material Control . The main source of radioactive waste is the existing spent fuel racks. The disposal process for the existing racks is briefly described below. The existing racks will be washed, with pool water, prior to being removed from the pool to . remove as much contaminants as possible. After removal from the pool the racks will be bagged, . scaled and placed into a special DOT approved shipping container. The racks will be braced inside the container, prior to sealing the container, to prevent shifting during transit. Health.

~~Physics will monitor the packaging to assess dose rates and prevent dispersal of contaminants.~~

The container and the enclosed existing rack will be shipped to a volume reduction facility for processing and later delivered to a burial site. In addition to the existing spent fuel racks, radioactive waste generated from the rack installation effort may include vacuum filter bags, miscellaneous tooling, and protective clothing. Vacuum filter bags may be removed from the pool and stored as appropriate in a suitable container in order to maintain low dose rates. Contaminated tooling shall be properly stored per Radiation Protection direction throughout the project. 'At project completion, an effon will be made to decontaminate tooling to the most < - practical extent possible. l i l-i I l l' - SHADED TEXT CONTAINS PROPRIETARY INFORMATION ' Holtec International , Il 9 Repon HI-98'2083 r 12.0 ENVIRONMENTALCOST/BENEFIl' ASSESSMENT '12.1 Introduction [: Article V of the USNRC OT Position paper [12.1.1] requires the submittal of a cost / benefit

analysis for the chosen fuel storage iapacity enhancement method. This section abstracts the l analyses and evaluations made by Commonwealth Edison before selecting rack replacement as the most viable altemative.

12.2_ Imperative for Rack Reolacement The specific need to increase the existing storage capacity of the Byron and Braidwood spent fuel pools is based on the continuing degradation of Boraflex and a lack of viable economic i l alternatives. The Silica release associated with Boraflex degradation results in spent fuel pool f water clarity problems. Although additional spent fuel pool water filtering has temporarily improved water clarity, Boraflex degradation is a continuing process without remedy. Water ' clarity problems have caused refueling outage delays and also increases the potential for a reactivity management error. 12.3 Aooraisal of Altemative Options . Comed has determined that spent fuel storage rack replacement is by far the most viable option for the Byron and Braidwood plants in comparison to other alternatives. The. key considerations in evaluating the altemative options are provided below.

~~a. Saf. sly: minimize the number of fuel handling steps~~
~~b. Economy: minimize total installed and O&M cost~~
~~c. Security: protection from potential saboteurs, natural phenomena~~
~~d. Non-intrusiveness: minimize required modification to existing systems~~
e. Maturity: extent of industry experience with the technology SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 12-1 Report HI-982083

~~rl. i~~

~~f. ALARA: minimize cumulative dose due to handling of fuel L~~

_71nc Iniection - Boraflex degradation is an industry wide issue. It is a non-reversibi 7roblem that is expected to last for the life of the spent fuel storage racks. The Electric Power Research Institute (EPRI) has studied the use of chemicals (zine) to slow the release of silica. A Zinc injection process was performed at a BWR plant with some limited success reported. Zinc cannot be used at a PWR plant due to the presence of soluble boron in the spent fuel poel water. Additional Spent Fuel Pool Water Filtering The current plant practice is to provide additional spent fuel pool filtering to maintain pool water clarity and to dilute the silica concentration to less than 1 PPM during plant start up. The dilution requires extensive coordination and effort. Continued reliance on this practice is not - considered prudent because it does not address all aspects of the Boraflex degradation problem and it is not cost effective. Filtering requires 3 spent fuel pool filters, routine change-out of filter ! media and significant attention to the dilution process prior to plant restart. Further, this option does not adequately resolve the loss of storage cells due to Boraflex degradation. The Net Present- '.Value of this option does not show an advantage over rack replacement. l f On-Site Cask Storage l D y cask storage is a method of storing spent nuclear fuel in a nigh capacity container. The cask h provides radiation shielding and passive heat dissipation. Typical capacities for PWR fuel range from 21 to 37 assemblies that have been removed from the reactor for at least five years. The casks, once loaded, are then stored outdoors on a seismically qualified concrete pad. The pad ' will have to be located away from the secured boundary of the site because of site limitations. SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 12-2 Report HI-982083 e a The storage location will be requimd to have a high level of security which includes frequent l . 4 I tours, reliable lighting, intruder deu:ction, and continuous visual monitoring. ) i i The ca *.;, as presently licensed, are limited to 20-year storage service life. Once the 20 years has expimd the cask manufacturer or the utility must recertify the cask or the utility must remove the spent fuel fmm the container. In the interim, the U.S. DOE has embraced the concept of multi- - ] purpose canister (MPC), obsolescirig all existing licensed cask designs. Work is also continuing by several companies to provide an MPC system that will be capable of long term storage, transport, and final disposal in a repository.. l For example, the plant must provide for a decontamination facility where the outgoing cask can , be decontaminated for release. l There are several plant modifications required to support cask use. Tap-ins must be made to the gaseous waste system and chilled water to support vacuum drying of the spent fuel and piping must be installed to return cask water back to the spent fuel pool / cask pit. A seismic concrete pad must be made to store the loaded casks. This pad must have a security fence, surveillance protection, emergency power, and video surveillance. - Finally, facilities must be provided to vacuum dry.the cask, back fill it with helium, perform leak checks, remachine the gasket surfaces if leaks persist, and assemble & cask on-site. . Presently, no MPC cask has been licensed. Because of the continued uncertainty in the government's policy, the capital investment to use a dry storage system is considered to be an inferior alternative for Byron and Braidwood at this time. l t SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec Intemational 12-3 Report HI-982083 r Modular Vault Dry Storage ' Vault storage consists of storing spent fuel in shielded stainless steel cylinders in a horizontal configuration in a reinforced concrete vault. The concrete vault provides radiation shielding and ' missile protection. It must be designed to withstand the postulated seismic loadings for the site. "A transfer cask is needed to deliver the storage canisters from the fuel pool. The plant must provide for a decontamination bay to decontaminate the transfer cask and connection to its gaseous waste system and chilled water systems. A collection and delivery system must be installed to retum the pool water entrained in the canisters back to the fuel pool. Provisions for canister drying, helium injection, handling and automatic welding are also necessary. The storage area must be designed to have a high level of security. Due to the required space, the - vault secpmd area must be located outside the secured perimeter. Consideration of safety and security requires it to have its own video surveillance system, intrusion detection, and an autonomous power source. ' Some other concems relating to the vault storage system are: the inevitable " repackaging" for L shipment to the DOE repository, the responsibility to eventually decommission the new facility, t the large " footprint" (land consumption), the potential fuel handling accidents, the potential fuel / clad rupture due to high temperatures, and the high cost. At the present time, no MPC technology based vault system has been licensed for fuel transport. The high cost and uncertainty make this option less prudent. Horizontal Silo Storage A variation of the horizontal vault storage technology is more aptly referred to as " horizontal silo" storage. This technology suffers from the same drawbacks that other dry cask technologies have, namely: SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 12-4 Report HI-982083 ~

~~a. No fuel with cladding defects can be placed in the silo.~~

~~F Concern regaiding long term integrity of the fuel at elevated temperature.~~

~~c. Potential for eventual repackaging at the site.~~
~~d. Potential for fuel handling accidents.~~
~~e. Relatively high cumulative dose to personnel in effecting fuel transfer (compared to reracking).~~
~~f. Compatibility of reactor / fuel building handling crane with fuel transfer hardware.~~
~~g. Potential incompatibility with DOE shipment for eventual off-site shipment.~~
~~h. Potential for sabotage.~~

Dry storage technology is not intended as a total replacement icr spent fuel storage pools at operating plants and therefore no dry storage system was considered technically or economically viable. New Fuel Pool Constructing and licensing a new fuel pool is not a practical alternative for either plant since such an effort may take up to 10 years. Moreover, the cost of this option is prohibitively high. I An estimate of relative costs in 1998 dollars for the aforementioned options is provided in the following: Rack replacement: $23 million Horizontal Silo (NUHOMS): $15 to 20 million Metal cask: $15 to 20 million Modular vault: $56 million New fuel pool: $150 million i Comed's estimate of comparative costs of various options is consistent with other published industry data [12.3.1,12.3.2]. SHADED 7 EXT CONTAINS PROPRIETARY INFORMATION Holtec Intemational 12-5 Report HI-982083 j To summarize, th:re are no acceptable alternatives to replacing spent fuel storage racks at Byron and Braidwood Nuclear Stations. First, there are no commercial independent spent fuel storage . facilities operating in the U.S. Second, the adoption of the Nuclear Waste Policy Act (NWPA) created a de facto throw-away nuclear fuel cycle. Since the cost of spent fuel reprocessing is not offset by the salvage value of the residual uranium, reprocessing represents an added cost for the nuclear fuel cycle which already includes the NWPA Nuclear Ware Fund fees. In any event, there are no domestic reprocessing facilities. 12.4 Cost Estimate Tne proposed construction contemplates the replacement of the Byron and Braidwood spent fuel 4 1 pool racks using free-standing, high density, Boral poisoned spent fuel racks. This will completely remove all of the racks containing Boraflex. The engineering and design work is completed for the project. The total capital cast for both sites is estimated to be approximately $23 million as detailed below. Engineering, design, project management: $1.5 million Rack fabrication: $15.5 million Rack installation: $ 6 million As described in the preceding section, several alternatives were considered prior to proceeding with rack replacement. Rack replacement provides a definite cost advantage over other technologies. 12.5 Resource Commitment The replacement of the spent fuel pool storage racks is expected to require the following primary - resources. SHADED TEXT CONTAINS PROPRIETARY INFORMATION Holtec International 12-6 Report HI-982083 m Stainless steel: 470 tons - Boral neutron absorber: 30 tons (25 tons Boron Carbide,5 ton aluminum) The mquirements for stainless steel and aluminum represent a small fraction of total world output of these metals (less than 0.001%). Although the fraction of world production of Boron Carbide required for the fabrication is somewhat higher than that of stainless steel or aluminum, it is unlikely that the commitment of Boron Carbide to this project will affect other related activities. Experience has shown that the production of Boron Carbide is highly variable and depends upon need and can easily be expanded to accommodate worldwide needs. 12.6 Environmental' Considerations Rack replacement will not increase the maximum bulk pool temperature above the previously licensed value. Therefore, the cooling water demand and the water vapor emission to the environment remains unchanged. ~ 12.7 References for Section 12 (12.1.1] SNRC Ietter to All Power Reactor Licensees transmitting the "OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications," April 14,1978, and Addendum dated January 18,1979. [12.3.1) Electric Power Research Institute, Repon No. NF-3580, May 1984. [12.3.2] . " Spent Fuel Storage Or Jons: A Critical Appraisal," Power Generation Technology, Sterling Publishers, pp. 137-140, U.K. (November 1990). ' SHADED TEXT CONTAINS PROPRIETARY INFORMATION - Holtec International - 12-7 Report HI-982083 /

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