ML20134F874
| ML20134F874 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 09/04/1996 |
| From: | Ohanlon J VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.) |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML19353D965 | List: |
| References | |
| 96-409, NUDOCS 9611080111 | |
| Download: ML20134F874 (125) | |
Text
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VIRGINIA ELzcTRIc AND Powza COMPANY !
Ricnwown,VIBODtIA 23261 September 4, 1996 U.S. Nuclear Regulatory Commission Serial No.96-409 Attention: Document Control Desk NAPS /GSS/ETS R0 Washington, D.C. 20555 Docket Nos. 50-338 50-339 License Nos. NPF-4 NPF 7 Gentlemen:
VIRGINIA ELECTRIC AND POWER COMPANY NORTH ANNA POWER STATION UNITS 1 AND 2 PROPOSED TECHNICAL SPECIFICATIONS CHANGE USE OF t FAD FUEL AgSsunLIES WITH ADVANCED CLADDING MATERIAL Pursuant to 10 CFR 50.12 and 10 CFR 50.90, \/irginia Electric and Power Company requests an exemption from certain requirements of 10 CFR 50.44,10 CFR 50.46, and Appendix K of 10 CFR 50, and changes to the license and Technical Specifications for Facility Ooerating License Numbers NPF-4 and NPF-7 for North Anna Power Station Units 1 and 2, respectively. The proposed exemptions, license and Technica' Specification changes will allow the use of four lead test assemblies fabricated by Framatome Cogoma Fuels (FCF). NRC permission is also requested to apply Virginia Electric and Power Company's standard reload design methodology to the North Anna cores in which these four lead test assemblies are irradiated. Previously application of Virginia Electric Power Company's standard reload design methodology was limited for use only with Westinghouse fuel.
Implementation of this Technical Specification change requires an exemption from certain requirements of 10 CFR 50.44, " Standards for combustible gas control system in light water-cooled power reactors," 10 CFR 50.46, ' Acceptance criteria for emergency core cooling systems for light water nuclear power reactors," and Appendix K of 10 CFR 50, 'ECCS Evaluation Models." The basis for the exemption from the requirements of 10 CFR 50.44,10 CFR 50.46 and 10 CFR 50 Appendix K is included in Attachment 1.
These lead test asserrm 'ios will use two advanced zirconium-based alloys which do not fit the design specifications for either Zircaloy or ZlRLO for the fabrication of the structural tubing (guide thimbles and instrumentation tube) and fuel rod cladding.
A discussion of the proposed Technical Specifications changes, including an evaluation of the safety significance of irradiating the lead test assemblies, is provided in Attachment 2. The proposed Technical Specifications changes are provided in Attachment 3.
96110so111 96o904 PDR P ADOCK 05ooo338 \' i ppy g
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The proposed exemptions and licenseffechnical Specifications changes have been reviewed and approved by the Station Nuclear Safety and Operating Committee and the Management Safety Review Committee. It has been determined that the proposed 4 exemptions, license and Technical Specifications changes, and the use of the lead test )
assemblies supported by this change do not involve an unreviewed safety question as )
defined in 10 CFR 50.59 or a significant hazards consideration as defined in 10 CFR I 50.92. The basis for our determination that the changes do not involve a significant l hazards consideration is provided in Attachment 4.
The lead test assemblies are currently scheduled to begin operation in North Anna Unit 1 in the spring of 1997. To support the planned operation of these assemblies, we request approval of the proposed license and Technical Specifications changes and issuance of the necessary exemptions by February,1997.
Attachment 2 contains information that is proprietary to Framatome Cogema Fuels.
This is supported by an affidavit (Attachment 5) signed by J. H. Taylor, Manager -
Licensing Services, Framatome Cogema Fuels, Inc. This affidavit sets forth the basis on which the information may be withheld from public disclosure by the Commission, and addresses the considerations listed in 10 CFR 2.790. Accordingly, it is requested that Attachment 2 of this letter, which contains information proprietary to FCF, be )
withheld from public disclosure in accordance with 10 CFR 2.790. In compliance with l the guidelines of NUREG-0390, a copy of Attachment 2 in which the proprietary information has been identified is also being provided as Attachment 6.
Should you have any questions or require additional information, please contact us.
1 Very truly yours, P % JL James P. O'Hanlon Senior Vice President - Nuclear Attachments
1 cc: U.S. Nuclear Regulatory Commission Region 11 101 Marietta Street, N.W.
Suite 2900 Atlanta, Georgia 30323 Mr. R. D. McWhorther NRC Senior Resident inspector North Anna Power Station Commissioner Department of Radiological Health Room 104A 1500 East Main Street I Richmond, VA 23219 l
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COMMONWEALTH OF VIRGINIA )
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COUNTY OF HENRICO )
The foregoing document was acknowledged before me, in and for the County and Commonwealth aforesaid, today by J. P. O'Hanlon, who is Senior Vice President -
Nuclear, of Virginia Electric and Power Company. He is duly authorized to execute and file the foregoing document in behalf of that Company, and the statements in the
, document are true to the best of his knowledge and belief.
Acknowledged before me this N day of $n,hs.1984_.
My Commission Expires: r M AC 8/
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Notary Public (SEAL)
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ATTACHMENT 1 l 4
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VIRGINIA ELECTRIC AND POWER COMPANY !
l BASIS OF EXEMPTION REQUEST l
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' Serial No.96-409 Advanced Fuel TS Change Package Page 1 of 1 bc:
Mr. D. A. Heacock - NAPS (without Attachments 5 and 6)
Mr. B. L. Shriver - SPS (without Attachments 5 and 6)
, Mr. P. A. Kemp - NAPS (without Attachments 5 and 6)
Ms. C. G. Lovett - SPS (without Attachments 5 and 6)
, Mr. G. L. Darden - 1 N-3SW (letter only)
Mr. N. P. Wolfhope - 1N-3SW (without Attachments 5 and 6)
Mr. J. B. Lee - IN2SE (without Attachments 5 and 6)
. Licensing File - GOV 02-54B (without Attachments 5 and 6) j MSRC Coordinator (without Attachments 5 and 6)
Records Management - GOV 02-54B (bc original) - IN-GW NOB Distribution (without Attachments 5 and 6)
Concurrence:
Mr. J. P. O'Hanlon Mr. M. R. Kansler M ?
Mr. R. F. Saunders /VM Mr. M. L. Bowling Af(1/ drNd Mr. R. M. Berryman W r4 4r 5 WJ4 8/W
Mr. W. R. Matthews 4*'4
- A M* c5r M//'
Mr. D. A. Sommers ud)- 8hk%
Verification of Accuraev:
- 1. Approved North Anna Technical Specification Change Request No. 337 dated August 28,1996.
Reauired Chanaes to the UFSAR or the Tooical Reoort:
- 1. Yes, UFSAR Change Request (ru %-v50 Action Plan / Commitments (Stated or imolied):
- 1. , Corporate Licensing to process UFSAR Change Request, upon issuance of the Technical Specification Amendment.
a REGULATORY BASIS FOR SPECIFIC EXEMPTIONS i
e Virginia Electric and Power Company plans to irradiate four (4) fuel assemblies fabricated by Framatome Cogema Fuels (FCF) at North Anna. Operation of these lead test assemblies is currently
- scheduled to begin in North Anna 1 Cycle 13, in the second quarter of1997. Rese fuel assemblies i
will be very similar to the FCF Mark-BW fuel assembly design that has previously been irradiated
! in other Westinghouse-designed reactors. However, the North Anna fuel assemblies will incorporate several new features, including use of two advanced zirconium-based alloys, Alloy 4 and Alloy 5, for the fuel rod cladding. %e majority of the fuel rods will have cladding fabricated from Alloy 5, j but two of the assemblies will also contain a small number of fuel rods with cladding fabricated from Alloy 4. These two alloys have previously been used as cladding materials for limited numbers of L ' fuel rods in demonstration assemblies in the McGuire Unit I and Three Mile Island Unit I reactors ,
j as well as in several European reactors. The North Anna lead test assemblies will differ from these i
demonstration assemblies in using advanced alloys as the cladding materials for all fuel rods in the assemblies.
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!' In support of the proposed irradiation of these lead test assemblies, exemptions are being requested to 10 CFR 50.46 and 10 CFR 50.44, which specifically refer to fuel with Zircaloy or ZIRLO
}_ cladding, and Paragraph I.A.5 of Appendix K to 10 CFR Part 50, which requires use of a specific model that was originally derived for Zircaloy clad fuel.
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10 CFR 50.12 states that the Commission may grant an exemption from requirements contained in j
10 CFR 50 provided that: 1) the exemption is authorized by law, 2) the exemption will not result
] in an undue risk to the public health and safety,3) the exemption is consistent with the common defense and security, and 4) special circumstances, as m def' ed in 10 CFR 50.12(a)(2), are present.
! The requested exemptions to allow the use of advanced zirconium based alloys other than Zircaloy or ZlRLO for the fuel cladding material in four lead test assemblies to be supplied to North Anna Power Station by Framatome Cogema Fuels satisfy these requirements as described below.
l 1. The requested exemntion is authnrized by law.
f Lead test assembly programs and irradiation ofnew materials are not precluded by any law.
l The FCF lead test assemblies to be irradiated at North Anna incorporate cladding materials l which do not conform to the cladding material designations explicitly defined in 10 CFR l
' 50.44 and 10 CFR 50.46 (i.e., Zircaloy or ZIRLO). However, the criteria of these sections will continue to be satisfied for North Anna cores incorporating these fuel assemblies.
Similarly, Appendix K of 10 CFR 50 requires use of the Baker-Just equation, which was developed for use with Zircaloy clad fuel. Although the lead test assemblies at North Anna j will use different zirconium-based alloys for the fuel rod cladding, the Baker-Just equation was determined to be appropriate for evaluation of these materials, and was applied to the loss of coolant accident (LOCA) analyses of the lead test assemi: lies. Therefore, issuance i ,
of exemptions to allow the use of cladding materials other than Zircaloy or ZIRLO in the North Anna lead test assemblies will not result in the violation of the criteria of the j applicable sections of10 CFR 50.
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The reauested exemntion does not present an undue risk to the public health and safe i The safety evaluation for the use of the lead test assemblies demonstrated that i
safety as defmed in the Bases to any North Anna Technical Specification is not reduced.
! of the lead test assemblies will not increase the probability of occurrence or the col 2 of an accident at the North Anna Power Station, and will not create the possibility for a or different type of accident which could pose a risk to public health and safety. Sa analyses which are based on full cores of Westinghouse fuel and which are suppo applicable North Anna Unit I and North Anna Unit 2 Technical Specifications will remain i applicable for cores incorporating the lead' test assemblies.
i ' For each applicable reload cycle, the lead test assemblies will be specifically e Virginia Electric and Power Company's standard reload design methods. This will inclu consideration of the core physics analysis peaking factors and core average linear heat r
- effects. Cores incorporating the lead test assemblies will be operated in accordance with operating conditions identified in the Technical Specifications. In the urMely event that cladding failures occur in the lead test assemblies during normal operauon of the co i
environmental impact will be minimal and bounded by previous environmental assessme 3.
The reonested exemntion will not endanoer the common defense and securit.y.
l The lead test assemblies are similar to normal reload fuel assemblies, and the specia j matenal used in these assemblies will be procured, handled and controlled in accordance with approved procedures. Use of the four FCF lead test assemblies will not affect the Operation of the North Anna Power Station or endarger the common defense and se 4.
Snecial circumstances are present which necess;,,,, the reone=t for an exemntion to the lations of 10 CFx 50.44.10 CFK S0.46. and pararanh 115 of Apnendiv Id to 10 CFR i
j j Pursuant to 10 CFR 50.12(a)(2), the NRC will not consider granting an exemption to the regulations unless special circumstances are present. The requested exemptions meet the l '
special circumstances ofparagraph (a)(2)(ii), in that application of these regulations in this particular circumstance is not rswory to achieve the underlying purpose of the regulations
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- The underlying purpose of 10 CFR 50.46 is to ensure that nuclear power facilities j
i have adequate acceptance criteria for their ECCS. The effectiveness of the ECCS at North Anna Units 1 and 2 will not be affected by the insertion of the four lead test
assemblies. Although these assemblies incorporate cladding materials other than
- those explicitly defined in 10 CFR 50.46, the criteria of this section will continue to be satisfied for North Anna cores incorporating the lead test assemblies. Safety i analyses based on the resident fuel design will remain applicable for cores which i
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- incorporate the lead test assemblies. Thus use of the advanced zirconium-based cladding materials will not have a detrimental impact on the performance of the i
North Anna cores under LOCA conditions.
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l e The intent of 10 CFR 50.44 is to ensure that there is an adequate means of controlling the hydrogen generated following a LOCA. The post-LOCA hydrogen source which
, is relevant to the lead test assemblies is the metal-water reaction between the fuel rod ,
cladding and the reactor coolant. The Baker-Just equation has been confirmed to j conservatively assess the metal-water reaction rate for the advanced zirconium-based j alloys. Therefore, the amount of hydrogen generated by metal-water reaction in l these materials will be within the design basis for the North Anna units.
e The intent of Paragraph I.A.5 of Appendix K of 10 CFR Part 50 is to apply an equation for rates of energy release, hydrogen generation, and cladding oxidation ,
from a metal-water reaction that conservatively bounds all post-LOCA scenarios.
Application of the Baker-Just correlation will continue to conservatively bound all '
post-LOCA scenarios for the use of the lead test assemblies due to the similarities between the compositions of the advanced zirconium-based alloys and Zircaloy-4.
Therefore, the intent of 10 CFR 50.46,10 CFR 50.44, and 10 CFR Part 50, Appendix K, will continue to be satisfied for the planned operation of the lead test assemblies. Issuance of a j temporary exemption from the criteria of these regulations for the irradiation of these four assemblies in the North Anna reactors will not compromise the safe operation of the reactors.
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t VIRGINIA ELECTRIC AND POWER COMPANY DISCUSSION OF PROPOSED TE'CHNICAL SPECIFICATION CHANGES NORTH ANNA UNITS 1 AND 2 4
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l DISCUSSION OF CHANGES 1
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l'. INTRODUCTION l . Virginia Electric and Power Company plans to insert four (4) fuel assemblies fabricated by 5
. Fr===tama Cogema Fuels (FCF) into the North Anna 1 Cycle 13 core, My =4= Mad to begin 5~
sp. Gen in the second quarter of 1997. 'Ibese fuel ataemblies will be very similar to the FCF Mark-BW fuel assembly design which has previously been irradiated in other WMy.c=> a- "
jesigned reactors.[
M ama l Although the lead test masambly program is currently planned for implementation only in North Anna Unit 1, Virginia Electric and Power C4-5 =y is r=q'W NRC approval to irradiate the lead assemblies in either North Anna Unit 1 or North Anna Unit 2 to marimim program flexibility.
Based on evaluations and analyses, no unreviewed safety questions exist as a result ofinserting the advanced cladding materials or fuel meeamblies of the FCF design into the North Anna Units 1 and 2 reactor cores. However, the Technical SpWcations for both North Anna Unit I and North Anna Unit 2 define the fuel rod cladding material as either Zircaloy-4 or ZIRLO. Use of a different cladding material in the lead test assemblies therefore requires changes to the Technical Specifications and a license condition to permit use of the lead test assemblies. Exemptions are also required to 10 CFR 50.46 and 10 CFR 50.44, which specifically refer to fuel with Zircaloy or ZlRLO cladding, and Appendix K of 10 CFR Part 50, which requires use of a specific model originally derived for Zircaloy clad fuel. In addition, the Safety Evaluation report for our standard reload nuclear design methodology (VEP-FRD-42 Rev.1-A) specified that, in its present form, our methodology was approved only for application to Westinghouse-supplied fuel in Westinghouse-supplied reactors. 'lherefore NRC concurrence is also required to apply Virginia Electric and Power Company's standard reload design methodology to cores containing these lead test assemblies. ,
LEAD TEST ASSEMBLY PROGRAM Framatome began an advanced alloy program in 1987 to develop fuel rod cladding and structural tube materials for high bumup application. This program has involved extensive testing of several candidate alloys, with two alloys being selected for further characterization on the basis of their superior performance in both in-core and out of core testing. Demonstration assemblies which included both alloys have been irradiated in three European reactors as well as in Duke Power
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i Company's McGuire Unit 1. Imdiation of these advanced alloys is continuing in additional
! European reactors and in Three Mile Island Unit 1 in the United States.
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! Framatome Cogema Fuels (FCF) and the Virginia Electric and Power Company have entered into l an agreement to irradiate four (4) lead test assemblies with advanced FCF fuel design features. His l program is awp-*~i to provide additional information on the behavior of the FCF advanced cladding i materials under commercial operating conditions to a lead rod burnup of approximately 55 to 60 i GWD/MTU in three 18-month operating cycles, as well as to demonstrate the performance of the l FCF fuel assembly advanced mechanical design features under in-reactor conditions. Use of the four i lead test amm mblies is currently planned to begin at North Anna Unit I starting with Cycle 13, which
- is currently scheduled for late spring,1997.
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- Post irradiation examinations of the lead test assemblies will be performed during the lead test assembly program as permitted by the North Anna refueling schedule. In addition to visual examinations, these examinations may include
- measurement of fuel antembly length and bow, holddown spring compression testing, functional testing of the quick disconnect locking =~haai-a; oxide measurements on fuel rods and guide thimbles, and measurements of fuel rod diameter and length. Depaading on the outcome of the cunent Waeinghause Owners Group evaluation of contml rod insertion in high burnup fuel assemblies, control rod drag testing may also be desirable once the lead test assemblies reach high burnups.
The current fuel in North Anna Units 1 and 2 is the North Anna Improved Fuel (NAIF) design, which is a Westinghouse 17x17 VANTAGE-5H design, into which additional debris resistance features and ZIRLO fuel rod cladding and skeleton components have-subsequently been incorporated. Descriptions of the fuel design can be found in our submittals to the NRC for the implementation of the NAIF design, dated January 15,1990 (Reference 1), and for the implementation of ZIRLO, dated October 4,1993 (Reference 2).
The North Anna lead test assemblies will be mechanically similar to, and fully compatible with, the resident Westinghouse fuel assemblies. The primary differences between the resident Westinghouse fuel design and the FCF the lead test assemblies include the use of the different zirconium-based alloys for fuel rod cladding and fuel assembly structural tubing, incorporation of the mid-span mixing grids into the lead test assemblies, and use of a higher nominal fuel pellet density in the lead test assemblies, which will result in a higher uranium loading than in the Westinghouse fuel design.
Incorporation of the quick release top nozzle design, the use of the fine mesh debris filter bottom nozzle, and the use of FCF's axially ' floating' grid design (versus the more rigid attachment of the Westinghouse grids) are not expected to affect the compatibility of the lead test assemblies with the resident fuel.
The areas assessed during the safety evaluation piucess included: chemicWmechanical properties, neutronic performance, thermal and hydraulic performance, cladding performance under non-LOCA conditions, and cladding performance under LOCA conditions. These evaluations and analyses have shown that the present safety related design bases and calculations are applicable for North Anna cores which incorporate the lead test assemblies.
The use of the FCF lead test assemblies does not alter the models end methods used for analyzing j- cycle specific reloads ofNorth Anna fuel (References 3 and 4). Analyses and evaluations performed j by Virginia Electric and Power Company, Framatome Cogema Fuels (FCF), and Framatome i Technologies Inc. (FTI, the division of Framatome which performed the LOCA analyses) to support this conclusion are described in the attached Safety Significance evaluation. Plant and cycle specific
! evaluations and analyses will continue to be performed for North Anna Units 1 and 2 core designs j" to demonstrate that the current design bases and limits remain valid for cores containing the lead test assemblies, i
i LICENSE CONDITIONS AND TECHNICAL SPECIFICATIONS CHANGES
- General
! Although the FCF lead test assembly program is currently planned for implementation in North
} Anna Unit 1, Virginia Electric and Power Company is requesting NRC approval to irradiate the lead
- assemblies in either North Anna Unit 1 or North Anna Unit 2. This flexibility will support a more j
timely completion of the program if, for example, after 1 or 2 cycles ofirradiation a more extensive characterization of the assemblies is desired than could be supported by the North Anna 1 refueling schedule. In such a case, the program could then be expedited by reinserting the lead test assemblies
! into Unit 2 after the testing is completed, rather than waiting until the next North Anna 1 refueling j outage. The license and Technical Specification changes described herein therefore apply to both
! North Anna Unit I and North Anna Unit 2.
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Licence Conditions t
. - License conditions are being requested to permit use of the FCF lead test assemblies in North Anna l Units 1 and 2. These license conditions will permit the use of up to four (4) assemblies with the l
advanced zirconium based alloys. The following license conditions are being proposed for both Unit '
- I and Unit 2.
- " Virginia Electric and Power Company may use up to four (4) fuel assemblies containing advanced zirconium based alloys as described in the licensee's submittal I
dated September 4,1996."
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! For North Anna Unit 1, the proposed license condition replaces existing licensing condition l 2.D.3.(d). For North Anna Unit 2, a new license condition,2.C.(24), is being created.
Technical Snecification 5.3.1 i
The Design Features section on the Fuel Assemblies (Technical Specification 5.3.1) will be changed j l to allow use of fhel rods with slightly different nominal dimensions or rods clad with materials other ;
. than Zircaloy-4 or ZIRLO with an approved exemption or license condition. This change acknowledges that fuel assemblies, such as those specifically approved for use as lead test 3
assemblies, may sometimes be irradiated that have slightly different features than the resident fuel (e.g., the Alloy 4 and Alloy 5 fuel rod cladding in the FCF lead test assemblies). i l
Technieml Snecifientinn 6.9.1.7 b l
Thorough evaluation of the performance of fuel that has features different from the resident fuel l assemblies may require the use ofNRC-approved models and methods beyond those normally used i to evaluate North Anna fuel. An example of this is the use of the Framatome Technologies Inc.
(FTI) NRC-approved models to perform the LOCA evaluation oflead test assemblies with Alloy 4 and Alloy 5 fuel rod ckdding. Therefore, t'ne identification of applicable references for the methods used to demi = the North Anna core operating limits (Technical Specification 6.9.1.7.b) is being modified to note that such additional approved methods may be used with an approved exemption or license condition. The specific applicable models and methods will be identified in the documentation supporting the request for the exemption or license condition.
REOUEST FOR FXFMPT10NS Title 10 CFR 50.46(a)(1)(i) states, "Each boiling and pressurized light-water nuclear power reactor fueled with uranium oxide pellets within cylindrical zircaloy or ZIRLO cladding must be provided with an emergency core cooling system (ECCS) that must be designed so that its calculated cooling performance following postulated loss-of-coolant accidents conforms to the criteria set forth in paragraph (b) of this section. ECCS cooling performance must be calculated in accordance with an acceptable evaluation model and must be calculated for a number of postulated loss-of-coolant accidents of different sizes, locations, and other ymysties sufBeient to provide assurance that the most sever postulated loss-of-coolant accidents are calculated." Section 10 CFR 50.46 goes on to delineate specifications for peak cladding temperature, maximum cladding oxidation maximum hydrogen generation, coolable geometry, and long~-
term cooling.
In addition,10 CFR 50.44 (a) states, "Each boiling or pressurized light-water nuclear power reactor fueled with oxide pellets with cylindrical zircaloy or ZIRLO cladding, must, as provided in paragraphs (b) through (d) of this section, include means for control of hydrogen gas that may be generated, following a postulated loss-of-coolant accident (LOCA)..."
Since 10 CFR 50.46 and 10 CFR 50.44 specifically refer to fuel with Zircaloy or ZIRLO clad, the use of fuel clad with zirconium-based alloys that do not conform to either of these two designations I in the North Anna lead test assemblies requires an exemption from these sections of 10 CFR 50.
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Further,10 CFR Part 50, Appendix K, paragraph I.A.5 states,
- "The rate of energy release, hydrogen generation, and cladding oxidation from the
- metal water reaction shall be calculated using the Baker-Just equation."
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Since the Baker-Just equation was originally developed for the use of Zircaloy cladding, the use of 1 j fuel with the advanced zirconium-based alloys also requires an exemption from this section of the i code.
l The FCF lead test assemblies to be irradiatM st North Anna will use fuel rod cladding fabricated l from two advanced zirconium-based alloys, which do not meet the definition of Zircaloy or ZIRLO.
As a result, a temporary exemption to 10 CFR 50.46,10 CFR 50.44, and 10 CFR Part 50, Appendix l K, is required for the irradiation of these four lead test assemblies. A safety evaluation has been j
, performed to demonstrate that the intent of these regulations are satisfied for the lead test assemblies
- and to confirm thu no unreviewed safety question exists. Specifically
- i ir j e The underlying intent of 10 CFR 50.46 is to ensure that nuclear power facilities have l
- adequate acceptance criteria for their ECCS. The effecdveness of the ECCS at North i Anna Units 1 and 2 will not be affected by the insertion of the four lead test l assemblies. Due to similarities in materials properties of the two advanced j zirconium-based alloys to Zircaloy-4, the ECCS performance in the North Anna i reactors will not be adversely affected by the presence of the lead test assemblies.
Thus use of the advanced zirconium-based cladding materials will not have a detrimental impact on the performance of the North Anna cores under LOCA conditions.
- The intent of 10 CFR 50.44 is to ensure that there is an adequate means of controlling l generated hydrogen. The post-LOCA hydrogen source which is relevant to the lead j test assemblies is the metal-water reaction between the fuel rod cladding and the l reactor coolant. The Baker-Just equation, which is used to assess the metal-water i reaction rate for Zircaloy-4, has been confirmed to conservatively assess the metal-water reaction rate for the advanced zirconium-based clloys as well. Therefore, the j amount of hydrogen generated by metal-water reaction in these materials will be j within the design basis for the North Anna units, and existing plant specific analyses
- for the total hydrogen generation following a LOCA will remain applicable for use of the lead test assemblies.
- The intent of Paragraph I.A.5 of Appendix K of 10 CFR Part 50 is to apply an j equation for rates of energy release, hydrogen generation, and cladding oxidation i from a metal-water reaction that conservatively bounds all post-LOCA scenarios.
- Application of the Baker-Just correlation will continue to conservatively bound all .
i post-LOCA scenarios for the use of the lead test assemblies due to the similarities l between the compositions of the advanced zirconium-based alloys and Zircaloy-4.
l Therefore, the intent of 10 CFR 50.46,10 CFR 50.44, and 10 CFR Part 50, Appendix K, will i
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f continue to be satisfied for the planned operation of the lead test assemblies. Issuance of a temporary exemption from the criteria of these regulations for the irradiation of these four assemblies in the North Anna reactors will not compromise the safe operation of the reactors.
1 l Finally, the NRC Safety Evalcation Report (SER) which approved Virginia Electric and Power !
! Company's reload nuclear design methodology (VEP-FRD-42 Rev.1-A) concluded that the report
! is "..mysble for referencing by Virginia Power in licensing Weinghause supplied reloads of Waeinghause supplied reactors." Since these lead test assemblies will be supplied by Framatome Cogema Fuels, NRC permission is requested to apply the Virginia Electric and Power Company i standard reload design methodology n North Anna cores containing the four lead test assemblies.
l j As documented in the attached Safety Significance evaluation, use of the lead test fuel assemblies i in the North Anna cores has been thoroughly evaluated and will have no discernible impact on the
- overall core performance. Incorporation of these assemblies into North Anna cores will not affect
] the ability of the reload methodology to predict the core performance or to conservatively assess the
- core response to accident scenarios.
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- SAFETY SIGNIFICANCE
SUMMARY
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{ The FCF lead test assemblies are mechanically and neutronically very similar in design to the Westinghouse fuel that comprises the remai* of the core. The reload core design for North Anna cycles which incorporate the lead test assemblies will meet all applicable design criteria, and will not result in any changes to the North Anna Units 1 and 2 operating and safety analysis limits. The
! existing safety analyses based on the resident Weinghause fuel design will remain applicable for i
cores incorporating the FCF lead test assemblies. Analyses or evaluations will be performed each cycle to confum that the criteria in 10 CFR 50.46 will be met. Use of the FCF lead test assemblies I will not result in an unreviewed safety question as defined in 10 CFR 50.59, and will not constitute a significant hazards consideration as defined in 10 CFR 50.92.
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! TECHNICAL AND SAFETY EVALUATION l
! ~1. Lead Test Assembly Design Description I
j ne lead test anamhlies to be irradiated at North Anna are very similar to the Mark-BW fuel
} assemblies that have been supplied by Framatome Cogema Fuels (FCF) to five other Wdaghause-
- ' designed operating units. The Mark-BW design, which is described in detail in Reference 5, is a l 17x17 standard lattice, Zircaloy spacer grid fuel assembly designed specifically for use in j WMagha'e units.
i l The lead test assemblies for North Anna, which are designated the Mark-BW17 design (Figure 1.1),
j ' also incorporate several advanced components and features, as described below.
! 1.1 QuickDisconnectTopNozzle l The Mark-BW17 fuel assembly design incorporates a reconstitutable, quick disconnect top nozzle i assembly. The quick di+ - ==d feature provides easy removal and reattachment of the top nozzle i with no loose parts. The design features a double-spline sleeve attached to the guide thimble and i .
a locking ring which is contained within the top nozzle. Rotation of this ring by 90' using reconstitution tooling allows the sleeve splines to be locked or unlocked.
l . The Mark-BW17 top nozzle also incorporates two additional changes from the standard FCF top j nozzledesign. OntheMark-BWdesign,fuelamenblyliftd
- operationisprecluded by foursets i of3-leafspringsmadeofQ , *ch are fastened to the nozzle Pwith[ jlamp screws ( '
~ '
The top nozzle plate also incoiprates a modified flow holipattern. This flow hole patted provi es an increased flow area, and thus a lower pressure drop, compared to the traditional FCF design, while satisfying the same strength requirements.
1.2 { huide Thimble Tubing The guide thimbles and instrument tube are dimensionally identical for the Mark-BW and Mark-BW17 designs. However, the Mark-BW fuel design uses guide thimbles and an instrument tube fabricated fro
. ] while in the Mark-BW17 des' these components are fabricated fro The composition of this material and i ysical properties are discussed in more detail in S n 2 of this assessment.
The guide thimbles have a relatively large diameter at the top to permit rapid insertion of the rod cluster control assembly (RCCA) during a reactor trip, and a reduced diameter at the lower end of the tube (the dashpot) to decelerate the control rods near the end of the control rod travel. The diameters of the Mark-BW17 fuel assembly guide thimbles are comparable to those found on older Westinghouse fuel assemblies used at North Anna (i.e.,17x17 LOPAR assemblies, with Inconel
l
} mid. grids). The Mark-BWl7 guide thimbles have four small holes located just above the dashpot to allow outflow of the water during RCCA insertion. These holes are identical to those on the guide thimbles of the Mark-BW fuel design.
2 . .
l De stainless steel. quick dire +=^ sleeve is attached to the upper end of the Mark-BW17, guide thimble by a ==F=Jcal swage for connection to the top n i a Zircaloy-4 lower end plug is welded onto the end of the guide thimble dashpot section. This lower
~
l end plug is intamally threaded for engagement with the guide thimble bolt, which connects the guide
, thimble to the bottom nozzle.
4
- i. The instrumentation tube diameters are comparable to those on the 17x17 LOPAR Wertinghouse i fuel assemblies used at North Anna, which helps ensure compatibility of the lead test assemblies '
. with the in-core instrinnentatian I
! l3 SpacerGrids J ,
i . . ,
3 The Mark-BW17 fuel assembly design incoyv..- a total of 11 grids. As shown in Figure 1.1, !
}. these include (from the bottom. ofthe assembly to the top): an Tacanal bottom end gnd, one vaneless '
l Zircaloy-4 intermediate grid, five vaned Zircaloy-4 inksi!** grids, three Zircaloy-4 mid-span mixing grids located beries the top four vaned intermediate grids, and an Inconel top end grid, l I
i 1J.1 Intermediate Grids i
l The Mark-BW17 intermediate spacer grids are fabricated ofZircaloy-4. Rese grids are identical j
to those used on the Mark-BW design with the exception ofa small modification to the vane pattem The grids are fabricated from strips of Zircaloy-4 which are assembled in an " egg crate" fashion.
A laser weld is performed at each strip intersection with the outer face of the grid to secure the The inner strip end tabs are also laser (bead) welded. A combination of springs and dimples act in two orthogonal planes to support each fuel rod. The standard FCF keyable features are maintained for the Mark-BW17 in6 eii=* grids to allow scratch free and stress free fuel rod insertion durin fuel assembly fabrication.
As in the Mark ,BW fuel design, two types ofinterme'diate spacer grids are beirg used on the lead test assemblies: vaned and vaneless (see Figure 3.1). The five intermediate grids in high hea4
& region of the fuel assembly incorporate vanes to promote miving of the coolant. ,
6
j i
i l
i To minimize the overall fuel assembly pressure drop. the first intermediate spacer grid (i.e.. the Zircaloy grid closest to the bottom of the fuel assembly) does not incorporate mixing vanes. With the exception of the lack of mixing vanes, this grid is identical in design to the intermediate vaned j -
', grids. The resident Waainahause fuel designs have a mixing vane grid at this location.
1 .- .
- Another. Mark-BW fuel assembly design is incorporated into the lead test assemblies, where the j interme4iate spacer grids on the Mark-BW17 design are allowed to move axially upward, follo l the movement of the fuel rods as they grow due to irradiation, until burnup effects have significal
~
relaxed the Zircaloy spacer grids. Restraining ferrules, which are short sleeves made of2.ircaloy, '
l are attached to eight selected guide thimbles above each intermediate grid to limit the amount of axial movement. A ferrule is also attached to the "" - ". ~ tube below the top end spacer grid and l
} below the intarmediate spacer grids to y.a downward motion ofthe grids. The distance that the i
j ferrules permit the grids to move increases slightly with elevation. Figure 1.2 illustrates the pr features of this grid restraint system, and identifies the guide thimble locations that are used to
[ restrain grid movement This floating grid system minimim both slip loads in the fuel assembly and stresses in the fuel rods. .
)
}- 13.2 End Grids i -
i The end spacer grids on the Mark-BW17 fuel design are fabricated from Inconel-718, and are '
^
i identical to the Inconel end spacer grids on the Mark-BW design. Short stainless steel sleeves are j
attacheA to weld tabs on the guide thimble locations of these' grids. On the upper end grids, these tabs are on the top of the grid strips, while on the lower end grids these tabs are on the bottom of the l grid strips. 'Ibe Mark-BW.17 end grids, like the k' a=~ grids, irangerate keying ' windows, to allow deflection of the soft stop springs during fuel rod insertion. This standard FCF feature minimizes fuel rod --As, cell hardstop and softstop damage, and fuel assembly residual stresses.
As on the Mark-BW fuel assembly, the bottom end grids on the lead test assemblies are connected to the guide thimbles by ==h=ia.ny crimping the stainless steel sleeves into grooves in the guide thimble bottom end plugs. This =t+=ehment prevents axial motion of the bottom grids.
Two top end grid restraint designs are being used on the lead test assemblies. Two of the assemblies will use the same type of restraint as the Mark-BW design, with the top end grid sleeves seated against the bottom surface of the quick disconnect sleeve. 'Ihis design prevents axial movement of the top end grid. The second design, to be used on the remaining two lead test assemblies, is a -
floating t nd grid design.
this case, slightly shorter top end grid sleeves are used, allowing a small ga o exist between the top of the sleeves and the bottom of the quick
. disconnect s eeve at beginnmg oflife. With this design, as the fuel rods grow (due to irradiatio the top end grid enoves upward with the fuel rods until the top grid sleeves contact the. quick disconnect sleeves By the time the top end grid is in contact with the sleeves, bumup effects will have significantly relaxed the force exerted on the fuel rods by the top crid grid. The floating to grid design has becn shown to result in reduced fuel assembly growth.
m.-
9 l -
t incorporation of the two top end grid restraint designs M
- mto the North Anna lead test assemblies wi permit quantificati on of the relative contributions of l the advanced alloy and the restraint desigrito the overall reduction in assembly growth.
l 133 HandlingIntsd.ces i
} -
Fmm the p..yecdve of handling interfaces, both the intennediate and end grid designs on the lead l test nuemblies are identical to those on the Mark-BW fuel design. The outer grid straps have
} genemus lead-in vanes that aid in guiding the grids and fuel anemblies pest projecting surfaces, to
' facilitate core onload and offload. Dese lead-in vanes are strengthened by e.lding them to the inner l
strip. The grid outer straps. also have press-formed stiffening dimples that pinvide added strength l to resist tearing. De recess of the stiffening dimple is also used as a weld land for the inner and outer strip connection, elimiaa%g any exposed edges. The outer strap comer joint is a welded, lappedjoint which is carefully dressed to remove weld buildup and minimin distortion. He outer I
grid comer also irs-po..iss a structural support column to increase comer stren
}
p The grid corner strength is designed to exceed normal h=adling equipment limits by more than ,
In more than 900 handling oppvdmities, not one Mark-BW fuel assembly incorporatingT2 features ims
! sneaineA knadling damage requiring either discharge or refurbishment.
i j 1.4 Mid-Span Mixing Orids i
L To provide additional flow mixing in the high heat flux region of the fuel assembly, each lead test
- assembly includes three mid-span mixing grids (MSMGs), located mid- between the upper fo
, intermediate vaned grids (Figure 1.1). nese grids are constructed fro :
- strips which are assembled and welded in a manner similar to the Zircalo termediate 'ds. ~ e j- gMGs use the same mixing vane pattem as the Mark-BWl7 intermediate vaned grid. l l very slight difference in the vane profile, as shown in Figure 13, but this has a negligible impact on
/ .
4 l '
i j
7the four cell walls prevent the fuel rods from contacting th force (or slip load) on the rods. Therefore, the MSMGs are designated as non-contacting grids. '
l The outer strip design of the MSMG includes a large lead-in feature to preclude grid hangup or 1
damage durin handling. A wrap-around comer design is also used to improve the handi' interfaces.
~
](The resident Westinghouse fuel assemblies o not have grids at the MSMG elevations.)
4 i
The MSMGs are attached to the comer guide thimble locations (Figure 1.2). Because these are non-contacting grids (i.e., they exert no load on the fuel rods). the MSMGs are rigidly attached to the
) guide thimbler. Axial movement is precluded by velding grid sleeves (ferrules) to the top of th
'MSMG stnps at the corner guide thimble locanons and then dimpling these sleeves to thimbles in a manner similar to that used to attach the restraining ferrules for the intermediat i To help distribute the hydraulic loads, the MSMGs are anached to different guide thimbles than the
. floating intermediate grids.
t
- 1 2
I a
u
'~
~
I 1.5 Advanced Debris Filter Bottom Nozzle l'
The Mark-BW17 fuel design incorporates a debris filter bottom nozzle consisting of a fine mes filter plate supported by a structural frame. 'Ibe structural frame consists of deep ribs o stainless steel which connect the guide thimble locations, and which are attached to conventiona
. gs that interface with the reactor internals. [ ~ ~
. l l
I eau a The filter plate is attached to the top of the structural frame prior to skeleton fabrication welded to the nozzle at the four corners. Upon skeleton assembly, the guide thimble bo through the bottom nozzle into the guide thimble lower end plugs also clamp the filter plate bottom nozzle structural frame at each guide thimble location.
1 i M 1
i .e- -
< A similar bottom nozzle (with a coarser mesh) has been incorporated into FCF M 1
gemblies supplied to Duke Power's McGuire and Catawba units since February,1996. L M .
m-t 4
i 1
l l.'6 Fuel Rod l The fuel rod design for the North Anna lead test assemblies is very similar to the fuel rods in both I i
the resident' Weninghause* fuel and the FCF Mark BW fuel design. The primary difference is the use of two advanced
- alloys, Alloy 4 and Alloy 5, as cladding materials. In two of
- tlielead test assembli rods will have c1=Ading fabricated from Alloy 4. These Alloy i 4-clad rods will be unif~ml buted around the j ;phsry of the two fuel assemblies, as shown l in Figure 1.6. The el Ading for the r===inde of the rods in these assemblies will be fabricated from i Alloy.5. In the other two lead test assemblies, all of the fuel rods will be fabricated with Alloy 5 l l claddiag A ehematie of the fuel rod design is shown in Figure 1.7, and the composition and ;
j, ivy & des of the advanced alloys are di==A in more detail in Section 2. Except for the use of l
l different cladding materials, the Alloy 4 and Alloy 5 fuel rod designs are identical. '
' Each fuelrod consists of ofUO 2pellets c+a-Miin a seamless Alloy 4 or Alloy
- 5 tube, with Zircaloy and r at each end. A stainless steel spring is k>cated in the upper i~
plenum of the fuel rod to prevent the formation of fuel stack gaps during shipping and h=adling.
} while also allowing for the erpaneian of the fuel stack during operation. The fuel stack rests on the i
lower end cap. By comparison, the Mark-BW fuel rod design has a two spring system, where one 4
spring resides in the plenum above the fuel stack and the second spring resides in a plenum space
- below the fuel stack. The spring system design for the fuel rods in the Mark-BWl7 lead. test j
assemblies is similar to that of the resident North Anna fuel. The fuel rod upper end cap has a j
grippable " top hat" shape that allows for the removal of the fuel rods from the fuel assembly if-l ., necessary. The upper end cap also has a laser-drilled hole which permits evacuadon of the fuel rod and haMilling with helium gas. The lower end cap has a bullet-nosed shape for reduced hydraulic
, resistance, that also facilitates reinsertion of the rods into the assembly if any rods are removed after the assemblies have been irradisted (e.g., during fuel exarninntion programs).
The fuel pellets are a sintered ceramic, high density2UO . Each pellet is cylindrically shaped with a spherical dish at each end. The comers of the pellets have an outward land taper (chamfer) that eases the loading of the pellet into the cladding. The dish and taper geometry also reduce the tendency for the pellets to assume an " hourglass" shape during irradiation. The design density of the pellet is 96% theoret.ical density (TD), which is the same as for the Mark-BW fuel but slightly 2 llets hig(her than the nominal ar ]n diameter. density of rod
'Ihe fuel thecladding UO inhas the resi' dent uter Westing diameter, with This configuration leaves a small c earance n the inside diame(ter of the cladding and the]outside diameter lead test assemblies use a larger pellet diameter and thinner cladding thickness than used on th'e Mark-BW fuel rod. However, the clearance between the pellet and the cladding is the same for both designs. The fuel stack length, pellet diameter, cladding diameter and cladding wall thickness for the lead test assembly fuel rods are comparable to the dimensions of the resident Westinghouse fuel.
The nominal enrichment of the fuel in the lead test assemblies,4.2 w/o U-235, is also typical of reload fuel assemblies for North Anna.
- ~- .. - - - - - - - - - . - - - - - - - . - - - . - --.
i
\
I
- 2. Advanced Materials i
l j Numerous componeris of the North Anna lead test assemblies are fabricated from standard materirJs i
i used for FCF 17x17 Mark-BW fuel assemblies. These materials include: fully recrystallized low-tin j Zircaloy-4 (intennMiste spacer grids, mid-span mixing grids, fuel rod end caps and guide thimble
! end plugs, and guide thimble ferrules), Inconel-718 (end grids, holddown spring leaves. holddown i spring clamp screws, and quick disconnect locking ring), and austenitic stainless steel alloys (top 4
and bottom nozzles, quick di=w=t sleeves, end grid guide thimble sleeves, bottom nozzle filter
! , plate, guide thimble bolts and fuel rod plenum springs).
I De guide thimble tubes, instrument tube, and fuel rod eluding are fabricated from two alloys tg developed for high burn-up, low corrosion, and low growth applications.C i .
~
e 2.1 compares the compositions of these alloys wittithe nominal compositions of Zircaloy. -4 i
and ZIR10, which have been used for fuel rod and structural tubing in North Anna's resident Westinghouse fuel assemblies. De mechanical emymies ofrecry*11iW Alloy-4 and Alloy 5, at I room tempe>ature and at various elevated %. ares, are summarized in Table 2.2. De l l appropriate specification limits and comparable data are also provided for FCF Zircaloy-4 for i ;
comparison. He limits shown for Zircaloy-4 apply to both fuel rod cladding and structural tubing.
Alloy 4 and Alloy 5 have both been used as cladding materials for limited numbers of fuel rods in
- demanstration ===embliae in several European r- e*==, as well as at McGuhe Unit I and Three Mile
} . Island Unit 1 United States. Table 2.3 summarizes the in-reactor irradiation experience l l these alloys.
{ The Alloy 4 material irradiated to date has all had the same nominal chemical composition and hE treatment. The nominal composition of All 5 has remained unchanged, but there has been an
{ g*
j a
[nnealing processes [ ~ During tubing fabrication, the alloy u '
j Theinitial .loy 5 tubing material, designated
} "
as "5R" in Table 2.3, was fabricated usm{ emperature I and a lower temperature for the final anneal. "
d 1
i I
2
- No differences in TEE i
(
E
I l ' e irradiation data to datch ' -~
]
i
]have shown thap i
- umform cormsion rate and irradiation growth for both Alloy 5 and Alloy Me approximately the cou+'-7.&- rates for low tin Zimaloy-4. " Hot cell examintions have shown that Alloy bhs
! 'a hydmgen ickisp fraction whi is
ofZircaloy-4. He Allo 4 hydrogen picku fraction i oy-4. owever, Wa= Alloy 4 _
Zircaloy , ydrogen accumulation in the Alloy 4 c1=A ino is]than m f
4
] 3. Mechanical Design Evaluations i The mechanical design of the lead test assemblies is supported by test programs, evaluations and analyses. De following dis ~~= ions summarize the test programs that have been conducted and j
describe the analyses performed by FCF to support use of the lead test assembly design at North i s Anna. The impact of the advanced design features, such as the mid-span mixing grids and the use j -
of advanced alloys, on the mechanical design of the assemblies is incorporated into these j evaluations. The physical campatibility of the FCF lead test assemblies with North Anna's resident j Westinghouse fuel and core internals is also addressed.
i -
j .
3.1 Test Programs .
4 i
Two comprehensive test programs have been cand"M by FCF which support the North Anna lead test assembly design (i.e., the Mark-B.W17 design). !
l '
- 3.1.1 Mark-BW Prototype Testing The first test program was conducted on a prototype Mark-BW fuel assembly, which incorporates several features also found on the Mark-BWl 7 design. This prototype was fabricated using spacer j grids that had been sized to simulate end oflife (EOL) conditions. The assembly was then subjected i to a series of thermal and hydraulic, environmental'and mechanical characterization tests. The i
assembly was hydraulically characterized by pressure drop and spacer grid laser Doppler velocimeter j ,'
(LDV) tests. The environmental, or " life and wear', tests consisted of exposing the fuel assembly to representative reactor temperature, pressure and flow conditions for two 500 hour0.00579 days <br />0.139 hours <br />8.267196e-4 weeks <br />1.9025e-4 months <br /> periods. The j
fuel assembly exhibited no signiScant corrosion or unusual wear. Control rod trip testing was also l performed utilizing a simulated Westinghouse-type control rod drive meelumism to determine the i
control rod drop times. Subsequent in reactor testing and operation have confirmed the Mark-BW i fuel assembly operational performance.
i I The mechanical and structural characterization testing on the Mark-BW prototype included assembly frequency and damping tests, assembly axial and lateral stiffness testing, spacer grid stiffness and i strength tests, guide tube buckling tests, and force-deflection testing of miscellaneous components.
1 Additionally, the prototype assembly was exercised in handling and storage equipment at Duke 4
e
~ , - - - - - - - ,,
L Power's Catawba Nuclear Station. Unit 2. demonstratmg compaubility with critical interfaces o i
Westinghouse. designed reactor. This exercise included functional tests with both a control ro assembly and a thimble plug assembly. , Much of the information obtained from these tests is 1
considered applicable to the Mark-BW17 design since the spacer grid and guide thimble the same, and the ferruled or
- floating" grid structure is also employed. De structural and functi testing on, this prototype has been verified by m==%1 in reactor operation of 15 batches (1096 fj
.: assemblies)ofMark-BW fuel. '
1
- 3.1.2 XI Pmtotype Testing -
i
- i For the second test program (the XI program), two additional fuel assembly prototypes with mid-span mixing grids (MSMGs) were fabricated to be used in r+t=: cal and thermal hydraulic 4 .
The ==hmical prototype was fully instrumented with strain gages and <Wphement sensors, and i
featured the advanced " quick 4==+f top tozzle connection. The tests performed on this '
assembly included a static axial compression test, :Aatic lateral bending tests, shaker tests from w
! natural frequencies and mode shapes are d-*=
- =', lateral pluck tests with and without imaam, 1
axial drop tests from various heights, and an axial tension test. The results of the XI pro meh=aical tests c+# ..+i the applicability of the licensed Mark-BW models (Reference 5) to t
{
1 .
Mark-BW17 design being used for the North Anna lead test assemblies.
- m.
s 4
m -
amme The thermal hydraulic prototype for the XI program utilized the fine mesh debris filter bottom nozzle, and had select spacer grid cells sized to EOL conditions. The EOL cell sizes were conservatively established based on grid relaxation estimates at 65,000 MWD /MTU burnup. H thermal hydraulic test scope included pressure drop (A,P), life and wear, and flow-induced vibration (FIV) testing. The life and wear test consisted of RCCA drop tests, RCCA stroking tests and the 1000 hour0.0116 days <br />0.278 hours <br />0.00165 weeks <br />3.805e-4 months <br /> endurance test in an environment representative ofin-reactor conditions. The Fl served to characterize the flow-induced behavior ofthe prototype fuel assembly adjacent to a Mark- -
BW fuel assembly for flow rates representative ofreactor startup and normal operation conditions.
3.1.3 Component Testing In addition to the full-scale prototype testing, various components were also characterized via testing. The spacer grid design was subjected to static buckling and dynamic crush tests. Static compression tests were performed on the holddown spring and clamp screw. The debris filter bottom nozzle assembly was tested individually for debris filtering efficiency and pressure d characteristics. The results of all prototype characterization testing have been incorporated into th various analytical models used to support the Mark-BW17 design.
1 i l l
3.2 Fuel Assembly Compatibility.
[ The Mark-BW17 lead test murmblies are designed for full compatibility with the North Anna
- . mechanicalinterfaces including
- e compatibility with core intemals, l ,
e compatibility with control and other insert components, i
e compatibility with resident fuel, and j
e shipping and handling compatibility.
I
! FCF uses encamar-supplied design information to perfonn =-w == dimentianni analyses which
- i. ensure the functional e - ;+fhility ofthe Mark-BW17 lead test assemblies in the North Anna reactor l
j enviranment. Direct measurements of Waaiaghause 17x17 standard (LOPAR) and VANTAGE SH l
! fuel assemblies made in support of other Mark-BW fuel assembly reload centracts are also used as
- input'for these compatibility analyses.
l .. -
l Cnmaa+4ility ofFCF-supplied fuel with resident Westinghouse fuel and core components, as well asWaeinghause da*igwcore ia**=a1=. has been demonstrated through meca=ful reload transition experience at five different W#p-E-M==%=d reactors. Similar compatibility issues were also addressed by FCF in providing discrete burnable abswibs3 for use with Weeinghause-supplied fuel at Virginia Electric and Power Company's North Anna and Surry units.
3.2.1 Assembly Compatibility with Core Intemals All interfaces between the Mark-BW17 lead test assemblies and the core plates, guide pins, and core baffles are designed with sufficient clearances for the proper interface and continuous capture of the Mark-BW17 lead test anemblies. Core intemal interface calculations include guide pin-to-top nozzle clearances, fuel assembly-to-core plate axial gap, minimum guide pin engagement, and minimum fuel assembly-to-baffle plate clearances.
He Mark-BW17 lead test assembly top and bottom nozzle envelopes are equivalent to t' hose of the resident North Anna fuel assemblies. To facilitate inaallation of the fuel assembly and reactor head, both the top and bottom nozzles on these lead testbssemblies incorporate tightly toleranced core plate alignment pin holes with generous chamfered surfaces. The top nozzle holddown spring is designed to propctly interface with the upper core plate, providing the necessary holddown force i during both normal operation and accident situations, while still meeting handling and storage .
interface requirements. '#
ne axial gap between the lead test assembly top nozzle and the reactor core plate was analyzed to verify that sufficient margin exists to accommodate fuel assembly growth for the operating life of the fuel assemblies. This analysis conservatively modeled the expected growth of the North fuel assemblies with ide thimbles fabricated from Alloy 5 (which exhibit (
lt was determined that solid contact between the fuel assembly top nozzle
[d the core plate wi an not occur during the design life of the lead test assemblies.
i A special case was also analyzed to consider the possible shrinkage of the fuel assembly early in life.
This possibility was considered because fuel assemblie fabricated by Framatome with zirconium-j j
niobium guide thimbles similar to Alloy 5 were fo ,
after two cycles of'
]
ation at the Ringhals reactor in Sweden.
l fora shrinkage ten times reported for the Framatome assemblies, fuel assembly holddown i will Si maintained for the North Anna lead test assemblies, i
j 3.2.2 Assembly CWbility with Control and Insert Components
! Evaluations have been performed to address the interfaces between the Mark-BW17 lead test
- assemblies and the North Anna control and insert =-q=
- . such as control rods (RCCAs) and
- discrete burnable absorbers (BPRAs). Many di==iane for the lead test assemblies that are critical to ensuring compatibility with the -4ag control comp =ana. such as the guide thimble pitch, j
innermes and length, the adapter plate elevation, and the top nozzle height, are similar to the j corresponding dimensions on the resident Wa*+inghause fuel. The control and insert component interfaces with the lead test -amblia* including absorber to active fuel overlap as well as both axial j and diametral clearances beta.c the guide thimbles and the control and insert rods, have been i
determined to be ==tidactory.
I i
j . The guide thimble geometry for the Mark-BW17 lead test assemblies is the same as for FCF's standard Mark-BW fuel product. On both fuel designs, the guide thimble dimensions were designed j to be similar to the guide thimbles on the Wantiaghause 17x17 LOPAR fuel to ensure control rod j .
drop time compatibility. Control rod drop tests were also performed on both the Mark-BW '
j prototype assembly and the Mark-BW17 prototype assembly. The results of these tests show that the control rod drop times for the North Anna lead test antemblies can be expected to be comparable '
l to those for Westinghouse 17x17 LOPAR fuel assemblies at North Anna.
1 3.2.3 Fuel Assembly Mechanical Compatibility with Resident Fuel The Mark-BW17 lead test assemblies have envelope dimensions comparable to those of the resident Westinghouse fuel at North Anna. The lead test assemblies will therefore have in-reactor lateral j pitch, reactor internal axial clearances, and equipment interfaces similar to those of the resident fuel.
- The grid and nozzle interfaces represent the primary locations for mechanical interaction between fuel assemblies. The Mark-BW17 as-built structural grid elevations (Table 3.1) are functionally -
l equivalent to those of the resident fuel assemblies at North Anna. Additionally, worst-case j operational grid elevations (i.e., including thermal expansion and irradiation effects) have also beeb
- evaluated to casure that grid elevations match those of the resident fuel assemblies sufficiently to 3 transfer any lateral loads which may occur. It has been confirmed that even for the worst-case j.
companson of grid elevations, sufficient overlap will remain to permit load transfer.
- 3.2.4 Fuel Assembly Shipping and 11andling Compatibility a
( Safe transport of the Mark-BW17 lead test assemblies to North Anna is assured by FCF's shipping j container design. The container supports for the fuel assemblies are adjustable to accommodate
l .
4 varying fuel assembly designs. These containers have been used to ship other Westinghouse compatible fuel assemblies (Mark'-BW) and are compatible with the Mark-BW17 design.
- i. 1
- The interface compatibility of the lead test assemblies with the North Anna handling and storage i j , equipment such as the fuel elevator, ip ad~, and fuel pool racks, is ensured by the mechanical and I dimensional similarity of the Mark-BW17 design to Westinghouse (LOPAR. VANTAGE SH) and other FCF Waeinghause-c+; ;41e (Mark-BW) fuel designs. The Mark-BW17 lead test assembly envelope dimensions and structural grid elevations are equivalent to those of resident fuel designs in North Anna (Tables 3.1 and 3.2., providing assurance of the compatibility of the Mark-BW17 l assemblies with the North Anna h=adling equipment. The Mark-BW17 grid design also possesses i~~
generous lead-in features to aak=aaa the fuel assembly h=adling characteristics during core onload and ofBoad.
1 A speciSc analysis was cand=*d by FCF to address the lead test assembly interface with the fuel
- h=adling tool. This analysis t-3 bd that the Mark-BW17 lead test assembly top nozzle (i.e., the i !
quick di c+ =ct top nozzle design)is fully c.-Ale with the fuel handling tool at the North Anna '
j Power Station. '
i, j .
33 StructuralIntegrity j
In order to insure safe and reliable operation of the Mark-BW17 lead test assemblies at North Anna, i
the structural integrity of the maembly design has been verified for the loading associated with both j .
' normal operation and faulted conditions. The evaluations performed to verify the structural integrity of the fuel assembly components under normal operating loads arer =-
- in Section 33.1, while
[ Section 33.2 addresses control rod drop times for accident evaluations of the Mark-BWI7 design.
! The structural integrity of the fuel assembly under faulted conditions (LOCA and seismic) is covered j in Section 333.
i 33.1 Normal Operation
(
i The FCF structural design requirements for the Mark-BW17 lead test assemblies are derived from I
past experience with the other Westinghouse-designed plants, as well as experience with other FCF designs, and were verified to conform with North Anna's plant specific design ~ requirements. The
} r formed to verify the structural integrity of the fuel assembly components are evaluations pe' presented in the following sections. *'
3.3.1.1 Fuel Assembly Holddown Springs '
Incorporation of the mid-span mixing grids causes the Mark-BWl7 fuel assemblies to have a higher pressure drop than the resident Westinghouse fuel assemblies at North Anna, which do not have intermediate flow mixing grids. { }n the Mark-BW17 fuel assembly are designed to: ,
accommodate irradiation growth of the fuel assembly, accommodate the differential thermal expansion of the fuel assembly and core internals,
l provide adequate holddown when compared to mechanical design flowrate lift loads, and prevent excessive spring set during pump overspeed conditions.
The springs are designed to provide adequate margin for both the fixed and floating top end grid lead l test assembly designs.
The top nozzle and springs are designed to provide positive retention of the holddown springs in the
! unlikely event of spring failure. The clamp screws, which secure the holddown springs to the top l nozzle, are also analyzed to confirm their structural adequacy.
It has been shown that the North Anna lead test assembly top and bottom nozzles will maintain engagement with reactor internals for all Condition I - IV events.
! 3.3.1.2 Spacer Grids L The design bases for the Zircaloy intermediate spacer grids, Zircaloy mid-span mixing grids and l Inconel end grids require that no crushing deformations occur due to normal operation. The Zircaloy L intermediate spacer grids and Inconel end grids must also provide adequate support to maintain the i
fuel rods in a coolable configuration under all conditions. The Zircaloy mid-span mixing grid is designed as a non-contacting grid, but is to provide adequate surface contact to prevent interference ,
between the vanes and the fuel rods.
The mechanical characteristics of the grids are confirmed through a series of tests, including:
Dynamic impact tests performed on the spacer grids and mid-span mixing grids to establish allowable impact loads, to characterize the plastic defonnation of the grids, and to determine
' the value of grid deformation at which localized distortion of the guid- thimble array would affect insertion of a rod control cluster assembly.
Static crush tests of the spacer grids and mid-span mixing grids, to establish allowable grid clamping loads during shipping.
L l .
Slip load measurements (i.e., the forces required to axially move the Zircaloy spacer grids l - and Inconel end grids relative to the fuel rods, guide thimbles, and instrument tube), for use as input to analytical models of the fuel assembly. l
- Grid corner hang up tests, conducted on the Zircaloy intermediate spacer grid design to detennine the elastic load limit and failure mode of the corner cell (simulating grid hang-up).
These tests showed that the failure mode of the corner is through weld fracture, with very ;
little outer strip and corner defonnation.
L
-* Handling tests, conducted using a full scale Mark-BW prototype, to determine the fuel I
assembly insertion and withdrawal loads. These results remain valid for the North Anna lead
[ test assemblies sincet 'he mid span mixing grids on the Mark-BW17 fuel design have a reduced envelope to resist interface with adjacent assemblies.
l )
i
.-. __- = _ _ _ _ _ _ . - . . . . - . -- . . . - . . . . . . . -
3.3.1.3 Top and Bottom Nozzles Finite element analyses have been performed to ensure the structural adequacy of the Mark-BW17 top and bottom nozzles subjected to conservative shipping and handling (4g). and normal operating loading conditions. The results of the analyses showed that the nozzles meet all design requirements under the specified loads.
. 33:1.4 GuideThimbles ne guide thimbles were analyzed under normal operating conditions (including mechanical design flowrates, pump overspeed and scram loads) for buckling, primary membrane stress and primary +
secondary membrane stress.
Loads were determined using a finite element axial model that is representative of both lead test assembly configurations (fixed and floating top end grids). Both assembly types were analyzed separately for beginning oflife conditions due to their differences. At end oflife conditions, the floating top end grid is in contact with the quick disconnect sleeves, so the different assembly types are equivalent. Several normal operating conditions were evaluated (including different temperatures and times in life), and effects due to guide thimble maldistribution and asymmetrical hydraulic lift loads were also accounted for in the analyses.
The allowable buckling loads were based on the maximum compressive yield strength and/or the guide thimble lateral deflection limit. All loading conditions at cold conditio d those which included Scram loads are compared to the allowable limits based on e strength (because the control components are already fully inserted). For hot without S loads {
which ensures that control rod insertion performance is not affected. Primary an rimary-pl econdary membrane stresses are compared to allowable limits based on the ASME
~
de,Section III (Reference 6). De limiting condition was found to be{
l demonstrated fo'r expected normal operating conditions.] Margin to the allowable lim 3.3.1.5 Connections He fenule to guide thimble interface is tested to determine the stiffness and strength of the interface.
The results of this test, coupled with the results of the guide thimble buckling test, are used to evaluate the floating intennediate and upper end grid restraint system. Sufficient margin exists' fdr the ferrule to grid interface under both operating and handling conditions.
He guide thimble upper and lower connections, such as the quick disconnect sleeve-to-end grid sleeve interface and the bottom end grid sleeve to-thimble plug crimp, are verified through testing and/or analysis. Process qualifications are also performed for the weld, swage, and crimp-type connections to ensure repeatability.
I
I l l \
l 3.3.2 Control Rod Drop Times l
'Ihe design bases for the fuel assembly states that the fuel assembly shall not experience any
. permanent deformation during either a Condition I or 11 event that would cause the control
' component drop time to increase beyond the allowable limits. The maximum allowable control rod drop time specified in the North Anna Technical Specifications (References 7 and 8)is 2.7 seconds, measured from the start of control rod spider movement until the control rod enters the dashpot region of the guide thimbles.
' Excore testing has been marktad by FCF on the Mark-BW an'd Mark-BW17 prototype assemblies at temperatures and flow rates that are representative of operating conditions. The maximum measured control rod drop times in these tests "espesively.
Incore drop at using the Mark-BW fuel have typically resulted m p times that were abo i higher, with no apparent dapaad-ce on assembly burnup.
An evaluation based on the resident North Anna control rod assemblies has detennined that there would be no restriction of rod insertion into the lead test assemblies under normal operating conditions. Comparison of the guide thimble di-nat- s for the resident Westinghouse fuel to those of the FCF Mark-BW17 lead test mmblies (refer to Table 3.2) shows that the guide thimble
' diameter for the resident NAIF fuel design (with either Zircaloy-4, and ZIRLO guide thimbles) is smaller than on the FCF lead test assemblies, which are directly comparable to the Westinghouse LOPAR fuel design. Thus, drop times for the FCF lead test assemblies are erp-e+fd to be faster than those for the most recent reload fuel at North Anna, and comparable to the North Anna drop times in the older LOPAR fuel assemblies.
3.3.3 Seismic and LOCA Evaluation The criteria for the fuel assembly seismic and LOCA analyses are consistent with the acceptance criteria of the Standard Review Plan (NUREG-0800), Section 4.2. (Reference 9). Specifically:
a.
The fuel assembly is designed to ensure safe operation following an Operational Basis Earthquake (OBE). For the Safe Shutdown Earthquake (SSE), the calculated spacer grid impact loads are within the elastic load limit. 'Because the magnitude of the OBE is usually about half the magnitude of the SSE, these results also satisfy the OBE requirements, that the assembly or component not exceed its yield limit. Hence a separate OBE analysis was not required for the North Anna lead test assemblies.
- b. The fuel assembly is designed to allow control rod insertion and to maintain a coolable geometry during a Safe Shutdown Earthquake. A separate Safe Shutdown Earthquake analysis was done to ensure that the fuel assembly does not sustain permanent damage that will impede control rod insertion and core coolability. For the North Anna lead test assemblies, there is no permanent deformation. Therefore the requirements of control rod insertion and coolable geometry are met.
i
i c.
The fuel assembly is designed to allow for the safe shutdown of the reactor system following a Loss of Coolant Accident (LOCA) or combined LOCA/SSE by maintaining the overall structural integrity and a coolable geometry. This criterion places limits on the permanent ;
deformations to which spacer grids may be subjected. For the North Anna lead test assembly I LOCA cases and the combined LOCA/SSE cases, the calculated spacer grid loads are less than the elastic load limit value. Therefore, the grids sustain no permanent deformations and no further calculation was required to ensure that the coolable geometry requirements were met. -
3.3.3.1 Mixed Core Horizontal Seismic and LOCA Loads A bounding analysis for a mixed core of the resident Waeinghause fuel and the FCF Mark-BW17 lead test assemblies (LTAs) was performed for the combined seismic and Loss of Coolant A ccident (LOCA) events. Core coolable geometry will be maintainM for all the faulted loads. Imeractive spacer grid impact loads were maanM to show the compliance with the core cookble geometry requirement. ..
The static (lateral stiffness) and dynamic (natural Aequercy and damping) characteristics of the i
Mark-BW17 lead test assemblies and podaction VANTAGE-SH fuel assemblies were determined ;
analytically and syerhsentally. These ch&scaistics .were found to be compatible due to '
similarities in assembly geometry and construction.
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r w w 333.2 VerticalLOCA Analysis j
.; One design difference between the Westinghouse and the FCF fuel assembly designs is in the i
method ofrestraining the spacer grids, as described in Section 1.3. Although the overall assembly vertical stiffness and the strength of the two designs should be the same, to confirm that the FCF
' guide thimbles will allow control rod insertion during a LOCA a separate vertical LOCA analysis 4
' was gifvuoed to marmn the structural integrity of the FCF guide thimbles. Only the LOCA cases were evaluated in the vertical din: tion as the upper and lower core plates move in phase for the seismic case and cause no fuel assembly loading. .
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' 3.4 Fuel Rod Design A series of analyses was performed for the lead test assembly fuel rod design to wrim that the fuel rods with the advanced swidum-alloy eiwi= will exhibit ==d%~y in. reactor mechanical performance. These analyses included: -1 ding stress and strain, c1 ding fatigue, creep collapse, fuel rod growth, corrosion, shipping and h=adling loads, and fretting wear. An evaluation of the end oflife fuel rod internal pressure was also F 1 =i The analyses in these areas, summarized below, j demonstrate that the lead test assembly fuel rods will perform ==tiehtorily to burnups ava-ling l
Virginia Electric and Power Caaaaay's NRC l'aW lead rod burnup limit of 60 GWD/MTU.
. 3.4.1 Fal Rod Clwing Stress 1
The operation inthmi stresses in the fuel rod cladding were evaluated using conservative values for physical parameters such as alwia: thickness, oxide layer buildup, mamal pressure, internal fuel -
rod pressure, diferential z r---w and unirradiated <1 ming yield strength. The ASME vessel stress intensity limits (Reference 6) are used as guidelines for this evaluation.
-m ..
The clad stresses for the Alloy 4 and Alloy 5 clad fuel rods in the North Anna lead test assemblies were evaluated in the same manner as the fuel rods in the previous demonstration programs. This
,i nvol ves dividing the stresses into compressive and tensile stresses according to criteria given in Reference 11. The stress level intensities were calculated in accordance with the ASME code, and include both normal and shear stress etYects. The stress intensities were then compared to the ASME criteria, as follows:
- 1. Primary geneml membrane stress intensities (P.) must not exceed S,.
i
II. Local primary membrane stress intensities (P) i must not exceed 1.5*S,. These include the contact stresses from spacer grid-fuel rod contact. The total of the local i primary membrane and bending stress intensities (Pi + Pi ,) must not exceed 1.5 *S,. l III. The sum of the local primary membrane, bending. and secondary stress intensities (P + P b+Q) must not exceed 3.0*S..
For the beeianing oflife tensile stresses, S, is defined as 2/3 of the minimum unitradiated yield i strength of the c1=AAlag; consistent with the ASME code. The worst case tensile stress condition occurs late in the life of the fuel rod, and is enveloped by the stress calculations for the hardened
{
cladding. For the compressive stresses, S, is set equal to the minimum unirradiated hoop yield i
strength of the c1=dding at operating e-y eurs, consistent with Reference 11. .
l It was determinad that the calculated stresses for the fuel rods in the North Anna lead test assemblies will be within the prescribed limits.
3.4.2 Fuel R6d Cladding Strain l 'Ibe fuel rod was analyzed to he b the maximum transient the fuel rod could experience before j the transient strain limit of 1% is exMd The transient strain limit evaluation uses cladding i
circumferential changes before and after a linear heat rate transient to determine the strain. This analysis is performed using the TACO 3 fuel rod thermal analysis code (Reference 12). Both the Alloy 4 and Alloy 5 =%1s have a relativel gh strength w,my d to Zircaloy-4, and the ;
creep rates of both materials are less than f Zircaloy-4. It was determined that the
, calculated linear heat rates which would resul 1% ding strain for rods with Alloy 4 and Alloy l: 5 cladding are much greater than the maximum transient the North Anna lead test assembly fuel rods l are expected to experience.
3.4.3 Fuel Rod Fatigue Usage The total fatigue usage factor for the North Anna Alloy 4 and Alloy 5 fuel rods was analyzed using the ASME pressure vessel code as a guideline. Testing conducted by Framatome Fuel Division in France has wn the recrystallized claddings to have fatigue endurance performance similar
~
Zircaloy-4, The Alloy d Alloy 5 data are well enveloped by the standard Langer-O'Donnell design fatigue '
curve for irradiated Zircaloy. All possible Condition I and Condition 11 events that would be experienced by the lead test assembly fuel rods over a (conservative) lifetime of 8 years were' l assumed, along with one Condition III event. Conservative inputs were also assumed for cladding thickness, oxide layer buildup, extemal pressure, intemal fuel rod pressure and differential pressure across the cladding. ysis for the North Anna lead test assembly fuel rod resulted in a fatigue usage factor o ch is well below the maximum allowable fatigue usage factor of 0.9.
i 3.4.4 Fuel Rod Cladding Creep Collapse i
b The fuel rod creep collapse analysis is performed in accordance with the NRC-approved 1
methodology described in Reference 13. Fuel rod failure due to collapse is predicted to occur when either:
- 1. the rate of creep ovalization exceeds 0.1 mil /hr, or l
- 2. the maximum fiber stress exceeds the unirradiated yield strength of the cladding.
The creep rate of Alloy 4 and Alloy 5 is approximatal of Zircaloy-4. Therefore an appropriate multiplier was used on the creep model in this ysis to represent the behavior of the
{
Alloy 4 and Alloy 5 materials. It was daw =W that North Anna lead test assembly fuel rod creep l
' collapse lifetime is greater than the w+3+1 in-core life of the fuel.
3.4.5 Fuel Rod Cladding Corrosion l
One of the major purposes for development of the advanced zirconium based alloys was to obtain a reduction in the amount of cladding corrosion relative to Zircaloy-4. Both Alloy 4 and Alloy 5 exhibit a strong resistance to co ' n, '
irradiation experience to date showing the corrosion of these materials to be less corrosion oflow-tin Zircaloy-4 claddings. An i evaluation based on t g tF f oy Alloy 5 predicted an upperlimit on cladding oxidation of whichisless f the limit.
f The hydrogen ick-up ra f the Alloy 4 and Alloy 5 cladding materials have been found to be l approximately 'vely. For the predicted conosion level, the upper limit for l hydrogen content of claddings at the end of the lead test assembly operating lifetime will be approximatel '
l f the FCF upper limit for hydrogen pick-up.
This level of oxidation and associated hydriding will not adversely affect the stmetural integrity of the North Anna lead test assembly fuel rod during its design lifetime.
l 3.4.6 Fuel Rod Shipping and Handling Loads The fuel rods in the North Anna lead test assembiles are designed to ,vithstand a 4g load during shipment and handling of the fuel assembly without the formation of gaps between pellets in the fuel stack. This design condition is achieved by using a stainless steel spring in the upper plenum of the fuel rod. The dimensions of this spring have been specified to ensure that the spring will maintain a 4g pre-load on the fuel stack, prohibiting the formation of gaps within the fuel stack. -
3.4.7 Fuel Rod Fretting Wear '
A life and wear test has been conducted at maximum reactor flow conditions for 1000 hours0.0116 days <br />0.278 hours <br />0.00165 weeks <br />3.805e-4 months <br /> t evaluate the fretting characteristics of the fuel rods and spacer grids. The preliminary results of this test showed no indication of fretting wear of the fuel rods.
~
Extensive operational experience with the Mark-BW design, using grid designs and fuel rod dimensions comparable to those of the North Anna lead test assemblies, has shown only one fuel rod fretting failure. This failure occurred in a comer grid cell of the lower end grid, and was determined
l 1
1 i to be caused by either manufacturing or opemtional damage to the cell. To improve the frettinc i performance of the grid. a design change was subsequently implemented to increase the force exerted on the fuel rod by the comer cell spring stop.
! ne modification to the mixing vane panem on the Mark-BW17 vaned grids, described in Section
!- 13.1, is very mmm11 and will not affect the flow iM vibration performance of the fuel assembly relative to that previously seen with the Mark-BW design. Flow induced vibration testing has also
! been performed with prototype naamblies to d==s.i. that fuel rod fretting wear will not occur j m Mark-BW17 assemblies is;ww.dsg mid-span mixing grids.
j Based on previous ~ operating experience and on both life and wear and flow induced vibration 1
testing, no fuel rod west due to fretting is q+ned in the North Anna lead test assemblies.
- 3.4.8 Fuel Rod Growth 1
) -
I he axial gap kr a the fuel rods and the lead test ====hly top nozzle was analyzed to verify that j {
sufficient margin exists to accommndste fuel rod growth for the op. Jug life of the fuel assemblies. !
! To calculate the closure ofthis gap, the upper tolerance growth of Alloy 4 and Alloy 5 is used with a conservative ===* ion ofzero fuel assembly growth. Contact Mr .ea the Alloy 4 and Alloy l 5 fuel rods and the fuel assembly top nozzle will not occur during the design life of the lead test i
mnemblies.
l M
M i
3 i ;
i -
f l 3.4.9 FuelRodInternalPressure
- i .
Fuel rod thermal performance analyses, including evaluation of the end oflife rod intemal pressure, were performed with the NRC-approved TACO 3 fuel rod thermal performance code (Reference 12).
l The TACO 3 code and intemal gas pressure methodology have been extended to address operation j
with intemal pressure greater than reactor coolant system pressure in the NRC-approved topical l report BAW-10183P-A (Reference 14).
l -
! Maximum fuel rod intemal pressure was d+ mined using a pin power history and axial flux shapes provided by the Virginia Electric and Power Company. The power history and axial flux shapes f (both steady state and transient shapes) were generated in a manner consistent with FCF's NRC-
{ approved methodologies, using Virginia Electric and Power Company's standard design codes. The j
enveloping power history used represents the peak fuel rod in core over the burnup history, and is i
a composite from four previous North Anna fuel cycle designs. His power history bounds the
! planned irradiation of the Mark-BW17 lead test assemblies at North Anna. It was determined that i
l i
1 the end oflife fuel rod intemal pressure will remain below the FCF criterion for operation above system pressure defined in Reference 14.
' Prior to each operating cycle, the cycle specific pin power histories for the lead test assemblies will
' be compared to the powerhistory envelope ====aad for this analysis. Should the cycle specific peak i pin p6wer for the lead test assemblies violate the envelope, then either the power history envelope
- can be adjustedI and an analysis gG#=' to demorou.m that the pin pressure remains below an acceptable le 9 C J 4.ThermalHydraulic Design Analyses of the ther==1 and hydraulic earna=+ibility of the four FCF lead test assemblies with the resident Weetinghause fuel at North Anna were pL. 4 by FCF using NRC-approved models and methods. nese analyses addressed areas such as unrecoverable core pressure drop, hydraulic lift -
forces, inter-bundle crossflow, and DNB performanca.
-4.1 Design Comparison l
The components ofthe Mark-BW17 lead test assemblies and resident Wu Ly-,use fuel assemblies are hydraulically similar with the exception ofa slightly smaller thimble tube dia=*ar for the recent Westinghouse fuel designs and higher grid pressure drop for the Mark-BW17 design. The Mark-BW17 lead test assemblies forNorth Anna also incorporate three additional mid-span mixing grids between the top four standard Zirealoy mixing vane grids for enhan~dt hermal margin. A similar component is not currently utilized on the resident fuel.
4.2 Calculational Methods The calculational methods currently used on the Mark-BW product for use in Westinghouse reactors are applicable to the evaluation of a core containing both Mark-BW lead test assemblies and the Westinghouse fuel products. The thermal and hydraulic (pressure drop, lift force, crossflow, and DNB) analysis of the FCF lead test assemblies was, performed by FCF using the LYNXT code (Reference 15). LYNXT is a single-pass code which employs crossflow methodologies to evaluate
'subchannel thermal-hydraulic conditions for both steady-state and transient analyses. Two LYNXT models of the North Anna core were used for the evaluations. An eighth core 26 channel bundle by- '
bundle model was used for the hydraulic analyses and a 12-channel model was used for the DNB analyses.
The DNB analysis of the Mark-BW17 lead test assemblies was performed using FCF's Statistical l
Core Design (SCD) methodology. The NRC has approved the SCD methodology for both B&W-and Westinghouse designed reactors in Topical Reports BAW-10187P-A and BAW-10170P-A l (References 16 and 17), respectively. DNB analyses to support use of the Mark-BW17 lead test i
assemblies at North Anna were performed using two NRC-approved critical heat flux (CHF) i correlations: the BWCMV-A CHF correlation (Reference 18), and the BWU-Z CHF correlation
{ (Reference 19). The Mark-BW17 fuel assemblies showimproved DNB performance relative to the
resident Westinghouse fuel in part due to the Mark-BWl7 mixing vane design. but also from the addition of three mid-span mixing grids. as described in Section 1.4. The enhanced DNB performance obtained from the use of the mid-span mixing grids was evaluated by means of a spacer
}d spacing factor incorporate.d into the BWCMV-A CHF correlation.[
Z correlation.)
43 Hydraulic Ovibility The pressme dmp of the Mark-BW17 lead test assemblies is higher than that for North Anna's resident Wda=h fuel annemblies, primarily due to the presence of the MSMGs on the Mark-BW17 design. [
~
The net effect is a flow iversion from the Mark-BW17 lead test ===emblies to the surro]u assemblies.
4.3.1 Grid Pressure Drop
\
l .
The grid pressure drops for the Mark-BW17 lead test assembly were sce.us4 in Frammtame's HERMES P loop in Cadarache, France. The HERMES P loop operates at PWR primary coolant conditions (600*F,2250 psia). Cc. ..;e===t form loss coefficients for analyses were determined from this pressure drop data.
~
! Core pressure drop data for the resident Wdaghause fuel assemblies were obtained from Westinghouse and used to analytically determine the grid form loss coefficients.
4.3.2 Core Pressure DrEp The core pressure drop across the fuel assemblies consists of contraction and expansion, friction. l elevation, and crossflow losses. The core pressure drop was determined for a full core of Mark-BW17 fuel and a full core of current North Anna Improved Fuel assemblies. )
The following results were obtained for 100% core power at a core flowrate of 330,000 gpm:
[
Mark-BW17 Core Westinghouse Core
== z \
l '
These results demonstrate that the predicted pressure drop for a full core of Mark-BW17 fuel l assemblies is within 6% of the resident fuel, which is considered negligible. 'Ihe core pressure drop of the mixed-core is bracketed by the pressure drop values for full cores of each design.
, 4.3.3 Lift i i
The fuel assembly hydraulic lift force for the Mark-BWI7 lead test assemblies was determined for j the limiting core configuration, which - because of their higher pressure drop - is a homogeneous j i
1
4 i
i
! core of Mark-BW17 fuel assemblies. Hydraulic lift forces were determined for the Mark-BW17 design at both isothermal and 'at power' conditions. Analyses were performed for core flowrate at both the Mechanical Design and the Pump Overspeed ('at power' only) conditions. The results of i
- these analyses were used in the evaluation of fuel assembly holddown springs in Section 3.3.1.1.
i The efIect of the Mark-BW17 lead test assemblies on lift of the resident fuel was assessed w f mixed core analysis. 'Ibe total predicted lift forces were cerupsed for the North Anna core with and j without the FCF Mark-BW17 anemblics. The results of the mixed core analysis indicatad that the
!- lift pairfonnance of the resident fuel was insignificantly affected by the pw ence of the lead test j assemblies. .
1 43.4 CrossflowVelocity
{
average crossflow velocities are used to consider the integrated effects over the total span. A average value is a conservative criterion that has been used historically on a variety of j i to preclude r==g -ble flow induced vibration of the fuel rods. Mixed core analyses j assuming a single Mark-BW17 fuel enemh in a core ofWaeinghan== fuel assemblies .hnr. w
].
span average crossflows are less than ]spm average) criterion.
i
- 43.5 ComponentBypassFlow I
Component bypass flow through the Mark-BW17 lead test assembly guide thimbles was
- conservatively evaluated assuming a full core of Mark-BW17 fuel assemblies at conditions j .
representative ofNorth Anna. 'Ibe effect on the overall core bypass was hminad to be negligible.
i 4.4 DNB Performance Evaluation i
! To demonstrate that the DNB performance of the Mark-BW17 lead test assemblies is non-limiting,
{ calculations were performed by the Virginia Electric and Power Company for the resident j Westinghouse fuel and by FCF for the mixed core configuration using the applicable statistical j
DNBR methodologies and CHF conelations. The DNBR results were compared to the fuel specific
! thermal design limits. These calculations, described below, demonstrate that the lead test assemblies
! have more margin to their applicable DNB limit tha6 the resident Weeinghause fuel has to its limit.
j Cycle specific and UFSAR non-LOCA analyses with DNB acceptance criteria which assume a full j core of Westinghouse fuel will therefore be com.ervative for North Anna cores containing the lead i test assemblies.
4.4.1 Statistical Core Design (SCD) a
{ The FCF Statistical Core Design (SCD) uses a statistical combination of uncertainties technique.
- j. In the SCD method, input uncertainties are arnlyzed using statistical methods and an overall DNBR uncertainty is determined. This overall uncertainty is then used to establish a design limit DNBR j known as the Statistical Design Limit (SDL). For added flexibility, margin is added to the SDL.
j This added margin defines an analysis limit termed the Thermal Design Limit (TDL). Once the TDL i
has been established, the calculated DNBR at a specific core state is corppared to the SDL to -
1 5
~
[
1 4
- determine if the DNB protection criterion is met.
i For the planned insertion of the Mark-BW17 lead test assemblies into North Anna. the plant specific
~
- variables listed in Table 4.1 were used to determine Statistical Design Limits for the BWCMV-A i ' and BWU-Z CHF. correlations. The ranges and uncertainties of these variables are consistent with those used for the implementation of the Virginia Electric and Power Company Statistical DNBR
- E Methodology (Reference 20) Anna (Reference 21). The resulting SDL values
! . BWCMV-A correlation, r the BWU-Z correlation.
i j 4.4.2 Retained DNB Margin
! Forapplication ofthe SCDmethod ton cores with the Mark-BW17 lead test assemblies, !
j the retained margin added to the SDL
- is defined by the following formula:
TDL-SDL
! Retained thermalmargin(%) =
. x 100 TDL This results in a hermal Design Limit of 1.58 for the BWCMV-A CHF correlation, and 1.49 for
- the BWU-Z correlation.
j 4.43 DNB Analysis and Results l
[ Statepoints were developed by Virginia Electric and Power Company to west points on the safety limit line, hWg axial flux shapes, highly peaked radial power distributions, and low flow conditions for the resident Westinghouse fuel at its thermal design limit of 1.46 (Reference 21).
For the FCF DNB analyses, the North Anna core was modelled with the LYNXT code using the 12 i
channel model. The core configuration consisted of a single Mark-BW17 fuel assembly in a core
! with the remaining assemblies being modeled as W@ghause fuel. This configuration is limiting j because of the higher pressure drop of the Mark-BW17 fuel assembly relative to the current North i
Anna fuel design, which results in flow being diverted from the lead test assemblies to the resident fuel. The predicted MDNBRs for the LTAs were obtained at the statepoints provided by Virginia Electric and Power Campany using the BWCMV-A and BWU-Z correlaticms, and were determined
{ to be greater than the applicable TDLs.
i (l I 4.5 Rod Bow '
i
{ The rod bow behavior in the Mark-BW17 lead test assemblies should be comparable to or less than that ofMark-BW fuel Because essentially the same grid designs and fuel rod dimet:sions are used
- in both fuel assembly designs, similar grid forces are exerted on the fuel rods. Each chsign is self-consistent in the use of materials for fuel rods and guide thimbles, with the Mark-BW fuel using i Zircaloy-4 cladding and guide thimbles, and the Mark-BW17 lead test assemblies using AUoy-4 and Alloy-5 (or all Alloy-5) cladding in skeletons with Alloy-5 guide thimbles. The low growth j characteristics of the Alloy 4 and Alloy 5 advanced materials, which have been demonstrated h
4
i
! i i through irradiation experience. are expected to contribute to reductions in fuel rod and fuel assembly j
bow. The Mark-BW17 fuel design can therefore be reasonably expected to exhibit rod bow behavior no worse than that previously seen for the Zircaloy-4 clad fuel rods in Mark-BW fuel assemblies.
! The same approach to irs.yo. wing the effect of fuel rod bow on DNBR that is used for Mark-BW .
i
- ' fuel assemblies is therefore also applicable to the lead test assemblies. !
' l l ThePO*aan offuel rod bowing is accounted in DNBR safety analysis by assessing aC j DNBR, penalty for burnups less No DNBR penalty is as=-s=-d for burnups
!, 24 GWD/MTU ainee ding be reachM (References 22 and 23).
l ty is accounted for by . owance that is incorporated into the 1
e spacmg parameter (shown
- able 4.1), is combined statistically with other
- uncertainties to establish the statistical design limit (SDL) DNBR.
f
- 4.6 FuelTmuym4 ares i
} The fuel temperatures were evaluated for the lead test n=mblies using FCF's TACO 3 fuel rod i
thermal performance code (Reference 12). The peak local power at which fuel melt is predicted to l begin was calculated as a function of burnup. The results of these calculations show that, over the i
planned cpm.Gug life of the lead test assemblies, the minimum linear heat generation rate at which fuel melting will occur aveaarie the minimum linear heat generation rate used by Virginia Electric j and Power Company in the reload safety analysis e*1~wions for North Anna to ensure that fuel j melt does not occur in the resident W-*+inghause fuel. Herefore, Virginia Electric and Power j
Company's standard reload design process criterion can be conservatively applied to cycle specific l designs forNorth Anna cares iscerycmdug the lead test assemblies to ensure that fuel rselt does not i occur in either the FCF or the retident Westinghouse fuel assemblies.
l :
i i
4.7 Impact on Reload Evaluation Methodology i
I Le analyses performed by FCF have demonstrated that use of the four lead test assemblies in a core of Westinghouse fuel will have a negligible impact on the core pressure drop, hydraulic lift forces j on the resident fuel, span average crossflow, and overall core bypass flow. The DNB performance '
of the lead test assemblies is bounded by the DNB performance of the resident Westinghouse fuel.
l Therefore, cycle specific thermal hydraulic evaluationsfor North Anna cores containing the four lead j test assemblies can conservatively be modeled as a homogeneous core of Westinghouse fuel i j assemblies. There is no impact on the models or methods normally used by Virginia Electric and
i Power Company to perform thermal hydraulic analyses of the core, and no transition core penalti.es
! must be applied. '
I a
i 5. Neutronic Performance 1.
3 The physical differences between the Mark-BW17 lead test assemblies and the resident j Westinghouse fuel are small. Cycle specific neutronic calculations will account for the effects of 1
j the composition of the Alloy 4 and Alloy 5 fuel rod cladding materials and the use of Alloy 5 for the r guide thimbles and instrument tube. The presence of the Zircaloy-4 mid-span mixing gridsand the
! impact of the higher nominal fuel density will also be incorporated into the analyses. As a result of
\
i i - .. ._
the general physical similarity to the resident Westinghouse fuel designs, the Mark-BW17 LTAs have essentially the same neutronic behavior as the resident fuel assemblies.
On an equal enrichment basis, the Mark-BW17 lead test assemblies initially exhibit reactivity similar to the resident Westinghouse fuel. Due to the higher uranium loading (primarily the result of a higher nominal fuel density), the rate of depletion of reactivity is slightly smaller for the lead test assemblies than for the majority of the fuel in the North Anna core. This difference will be explicitly modeled in the cycle specific neutronic calculations and will not have any adverse impact on the operation of the plant.
To ensure that meaningful data are obtained on the performance of the lead test assembly new design features and advanced materials, these assemblies will be placed in high power locations, particularly in the first two operating cycles, and are expected to achieve assembly average burnups over 50,000 MWD /MTU in three operating cycles. Although the fuel will experience moderately severe duty typical of normal reload fuel, to ensure that the existing safety analyses based on the resident Westinghouse fuel remain applicable, the core locations will be selected so that the FCF assemblies are not placed in the highest fuel rod power density locations in the core. The lead test assemblies will not be limiting with respect to any safety analysis limit, meaning thatqF and Fa margins will ,
be preserved for the lead test assemblies and they will not set any safety or operating limits. In '
addition, the lead test assemblies will not have the highest cycle-averaged assembly average power ;
for any given cycle. 1 The North Anna core reactivity coefficients and nuclear performance for cores containing the FCF j lead test assemblies will not be noticeably different from recent reload cores consisting of all i Westinghouse fuel. Cores containing fuel with a variety of Westinghouse design features (e.g., l different cladding materials; Inconel, Zircaloy-4 and/or ZlRLO mixing vane grids; minor changes i to fuel loading due to pellet dish and chamfer dimension changes as well as normal manufacturing i variations in pellet density; small changes to grid and stack axial elevations; and fuel assemblies with and without protective grids) have shown aceeptable power peaking and reactivity behavior. Past changes made to the neutronic model inputs to incorporate these material and design changes to the Westinghouse fuel p oducts, as well as other more (neutronically) significant product changes such as burnable poison design changes, use of vibration suppression inserts, and use of flux suppression inserts at Surry Unit 1, have been incorporated with fully acceptable predicted-to-measured power distributio.n and reactivity parameter agreement. Changes to the neutronic model inputs necessary to model the physical differences between the lead test assemblies and the resident Westinghouse fuel assemblies are similar to those used for previous Westinghouse fuel product changes, and are of a smaller magnitude than was necessary for many of the Westinghouse fuel product changes.
Therefore it is concluded that the Nuclear Design Reliability Factors specified in Reference 24 remain applicable for use with the lead test assemblies. It is also concluded that the methods and models used to verify local rod powers for Relaxed Power Distribution Control analyses (Reference
- 25) remain valid for use with the lead test assemblies. Use of the Mark-BW17 lead test assemblies in conjunction with the Westinghouse fuel in the North Anna cores will not adversely affect plant operation or neutronic parameters.
The use of the lead test assemblies will also have no significant impact on spent fuel pool
calculations. Margin in existing North Anna spent fuel pool analyses will more than account for the ,
additional amount of uranium present in the FCF lead test assemblies. Therefore the North Anna spent fuel pool analyses will remain bounding for the lead test assemblies. ,
6.'Non LOCA Safety Evaluations l
The performance of the Mark-BW17 lead test assemblies under postulated non-LOCA accident l
conditions was evaluated by the Virginia Electric and Power Company. The intent of the evaluation l was to assess the applicability of the existing non LOCA safety analyses, which assume a full core l of North Anna Improved Fuel (NAIF) assemblies, to cores containing the FCF lead test assemblies.
6.1 Assembly Design Comparison in this evaluation, the design features used as key parameters in accident analyses were compared between the Mark-BW17 and North Anna Improved Fuel (NAIF) designs. The features reviewed included dimensional data for the fuel pellets, fuel rod, and fuel assembly features. In addition, L material properties for the Alloy 4 and Alloy 5 were compared with the Zircaloy-4 and ZIRLO materials used in the Westinghouse fuel rods. The majority of design features are identical or comparable between the Mark-BWl7 and NAIF designs. Only a limited number of characteristics that are relevant for NSSS accident analysis exhibit any difference from the existing NAIF design.
E The following Mark-BW17 features are different in the manner indicated relative to the NAIF design:
, 1. Increased fuel assembly pressure drop l
- 2. Reduced fuel average temperature
- 3. Presence of Mid-Span Mixing Grids (MSMGs)
- 4. Increased guide thimble tube inner and outer diameter
- 5. Reduced cladding alpha-beta phase shift temperature 6.2 Events with DNB Acceptance Criterion These differences for the FCF Mark-BWl7 fuel assembly design were evaluated for potential impact upon the existing analysis of record for non-LOCA events. For this purpose, the events have been divided into two categories: (1) events with a DNB acceptance criterion, and (2) events with all other acceptance criteria. The DNB performance of the Mark-BW17 fuel assemblies for operation in the Nonh Anna cores is documented in Section 4.4. These DNB analyses have accounted for the effects of the Mark-BWl7 pressure drop, presence of MSMGs and guide thimble tube dimensions (Items 1,3 and 4 above). These features affect key parameter inputs for core and bypass flowrates which are relevant only for DNB analyses. The analysis results demonstrate that the DNB margin for the FCF Mark-BW17 fuel assemblies is greater than that of NAIF, with respect to their applicable DNB correlations and limits, for bounding mixed core configurations. The assessment demonstrates that
- DNB analyses performed for full cores of NAIF provide bounding results for application to the Mark-BWl7 design. These analyses thus verify that existing non-LOCA event analyses for North Anna with a DNB acceptance criterion are conservative licensing analyses for the Mark-BW17 lead j test assemblies. This DNB-related assessment applies to the following NSSS events, which have l
l l
i i
j a DNB acceptance criterion:
i Uncontrolled RCCA Bank Withdrawal from a Suberitical Condition ,
Uncontrolled RCCA Bank Withdrawal at Power Rod Cluster Control Assembly Misalignment Uncontrolled Baron Dilution l -
Startup of an Inactive Reactor Coolant Loop i -
Spurious Operation of the Safety Igjection System at Power Loss ofExternalElectrical Load Excessive Heat Removal Due to Feedwater System Malfunctions F -
Excessive ImadIncreaseIncident
! . Accidantal Ardu.iian of the Reactor Coolant System
} -
Accidental Lr arization cf the Main Steam System
{ -
Main StearnHa* Rupture i - Complete Loss ofReactor Coolant Flow Single Reactor Coolant Pump Locked Rotor j - Singlei RCCA Withdrawal at Full Power
)
l .
In addition to physicali.a 4 a separate ==ma==rnant of the features that govern neutronic behavior i as modeled in Virginia Electric and Power Cn=a ay's core reload methodology concluded that there are no differences between the Mark-BW17 and NAIF designs which either invalidate an existing l key physics parameter limit value or require the introduction of a new parameter for accident 1 analysis. A more dann%d 4 44;on concerning the ==eament ofneutronic mi:nitarity bere.s the j Mark-BW17 and NAIF fuel assemblies was presented in Section 5 of this Safety Significance
- evaluation.
4
!- 6.3 Events with Non-DNB Acceptance Criteria i
ne fuel design feature differences noted above also have the potential to affect results of the non-i LOCA events which do not have a DNB acceptance criterion. These events are listed below.
I -
Loss ofNormal Feedwater Loss of Offsite Power to the Station Auxiliaries Spurious Operation of the Safety injection System at Power j -
Steam Generator Tube Rupture '
l Single Reactor Coolant Pump Locked Rotor j -
Fuel Handling Accident (Inside and Outside Containment)
i -
Major Rupture of a Main Feedwater Line l Rupture of a Control Rod Drive Mechanism Housing (Rod Ejection) 1 j Virginia Electric and Power Company conducted a review to determine the impact on the non-DNB NSSS accident analyses. It was concluded that the only key parameters affected are those used as i inputs for the two events that involve a detailed A=Ading temperature calculation: the Locked Rotor j and Rod Ejection events.
i
_I
i i
1 The reduced fuel average temperature of the Mark-BWlf design relative to the NAIF fuel is a l benefit for cladding temperature analyses. and requires no further evaluation. The difference it the
! phase shift temperattre of the Mark-BW17 advanced alloy claddings impacts the cladding heat l .
capacity at temperatums within the alpha-beta phase shift region. This behavior has previously been
- evaluated for the use cf Westinghouse's ZIRLO alloy, which exhibits similar characteristics.
l
- FCF has demninad that the thermal peysses of Zircaloy-4 and Alloy 4 are the same, and that the use of Zircaloy-4 properties is acceptable for madaling Alloy 4 in analyses. He Alloy 5 and
- Zircaloy-4 r ywdes e are also comparable, with the exception of the difference in the alpha-beta phase change temp um. Alloy 5 and Zircaloy-4 have e=*entially identical heat capacities up to l .g r.# =^=1 the Alloy 5 material undergoes an alpha-beta phase change. He Alloy ;
- 5 specific heat to the value applicable to the beta phase. Zircaloy-4 exhibits similar
! behavior, but the onset of the phase change occurs at a higher ^w are (approximately 1500'F),
i and the peak specific heat value reached during the transition is higher than for Alloy 5. Both Alloy 5 and Zircaloy-4 t+ sE= pure beta at approximately 1800*F. Since the total energy required to alter j the crystalline structure of the swo alloys is enentially the same and the Alloy 5 phase shift begins at a lower sym.ame, Alloy 5 has a broader functional relationship krea the heat capacity and j temperature than does Zircaloy-4. Dese differences in heat capacity kras Alloy 5 and Zircaloy-l 4 can have a small effect on calm 1*d- cladding temp- for transients with limiting temperatures in the phase transition zon(
l .
A review of the North Anna non-LOCA licensing basis analysis results indicated that y l' cladding temperature calculations for the lwd Rotor and Rod Ejection events ex -
For each of these events, the peak cinMing tempmaurs rapidly passes through the i transition zone, reaching values well into the beta phase. For such transients, there is a negligible j effect on the calculated cladding temperature. It is therefore concluded that the existing analyses
- which are based on the thermal characteristics ofZircaloy-4 are acceptable for evaluation of both I Alloy 4 and Alloy 5.
[
i l
l For the Locked Rotor and Rod Ejection events, transient behavior is also dependent upon detailed 4
fuel assembly design features such as pellet and cladding geometry. Virginia Electric and Power l Company performs analyses ofcladding temperature, behavior for these events using the RETRAN j Hot Spot Model (Reference 26).
i i In addition to the cladding material thermal properties, the REIRAN Hot Spot Model incorporates '
- numerous items which are a function of the fuel rod design. These include
- 1. Fuel pellet diameter
- 2. Fuel rod cladding inner and outer diameter
- 3. Fuel rod pitch *
{ 4. Fraction of heat generated in fuel q 5. Fuel melt temperature as a function of burnup l 6. Initial fuel density (percent theoretical) and enrichment
- 7. Pellet-cladding gap initial gas composition and backfill pressure f
1 i
i i
l
l l
Each of these items has been compared to determine whether the existing analysis parameter values i for the NAIF design bound the features of the Mark-BW17 assemblies. This review has concluded l that all key parameter values are either bounded by the NAIF design or are insignificantly different I between the two fuel designs. The cladding temperature analysis results for the Locked Rotor and l Rod Ejection events obtained for use of the NAIF design at North Anna will thus be conservative
! for the Mark-BW17 lead test assemblies.
6.4 Conclusions A review of fuel design feature differences between the Mark-BWl7 and resident NAIF designs has l indicated that only a limited number of features used as key NSSS accident analysis inputs are l different. Assessment of these specific differences for both events which have DNB and non-DNB acceptance criteria has concluded that the existing licensing analyses performed for the NAIF design will be applicable for the Mark-BW17 lead test assemblies in North Anna.
It should also be noted that there will only be four lead test assemblies with the Alloy 4 and Alloy 5 cladding materials, and that these fuel assemblies will not be placed in the highest fuel rod power density locations in the North Anna cores in which they are irradiated. This will further ensure the existence of margin to the safety analysis limits for these fuel assemblies. The conclusions of the North Anna analyses of record for the Chapter 15 non-LOCA accidents will remain valid for cores containing the four FCF lead test assemblies.
- 7. LOCA/ECCS Evaluation The North Anna plants are fueled with North Anna Improved Fuel (NAIF) supplied by Westinghouse Electric Corporation. The NAIF fuel design is similar to and compatible with the Westinghouse VANTAGE SH design. Compliance with 10 CFR 50.46 has been demonstrated by l calculations performed by the Virginia Electric and Power Company using the Westinghouse NRC-approved evaluation model and methods, and documented in the North Anna Units I and 2 UFSAR (Reference 27). This section documents calculations performed by Framatome Technologies, Inc.
(FTI) that demonstrate tha: the existing North Anna calculations based on the NAIF assemblie l
conservatively bound the LOCA performance of the Mark-BWl7 lead test assemblies. Several confirmatory large break LOCA (LBLOCA) calculatiohs were performed for the Mark-BWl7 fuel design. The results demonstrate that the NAIF design, assuming operation under identical constraints produces key results that bound those obtained for the Mark-BWl7 fuel. Thus, the NAIF LBLOCA licensing record can serve to demonstrate that the Mark-BW17 lead test assemblies meet the criteria of 10 CFR 50.46. For the Small Break LOCA, compliance with 10 CFR 50.46 is shown by validating that the calculations performed in support of the NAIF design are wholly applicable to the Mark-BWl7 test assemblies. This approach i:: pc~:ible because small breaks are plant system determinant and not dependant on fuel assembly design for reasonably equivalent designs. Thus, for the entire LOCA spectrum it is shown that the NAIF licensing calculations represent a conservative analysis for licensing the operation of the Mark-BWl7 lead test assemblies.
l
i l 7.1 CalculationalInputs and Assumptions l l
! The LBLOCA analysis supporting the l'icensing of the lead test assemblies was performed in !
l
~accordance with Revision 3 of the FTI recirculating steam generator LOCA evaluation model *
! (Reference 28). The evaluation of cladding temperature transients and local oxidation used with l three computer codes, intercoanae'aA as shown in Figure 7.1. The RELAP5/ MOD 2-B&W code j calculates system thermal hydraulics and al-Ading to.-y .iare ispas==, including the hot channel, l '
during blowdown. The thermal hydraulic transient calculations are continued within the
, REFLOD3B code to determine refill time and core reflooding rates. The BEACH code determines l
the hot pin al-Adia:i y.. are response during refill and reflood.
7.1.1 Inputs and Assumptions Because the purpose of the comparative LBLOCA calculations is to demonstra:e that the NAIF l licaa*ing calania+iaae for North Anna bound the results for the Mark-BW17 lead test assemblies, the i major plant operating p et ; used in the calculations curiospond to those used in the NAIF j calculations. The key p ucters and their values, also summarized in Table 7.1, are:
l l '
- 1. Power Level - The plant is assumed to be operating in steady-state at 2951 MWT (102% of j 2893 MWT).
i l
b 2. Total System Flow - The initial RCS flow is 288,000 gpm. l l .
I j .
- 3. Fuel Parameters - Studies di*en**ad in Section 7.2.2 of this Safety Significance mueament
- . show that fuel conditions at the beginning oflife are the most severe for the LBLOCA 5 evaluation of the Mark-BW17.
- 4. ECCS - The ECCS flows are based on the assumption of a single active failure that takes one i complete train of ECCS out of service. This is the worst case ==mation for the North Anna j NAIF fuel LOCA calculations and is preserved in the Mark-BW17 comparative calenwiaa*.
i 5. A value of F =n 2.19 was used for the total peaking factor.
j 6. The Steam Generator tube plugging level was set at 7 percent per generator.
i
) 7. The FTl LOCA evaluation model does not require fixing or controlling the relationshi~p
{ between the axial and radial peaking factors. Under the FTI model, control of the maximurn i local heating rate is considered sufficient to assure a conservative prediction of peak cladding i temperature. Furthermore, the model does not differentiate the hot pin from other pins in the i hot assembly. Thus no separate Fag value is employed. To make the comparison between
- the Virginia Electric and Power Company LOCA calculations and the Mark-BW17 calculations, the hot assembly radial p aking from the NAIF calculations was used with a i
revised axial peaking that would generate the total peak, i.e. preserve Fq = 2.19. This provides a correspondence between the models that matches both the energy deposition to
't
-mm,
the fluid cooling the hot pin and the maximum local power at the peak in the hot pin.
Although FTl LOCA methods do not require control of bundle power or Fan. the Mark.
BW17 assemblies will be controlled to the same criteria as the NAIF assemblies.
'8. Because it is anticipated that the Mark BW17 test assemblies will be irradiated in North Anna Unit 1, the calculations were performed for the upflow baffle gap configuration of
. North Anna 1 and the calculational results are compared to Virginia Electric and Power
. Cnmaaay's NAIF e=Imiteiaae for this configuration. Should the assemblies be irradiated in North Anna Unit 2, which has a downflow befDe gap canA miration, the differential peak clad ta==are== (APCT) as calailmed for the NAIF assemblies will be added to the tw.are results for the Mark-BW17. His is appropriate haa-a , as will be seen, the LOCA results for the assemblies are similar, with the NAIF results being alightly conservative.
- 9. The mod .ter density reactivity coefBeient is based on beginning-of-cycle conditions to
~
rninimim negative reactivity.
- 10. Clad swelling and rupture is evaluated with the FTI RSG evaluation model rupture model for Zircaloy-4.' He model is based on NUREG-0630 (Reference 29). Section 7.1.4 of this l
===eament presentsjneina* ion for the application of this model to the advanced al Adiag material ismym.ie.d into the Mark-BW17 design
- 11. Both the structural grids and the mid-span mixing grids (MSMGs) are explicitly modeled using the .yy. h set forth in the FTl evaluation model.
7.1.2 RELAP5 MOD 2-B&W and BEACH Modeling The RELAPSMOD2-B&W computer code is used to analyze RCS dwrmal hydraulic behavior and cladding temperatum response during the blowdown phase of a LOCA. In its BEACH implementation, RELAPSMOD2-B&W is also the hot channel calculation for refill and reflood.
RELAPSMOD2-B&W, a modified version of the RELAPSMOD2 code, is documented in BAW-10164 (Reference 30). De BEACH implementation of RELAPS is documented in BAW-10166 (Reference 31).
' RELAPS permits the user to select model representation that results in a suitable f* mite difference model for the fluid system being analyzed. Control volume inputs generally consist of geometry (area and height), flow-related parameters (resistance, hydraulic diameter, and surface roughaaa),
and initial conditions (pressure, temperature, and flow). He non-equilibrium /non-homogeneotis option is used throughout the model, except for the core region where the equilibrium / homogeneous option is selected because the core heat transfer package is based on such an assumption. Flow paths
- l. are defined between control volume geometric centers. The model is run in a steady-state mode to assure properinitialization.
BEACH and RELAPSMOD2-B&W employ the same core model; BEACH is actually a restart of l RELAPS with controlled heat transfer logic and without the loop modeling. Two core nodalization
- schemes were used for the comparative calculations. De preliminary sensitivity studies, break t
s-r --,-,- - , -- - - .- - ----w--
i i
- i spectrum and burnup sensitivity were performed using a model that incorporated all features of the j Mark BW17 fuel assembly except the mid-span mixing grids (MSMGs). This lack of detail limits
.the application of the model for fmal predictions but does not compromise its application for
'detennining sensitivity study trends. For the actual comparison runs used to confirm the
- i applicability of the Virginia Electric and Power Company LOCA calculations and the North Anna 3 operational limits to the Mark-BW17 anemblies, a spatially refined model which includes the MSMGs was used. The nodalintion for the sensitivity studies is shown in Figures 7.2 and 73. ne core node distribution is based on three nodes per grid span in accordance with Appendix C of the
! BEACH topical (Reference 31). .
I i ne core node distribution for the final comparison calculations adds more detail. The three node
! per grid span model remains but, with the addition of MSMGs after the fourth, fifth, and sixth i structural grids, the total axial node count for the core increases to 29. The resulting arrangement
! is shown in Figure 7.4. ne nodallength is defined so that each grid is located at or near the bottom j of a node and three nodes are used to cover a grid span. This eyymsch is used with both the
- structural and the mid-span mixing grids. Cross-flow is modeled at each axial elevation. Le cross-
! flow resistance is kept at the value specified by the evaluation model but path areas are reduced to
'ceispond to the new nodalheights.
! . Both of the FH models simulated a full core of the Mark-BW17 fuel. Section 7.8 of this assenment l dim the application of the results within a core made up mostly of NAIF assemblies.
l l . 7.13 REFLOD3B Mad ling -
i l The REFLOD3B code simulates the thermal hydraulic behavior of the primary system during the' j core refill and reflood phases of the LOCA. The noding, shown in Figure 7.5, consists of reactor vessel and loop models. RELAPS results at the end of blowdown (EOB) define the starting point
] for the REFLOD3B calculations. Durin lent, the primary metal surface heat transfer
- coefficient for regions with flow is set t TU/hr-ft2 F, insuring that the fluid leaving the l steam generator is continuously dry steam, superheated to the secondary side temperature. He i pump rotor resistance is based on the locked rotor condition for the North Anna reactor coolant i pumps (Westinghouse Model 93A pumps). ( hsi pressure drop is imposed on cold leg pipe i junctions to account for momentum losses'd'ue to steam ECC water interaction during the l accumulator injection phase. This value is reduced t umped injection only) once the j accumulators have fully discharged. Hes as containment b(ackpress a function of time from the j North Anna UFSAR LBLOCA minimum containment pressure calculations are used in the
{ REFLOD3B calculations, ne use ofUFSAR minimum containment pressure predictions has becin j employed in previous fuel reload licensing and is specifically approved within the FTl evaluation l model. -
l
}
I 7.1.4 Cladding Oxidation and Swelling and Rupture Models l \
j As described in earlier sections, the Mark-BW17 fuel design incorporates two advanced cladding i alloys designated as Alloy 4 and Alloy 5. These alloys do not fall within the Zircaloy specification,
- necessitating confirmation that certain Zircaloy-based LOCA fuel performance models can be f
J I
i 4
. reasonably applied to the new materials. Validation was performed by Framatome for high j temperature oxidation, cladding brittle fracture, and clad swelling and rupture modeling. Because Alloy 5 comprises the majority of the cladding in the lead test assemblies and because Alloy 4 i closely perpliels the Zircaloy 4 specification, the brittle fracture testing (cold water plunge tests) and l the high temperature oxidation rate testing were conducted only for the Alloy 5 material. The results show that the Baker-Just oxidation correlation is conservative for the advanced zirconium alloys. that
. the oxidation limit for brittle fracture is the same for these alloys as for Zircaloy, and that the
- NUREG-0630 model can be applied for the prediction of cladding swelling and rupture.
~
! Brittle Fracture and Oxidation Tests '
l i
i' High k.or e- oxidation imfo.u-is for Alloy 4, Alloy 5 and Zircaloy can be expected to be j nimilar because these alloys are p- h---In;ndy zirconium and the oxidation ofinterest occurs after j the change to the beta phase. For this reason, the early testing of the Alloy 5 high temperature
[
oxidation rate was combined with the brittle fracture testing. Piahi fuel pin samples were
~~adad above.a cold water pool and heated in a steam environment until oxidations of 20 to 30
' percent were achieved. He samples were then plunged into the water bath and quenched. ne occurrence of brittle fracture is indiced if the sample can not continue to hold pressure during and after the q-Mag process. The k.or e- at which the oxidation took place was measured by l optical devices and fed back to the beating mechanien, maintaining a constant temperature during i oxidation.
J l
!' , Post-quench examination of the sample provides the oxidation thickness actually achieved during l
! the testing. By comparing the time at temi en to the measured oxidation, the rate of oxidation i can be determined. De results indicate that the rate of oxidation at high temperatures for Zircaloy
} and Alloy 5 are similar and that the advanced zirconium-based alloys are easily bounded by the l Baker-Just correlation. Thus, the Baker-Just correlation can be conservatively applied for the computation of cladding oxidation as required by Appendix K of 10 CFR 50. The result of the
- quenching tests showed that the threshold for brittle fracture occurs atCpercent oxidation.
- NUREG/CR-1344 (Reference 32) indicates that the threshold for Zircaloy occurs at approximately
- 20 percent. Thus, the brittle fracture of the advanced alloy cladding material is no more likely than
[
the fracture ofZircaloy provided the 17 percent local oxidation limit of 10 CFR 50.46 is met.
i i Claddino Swelline and Runture Tests f
l The analyses performed to support the licensing of the Mark-BW17 lead test assemblies used the L FTl implementation of the NUREG-0630 cladding swelling and rupture models. To demonstrate i , the applicability of these models, single-pin rupture tests were conducted at the French l Govemment's EDGAR test facility. The requimment, as sym.s.w.d in Appendix K, for clad swelling l and rupture modeling is that "... the degree of swelling and the incidence of rupture are not
- underestimated." NUREG-0630 established models that meet these criteria for Zircaloy and thereby j the degree of conformity required between the experimental result of rupture testing and the predictive capability of a rupture model. To date, the testing done on Alloy 4 and Alloy 5 shows that i rupture test asults lie within the dispersion of the NUREG-0630 experimental data base and can be l expected to correlate equally well with the NUREG-0630 rupture models. Therefore, it is reasonable j
i i
1 t_._____ .-%_ _
i I
to apply the NUREG-0630 rupture models within the LOCA models for the Mark-BWl7 lead test assembly calculations.
The EDGAR test facility is comprised of a tank within which a pressurized tubing sample can be heated at various rates until rupture occurs. A schematic of the test facility is shown in Figure 7.6.
Both creep and ramp testing have been conducted for the advanced alloy materials. Creep results, however, are only of limited interest in generating LOCA models because ruptures occur within ramps during LOCA. Further, for the North Anna LOCA transients (see Section 7.3 of this l
assessment), rupture occurs during reflood and the cladding heatup rate is limited to approximately 1 20*F (10 to 12'C) per second. Therefore, the slow ramp correlations and supporting data from NUREG-0630 are of the most interest in determining the applicability of the model to the Alloy 4 and Alloy 5 material. Figures 7.7 and 7.8 show the results of four Alloy 5 tests done at a ramp rate of 10 C/sec in comparison with the NUREG-0630 data base and the NUREG-0630 slow ramp i correlations (heating rates of 9 to 11 C/sec). '
As can be seen in the figures, the data lies within the span of dispersion of the data upon which the NUREG-0630 models were based. As shown in Figure 7.7, the Alloy 5 rupture temperature versus stress results follow the general trend of the NUREG-0630 data base and correlation. Figure 7.8 shows wide dispersion in the cladding strain data for Zircaloy. The dispersion is due to a variety of differences in the actual testing, but mainly to the stochastic nature of the phenomena being tested.
As with the rupture stresses, the Alloy 5 strain results fit within the experimental data base for the NUREG-0630 correlations. The results of preliminary rupture testing on Alloy 4, not shown, also fit within the data result range upon which the NUIEG-0630 correlations were based. The overall conclusion is that the NUREG-0630 models fit or bound the cladding swelling and rupture characteristics expected for the Alloy 4 and Alloy 5 materials. It is thereforejustifiable to apply the NUREG-0630 models to the analysis of the LOCA performance of the Mark-BWl7 lead test assemblies for North Anna.
7.2 Sensitivity Studies Although a considerable portion of the analysis inputs and assumptions for the North Anna lead test assemblies are set by the FTI evaluation model, some parameters are dependent on plant-specific or fuel design values. Two sensitivity studies were perforriled for the Mark-BWl7 calculations in order to assure a proper comparison i etween the FTI evaluation model predictions and the Virginia Electric and Power Company NAIF calculations. A break size spectrum was run to assure that the worst case break for the FTI evrduation model was used in the comparative cases, and a bumup study was conducted to assure that the calculations for the Mark-BW17 were done at the most limiting time in life. The core power distribution selected for these studies was peaked near the center of the core. The vessel modeling used is depicted in Figure 7.3.
7.2.1 Break Spectrum Analysis A discharge coefficient study, with coetilcients of 1.0, 0.8, 0.6, and 0.4, was conducted for a guillotine break of twice the piping area and located in the pump discharge piping. Table 7.2 provides the results of the study. Although there was very little difference in the results from one
case to another, the worst case was identified as the 0.6 discharge coefficient. Parameters ofinterest for the worst case are shown in Figures 7.9 through 7.13. Figure 7.14 compares the peak cladding
' ternymmre response for each discharge coefficient case. There are no major difTerences among the seluences of events for the four cases that make up the discharge coefficient study. Table 7.3 presents the sequence of events for the Cd = 0.6 case.
4 As shown in Table 7.2, the peak cladding temperatures differ by less than 65 F. This result is expected and consistent with the app!ication of the FTI evaluation model to other plant types.
Although the differences in cladding temperature response are small, the Cd of 0.6 has been incorporated as the worst case within the remainder of the FTI calculations and will be the case compared to the NAIF worst case calculation. For the Virginia Electric and Power Company NAIF 3
calculations, the Ca of 0.4 was identified as the worst case.
7.2.2 Time in Life Study Burnup sensitivity studies are conducted for the purpose of determining if the combination ofinitial fuel stored energy and initial internal fuel pin pressure selected for the LOCA evaluation comprise
, a worst case combination. Based on generic sensitivity studies (Reference 28), the FTI evaluation model concludes that the worst case combination will be that with the highest initial fuel stored j energy so long as no other combination of conditions can lead to a cladding rupture during blowdown. Although the creep performance of the advanced cladding materials used in the Mark-BWl7 lead test assemblies differs from that of Zircaloy, the highest fuel stored energy still occurs at beginning-of-life conditions. However, it is not obvious from inspection of the initial fuel conditions as functions of burnup that an irradiated condition will not lead to a fuel cladding rupture during blowdown. Therefore, a separate blowdown calculation was performed using the end-of-life fuel stored energy and internal pin pressure. The result showed that no blowdown rupture would occur for the North Anna calculations. Therefore the beginning-of. life fuel temperatures and intemal pin pressures are used for the comparative calculations.
, 7.3 Comparison Calculation Results and Verification of K, Curve Typically the LOCA evaluation is completed with a set of analyses to show compliance with 10 CFR i
50.46 for the core power and peaking that will limit plant operation. For the comparative analyses, this entails a demonstration of the Mark-BWl7 LOCA performance for the K, and Fn combination approved for the NAIF fuel. To accomplish the K, validation and to demonstrate that the NAIF calculations provide a conservative bound for the Mark-BWl7 LOCA performance, analyses were conducted for three different axial power shapes. The results of these calculations are presented below and compared with the NAIF worst case calculation results.
Figure 7.15 shows the axial distribution of normalized power peaking applicable to the North Anna plants. To determine the applicability of Figure 7.15 to the Mark-BW17 lead test assemblies and determine a worst case power distribution from the distributions allowed by Figure 7.15, the LBLOCA results for three different axial distributions were calculated. For full applications of the FTI LBLOCA evaluation model (Reference 28), five core elevations would have been calculated.
However, for this application, where a K, distribution is already established, it is necessary only to
f demonstrate that the worst case can be confirmed for the approved K, distribution. The Westinghouse evaluation model, used by the Virginia Electric and Power Company, demonstrates that the 6 foot peak is the worst case so long as the K, of Figure 7.15 is applied. Through calculation of a 4.6,6.7, and 10.1 foot peak with the FTl model, the worst case for the Mark-BWl7 lead test assemblies is established as an axial power shape peaked near the 6 foot elevation. A comparison of the FTl worst case to the NAIF worst case shows that the Virginia Electric and Pover Company '
K, curve is appropriate for the Mark-BWl7 design and that the NAIF calculation results are comparable to, and slightly bound, the Mark-BWl7 results.
The calculations were performed using the core model depicted in Figure 7.4 which incorporates the MSMGs. Figure 7.16 shows the power shape for each of the runs. Even though K, has decreased I slightly by the 6.7 foot level, both the 4.6 foot und the 6.7 foot peaks used an nF of 2.19. For the l case with an axial peak at 10.1 feet the Fq was reduced in accordance with K, to a value of 2.08.
The results of the calculations are tabulated in Table 7.4. The worst case PCT,1966 F, is for the peak at 6.7 feet. The 4.6 foot peak is somewhat lower,1935 F, because ofits proximity to the advancing quench front. The 10.1 foot elevation produces a lower cladding temperature because of the reduced Fn,2.08, and because the MSMGs are fully effective at this elevation. Figures 7.17 through 7.19 show the key parameters for the 6.7 foot peaked case. Figure 7.20 displays the cladding temperature responses versus time for the three separate axial power distribution cases.
7.4 Mark BWl7 Compliance with 10 CFR 50.46 Peak Claddine Temoerature The worst case LOCA result published for North Anna. 2013 F,is for a 6 foot axial peak. However, that result is for a downtlow baffle gap configuration which is representative of North Anna Unit 2.
The comparative case result for the upflow configuration at North Anna Unit 1 is 1975 F. This would compare to the 6.7 foot result of 1966 F for the Mark-BWl7 lead test assemblies. Table 7.5 presents a comparison of key results from the Virginia Electric and Power Company NAIF calculations and the FTI Mark-BWl7 calculations for both North Anna units. The results are essentially the same with the NAIF results providing a slight bound of the cladding temperatures expected for the Mark-BW17 fuel. This means that evaluations of the NAIF can be conservatively applied to the Mark-BWl7 lead test assemblies so lorig as the operating conditions for the Mark-BWl7 assemblies are limited in the same fashion as for the NAIF assemblies. Because the two fuel assembly designs will be operated under the same Technical Specification limits and operating constraints, the North Anna licensing calculations for the NAIF assemblies can be applied to the licensing of the Mark-BWl7 lead test assemblies and supply the necessary demonstiation of compliance with the criteria of 10 CFR 50.46.
Local Claddine Oxidation Table 7.5 compares the amount oflocal oxidation between the Virginia Electric and Power Company calculations and the Mark-BWl7 results for the centrally peaked power distribution cases. The FTl results show a maximum oxidation of 3.5 percent and the NAIF results show e maximum of 4.5 percent oxidation. The most severe local oxidation for the Mark-BW17 fuel,4.7 percent, occurs
i when the core power is peaked at 10.1 feet (see Table 7.4). This value is slightly higher (0.2 percent) than the maximum oxidation for the NAIF assemblies. However, the results are all well below the 17 percent criterion of 10 CFR 50.46, assuring that assemblies of both designs are in compliance with this criterion.
Maximum Hydrogen Generation The North Anna calculation for core wide oxidation and hydrogen generation will not change because of the inclusion of four FCF supplied lead test assemblies regardless of the assemblies' ;
LOCA performance. Additionally, a comparison between the calculated results for the Mark-BWl7 l
and the NAIF fuel designs shows that the peak local oxidation for the NAIF is comparable to that j for the Mark-BWl7. Therefore, the prediction of acceptable core wide oxidation for the full NAIF l core also demonstrates that North Anna cores with the four FCF-supplied lead test assemblies will i
also meet this 10 CFR 50.46 criterion. '
Coolable Geometry The fourth acceptance criterion of 10 CFR 50.46 states that calculated changes in core geometry shall be such that the core remains amenable to cooling. The calculations in Section 7.3 of this assessment directly address the alterations in geometry for the Mark-BWl7 that result from the worst case LOCA. These calculations demonstrate that the fuel pin is cooled successfully. As discussed in Section 7 of the FTl evaluation model report (Reference 28), clad swe!!ing and flow blockage due to rupture can be estimated based on NUREG 0630. For the Mark-BW17 calculations, the hot assembly flow area reduction at rupture is less than 40 percent. Furthermore, the upper limit of possible channel blockage, based on NUREG-0630, is less than 90 percent. Neither 90 percent blockage nor 40 percent blockage constitutes total subchannel obstruction. Since the position of rupture in a fuel assembly is distributed within the upper part of a grid span, subchannel blockage will not become coplanar across the assembly. Therefore, the assembly retains its pin coolant-channel pin coolant-channel arrangement and is capable of passing coolant along the pin to provide cooling for all regions of the assembly.
The effects of fuel rod bowing on whole-core blockage are considered in the FCF fuel assembly and fuel rod designs in such a way as to minimize the pote'ntial for rod bowing. Minor adjustments of fuel pin pitch due to rod bow do not substantially alter the fuel assembly flow area and the average subchannel flow area within an assembly is preserved. Therefore, due to the axial distribution of blockage caused by rupture, no coplanar blockage of the fuel assembly will occur and the core will remain amenable to cooling. Effects upon the fuel pin lattice from the combined mechanical loadings of the LOCA and a seismic event have been calculated separately for the Mark-BWl7 design. The loadings remain within the Mark BWl7 elastic limits regardless of the core locations for the lead test assemblies. Therefore, there is no permanent deformation of the Mark-BW17 fuel for combined LOCA and seismic loads.
The consequences of both thermal and mechanical deformation of the Mark-BW17 lead test assemblies in the North Anna core have been assessed and the resultant deformations have been shown to maintain coolable core configurations. Therefore, the coolable geometry requirements of L _ _ _ __ _ _ _ _ -
10 CFR 50.46 have been met for the Mark-BW17 assemblies and the assemblies have been shown j' to remain amenable to core cooling following a LOCA.
i 'inno-Term f% aline l ,
l The fiAh = - y---- criterion of10 CFR 50.46 states that the calculated core temperature shall be j . maintainari at an ==q -bly low value and decay heat shall be removed for the extended period of i time required by the long-lived radioactivity remaining in the core. This criterion is a system level 4- criterion and k'-j= ' M of the fuel design. Dere have been no system level changes introduced i with this reload that would alter the long-term cooling process. Therefore, the calculations and
!' argumentsr- 14 to license Forth Anna remain valid with four FCF-supplied Mark-BW17 fuel
- assemblies operationalin the core.
1 7.5 SmallBreakLOCA l
' e current licensing bases forNorth Anna comprise a spectrum oflarge and small break loss-of-1
,lant accidents (LOCAs)- lyM by Virginia Electric and Power Company and documented in gdad final safety analysis report (UFSAR). For operation ofNorth Anna with four FCF lead l
a assemblies, FTI has reanalyzed part of the large break LOCA transients as pra=antad in the !
. faregoing sections. Reanalysis of the small break LOCA for operation with FCF test fuel ===amhlies .
is not required since SBLOCA evaluations are unaffected by the design differences bere. c.a the l Mark-BW17 and the Waaiaghause fuel assemblies. Hus, the reference UFSAR analyses remain !
.. the bases for plant licensing and h- ama the basis for licensing use of the Mark-BW17 lead test -
assemblies in North Anna.
Fuel r.ssembly design influences the S.BLLYA calculations only to the extent that it affects overall !
system behavior. In that regard, differences between the Mark-BW17 fuel assemblies and the I resident NAIF assemblies should not materially affect the bounding SBLOCA sequences of the reference UFSAR. He FCF lead test assemblies and resident Waeinghause assemblies differ in the following areas: unrecoverable pressure drops across the assemblies, initial fuel temperatures, and initialpin internal gas pressure.
The Mark-BW17 fuel assemblies have unrecoverable pressure drops that are approximatel( )
higher than those of the NAIF assemblies at the North Anna design flow. The inclusion of four of these assemblies within the core will not affect the overall loop or core pressure drop. Thus, the
-initial subcooled depressurization phase of the SBLOCA will be unaltered. The reactor trip signal ~
and pump trips will occur at the same time in the transient as in the reference UFSAR calculations.
For the same reason the pump coastdown and natural circulation phases will be unaffected. The CHF performance of the Mark-BW17 fuel design exceeds that of the NAIF assemblies due to the inclusion of the mid-span mixing grids. Thus, core resistance variations will not change the fuel thermal transient or impact existing evaluations.
- Changes in the initial fuel temperature add or subtract overall energy from the RCS. However, the inclusion of only four assemblies will not alter the fuel energy removed from the reactor coolant system during the reactor coolant pump coastdown phase. Because all of the initial fuel pin energy
is transferred from the fuel pin during the early phases of the SBLOCA transient. the initial fuel enthalpy at operation has virtually no impact beyond the loop coastdown period. The core energy content during loop draining and core boildown will be identical to the current licensing base because it is solely depaad=at on the decay heating rate.
De fuel pin intemal fill gas pressure for the Mark-BW17 lead test assemblies is lower than that for the resident NAIF assemblies. The difference will result in lower intemal pressure both initially and during the LOCA transient. His will lead to a rupture of the Mark-BW17 fuel assembly at a somewhat later time than would occur for the Wa*iaghause fuel. An alteration of rupture timing can affect the result of a small break LOCA calculation ifit is possible for rupture to occur at very high temperatures. At L -- r m appra ahing 2000*F, the oxidation rate of unoxidized cladding is significant. If the fuel pin ruptures at these e-, .ozres, the oxidation on the inside of the c1=dding will drive the eladding temperature significantly higher. This is not a concem at North Anna haa='i- the peak SBLOCA al=Ading =+;e- s is approximately 1700*F and there is, thus, no possibility of a fuel pin even getting to the 2000*F range. Other than the possibility for high temperature rupture, the effects of rupture and rupture timing on SBLOCA are benign and not of concem for LOCA calc, ulational reruits.
As a final point, the T=V-J =J Spreifientions for a%wable local power levels, core pa leing. for core elevations at or above 6 feet wn! not be changed due to the inclusion of the four FCF lead test assemblies. Thus, the axial power profile used by Virginia Electric and Power Company in the current SBLOCA analyses remains hamding. This assures that the thermal power imposed on the fuel during a temperature excursion remmim conservatively modeled. The cladding temperature results for the current UFSAR evaluations are, therefore, conservative for the Mark-BW17 lead test assemblies.
In summary, the cost resistance variations will not affect the system flows such that the controlling hot leg temperature or CHF points are altered. The steam generator heat removal rate during the )
flow coastdown period will compensate for any initial fuel stored energy fluctuations. All but one controlling parameter in the phases following the pump coastdown and natural circulation phase will be unchanged and that one, rupture timing, does not affect the North Anna SBLOCA' cladding temperature prediction. Since the overall RCS geometry, initial operating conditions, licensed power, and goveming phenomena are effectively unchanged, the existing UFSAR calculations remain bounding for operation of North Anna with four FCF-supplied Mark-BW17 lead test assemblies. Therefore, the present SBLOCA evaluation calculations are applicable to the Mark-BW17 fuel for demonstrating compliance with the criteria of 10 CFR 50.46.
7.6 Mixed Core Considerations The Mark-BW17 test assemblies will reside in a mixed core configuration with the NAIF assemblies
, throughout their irradiation. The LOCA analyses in support of the Mark-BW17 lead test assemblies
( have been performed assuming that the entire core is comprised of Mark-BW17 assemblies. This
! standard FTl approach for reload fuel licensing calculations provides an adequate evaluation of the i fuel in both the mixed and unmixed core configurations.
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Differences between the mixed and pure core conditions arise because of the different pressur ,
of the two fuel assembly designs. At the North Anna design flow. the Mark-BW17 design has '
approximately gher pressure drop than does the NAIF assembly. Thus, flow can be
. expected to di sh)ghtly from the Mark-BW17 to the NAIF asse !
l involved in most reload analyses where the utility has switched fuel suppliers. Studies performed by FTl for other utility transitions from Westinghouse fuel to FCF-supplied fuel have shown that i
i the effects ofreasonable fuel assembly pressure drop differences on cladding temperatures durin blowdown and reflood s.a M1=nce each other, resulting in comparable peak cladding temperature predictions be.;.c e the LOCA results of mixed and pure core configurations.
ForNorth Anna WW1y, with only 4 of the 157 fuel assemblies in the core being of the Mark-BW17 design, there will not be mdficier* diversion ;-w At ot affect the cunent LOCA calculations based on NAIF assemblies. 'Ibe effects that would be imposed on the Mark-BW17 calculations b direct i.r= = Mon of the mixed core configuration are:
(1) Some flow will divert to the NAIF assemblies during blowdown resulting in a alight increase in the predicted cladding : -ym.im at the end ofblowdown, and (2) The core reflooding rate, being controlled by the average channel flow resistance,
' will be increased above the reference Mark-BW17 calculation resulting in better reflood cooling. '
Because peak c1=dding 'w_=.5 occur at **= dad imes t during reflood, the reflooding rate *'
impvement is likely to be dominant. However, the two effects essentially trade off against each other making the reference calculations appropriate for either a mixed or full core condition.
Therefore, the pure core calculations performed to support either the Mark-BW17 or the NAIF remain valid during the mixed core configuration. 'Ihe licensing position for the Mark-BW17, that the calculations done for the NAIF assemblies bound the Mark-BWI7 assemblies and c for licensing the Mark-BW17 test assemblies, is valid for mixed core configurations.
l 7.7 LOCA/ECCS Summary and Conclusion Calculations have been performed to demonstrate the LOCA performance of the Mark-BW17 lead test fuel assemblies in North Anna Unit 1 or 2. The calculations, performed with the NRC-approv FTl LOCA evaluation model, and the other supportive material referenced demonstrate that the five criteria of 10 CFR 50.46 are met. Specifically, it has been shown that for the operation of No'rth Anna Unit 1 or 2 with four FCF Mark-BW17 lead test assemblies:
- 1. The calculated peak cladding tempm.ims for the limiting cases are less than 2200'F.
4-
- 2. The maximum calculated local cladding oxidation is less than 17.0 %.
i i
) 3. The maximum amount of core-wide oxidation does not exceed 1.0 % of the fuel cladding.
- 4. The cladding remains amenable to cooling.
- 5. Long-term cooling is established and maintained after the LOCA.
Further, it has been demonstrated that the existing Virginia Electric and Power Company calculations for the NAIF assemblies produce results which are comparable to and which slightly bound the FTI results for the Mark-BW17 lead test assemblies. This allows the licensing calculation for the NAIF assemblies, in conjunction with the calculations documented herein, to serve as the licensing basis for the Mark-BWI7 lead test assemblies in North Anna. Therefore, ifin the future, additional LOCA calculational justification is required for North Anna, in mest cases it will be sufficient for Virginia Electric and Power Can=ay to perform those calculations only on the NAIF assemblies using the Wa*iaghause evaluation model. I
- 8. Applicability of Standard Reload Design Methodology Virginia Electric and Power Company performs a reload safety evaluation using a bounding analysis method as described in Topical Report VEP-FRD-42 Rev.1-A (Reference 3). This methodology defmes a set of key analysis parameters that fully describe a valid conservative safety analysis
(" reference analysis"). If all key analysis parameters for a reload core are conservatively bounded by the corrapaadiag parameters in the reference analysis, the reference safety analysis is bounding, and further evaluation is not aa~~ y. When a key analysis parameter is not bounded, further review is considered nece-ry to ensure that the required safety margin is maintained. 'Ihis last determination is made through either a complete reanalysis of the accident, or through a simpler, though conservative, evaluation process using known p--octs sensitivities.
The NRC Safety Evaluation Report.(SER) approving use of the normal reload nuclear design methodology in Reference 3 concludes that the report is "... acceptable for referencing by Virginia Power in licensing Westinghouse supplied reloads of Westinghouse supplied reactors." At least two
- additional statements in the summary of the NRC's evaluation of the report also specifically confine the applicability of this methodology to only Westinghouse fuel. Other than noting the similarity of our methodology to that of our current fuel vendor, no specific details are presented to clarify what concems the NRC may have regarding application of our current reload methodology to fuel supplied by other vendors. No similar statements are found in the SERs for our topical reports on specific analytical models or methods.
\1 Virginia Electric and Power Company recognizes that fuel products supplied by different vendors could conceivably differ dramatically in design from our current fuel. In such a case, it would lie desirable to benchmark our models and methods to either actual irradiation data or to the new f supplier's approved models. Such beachmarking would demonstrate that we can accurately predict the behavior of the new fuel design, and ensure that its predicted behavior can be satisfactorily incorporated into both reload design and conservative accident analyses. This is particularly important when full reloads of the new design are to be implemented, where the presence of the new
{ fuel design can have a significant effect on the performance of the fuel remaining in the core, and
! where use of the new design may require new limits for accident analyses.
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1 l The lead test assembly program at North Anna. while using fuel assemblies provided by a new fuel i vendor. will not result in a significant change to the North Anna cores. Several factors support this conclusion. including: similarity of the Mark-BWl7 fuel design to the resident Westinghouse fuel, j use of a limited number (four) of the FCF fuel assemblies, and exclusion of these assemblies from i core locations where they would experience the highest fuel rod power density to ensure that existing
{ safety analyses remain applicable. ,
- The North Anna lead test menembly fuel design (FCF's Mark-BW17 design)is an extension of FCF's i Mark-BW fuel design, which was designed specifically for use in Westinghouse units and for l ' compatibility with the resident fuel in those units. The physical dimensions of this fuel are very P
similar to those of Westinghouse fuel, particularly the older Wa*iaghause LOPAR (Inconel grid) design, as can been seen from Tables 3.1 and 3.2. The meeambly envelope and fuel rod pitch within
- the envelope are comparable for the FCF and Wa*iaghause designs. Individual fuel rod dimensions (including fuel stack length, fuel pelletdiameter, pellet-to-clad gap size, and clad thickness) as well as guide thimble dimensions are comparable to the Weeinghause fuel used in North Anna.
Evaluations of the mechanical fuel performance are p fm-ed by the fuel vendor (FCF for the lead test anemblies, and Waeingha=- for the normal reload fuel). 'Ibe FCF evaluation for the lead test assemblies has also addressed co-rdbility with the resident Waeinghouse fuel.
t 1 ' Physical differences in the lead test assembly design that coul,d affect core reload neutronic calculations such as the slightly higher uranium loading, the chemical composition of the cladding; j -
and the presence of additional material in the active fuel regions (the mid-span mixing grids) can be easily incorporated into our core design models. Past changes made to the model inputs to j
incornweneutronicallysignificantd=9 to the Waeiaghause fuel designs have resulted in fully i .
i acceptable predicted-to-measured power distributions and reactivity parameter agreement. Changes to the neutronic model inputs we==y to model the physical differences between the lead test
- assemblies and the resident Westinghouse fuel assemblies are similar to those used for previous j
. Westinghouse fuel product changes, and are of a smaller magnitude than was necessary for many i
of the Westinghouse fuel product ehmages. With only four FCF lead test assemblies in use at North l Anna, there will be a very limited impact on the overall core performance, with only a minor effect
- j. on core depletion, core reactivity parameters, or core reactivity control.
! Thermal hydraulic analyses of the lead test assemblies were performed by FCF using their NRC-i approved models and methods. It was determined that use of the four Mark-BW17 lead test j assemblies will have a negligible impact on core thermal hydraulic evaluations, and that North Anna cores containing these assemblies can conservatively be modeled as a homogeneous core of Westinghouse fuel. Therefore use of these lead test assemblies will have no impact on Virginia
- Electric and Power Company's standard reload thermal hydraulic evaluations.
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{ There are no differences bet =ca the FCF and Westinghouse fuel designs that would result in a different set of key analysis parameters than those already defined for the current safety analyses
{ (Reference 3). Sensitivity of core transients to changes in the key analysis parameters will remain
} the same for the FCF fuel design due to the small difference from the Westinghouse fuel design. The j impact of the lead test assembly design on both LOCA and non-LOCA accident analyses has been j considered, by Framatome Technologies Inc. (FTI) and by Virginia Electric and Power Company 4
i
i i using appropriate input provided by FCF. respectively. The analyses of record. which are based on i
the Westinghouse fuel design. will remain applicable for the lead test assemblies. Cycle specific evaluatiorc will continue to verify that the assumed values for any key analysis parameters are not exceeded for cycles in which the lead test assemblies are irradiated. There are no differences between the Mark-BW17 and Westinghouse fuel designs that could result in new failure red =!== that would increase the consequences ofpreviously considered accident scenarios, or that would interfere with safe operation of the core. %erefore, incorporation of these assemblies into Nor'th Anna cores will not affect the ability of our current reload methodology to conservativ assess the core response to accident scenarios. -
, 9. Aeneament ofUnreviewed Safety Question he' four FCF Mark-BW1*/ lead test assemblies are very similar in design to, and will exhibit performance cou, .ble to, the resident Wa=daghause fuel assemblies at North Anna. There will be nd reduction in the design margin of safety. It is concluded that neither the use of the four FCF Mark-BW17 lead test assemblies at North Anna, nor the use of Virginia Electric and Power
. Company's standard reload design methodology to evaluate cores in which these assemblies are irradiated, results in the acceptable safety limits for any incident being exceeded or in any unreviewed safety questions as define ( by 10 CFR 50.59 (a)(2). The basis for this determination is delineated below. .
9.1 Probability ofPreviously Evaluated Accidents This safety significance mae== ment dem that the probability of an accident previously evaluated in the North Anna Units 1 and 2 UFSAR is not incre. sed. The designs for cycles at both units which incorporate the lead test assemblies will meet all applicable design criteria and ensure that all pertinent licensing basis =ep-= criteria are met. The demore.ied adhe-ce of the fuel !
and core designs to applicable standards and acceptance criteria precludes new challenges to components and systems that could increase the probability of occurrence of any previously evaluated accident. Specifically, neither the use ofFCF fuel assemblies (with mid-span mix' grids and
{ small mechanical design differences from the resident fuel) nor the use o loy}s abricated from the advanced zirconium based (Alloy 4 and Alloy 5), will increase the pro bility of' occurrence of an accident previously evaluated in the North Anna Units 1 and 2 UFSAR. The FCF fuel assembly design is mechanically and neutronically very similar to that of the resident fuel. The advanced alloys improve corrosion *
' performance, while the mid-span mixing grids provide additional DNB margin. The use of the four lead test assemblies will not cause the core to operate in excess of pertinent design basis operatin limits. Therefore, the probability of occurrence of an accident previously evaluated in the UFSAR has not increased.
9.2 Consequences of Previously Evaluated Accidents This safety significance assessment documents that the consequences of an accident previously evaluated in the North Anna Units 1 and 2 UFSAR are not increased. The reload core design for cycles which incorporate the lead test assemblies will meet all applicable design criteria and ensure
[ that all pertinent licensing basis acceptance criteria are met. The demonstrated adherence to these
! standards and criteria precludes new challenges to components and systems that could (a) adversely j affect the ability of exieting components and systems to mitigate the consequences of any accident.
] and/or (b) adversely affect the integrity of the fuel rod cladding as a fission product barrier.
i Furthermore, adherence to applicable standards and criteria ensures that these fission product barriers
} maintain design margin to safety limits. Specifically, safety analyses based on the resident fuel 1 design will remain applicable for cores which incorporate the four FCF lead test assemblies.
l Therefore the use of these assemblies will not increase the consequences of an accident previously
- evaluated in the North Anna Units 1 and 2 UFSAR. Similarly, the radiological consequences of j accidents previously evaluated in the North Anna Units 1 and 2 UFSAR do not increase.
j 9.3 Possibility of Accidents Not Previously Evaluated i
} This safety significance nemensment h-nants that the possibility of an accident which is different j from any already in the North Anna Units 1 and 2 UFSAR is not created. The FCF lead test
! assemblies are very similar in design to the resident Waeinghause fuel. Cores irsim iing the lead j
test assemblies will meet all applicable design criteria and ensure that all pertinent licensing basis i acceptance criteria are met. The demonstrated adherence to these standards and criteria precludes j new challenges to components and systems that could introduce a new type of accident.
} Specifically, the design of cores which irsiyorate the FCF fuel assemblies using Virginia Electric and Power Company's standard reload design methodology.will not create the possibility of an j accident of a different type than any previously evaluated in the North Anna Units 1 and 2 UFSAR.
Safety analyses based on the resident fuel design will remain applicable for cores which irwiyer.e
, the four FCF lead test assemblies. No new single failure machanisms have been created, nor will j use of these assemblies cause the core to operate in excess of gid-ut design basis cg..dng limits.
j Therefore, the possibility of an accident of a different type than any previously evaluated in the UFSAR has not been created.
9.4 Probability of Previously Evaluated Malfunction of Equipment Important to Safety This safety significance assessment documents that the probability of a malfunction of equipment
! important to safety previously evaluated in the North Anna Units I and 2 UFSAR is not increased.
The design of cores which incorporate the lead test assemblies will mee%11 applicable design criteria and ensure that all pertinent licensing basis acceptance criteria are met. ,monstrated adherence to applicable standards and acceptance criteria precludes new challenges to components and systems that could increase the probability of any previously evaluated malfunction of equipment important to safety. Specifically, the use of FCF fuel assemblies with mid-span mixing grids, advanceid zirconium-based alloys, and minor mechanical differences from the resident fuel will not increase i
the probability of occurrence of a malfunction of equipment important to safety previously evaluated 3
in the North Anna Units 1 and 2 UFSAR. No new performance requirements are being imposed on i
any system or component such that any design criteria will be exceeded nor will the core be operated
! in excess ofpertinent design basis operating limits. No new modes or limiting single failures have j been created with the lead test assembly design. Safety analyses based on the resident Westinghouse l fuel design will remain applicable for cores incorporating the lead test assemblies. Therefore, the
! probability of occurrence of a malfunction of equipment important to safety previously evaluated i
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in'the UFSAR has not increased.
i i 9.5 Consequences of Previously Evaluated Malfunction of Equipment important to Safety l
This safety significance assessment documents that the consequences of a malfunction of equipment important to safety previously evaluated in the North Anna Units I and 2 UFSAR are not increased.
Cycle designs for cores which incorporate the lead test assemblies will meet all applicable design i criteria and ensure that all pertinent licensing basis acceptance criteria are met. The demonstrated
{
adherence to these emndards and criteria precludes new challenges to components and systems that I could adversely affect the ability of existing +,w= and systems to mitigate the consequences I
of any accident. Furthermore, adherence to applicable standarc's and criteria ensures that these fission product barriers maintain the design margin of safety. Specifically, the use of four FCF Mark-BWI7 lead test assemblies very nimilar in design to the Waeinghause fuel that comprises the remain & of the core will not increase the consequences of a malfunction of equipment important to safety previously identified in the North Anna Units 1 and 2 UFSAR. The use of these assemblies i does not change the performance reqtimsits on any system or component such that any design l criteria will be exceeded and will not cause the core to operate in excess of pertinent design basis l c> lug limits. No new modes or limiting single failures have been created with the Mark-BW17 fuel assembly design. ' 'Iberefore, the consequences of a malfunction of equipment important to safety previously evaluated in the North Anna Units 1 and 2 UFSAR have not increased.
9.6 Possibility ofMalfunction of Equipment Important to Safety Not Previously Evaluated This safety significance assessment documents that the possibility of a malfunction of equipment important to safety different from any already evaluated in the North Anna Units 1 and 2 UFSAR is not created. The design for North Anna cycles which incorporate the four FCF lead test assemblies will meet all applicable design criteria and ensure that all pertinent licensing basis acceptance criteria are met. The demonstrated adherence to these standards and criteria precludes new challenges to components and systems that could introduce a new type of malfunction of r equipment important to safety. Specifically, the use four FCF Mark-BWl7 fuel assemblies that are l very similar in design (both mechanical design and material composition) to the Westinghouse fuel l assemblies that constitute the remainder of the core will not create the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the North Anna Units 1 and 2 UFSAR. No new failure modes have been created for any system, component, or piece of equipment. No new single failure mechanisms have been introduced, nor will the core operate in excess of pertinent design basis operating limits. Therefore, the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the UFSAR has not been created.
9.7 Margin of Safety
- This safety significance assessment documents that the margin of safety as dermed in the Bases to i any North Anna Technical Specification is not reduced. Safety analyses which are based on full i
cores of Westinghouse fuel and which arc ..upported by the applicable North Anna Unit I and North Anna Unit 2 Technical Specifications will remam applicable for North Anna cores incorporating the l
l
l four FCF lead test assemblies. The use of the four Mark-BW17 lead test assemblies will not reduce the margin of safety as defined in the basis for any Technical Specification. The use of these fuel assemblies will take into consideration the normal core operating conditions allowed in the Technical Specifications. For each cycle reload core, these fuel assemblies will be specifically evaluated using Virginia Electric and Power Company's standard reload design methods. This will 1 include consideration of the core physics analysis peaking factors and core average linear heat rate l effects. Derefore, the margin of safety as defined in the Bases to the North Anna Unit I and North ;
l Anna' Unit 2 Technical Spah*!ons has not been reduced. l l
l -
- 10. Conclusions The North Anna Units 1 and 2 Technical Specifications (References 7 and 8) ensure that the plants operate in a manner that provides myiable levels of protection for the health and safety of the l
public. The Technical Specifications are based upon assumptions made in the safety and accident l analyses, including those relelog to the core design. His ensures adequate margin to the regulated i
' ==p-.* criteria for the accident analyses. Since it has been concluded that the North Anna safety analyses which are based on full cores of W@ghause fuel will remain applicable for cores which irs,ry .e four FCF Mark-BW17 lead test assemblies, the conclusions in the North Anna Units 1 and 2 UFSAR (Rderence 27) are valid. Therefore the regulated margin of safety as defined in the Bases of the Technical Specifications is not affected by the use of these lead test assemblies in North Anna Units 1 and 2. '
l, Based on the evaluations and analysis results presented in the foregoing safety significance evaluation, it has been demonstrated that neither the use of.the four FCF Mark-BW17 lead test assemblies at North Anna, nor the use of Virginia Electric and Power Company's standard reload i
design methodology to evaluate cores in which these assemblies are irradiated, results in the '
i acceptable safety limits for any incident being exceeded or in any unreviewed safety questions as defmed in 10 CFR 50.59. l
SUMMARY
i The foregoing analyses and evaluations demonstrate that the conclusions of the accident analyses in the North Anna Units 1 and 2 UFSAR remain apjilicable for the proposed use of the four Mark-BW17 lead test assemblies supplied by FCF. Each pertinent design and safety criterion was evaluated for the impact of both the material and mechanical design differences from the resident Westinghouse fuel, and the evaluation results were found to be acceptable. It has also bee'n determined that the use of the lead test assemblies will not affect the ability of our standard reload design methodology to accurately assess the normal core performance nor affec our ability to conservatively model the core response to accident scenarios.
i 4
Table 2.1 Nominal Compositions (wt%) of Zirconium-Based Alloys for North Anna Fuel Element Zirealoy-4 ZIRLO Alloy 5 Alloy 4
, . Sn 1.45* 1.0 Fe 0.21 0.1 Cr 0.1 -
Nb - 1.0 V - -
Zr .. Balance Balance ., .
l l
This value represents the mid-point of the ASTM specification for tin in Zircaloy-4. Recent Zircaloy-4 cladding fabricated by both Westinghouse and FCF has been manufactured under tighter specifications on the concentration of tin to improve corrosion resistance. These low-tin !
materials still fall within the ASTM specification for Zircaloy-4.
a i l 1 i i Table 2.2 Mechanical Properties of ;
} Alloy 4, Alloy 5 and FCF Zircaloy-4 Tubing 3 i
! 20*C 343*C 385*C Tube Material Alloy 8.2% % 0.2% % 0.2%
l YA U.T.S. Elseg Yi U.T.S. Eloeg Y.S. U.T.S. Elong j Alley 5
] sHaemia criesria(kai) i .
(MPa)
Typicalvalues(ksi)
(MPa) m
{ Alley 4 -
specirmation crheria(ksi)
(MPa)
- . Typicalvalues(ksi)
(MPa) m Ir.4 (recrystallized tabe)
Specification crheria(ksi) i (MPa)
(major diam.) Typical values (ksi)
(MPa) m Zr-4 (stress-relieved tube)
Specification criteria (ksi)
(MPa)
Typical values (ksi)
- f 9 I f
i j Table 2.2 j Alloy 5 and Alloy 4 Irradiation Experience i
Number of Burnup, Corrosion, j Alloy Reactor Cycles GWD/MTU m
{
l 1
= - -
j ' Alloy 5 (SR)
El i 1
1 E2 1 i, E3
! USl i Alloy 5 (MS) E4 L ES
! E6
' Alloy 4 El E2 i .
U j USl j FCF Zr-4 (composite) 3 (Iow Sn) l
= = =>
, Cunently in 4th cycle with eWM burnup of 50 GWD/MTU in 1996 i " Currently in 5th cycle with ed burnup of 55 GWD/MTU in 1997 1,
1 1 i
i i
l
\
l I
t i
l
l Table 3.1 Comparison of BOL Nominal Grid Elevations for Lead Test Assemblies and North Anna Resident Fuel Types
. Mark- .E NAIF- NAIF-
. BW17 LOPAR Zircaloy ZIRLO 159.915 159.975 159.775 Tnn Nnnte 153.60 , 153.60 153.40 Tnn Fnd .
l 133.01 133.10 133.10 inter Grid ei N/A N/A N/A MRhin't -
112.46 112.55 112.55 Inter Grid 5 N/A N/A N/A MRMG?
91.91 92.00 92.00 .
inter Grid 4 l
N/A N/A N/A MRMG1 -
71.36 71.45 71.45 Inter Grid 1 50.81 50.90 50.90 inter Grid ?
30.26 30.35 29.70 inter Grid 1 5.835 5.835 6.535 Atm Fnd ,
N/A N/A 3.093 Prnt Grid 2.383 2.383 2.383
- Rim Nnnie 00 00 00 j =
i 1
- Distances are from bottom of bottom nozzle to top of grid assembly inner strap, in inches.
4 I
l l
i 1
3 Table 3.2 l Comparison ofMark-BW17 and Resident North Anna Fuel Designs j
l FCF ELOPAR H NAIF E NAIF i
Mark-BWI7 Fuel Fuel Fuel w/
! Parameter 'Fael Assembly 4
Assembly Assembly ZIRLO
~
} Fuel Assembly Lea =th,in. 159.915 159.975 159.775
?
j Assembly Envelope,in.
j TopNozzle ,
8.405 8.400 8.400 End Grids 8.426 8.426 8.426 j Intermediate Orids 8.417 8.418 8.418 j MSMOs - - -
l, Bottom Nozzle 8.426 8.425 8.426 l FuelRods -
j ,
Number ofFuel Rods /Assy . ; 264 264 264 Active Fuel Leagth, in. 144.0 144.0 144.0 Fuel Rod Pitch,in. 0.496 0.496 0.496 Fuel Clad Material Zircaloy-4 Zircaloy-4 ZIRLO l
! Fuel Rod Clad O.D., in. 0374 0374 0374 Fuel Rod Clad 'Ihickness,in. 0.0225 0.0225 0.0225 Fuel Pellet Diameter,in. 03225 03225 03225 Fuel Pellet Density, % TD 95 95 95 l Fuel Pellet Length,in. 0387 0387 0387 i Effective Dish, Percent 1.207 i 1.207 1.207 j' Guide Thimbles j Number ofGTs/Assy 24 24 24 ..
Guide Thimble Material Zircaloy-4 Zircaloy-4 ZIRLO
- GT Length (incl. End Plug) 153.10 153.215 153.015 Upper Portion Length to mid-transition,in. 129.160 129.275 129.075 Outer Diameter, in. 0.482 0.474 0.474 Inside Diameter, in. 0.450 0.442 0.442
l
\
j Table 3.2 i
Comparison ofMark-BW17 and l Resident North Anna Fu'el Designs 1
[ FCF ELOPAR
- E NAIF E NAIF Mark-BWI7 Fuel Fuel Fuel w/ i Parameter Fuel Assembly Assembly Assembly ZIRLO i Lower Portion - ==
- Length to mid-transition,in. 23.940 23.940 23.940
! Outer Diameter,in. 0.430 0.430 0.430
)
Inside Diameter,in. 0397 0397 0397 4
Instrumentation Tabe
! Number / Assembly 1 1
- , 1
! Instrumentation Tube Material Zin:aloy-4 Zircaloy-4 ZIRLO i
Instnanentation Tube O.D, in. 0.482 0.474 0.474 i .
Instrument Tube I.D.,in. 0.450 0,442
- 0.442 j ,
Spacer Grid j -
Axial Positioning Table 5.1 Table 5.1 Table 5.1 j .
Top and Bottom End Grids I
Grid Material inconel 718 Inconel718 Inconel 718 l Grid Sleeve Material 304 SS 304 SS 304 SS
- Strip Width (Height),in. 1322 1.522 i
-. 1.522 ;
j Intermediate Grids l Grid Material t Inconel 718 Zircaloy-4 ZIRLO l Grid Sleeve Material 304 SS Zircaloy-4
' , ZIRLO i
i Strap Width (Height),in. 1322 1.500 i 1.500 m
{ Mid Span Mixing Grids
! Grid Material
- N/A N/A N/A j Grid Sleeve Material N/A N/A N/A Strap Width (Height), in. N/A N/A N/A 1
1 1
I i
l 1
1
1 1
i i
Table 3.2 l Comparison of Mark-BW17 and Resident North Anna Fuel Designs 1
FCF. E LOPAR E NAIF E NAIF
- Mark-BW17 Fuel Fuel Fuel w/
I Parameter Fuel Assembly i
- . Assembly Assembly ZIRLO
} Protective Grid i
- Grid Material N/A N/A Inconel 718 l
Grid Sleeve Material N/A N/A~ 304 SS
) Strap Width (Height),in. N/A N/A 0.690 1
1 Top Nozr.le Holddown Spring l Type . . . 3 Leaf 3 Leaf 3 Leaf i Spring Material , ., Inconel718 Inconel 718 Inconel718 i .
i i
i I
k I .
i d
I l
i i . l t
I
\
I 1-f t
1 i
1 4
1 i
Table 4.1
_ Parameter Ranges & Uncertainties used for SDL Determination
}
Parameter Ranges for SDL Determination
?
Variable Range i
Core Power - 70% to 130% '
il Core Flow 65% to 125%
Core Pressure 1800 to 2600 psia 1 .
CoreInlet Subcooling 40'F to 110*F i
RadialPowerFactor(Fau) 1.1 to 1.9 2
Axial Peaking Factor 1.05 to 2.35 1 Axial Peak Location 0.14 to 0.90 Parameter Uncertainties for SDL Determination
- i l
! Variab'le Uncertainty Distribution Core Power *2.2% at 2a Normal
, Core Flow *2.0% at'2a
- Uniform j Core Pir *36.0 psi at 2e Uniform CoreInlet Temp. *3.7'F at 2e Uniform i -
Core Bypass Flow *l.0% Uniform j Measured FN AH *4.0% at 2o t -
Normal * .
Hot channel Factor 2.0% Normal
~
, DNB Correlation Unc. ~
- LYNXT Code Unc.
RSM to LYNXT Fit i Bundle Spacing
, Axial Peaking Factor i Axial Peak Location
. =
- i i
}
4 l
1 l
. - . . . - . - . _ - - . . . . - . - - . . - - _ . - . . - . ~ . - . . . . - - . - . . - - . . _ . . - . . -
Table 7.1 Plant Parameters and Operating Conditions Reactor Power 102% of 2893 MWT Nominal Pressurizer Operating Pressure 2250 psia
. . System Flow 288,000 gpm Hot Leg Temperature 620*F
! Cold Leg Temperature 552*F Steam Generator Operating Fiw 850 psia Fuel Pin Outside Diameter 0374 inch l Average Linear Power Generadon Rate 5.9 kW/R Highest Allowable Total Peaking (Fo) 2.19 i
k e
l .
1.
I i
i
i i
- Table 7.2 Discharge Coefficient Study Comparison l
Break Discharge Coefficient 1.0 0.8 0.6 0.4 l
l Peak Cladding Temperature Data.
i i Peak Cladding Temr..im, *F 1915 1944 1974 1910 1
l PCT Location, feet 63 6.9 6.9 8.6 l
j Rupture Node Data l
l Location, feet 6.9 63 63 6.9 I
Rupture Time, seconds 59 55 54 78 i
Peak Temperature at Rupture Location, *F 1756 1770 1848 1668 Oxidation Data i Local Maximum Oxidation, % 3.4 3.5 4.0 2.9 Location ofPeak Oxidation, feet 6.9 63 63 . 8.6 l
l t
i I
a i
- 1 i
i 1
ii i
1
l Table 7.3 Sequence ofEvents for DECLPDB Cd=0.6 (Time in seconds)
Leak Initiation 0.00 SafetyIgjection System Trip 3.2 Ace =d=*ar Injection Begins 12.0 End ofBlowdown 19.0 Bottom ofCore Recovery 32.5 Accumulator Empty 50.3 PCTTurnaround 96
~
. b Quench 679 l
1 l
t 1
i l
l
~
1 I
I I
I l
h 4
l
.l '
i
- Table 7.4 l Results for Variant Axially Peaked LBLOCA Cases l
1
! Axial Peak Position, feet 4.6 6.7 10.1 i
i Peak Cladding Temperature Data l
} Peak Ciuding Temperature, 'F 1935 1966 1928 1
Location of Peak Cladding T+. rare, feet 6.1 7.0 8.7 l
{ Rapture Nede Data
(
Peak Temperature, Ruptured Node, 'F 1846 1765 1841 a
j Location ofRupture, feet 5.1 6.1 9.6
~
Rupture Time, h=d=
{
1 52 56 62 i Oxidation. Data.
I '
j Maximumlocal Oxidation, % 3.9 3.5 4.7 i
j Location ofPeak Oxidation, feet 5.1 7.9 9.6 i , l F
1 l
l I
Table 7.5 j
i Comparison of Virginia Electric and Power Company and FT1 LOCA Calculation Results !
i l North Anna NAIF FT1 Mark-BW17 l ' Results Results i
Unit 1 Unit 2 Unit 1 Unit 2' 4
l Peak Cladding Temperature,'F 1975 2013 1966 2004 1 Maximum Local Oxidation, % 4.5 5.7 3.5 <5.7 jl
- The upfloE-to-downflow APCT from the NAIF calculations, as described in Section 7.1.1, has
} been used to determine the cladding temperature value. For the local oxidation it is merely l recognized that the value willh less than the corresponding NAIF value.
9 i
i _ , _ _ _ _ - . -.
i 1
i ,
j j a 1 i Figure 1.1 j Mark-BWl7 Lead Test Assembly Design 1 1
J l l
4 j -
- 4-trAr woLooO w s, l.
, ~ ASSEMBLY (WCONEL)
I s
EX5m) /* ***"
ALLov s GutoE THaeLEs FLOATWO # NCON )
)
temens Or- ) ,
j MRi& Ele i
1 temens l 1I111115 MIM. .
HilIIIB IIIII~
- : : l C WNED ......._......,"
(2 4)g , llIllilm iiiii-i IllllllR'IIIII-i 11lll11E Ell 1
~ __
i , IIIIIlle 11:11 '
i 1111111E MlM i
ALLOY 4 ANO .
ALA.Of 5 Fuel. / IIIIIIIE?'
- 1 i RCD CLADDING FLOATING VANELESS a
lilllllE NIN f(lNTERMEDIATE ZlRCALOY-4) GRID
'j '
.f...._-....../
l llllllll NltF i
WTE8* "*
1 -
i SEATED FUEL ROD 5
/ FINE WESH DEBRIS
.fFILTER l e s BOTTOW NOZZLE r
1 i
j j
i
}'
i i
4 Figure 1.2 Spacer Grid Restraint System i
1
- (IntermaAia'a Grid Shown) f i
a 1
i g.
- msTnuusur Tusa .
j pannutes nastamane euros tsausta j l TvP a LOCATNWis ompts
. u. .. . , .
.a arranuso ATs spacan anno
[.
g . . . 1. .
9 o ,
i " g j
.A ,
t e . E ,A "
- < w A . , . < *
\ W gg3gp W 9_M W yW W W ,MvW A d_ 3 .- g
, 1 WA R- LR E .R L s L ax.s g
\
l Y W n e m y c q K K K F &..
4 1
i
.i t i l j u =u. m. m u. m.m. i i i i i i 1 i l l
1 RESTRAINING GUIDE THIMBLE LOCATIONS DENOTED BY @
X N X x x X X % M X ~
i X K % k X 4'
4 X X 1
X N X 1
4 k
i
.I MsMa arsraAzwzwo LocATzows nanoTzo av E
}
j f
4 o
i
)
4 l
i
i l Figure 1.3 Comparison ofMixing Vanes for -
Intermediate Spacer Grids and i Mid-Span Mixing Grids i
i i.
l i
. m s
=p -
t I
i '
t l
i i
i i
i 1
! w l YANE CCNPARISCN OVERIAT Mark-BW17 Intermediate Grid Vane Shown solid l
i MSMG vano shown in Ridden Line .
t i
)
8
'1 I
i
)
i
}
l.
4 i
4 4
l i
l Figure 1.4 i Mid-Span Mixing Grid Design j i
i -
i l
A A A ver /'N 1
i
>!+!+!+!+!+!+!+!+!+y !+!+!+!+l+ Ag /N+!f Ar is es es es is is es es is e s fx f s is ex es f s f ;
l --
l .2.. um fv u.
~
. f ,9p :>
i i
TYPICAL NWt STRAP
-n u ._
! % %- A L.L M h,@Ak,8 j s' ,8,'r', s ,8, <
,,s.s
,, e <= s,s l
1 A- _L Nio. m es l A- L A- A- J-j
- r 7- r y-- v- r , -
v-l l
l l
1 i
l l'
{
I 1
.I 5
l i
i l Figure 1.5 l Bottom Nozzle Filter Plate Geometry i
l l -
i l
~
l ,
,e- ff -
y
< N
. M & m
,{
- m m 3 Z m
Hj SECTION X-X i
1 r -'
W v--
d 4 jv r < >
i
\.. 7, .,
l --l
,,r . .. ' /g i
- r. l
. . qv x' . . xv f,x - - f,) , -
) ~
.) - .
g X,
- ? :
'\ h-~' OM j
ssCrion z-z
. .. ~ l
- - J i1 - __
J FINE MESH FILTER PLATE i
i
)
i i
4 4
1 4
i
{ Figure 1.6 j Alloy 4 Locations in Lead Test Assemblies
- j. . . (2 assemblies only) 1 n
1 j 4m
! M f
1 h
?
}
1 i
i h
i, l
i 1 .
1 i
- 1 k
f i
h i
t 1
f .
l 1
4 3
I e
- 2 musme
?
- g.
s i
l .
1 4
4 i
l f
i
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- 4
\
- I
?
1 -
l 1 l 3 Figure 1.7
- 1 Mark BW-17 Fuel Rod Assembly i
- h a
i l j l i I t .
{ .
t
. 1 1
4 I
' .m_ _ __
g ~ = .
b g e==
j f" 1 1
i 8'
. . . . . . _b .4 it
\ if .
RI . N ig pg E
GM p . * ,
i wm een se su
- nm
~H~
, r7 -
i i
e L J !
i 1 .
- e i
i
- , I I
4 5
1 4
4 k
1 1 *?
d 3
i) i i
i i t f
i l
1 1
Figure 7.1 RSG LOCA EM Computer Code Interface Diagram
. l W ACJipsene mut l i
QweAusmedes i-I RELAP5/ MOO 2-B&W 0 < TIME < EOS l
caeakmedsnerw coreshows annem uses and snow nenhan Ace usesand enerw andgmesoure aG Mese endenery e
REFLOD3B i Pumped accPione EOS < Time < EOE !
l i
Pleading these I core Parametere At EDS BEACH -
EOS < Tune < EOE l
MetalWaterReeselon Hot and Average PinThermalResponse Peak cieddheTanpmeime 1
i t
. . . . - . . . . . _ . - - .-. - . ~ - - - . . _ - . . . ~ . - . . . - - . - . - . . . .-_. . . . - - . . - - - . . ..
I e-b -
i 1
Figure 7.2
! RELAPS/ MOD 2 LBLOCA Loop Noding Arrangement '
5 m 1 t I se 1 1
' 1 l l I ~
Y 3 I m m t i 3
i I ses I ses g 9 i
1- E.,
- ess i ese 1 j m 1 l Im 3 m I E sus f e 1 f f!s, E I
la d tas 9 , , t 3 ; as @ $1 @ ses aus 53' 35 lE E: E l
'E E l E lE E l lE ms >
see E El '
E -
E ses
- E s im as = - m 5
E Et m E ens m ;g aus -
gp E m e t. ,
E as d sus a 88 sus W
= Es i
. 1 $ sus ses l~ml W
1 .
.)
m
~
.)
T m
wm eeuuse
- =
I Sm I 4""*""" T' I ses l NTACTLDOP N N
- BROBEN LOOP P
O e
O
..s - =-,,r - ,,-,.,s ,e --,1
i f
i i
Figure 7.3 l RELAPS/ MOD 2 LBLOCA Reactor Vessel Noding Arrangement i , *
- e i -
=
! I e
i I =
, l s
i t
= l I ,
l.l I = 1 1=l j =
- Y_ -
=
i
_Aase j pescas ._
j 8cEE
- d - ,
pseen
= m acte
] m l '
Je lmei , t.w
. ,, 1.=
I
. , d d
Sid - 444 e
m N
- P
- 3. - 4 i
gag 'Y -# Y f r.-- -
> a 4
i -- y .-- .
, hME B
- g l
ass -- ass 8 v Y == IY !
l sY -
m- I
-. .=. #
l lf,"== Y
' lI
- .g g [ ,' V# ar 43.
I
- t
- tm bB g E "" O
- sei -- ess E I 1 .a.
) " m- - m-Y e. '
T PY e- Y g" e. Ya"
{ , ,,,,,j,, , t f.. t .
f l m
! m l t
su i t
- b. _
sw j m 6
Figure 7.4 RELAPS/ BEACH Core Noding
, with Mid-Span Mixing Grids Modeled 1
\
HSNumber Bowesen
- 12.0 M m g its 344 m 444 ao g to.se*
448 so m 1025 442 g " 8.30 341 se s.70s 441 g e.42 44o am das We g " 9.13s m 4as ri , , ,a.a,s,'
ms- , , ,,7 das as " 8.as das ans a., .". 7.77.se,ss .a.
'7" an 7 Ass' 4ss i,
mi ,g 7.14
- 4ss 431 ao ,g s. ass --
dao ans u 8.s7 as ,g 8. ass: 4as aar 8.00 4as I 3, 4 ar7
,," 5.71s ass s.43 e 42s
@w g
- 425 <
4.as s 424 "4.as 323 7 423 g - 3.72 *
- 422 3.1s 321 8 421 2.se ,
320 d !
420 2.01 li 319 8 419 1#
318 1 418 -
o.s7 317 i
..o,o.
417
- GridLocanon X MSMGLoca6on j
i
l l
t Figure 7.5 REFLOD3B Noding Diagram l .
l 1
l l
i 815AM N1CR1 STEAM NTOR2 sumamme suzzeer L
I n.
e r
e
. A.
a m
! @. .. "" .' 4 .
""* ,4 .
e 4 . m .
.4 p g REACTOR
%.. vesset .. J.
- a . er =
_ _ _ q 3<
l- ,,
e: ,e l n '
e M
! ta NTACTLOOP topor # BROMEN LOOP l 14 Acmq 88 CINE 9
- / m "
l Cone A 13 EVPnN IF BOTTOM CF . EN l AcfME conu l
l l
l i
i i
i l
Figure 7.6 Schematic ofEDGM'Tm FuiI;q s
S f unus I
l I
e 9
e l
f I
' w ame e
t o
1 i
Figure 7.7 Comparison of NUREG-0630 Correlations and Data to Alloy 5 Testing: Rupture Temperature numan emme l
l-l l
l l
l l
l M
l-NUREG 0830 model a NUREG Date,9 to 11 CAs e EDGARDets.845l O
i i
)
l 1
i Figure 7.8 Comparison ofNUREG-0630 Correlations and Data ;
to Alloy 5 Testing: Rupture Strain !
a A
- l
\ 1
, 1 l
I t
i I
I i
i 1
l l
l l
r 1
M examma >
' t ,.
I 1
L.
4
=v
. . -. .- = . . _._. - _ .__~ _
2 Figure 7.9 Pressure
. (Double Ended Cold Leg Pump Discharge Break Co = 0.6) l 2400 2000- .
1600-1200<
~
- E 800-400<
0 4 8 12 16 20 -
24 28 32 TIME SECONDS i
d i
Figure 7.10
.C!sdding Temperannes l (Double Ended Cold Leg Pump Discharge Break Cd = 0.6)
- e e
1
- i 2400 1
] 2000< <
)
4
- 1g00 '.' -
l .
t . .
2 -
l 9
800-9 h
n.... .. . .. , , _ ,
0 0 100 200 300 400 500 600 700 800 'l TWE, SECONDS e
Figure 7.11 Reflood Rate (Double Ended Cold Leg Pump Discharge Break dC = 0.6) '
j I
I
- 8 1 7 <
i 6<
5 1
4-
\
, d 3 j <
u g, !
4 1-I l
l 0
0 100 200 300 400 500
- 600 700 800 TIME, SECONDS l
l l
l l
l-I l
.- . . - - - . _ - .... . = - . . . .- . - . . . . .. _.
, Figure 7.12
{ Vapor Temperature
, (Double Ended Cold Leg Pump Discharge Break C4 = 0.6) i e
- 2400 i <
l 2000- - - -
,1800< f
\s i F-800< p i s
1 0
0 100 200 300 400 500 600 700 800 TIME. SECONDS 4 .
l 1
l
i i
j Figure 7.13 i Heat Transfer Coefficient
] (Double Ended Cold Leg Pump Discharge Break Cd = 0.6) j .. .
4 e
1 . ,
s 4 .
)
- i. 10';
4 1
1 i
10') !
1 . . "
1 w j 10 g .
]
-f .
l l 3 ' l gr I 4
10 % <
i j .
. 10* '
j 0 100 200 300 400 500 600 700 800 TWE, SECONDS 1
4
)
i I
i i -
i i
i i
9 1
4 d
i
. . - . __. . _. __ . . . . . ~ . _ . - _ . . _ _ - _ . - _ _ . . . _ _ _ _ _ _ _ . _ _ _ . . . _ _ _ . _ _ . . _ . . . ._
i I
- Figure 7.14 Cladding Temperature Comparisons for Cd= 1.0,0.8,0.6 and 0.4 1
i ,
i 2400 LEGEND 4
cd-0.s l 2000- - -
cd- 1.0
- cd-0.s j
4 4
wg -
- cd -0.4 4
-. g
_N-
- 1800- 3 i
4 mr T
\ .
E s gw%
! 1200' \ %~
1 m I .
! i i
400 l .
(_ m_. ., n.. .. .. ,
l 0
} O 100 200 300 400 500 600 700 800
?
I
< M l
l i
i l~
4 1
l l
I l
{ l a !
i I t
e i
i
Figure 7.15
- Normabd Core Axial Power Limit, &
1 .
I I l i .
1 e' i 2
12 Y
4 i
1 i -
. \
i E 0.8 -
o i i 1 0.8 - l i )
I l
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. 0.4 -
/
7 4
02 -
(
l 0 ,
0 2 4 6 8 10 12 CORE HSGHT FT i
a i
4
]
t f
i l
i
4 a
Figure 7.16
s S
4 e
0 i
J 2
4 4Mt Peak
---. 8,7.Pt Peak 1A -
10.1 Ft Peak
< j 1
1.s -
l i *
~
.... . l 1
15 -
,#<# 's*
i s' %
4
,/ Ns ;
i
. 1,2 -
e $
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4 %
f %
' 0 1 - ,' %
N%
l l %
, u -
,e 4 %
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, e*
, . N s '%
! 0.4 -
t .
0.2 -
1 9 9 I t t I t I t I t !
1 O 0 2 4 6 4 10 12 l
consHeerr. rr i
_- . . - . _ _ - . _ . _ . ._. - . . . . . . - - = - - - .
4 e
Figure 7.17
' Cladding Temperatures for the Worst Case LOCA with the Core Peak Power at 6.7 Feet 1
4 1
2400 l
. 2000<
"e '
, 1600<
I 1200<
i
. l 800<
i 400<
'f 0-0 100 200 300 400 500 '
a m g
. TIME, SECONDS 1
i 1
1 J
f fi a
- Figure 7.18 Vapor Temperature at the Hot Spot for the Worst Case LOCA
- with the Core Peak Power at 6.7 Feet i
I i .
1 2400 1
2000< .
- l i ..
~
- , 1606-1200< f 800<
400<
4 O
i 0 100 200 300 400 500 600 700 800 TIME, SECONDS tl l
b
- Figure 7.19
- Hot Spot Heat Transfer Coefficient for the Worst Case LOCA l
4
' with the Core Peak Power at 6.7 Feet t i
)
4 I
I i i
10'.
t l
- 10' - l l
l 1
1' 4 ,
n 10*:
E -
a I
a tal.fA ..attL.ataiu _ tr. .L . imhl <
d 10*: _
Z 7C*. '
J 70 O 100 200 300 400 500
- 600 700 800 TIME, SECONDS 4
.i s
6 j
)
i 4
4 1
t l
l i Figure 7.20 l Peak Cladding Temperatures for
, Axial Peaking at the 4.6,6.7, and 10.1 Foot Elevations 2@ LEGEND 6.7-Ft Peak 2000
- 4.6-Ft Peak 10.1-Ft Peak jy
. \.
1600< % ~
is. .._ g
.N 800<
\'
\. . \
)
goo .-
l
.- t 1 0
0 100 200 300 400 500 600 700 800 TIME, SECONDS e
+
l
l REFERENCES 1
1 l ~1. Letter from W. L. Stewart (Virginia Elecr *c and Power Company) to Leon B. Engle (NRC).
l
" North Anna Power Station Units 1 and 2 - Proposed Technical Specifications Change -
' North Anna Fuel Assembly Design Change," Serial Number 89-795. January 15.1990.
- 2. Letter from W. L. Stewart (Virginia Electric and Power Company) to U. S. Nuclear Regulatory Commimmion, " North Anna Power Station Units 1 and 2 - Proposed Technical Specifications Changes - Implementation of ZlRLO Cl=dding;" Serial Number 93-614 October 4,1993.
- 3. " Reload Nuclear Design Methodology," VEP-FRD-42, Rev.1-A, September 1986.
- 4. Bordelon, F. M., et al., " Westinghouse Reload Safety Evaluation Methodology,"
WCAP-9272-P-A (Proprietary) and WCAP-9273-A (Non-Proprietary), July 1985.
- 5. " Mark-BW M~hanical Design Report," BAW-10172P, July 1988.
- 6. ASME Code,Section III, Division I, Subsection NG, Boiler and Pressure Vessel Code,1980.
- 7. Technical Specifications - North Anna Power Station, Unit Number 1, Docket 50-338,
. through Amendment No. 200, April 1,1996. .
- 8. Technical Specifications - North Anna Power Station, Unit Number 2, Docket 50-339, through Amendment No. I81, April 1,1996.
- 9. Standard Review Plan, Section 4.2, NUREG-0800, Revision 2, U. S. Nuclear Regulatory Commission, July 1981.
- 10. North Anna Power Station Updated Final Safety Analysis Report, Section SA, " Analysis and Effects of LOCA on Reactor Coolant System Supports," Amendment Number 43, December 1,1975. -
- 11. " Mark-BW Advanced Claddings Fuel Rod Evaluation," BAW-2133P, March 1991.
- 12. " TACO 3 - Fuel Pin Analysis Computer Code," BAW-10162P-A, October 1989.
l 13. " Program to Determine in-Reactor Performance of BWFC Fuel Cladding Creep Collapse,"
BAW-10084P-A, July 1995.
I
- 14. " Fuel Rod Gas Pressure Criterion (FRGPC)," BAW-10183P-A, July 1995.
, 15. "LYNXT: Core Transient 'Ihermal-Hydraulic Program," BAW-10156-A, Revision 1, August i 1993.
i
i l 17. " Statistical Core Design for Mixing Vane Cores." BAW-10170P-A. December 1988.
I e
- 18. "CHF Testing and Analysis of the Mark-BW Fuel Assembly Design." BAW-10189P-A.
! January 1996.
l
- 19. "The BWU Critical Heat Flux Correlations," BAW-10199P, December 1994.
l l 20. R. C. Anderson, " Statistical DNBR Evaluation Methodology," VEP-NE-2-A, June 1987.
I
{ ,
- 21. Letter from W. L. Stewart (Virginia Electric and Power Company) to U. S. Nuclear i Regulatory Ca==i=*ian, " North Anna Power Station Units 1 and 2, Proposed Technical I- Specficiations Change," Serial Number 87-231, June 17,1987.
)
i 22. " Fuel Rod. Bowing in NM & Wilcox Fuel Designs - Revision 1," B AW-10147PA-RI, 1
May 1983.
- 23. "Evtandad Burnup Evaluation," BAW-10186P, November 1992.
- 24. J. G. Miller, "VEPCO Nuclear Design Reliability Factors," VEP-FRD-45A, October 1982.
)
25.
K. L. Basehore, et al., " Virginia Power Relaxed Power Distribution Control Methodology and Associated FQ Surveillance Technical Specifications," VEP-NE-1-A, March 1986.
f J
- 26. "VEPCO Evaluation of the Control Rod Ejection Transient," VEP-NFE-2-A, December j 1984.
j 27. Updated Final Safety Analysis Report - North Anna Power Station, Units 1 and 2, Docket
! Nos. 50-338 and 50-339.
l 28.
"RSG LOCA-B&W LOCA Evaluation Model for Recirculating Steam Generator Plants,"
- BAW-10168 Revision 3, B&W Nuclear Technologies, October 1993.
{ 29.
D. A. Powers and R. O. Meyer, " Cladding Swelling Models for LOCA Analyses," NUREG- -
0630, April 1980.
- 30. "RELAPS/ MOD 2-B&W, An Advanced Program for Light Water Reactor LOCA and Non-l LOCA Transient Analysis," BAW-10164 Revision 3, B&W Nuclear Technologies, October, i 1992.
[;
- 31. " BEACH, Best Estimate Analysis Core Heat Transfer," BAW-10166 Revision 4, B&W Nuclear Technologies, October,1992.
i i
i 1
i 32.' H. M. Chung and T. F. Kassner, " Embrittlement Criteria for Zircaloy Fuel Cladding i Applicable _- to Accident Situations in Light-Water Reactors: Summary Report."
! NUREG/CR-1344, ANL-79-48, Argonne Nation Laboratory, Argonne IL, January 1980.
i 1
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