ML20205S433

Application for Amend to License NPF-30,authorizing Use of Westinghouse 17x17 Vantage 5 (V-5) Fuel Assemblies for Reload During Cycle 3.Subsequent Cycles Will Use Increasing Amount of V-5 Fuel Assemblies.Safety Evaluation Encl

ML20205S433
Person / Time
Site:	Callaway
Issue date:	03/31/1987
From:	Schnell D UNION ELECTRIC CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20205S435	List: ML20205S433 ML20205S505
References
ULNRC-1470, NUDOCS 8704070030
Download: ML20205S433 (47)

Text

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1901 Gratiot Street. St. Louis Donald F. Schnell Vce President U.S. Nuclear Regulatory Commission ATTN:

Document Control Desk Washington, D.C.

20555 Gentlemen:

ULNRC-1470 DOCKET NUMBER 50-483 CALLAWAY PLANT APPLICATION FOR RELOAD LICENSE AMENDMENT USING WESTINGHOUSE VANTAGE 5 FUEL ASSEMBLIES This letter and its attachments contain information in support of the reload license application for Callaway Plant using Westinghouse 17x17 VANTAGE 5 (V-5) fuel assemblies.

The first cycle utilizing V-5 fuel is Cycle 3.

Subsequent cycles will utilize increasing numbers of V-5 fuel assemblies until a complete V-5 core is obtained.

Cycle 3 criticality is currently scheduled for October 1987.

Cycle 3 will consist of 28 17x17 Westinghouse V-5 fuel assemblies in Region 5A (3.6 w/o U-235), 32 V-5 assemblies in Region 5B (3.8 w/o U-235), and 36 V-5 assemblies in Region 5C (4.2 w/o U-235), in addition to the 13 Low Parasitic (LOPAR) and 84 Optimized Fuel Assemblies (OFAs) remaining in the core.

A number of the V-5 assemblies will employ Integral Fuel Burnable Absorbers (IFBAs).

The Westinghouse V-5 design features, including the IFBAs, Intermediate Flow Mixers, Reconstitutable Top Nozzles, and Extended Burnup Capability, and methodology have been generically approved by the NRC via the review of Westinghouse Topical Report WCAP-10444-P-A,

" Reference Core Report VANTAGE 5 Fuel Assembly".

The accident analyses have been performed at a core thermal power of 3565 MWt, 15% steam generator tube plugging, full power enthalpy rise hot channel factor (F delta-H) of 1.65 for V-5 assemblies, maximum LOCA heat flux hot channel factor (F sub-0) of 2.5, and a positive moderator temperature coefficient of +5pcm/ F from 0-70%

power and decreasing linearly to Opcm/ F at 100% power.

This submittal does not request any licensing changes for the above parameters; however, we do expect these parameters to change plant technical specifications for future cycles.

Five attachments to this letter are included in support of the reload license application.

A safety evaluation summarizing the results of the Westinghouse analyses, which justify the transition from a mixed OFA/LOPAR core to an all VANTAGE 5 core, is presented in to this letter.

Table 7.1 of Attachment 1 provides a N

8704070030 870331 I

PDR ADOCK 05000483 gO\\ ) g, f0 s

P PDR Maihng Address: P.O. Bnx 149. St. Lows, MO 63166 description of the Technical Specification changes necessary for Cycle 3 operation.

These proposed Technical Specification changes, indicated by a vertical bar in the right-hand margin of the page, are contained in Attachment 2 to this letter.

These proposed changes have been approved by the Callaway Onsite Review Committee and Nuclear Safety Review Board. to this letter contains an environmental evaluation of the extended fuel burnup capability of the V-5 fuel.

to this letter contains our evaluation, pursuant to the standards in 10CFR50.92, of significant hazards considerations in this request for a license amendment.

This evaluation is submitted as required by 10CFR50.91(a)(1). to this letter contains four appendices.

Appendices A and B present the non-LOCA and LOCA accident analyses, respectively.

The approved Westinghouse Improved Thermal Design Procedure and the WRB-1 and WRB-2 DNB correlations have been used in the analyses of LOPAR, OFA, and V-5 fuel assemblies for most DNB-limited accidents.

The BASH and NOTRUMP Codes have been used in the analyses of the large break and small break LOCAs, respectively.

These codes have received NRC approval.

Appendix C of Attachment 5 contains an analysis of the spent fuel pool cooling and HVAC systems' capability to accommodate spent V-5 assemblies.

Appendix D of Attachment 5 presents the radiological consequences for all analyzed accidents.

Approval of Cycle 3 operation with the Westinghouse 17x17 VANTAGE 5 fuel assemblies is requested by October 1, 1987.

In accordance with 10CFR170.12, a check is enclosed for the $150.00 application fee.

Very truly yours, s

Donald F.

Schnell GGY/ tar l

E STATE OF MISSOURI )

)

S'S CITY OF ST. LOUIS )

i Donald F.

Schnell, of lawful age, being first duly sworn upon oath says that he is Vice President-Nuclear and an officer of Union Electric Company; that he has read the foregoing document and i

knows the content thereof; that he has. executed the same for and on behalf of said company with full power and authority to do so; and that the facts therein stated are true and correct to the best of his knowledge, information and belief.

I f

By

~ hnell Donald F.

Sc i

Vice President Nuclear SUBSCRIBED and sworn to before me this.3/# day of ht-c/v, 1987 i

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.PFAFI BARBARA NOTARY PUBUC, STATE OF MISSOURI MY COMMIS$10N DPIRES AFRil 22, 1983 I

ST LOUIS COUNTY i

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Gerald Charnoff, Esq.

Shaw, Pittman,-Potts & Trowbridge 2300 N.

Street, N.W.

Washington, D.C.

20037 J. O. Cermak CFA Inc.

4 Professional Drive (Suite 110)

Gaithersburg, MD 20879 W.

L.

Forney Division of Projects and Resident Programs, Chief, Section lA U.S.

Nuclear Regulatory Commission Region III 799 Roosevelt Road Glen Ellyn, Illinois 60137 Bruce Little Callaway Resident Of fice U.S.

Nuclear Regulatory Commission RRil Steedman, Missouri 65077 Tom Alexion (2)

Of fice of Nuclear Reactor Regulation U.S.

Nuclear Regulatory Commission Mail Stop 316 7920 Nor folk Avenue Bethesda, MD 20014 Ron Kucera, Deputy Director Department of Natural Resources P.O. Box 176 Jefferson City, MO 65102 Manager, Electric Department Missouri Public Service Commission P.O.

Box 360 Jefferson City, MO 65102

ULNRC-1470 TABLE OF CONTENTS ATTACHMENT 1 Safety Evaluation ATTACHMENT 2 Technical Specification Changes ATTACHMENT 3 Environmental Evaluation ATTACHMENT 4 Significant Hazards Evaluation i

ATTACHMENT 5 Appendices A Non-LOCA Accident Analyses B

LOCA Accident Analyses i

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Spent Fuel Pool Analyses D

Radiological Consequences I

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ULNRC-1470 1-3 ATTACHMENT 1 i

i 1-SAFETY EVALUATION MR WE t

CALLAWAY PLANT TRANSITION TO WESTINGHOUSE 17 x 17 VANTAGE 5 FUEL a

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TABLE OF CONTENTS Section Pm l

1.0 INTRODUCTION

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2.0

SUMMARY

AND CONCLUSIONS 4

3.0 MECHANICAL EVALUATION 6

4.0 NUCLEAR EVALUATION 14 5.0 THERNAL AND HYDRAULIC EVALUATION 16 6.0 ACCIDENT EVALUATION 21 i

7.0

SUMMARY

OF TECHNICAL SPECIFICATION CHANGES 34 8

8.0 REFERENCES

38 1

LIST OF TABLES Table No.

Titie Pm 3.1 Comparison of 17X17 0FA, 17X17 LOPAR, and 12 1

17X17 VANTAGE 5 Fuel Assembly Design Parameters 5.1 Callaway Thermal and Hydraulic Design Parameters 18 7.1 Summary and Justification for the Callaway Plant 35 Technical Specification Changes for VANTAGE 5 Fuel i

LIST OF FIGURES l

Figure No.

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3.1 17x17 VANTAGE 5/0FA/LOPAR Fuel Assembly Comparison 13 j

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1.0 INTR 000CTION t

The Callaway Plant was operated in Cycle 1 with a Wastinghouse 17x17

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low parasitic (LOPAR) fueled core and is operated in Cycle 2 with a transition i

fueled core containing Westinghouse 17x17 Optimized Fuel Assemblies (OFA).

j For subsequent cycles, it is planned to refuel and operate the Callaway Plant I

with the Westinghouse VANTAGE 5 improved fuel design. As a result, future

]

core loadings would range from approximately 7% LOPAR, 43% OFA, and 50%

VANTAGE 5 transition cores (Cycle 3) to eventually an all VANTAGE 5 fueled core in Cycle 5.

The VANTAGE 5 fuel assembly is designed as a modificaticn of I

the current 17x17 LOPAR and 0FA fuel assembly designs.

The VANTAGE 5 design I

featuras include:

(1)IntegralFuelBurnableAbsorbers(IFBA),

(2) Intermediate Flow Mixer (IFM) grids, (3) Reconstitutable Top Nozzles, (4) Extended Burnup, and (5) Axial blankets (not applicable to Cycle 3).

These design features were generically approved in Reference 1.

l The 17x17 0FA design has been generically approved by the NRC via their review

!I of the W 17x17 0FA Reference Core Report, Reference 2.

The Callaway Cycle 2 transition from LOPAR to 0FA was reviewed and approved by the NRC via the l

licensing submittal given in Reference 3.

This report is to serve as a i

reference safety evaluation / analysis report for the region-by-region reload l

transition from the present Callaway LOPAR/0FA fueled core to an all VANTAGE 5 i

fueled core.

This report examines the differences between the VANTAGE 5 and LOPAR/0FA fuel assembly designs and evaluates the effect of these differences l

on the cores during the transition to an all VANTAGE 5 core. The VANTAGE 5 core evaluation / analyses were performed at a core thermal power level of 3565 I

MWt with the following conservative assumptions made in the safety evaluations:

a full power FAH of_1.65 for the VANTAGE 5 fuel, an increase I

in the maximum F to 2.5,15% plant total steam generator tube plugging, and 0

a positive moderator temperature coefficient (PMTC) of +5 pcm/ degree F from 0%

to 70% power and decreasing linearly to O pcm/ degree F at 100% power.

Certain non-LOCA transients were evaluated for VANTAGE 5 fuel and 3565 MWt, requiring no reanalysis.

These transients were analyzed for Cycle 2 at 10% steam

)

generator tube plugging.

Therefore, the current licensing basis for the l

Callaway Plant for these parameters, i.e., FAH = 1.55, Fg = 2.32, 10%

l steam generator tube plugging, and a non positive moderator temperature coefficient will continue to be in effect for Cycle 3.

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4 This report utilizes the standard reload design methods described.in Reference 4 and will be used as a basic reference document in support of j

future Callaway Reload Safety Evaluations (RSE) for VANTAGE 5 fuel reloads.

This report also considers the transition effects described in Chapter 18 of i

I Reference 2.

Sections 3.0 through 6.0 of this report summarize the Mechanical, Nuclear, Thermal and Hydraulic, and Accident Evaluations, respectively. Section 7.0 gives a summary of the Technical Specification i

changes needed for Cycle 3. contains the Technical Specification change pages. Attachment 3 presents the environmental evaluation of the I

extended fuel burnup capability of the VANTAGE 5 fuel. Attachment 4 contains i

an evaluation of Significant Hazards Considerations associated with Cycle 3. has four Appendices. Appendix A contains the non-LOCA safety analyses, Appendix B contains the large and small break LOCA safety analyses, i

Appendix C contains an analysis of the spent fuel pool cooling and HVAC system's capability to accommodate spent VANTAGE 5 assemblies. Appendix 0 contains the radiological consequences resulting from the use of VANTAGE 5 1

fuel.

1 In order to demonstrate the early performance of the VANTAGE 5 design product j

j features in a commercial reactor, four VANTAGE 5 demonstration assemblies l

(17x17) were loaded into the V. C. Summer Unit 1 Cycle 2 core and began power i

production in December of 1984. These assemblies completed one cycle of irradiation in October of 1985 with an average burnup of 11,357 MWD /MTU.

Post-irradiation examinations showed all four demonstration assemblies were of good mechanical integrity.

No mechanical damage or wear was evident on any of i

the VANTAGE 5 components.

Likewise, the IFM grids on the VANTAGE 5 demo's had no effect on the adjacent fuel assemblies.

All four demonstration assemblies were reinserted into V. C. Summer 1 for a second cycle of irradiation.

This cycle was completed in March of 1987, at which time the demonstration assemblies achieved an average burnup of about 30,000 MWO/MTU.

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In addition to V. C. Summer, individual VANTAGE 5 product features have been demonstrated at other nuclear plants.

IFBA demonstration fuel rods have been irradiated in Turkey Point Units 3 and 4 for two reactor cycles. Unit 4 contains 112 fuel rods equally distributed in four demonstration assemblies.

The IFBA coating performed well with no loss of coating integrity or adher-ence. The IFM grid feature has been demonstrated at McGuire Unit 1.

The demonstration assembly at McGuire has been irradiated for two reactor cycles and is of good mechanical integrity.

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SUMMARY

AND CONCLUSIONS j.

f Consistent with the Westinghouse standard reload methodology for analyzing cycle specific reloads, Reference 4, parameters were selected to j

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conservatively bound the values for each subsequent reload cycle and to i

facilitate determination of the applicability of 10CFR50.59. The objective of f.

subsequent cycle specific reload safety evaluations will be to verify that applicable safety limits are satisfied based on the reference evaluatica/

jl analyses established in this report.

The mechanical, thermal and hydraulic, I

nuclear, and accident evaluations considered the transition core effects described for a VANTAGE 5 mixed core in Reference 1.

The summary of these evaluations for the Callaway core transitions to an all VANTAGE 5 core are given in the following sections of this submittal.

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l The transition design and safety evaluations in Section 6.0 consider the j

following conditions:

3565 MWt core thermal power, 2250 psia nominal system pressure (pressurizer pressure), 556.8'F core inlet temperature (HFP) at 2250 1

j psia, 374,400 gpm RCS thermal design flow, and minimum measured flow of 382,630 gpm (557.4*F core inlet temperature HFP at 2280 psia).

These conditions are used in core design and safety evaluations to justify safe

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operation with the conservative assumptions noted in Section 1.0.

f i

The conditions summarized in the SER for the VANTAGE 5 reference core report, i

j WCAP-10444, have been considered in the Calloway plant-specific safety l

evaluations contained herein. Pertinent conditions were noted in the SER, i

which stated, for plants using the Improved Thermal Design Procedure (ITDP),

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that the values of the mean, the uncertainties, and the sensitivity factors of j

j the measured plant parameters should be provided. This information was submitted to the NRC staff for the Callaway Plant in ULNRC-1227, dated l

December 13, 1985.

l The results of the evaluation / analysis described herein lead to the following I

conclusions:

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a.

The Westinghouse reload VANTAGE 5 fuel assemblies for the Callaway Plant are mechanically compatible with the LOPAR and 0FA fuel assemblies,-

control rods, and reactor internals interfaces. The VANTAGE 5 fuel i

assemblies satisfy the current design bases for the Callaway Plant.

i b.

Fuel performance evaluation:: at extended burnups for the VANTAGE 5 fuel have not identified any safety issues with operation to the burnup values i

given in the Westinghouse Extended Burnup Topical, WCAP-10125-P-A, i

Reference 12.

Changes in the nuclear characteristics due to the transition from LOPAR/

j c.

OFA to VANTAGE 5 fuel will be within the range normally seen from cycle to cycle due to fuel management effects.

d.

The reload VANTAGE 5 fuel is hydraulically compatible with the LOPAR and l

OFA fuel assemblies from the Cycle 2 core, a

i The analyses for the transition to a VANTAGE 5 core provide acceptable j (

e.

i results by meeting the applicable criteria, such as, minimum DNBR, peak i

pressure, and peak clad temperature, as required. All licensed safety i

limits are met. Analyses in support of this safety evaluation establish a l

reference design on which subsequent reload safety evaluations involving VANTAGE 5 reloads can be based.

i f.

Plant operating limitations given in the Technical Specifications will be satisfied with the proposed changes noted in Attachment 2 of this report.

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3.0 MECHANICAL EVALUATION j

This section evaluates the mechanical design and the compatibility of the

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17x17 VANTAGE 5 fuel assembly with the current 17x17 0FA and LOPAR fuel assemblies during the transition through mixed-fuel cores to an all VANTAGE 5 I

core. The VANTAGE 5 fuel assembly has been designed to be compatible with the 0FA and LOPAR fuel assemblies, reactor internals interfaces, the fuel handling j

equipment, and the refueling equipment. The VANTAGE 5 design is intended to replace and be compatible with fuel cores containing fuel of the LOPAR or OFA

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designs. The VANTAGE 5 design dimensions as shown on Figure 3.1 are j

essentially equivalent to these designs from an exterior assembly envelope and

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reactor internals interface standpoint. The design basis and design limits are essentially the same as those for the OFA and LOPAR designs. -As such I

compliance with the "Acceptanca Criteria" of the Standard Review Plan (SRP, NUREG 0800) Section 4.2 Fuel System Design was fully demonstrated.

i i The significant new mechanical features of the VANTAGE 5 design relative to 3

the current fuel designs (LOPAR and 0FA) in operation include the following:

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Table 3.1 provides a comparison of the VANTAGE 5, OFA and LOPAR fuel assembly design parameters.

4 Fuel Rod Performance i

Fuel rod performance for all fuel rod designs is shown to satisfy the SRP fuel f

rod design bases on a region by region basis.

These same bases are applicable to all fuel rod designs, including the Westinghouse LOPAR, OFA, and VANTAGE 5 l

fuel designs, with the only difference being that the VANTAGE 5 fuel is designed to achieve a higher burnup consistent with WCAP-10125-P-A "$xtended Burnup Evaluation of Westinghouse Fuel," and VANTAGE E fuel is designed to I

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ys; operate with a higher FAH limit. However, the FAH for the OFA fuel during the transition cycle remains consistent with the current licensing basis, i.e., FAH = 1.55.

The design bases'for Westinghouse VANTAGE 5 fuel are discussed in Reference 1, Section 2.4.

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There is no impact from a fuel rod design standpoint due to having fuel with j

more than one type of geometry simultaneously residing in the core during the j

transition cycles.

The mechanical fuel rod design evaluation for each region

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incorporates all appropriate design features of the region, including any changes to the fuel rod or pellet geometry from that of previous fuel regions.

4 i

The IFBA coated fuel pellets are identical to the enriched uranium dioxide l

pellets except for the addition of a thin boride coating on the pellet i

cylindrical surface. Coated pellets occupy the central portion of the fuel i

column. The number and pattern of IFBA rods within an assembly may vary

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depending on specific application. The ends of the enriched coated pellets i

j and enriched uncoated pellets are dished to allow for greater axial expansion l(

at the pellet centerline and void volume for fission gas release.

Analysis of i"

IFBA rods includes any geometry changes necessary to model the presence of burnable absorber, and conservatively models the gas release from the coating. An evaluation and test program for the IFBA design features are l

given in Section 2.5 in Reference 1.

l 1

Fuel performance evaluations are completed for each fuel region to demonstrate i

that the design criteria will be satisfied for all fuel rod types in the core under the planned operating conditions. Any changes from the plant operating conditions originally considered in the mechanical design of a fuel region (for example, a power uprating or an increase in the peaking factors) are addressed for all affected fuel regions.

Fuel rod design evaluations are currently performed using the NRC approved models in Peferences 6, 7, and 8 to i

demonstrate that the SRP fuel rod design criteria (including the rod internal I

pressure design basis in Reference 9) will be satisfied.

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Grid Assemblies s

The top and bottom Inconel (non-mixing vane) grids of the OFA and VANTAGE 5 fuel assemblies are nearly identical in design to the Inconel grids of the Cycle 1 LOPAR fuel assemblies.

The only differences are:

1) the grid spring and dimple heights have been modified to accommodate the reduced diameter fuel rod, 2) the grid spring force has been reduced, and 3) the OFA top grid uses 304L stainless steel sleeves instead of 304 stainless steel sleeves used for the LOPAR top grid. The six intermediate (mixing vane) structural grids are made of Zircaloy material rather than the Inconal used in the LOPAR design, the straps are thicker and the grid height is greater compared to the LOPAR design.

The Intermediate Flow Mixer (IFM) grids shown in Figure 3.1 are located in the three uppermost spans between the Zircaloy mixing vane structural grids and incorporate a similar mixing vane array.

Their prime function is mid-span 2

flow mixing in the hottest fuel assembly spans.

Each IFM grid cell contains four dimples which are designed to prevent mid-span channel closure in the spans containing IFNs and fuel rod contact with the mixing vanes.

This simplified cell arrangement allows short grid cells so that the IFM grid can accomplish its flow mixing objective with minimal pressure drop.

The IFM grids are not intended to be structural members. The outer strap configuration was designed similar to current fuel designs to preclude grid hang-up and damage during fuel handling.

Additionally, the grid outer dimensions are smaller which further minimizes the potential for damage and reduces calculated forces during seismic /LOCA events.

A coolable geometry is, therefore, assured at the IFM grid elevation, as well as at the structural grid elevation.

Reconstitutable Top Nozzle and Bottom Nozzle The reconstitutable top nozzle for the VANTAGE 5 fuel assembly differs from the OFA/LOPAR design in two ways: a groove is provided in each thimble thru-hole in the nozzle plate to facilitate removal; and the nozzle plate thickness is reduced to provide additional axial space for fuel rod growth, 8

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To remove the top nozzle, a tool is first inserted through a lock tube and expanded radially to engage the bottom edge of the tube. An axial force is then exerted on the tool which overrides local icek tube deformations and withdraws the lock tube from the insert.

After the lock tubes have been f

withdrawn, the nozzle is removed by raising it off the upper slotted ends of the nozzle inserts which deflect inwardly under the axial lift load.

With the top nozzle removed, direct access is provided for fuel rod examina-tions or replacement. Reconstitution is completed by the remounting of the nozzle and the insertion of new lock tubes.

Additional details of this design feature, the design bases and evaluation of the reconstitutable top nozzle are

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given in Section 2.3.2 in Reference 1.

i The VANTAGE 5 bottom nozzle design is similar to the OFA design except it is i

shorter and has a thinner top plate to allow for fuel rod growth.

The design I

bases and evaluation of the VANTAGE 5 bottom nozzle are given in Section 2.3.1 in Reference 1.

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Axial Blankets Although noted as a new mechanical feature of the VANTAGE 5 design and i

licensed in Reference 1, axial blankets have been and are currently operating j

in Westinghouse plants. Axial blankets will not be implemented for the j

upcoming transition reload (Cycle 3), but will be implemented in future j

reloads. A description and design application of this feature are contained in Reference 1, Section 3.0.

J Mechanical Ccmpatibility of Fuel Assemblies Based on the evaluation of the VANTAGE 5/0FA/LOPAR design differences and l

hydraulic test results (References 1, 2), it is concluded that the three designs are mechanically compatible with each other.

The OFA/ VANTAGE 5 fuel l

rod mechanical design bases remain unchanged from that used for the LOPAR fuel assemblies in the Cycle 1 core, except for a change in the non-operational 6g loading basis. For VANTAGE 5, and the current OFA and LOPAR assemblies, the j

l design basis is changed to 4g axial loading with dimensional stability.

The i

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lateral loading design basis remains at 6g for all designs.

Westinghouse has generically incorporated this change, and the justification showing no un-reviewed safety question was previously transmitted to the NRC via Reference 4.

Rod Bow 1

i The amount of fuel rod bow for VANTAGE 5 fuel is predicted to be no greater j

than that for OFA rods, since both fuel designs have the sat,e fuel rod diameter, Zircaloy grid spacings, and' grid designs. This evaluation is based j

on the application of rod bow scaling factors given in Appendices C and D of j

Reference 11.

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Rod bow in fuel rods containing IFBAs is not expected to differ in magnitude i

i or frequency from that currently observed in OFA fuel rods under similar i

operating conditions. No indications of abnormal rod bow have been observed i

on visual or dimensional inspections performed on the test IFBA rods.

Rod l

growth measurements were also within predicted bounds.

k Fuel Rod Wear Fuel rod wear is dependent on both the support conditions and the flow environment to which the fuel rod is subjected. Due to the OFA, LOPAR, and a

j VANTAGE 5 fuel assembly designs employing different grids, there is an unequal j

axial pressure distribution between the assemblies. The VANTAGE 5 fuel l

assembly was flow tested adjacent to a 17x17 0FA, since vibration test results l

indicated that the crossflow effects produced by this fuel assembly combination would have the most detrimental effect on fuel rod wear.

Results of the wear inspection and analysis discussed in Reference 1, Appendix A.1.4, revealed that the VANTAGE 5 fuel assembly wear characteristic was l

l similar.to that of the 17x17 0FA when both sets of data were normalized to the l

i test duration time.

It was concluded that the VANTAGE 5 fuel rod wear would be less than the maximum wear depth established, Reference 10, for the 17x17 i

0FA at EOL.

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Seismic /LOCA Impact on Fuel Assemblies An evaluation of VANTAGE 5 fuel assembly structural integrity considering the lateral effects of a LOCA and a seismic accident was performed. The safe shutdown earthquake and the LOCA comparative analyses indicated that the IFM grids will. share the grid load among the structural grids. The resultant loads acting on the structural grids and the flow mixers were compared with the experimental data obtained from grid impact tests at operating temperature.

j

.The grid load comparison study results, Reference 1, showed that the VANTAGE 5 fuel assembly has more margin in withstanding the faulted condition transient load than the optimized fuel assembly design, Reference 10.

The LOCA and seismic induced grid impact forces were combined in accordance with the NRC requirements as given in the Standard Review Plan 4.2 Appendix A and the core coolable geometry requirement was met.

The reactor can be safely shutdown under the faulted condition impact loads.

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TABLE 3.1 COMPARISON OF 17x17 0FA, 17x17 LOPAR and 17x17 VANTAGE 5 FUEL ASSEMBLY DESIGN PARAMETERS 17x17 17x17 17x17 PARAMETER OFA DESIGN LOPAR DESIGN VANTAGE 5 DESIGN Fuel Assy length, in 159.765 159.765 159.975 Fuel Rod Length, in 151.56 151.56 152.255 Assembly Envelope, in 8.426 B.426 8.426 Compatible with Core Internals Yes Yes Yes Fuel Rod Pitch, in

.496

.496

.496 Number of Fuel Rods /Assy 264 264 264 Number / Guide Thimble Tubes /Assy 24 24 24 Number / Instrumentation Tube /Assy 1

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Fuel Tube Material Zircaloy 4 Zircaloy 4 Zircaloy 4 Fuel Rod Clad 00, in 0.360 0.374 0.360 Fuel Rod Clad Thickness, in

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T 17 X 17 ECONSTITUTABLE LDPAR n.E1. Ah Y FIGl%E 3.1 17 X 17 VANTAGE 5 / OFA / LOPAR FUEL ASSEMSLY COMPARISON DIWD6M IN INDES 13 n

4.0 NUCLEAR EVALUATION The transition from the current mixed LOPAR/0FA core to VANTAGE 5 fuel will not result in changes to the current nuclear design bases given in the Callaway Plant FSAR. The evaluation of transition and equilibrium cycle VANTAGE 5 cores presented in Reference 1 as well as the transition and equilibrium core evaluations for the i

Callaway Plant demonstrate that the impact of implementing VANTAGE 5 causes no t

significant change to the physics characteristics of the Callaway core beyond the normal range of variations seen from cycle to cycle.

The methods and core models used in the Callaway reload transition core analysis are described in References 1, 4, and 13.

These licensed methods and models have been used for Callaway and other previous Westinghouse reload designs using LOPAR, OFA, and VANTAGE 5 fuel.

(Improved nodal analysis methods described in Reference 14 will also be used in future cycle specific Callaway reload design analyses.) No changes to the nuclear design philosoph!, methods, or models are necessary because of the transition to VANTAGE 5 fuel.

Increased emphasis will be placed on the use of three-dimensional nuclear models because of the axially heterogeneous nature of the VANTAGE 5 fuel design when axial blankets and reduced length IFBAs are used in future Callaway cores (not applicable to Cycle 3).

One Technical Specification change that will be implementad at the same time as the transition to VANTAGE 5 fuel revises the surveillance requirements on the heat fiux hot channel factor, F (z). This change has been proposed to increase plant n

operating flexibility while more directly monitoring F (z).

Rather than 0

performing surveillance on Fxy(z), the radial component of the total peaking factor, surveillance is performed directly on F (z). The steady-state F (z) is g

g measured and increased by applicable uncertainties.

This quantity is further increased by an analytical factor called W(z) which accounts for possible increases in the steady-state F (z) resulting from operation within the allowed axial flux g

difference limits. The resulting F (z) is compared to the F (z) limit to g

0 demonstrate operation below the heat flux hot channel factor limit.

14 371# 6-870327

Power distributions and peaking factors show slight changes as a result of the incorporation of axial blankets, reduced length IFBAs, and increased peaking factor limits, in addition to the normal variations experienced with different i

loading patterns.

These changes are small, and the usual methods of enrichment variation and burnable absorber usage will be employed as Callaway transitions to VANTAGE 5 fuel to ensure compliance with Technical Specifications peaking factor limits.

The key safety parameters evaluated for Callaway as it transitions to an all VANTAGE 5 core show little change relative to the range of parameters experienced for the mixed LOPAR/0FA core.

The changes in values of the key safety parameters are typical of the normal cycle-to-cycle variations experienced as loading patterns change. The analysis of key safety parameters also included the additional conservative assumptions of increased peaking factor limits and a positive MTC limit at low power levels. As is current practice, each reload core design will be evaluated to assure that design and safety limits are satisfied according to the reload methodology.

The design and safety limits will be documented in each cycle specific reload safety evaluation (RSE) report which serves as a basis for any significant changes which may require a future NRC review.

15 ma s-com

5.0 THERMAL AND HYDRAULIC EVALUATION The analysis of the VANTAGE 5 fuel includes the use of the WRB-2 DNB correlation, Reference 1 and the Improved Thermal Design Procedure, Reference

15. The analysis of both LOPAR and 0FA fuel, as in Cycle 2, includes the use of the WRB-1 DNB correlation, Reference 19, and the Improved Thermal Design Procedure, Reference 15.

The WRB-2 DNB correlation takes credit for the significant improvement in the accuracy of critical heat flux predictions over previous DNB correlations and the VANTAGE 5 fuel assembly mixing vane design.

A DNBR limit of 1.17 is applicable for both the WRB-1 and WRB-2 correlations.

The thermal-hydraulic design of this core is analyzed at the 3565 MWt core thermal power level with a 588.4'F vessel average temperature. Table 5.1 summarizes the pertinent thermal and hydraulic design parameters.

The methodology employed to meet the DNB design basis is the Improved Thermal Design Procedure (ITDP) which has been approved by the NRC, Reference 16.

Uncertainties in plant operating parameters, nuclear and thermal parameters, and fuel fabrication parameters are considered statistically such that there is at least a 95 percent probability with a 95 percent confidence level that the minimum DNBR will be greater than or equal to 1.17 for the limiting power rod. Plant parameter uncertainties are used to determine the plant DNBR uncertainty.

This DNBR uncertainty, combined with the DNBR limit, establishes a DNBR value which must be met in plant safety analyses.

Since the parameter uncertainties are considered in determining the design DNBR value, the plant safety analyses are performed using values of input parameters without uncertainties. For this application, the minimum required design DNBR values are 1.32 for thimble coldwall cells (three fuel rods and a thimble tube) and 1.34 for typical cells (four fuel rods) for LOPAR/0FA fuel and 1.32 for thimble and 1.33 for typical cells for the VANTAGE 5 fuel.

In addition to the above considerations, a plant-specific DNBR margin has been considered in the analyses.

In particular, safety analysis DNBR limits of 1.42 for the thimble and 1.45 for the typical cells for LOPAR/0FA fuel, and 1.61 and 1.69 for thimble and typical cells respectively for the VANTAGE 5 fuel, were employed in the safety analyses.

The DNBR margin between the DNBRs used in the safety analyses and the design DNBR values is broken down as 16 ma s-mw

O follows. A fraction of the margin is utilized to accommodate the transition core DNBR penalty (2% for 0FA fuel and 11% for VANTAGE 5 fuel) and the appropriate fuel rod bow DNBR penalty, Reference 11, which is less than 1.5%.

The existing 7% margin in the OFA fuel and 18% margin in the VANTAGE 5 fuel between the design and safety analysis DNBR limits also include a greater than 3% DNBR margin in the OFA fuel and a greater than 5% DNBR margin in the VANTAGE 5 fuel reserved for flexibility in the design.

The LOPAR, and 0FA, and VANTAGE 5 designs have been shown to be hydraulically compatible in Reference 1.

The phenomenon of fuel rod bowing, as described in Reference 11, must be accounted for in the DNBR safety analysis of Condition I and Condition II events for each plant application. Applicable generic credits for margin resulting from retained conservatism in the evaluation of DNBR and/or margin obtained from measured plant operating parameters (such as F g or core flow) which are more restrictive than those required by the plant safety analysis, can be used to offset the effect of rod bow.

The safety analysis i

for the Callaway Plant maintains sufficient margin between the safety analysis limit DNBRs and the design limit DNBRs to accommodate full-flow and low-flow DNBR penalties.

The transition core DNB methodology given in References 2 and 17 has been approved by the NRC via Reference 18. Using this methodology, transition cores are analyzed as if they were full cores of one assembly type (full 0FA or full VANTAGE 5), applying the applicable transition core penalties of two percent for 0FA fuel and eleven percent for VANTAGE 5 fuel. The safety analyses for the Callaway Plant maintains sufficient margin between the safety analysis limit DNBRs and the design limit DNBRs to accommodate the transition core DNBR penalty and the appropriate rod bow DNBR penalty.

1 The fuel temperatures for use in safety analysis calculations for the VANTAGE 5 fuel are the same as those used for the OFA fuel. Westinghouse uses the PAD f

fuel performance code, Reference 6, to perform both design and licensing calculations. When the code is used to calculate fuel temperatures to be used f

I as initial conditions in safety analyses, a conservative thermal safety model, v

Reference 7, is used.

17 3754P6-470327

r'~

TABLE 5.1 CALLAWAY THERMAL AND HYDRAULIC DESIGN PARAMETERS Design Thermal and Hydraulic Design Parameters Parameters (Using ITDP)

Reactor Core Heat Output, MWt 3565 6

Reactor Core Heat Output, 10 BTU /hr 12,164 Heat Generated in Fuel, %

97.4 Core Pressure, Nominal, psia 2280 Radial Power Distribution (LOPAR/0FA):

1.49(1+0.3(1-P]

(V5):

1.59[1+0.3(1-P)]

s Design DNBR for Design Transients Typical Flow Channel 1.45 (LOPAR/0FA) 1.69 (VS)

Thimble (Cold Wall) Flow Channel 1.42 (LOPAR/0FA) 1.61 (V5)

DNBR Correlation WRB-1 (LOPAR/0FA)

WR3-2 (VS) e 18 375 4 6-470327

TABLE 5.1 (Continued)

CALLAWAY THERMAL AND-HYDRAULIC DESIGN PARAMETERS HFP Nominal Coolant Conditions

• VesseiMinimumMeasuredFlow+

6 Rate (Including Bypass), 10 lbm/hr 141.9 GPM 382,630 Vessel Thermal Design Flow +

6 Rate (Including Bypass), 10 lbm/hr 138.8 GPM 374,400 Core Flow Rate (Excluding Bypass, based on TDF),

6 10 lbm/hr 130.0 GPM 350,810 Fuel Assembly Flow Area for Heat 54.13 (OFA/V5)++

2 Transfer, ft 51.08 (LOPAR) 6 Core Inlet Mass Velocity, 10,

2.40 (OFA/V5) f lbm/hr-ft2 (Based on TDF) 2.55 (LOPAR)

Design Thermal and Hydraulic Desian Parameters Parameters (Based on Thermal Design Flow)

Nominal Vessel / Core Inlet Temperature, *F 556.8 Vessel Average Temperature, *F 588.4 Core Average Temperature, *F 592.2 Vessel Outlet Temperature, 'F 620.0 Average Temperature Rise in Vessel, 'F 63.2 Average Temperature Rise in Core, 'F 66.9

'a 19 me - 7n27 j

TABLE 5.1 (Continued)

CALLAWAY THERMAL AND HYDRAULIC DESIGN PARAMETERS Heat Transfer 2

Active Heat Transfer Surface Area, ft '

57,505 (OFA/V5)++

59,742 (LOPAR) 2 Average Heat Flux, BTU /hr-ft 206,090 (OFA/V5) 198,370 (LOPAR)

Average Linear Power, kw/ft 5.69 Peak Linear Power for Normal Operation +++, kw/ft 14.23 i

560.1*F and Pressure = 2280 psia Based on F5AR Safety Analysis T Includes 15% steam generator tub p= lugging

+

++

Assumes all 0FA, V5 or LOPAR Core

+++ Based on 2.50 F Peaking Factor g

s 20 m a s-ero w

m

,Q/

6.0 ACCIDENT EVALUATION 6.1 Non-LOCA Accidents This section addresses the impact on non-LOCA accident analyses of the following proposed changes and design safety analysis assumptions for the Callaway Plant.

Proposed Change to the Licensing Basis:

VANTAGE 5 Fuel Design Design Safety Analysis Assumptions:

Positive Moderator Temperature Coefficient Increased Design Enthalpy Rise Pecking Factor and Fq Increase in Allowable Steam Generator Tube Plugging Level for Certain

(

Transients (design assumption)

VANTAGE 5 The design features of VANTAGE 5 fuel considered in the non-LOCA analysis are:

VANTAGE 5 Fuel Rod Dimensions Axial Blankets (not applicable to Cycle 3)

Integral Fuel Burnable Absorbers (IFBAs)

Intermediate Flow Nixer Grids (IFMs)

Reconstitutable Top Nozzle Fuel Enrichment Extended Burnup Fuel Assembly Design Fuel Rod Dimensions The VANTAGE 5 fuel rod dimensions which determine the safety analysis temperature versus linear power density relationship are identical to the OFA s

rod design.

These dimensions include rod diameter, pellet diameter, initial s

21 3754F 6-870327

- -. ~

r pellet-to-clad gap size, and stack height.

Therefore, the non-LOCA safety 6

analysis fuel temperature and rod geometry assumptions are identical to those assumed in the Callaway 0FA transition core submittal, Reference 3.

As described in the OFA transition submittal, the fuel rod diameter reduction from the LOPAR to 0FA design results in higher fuel rod temperature, higher surface heat flux, and a DNB penalty. The OFA DNB penalty is offset through the use of the WRB-1 correlation, Reference 19, and the Improved Thermal Design Procedure, Reference 15. The VANTAGE 5 fuel DNB analysis uses the Improved Thermal Design Procedure and the WRB-2 correlation, which is described in Appendix A of Reference 1.

Axial Blankets and IFBAs Axial blankets reduce power at the ends of the rod which increases axial peaking in the interior of the rod. Used alone, axial blankets reduce associated DNB margin, but the effect may be offset by the presence of Integral Fuel Burnable Absorbers (IFBAs) which flatten the power distri-bution. The net effect on the axial shape is a function of the number and configuration of IFBAs in the core and time in life. The effects of axial blankets and IFBAs on the reload safety analysis parameters are taken into account in the reload design process. The axial power distribution assumption in the safety analyses kitetics calculations has been determined to be applicable for evaluating the introduction of axial blankets in the Callaway Plant for Cycle 4 and part length IFBAs in subsequent cycles.

IFM Grids and Reconstitutable Top Nozzle The IFM grid feature of the VANTAGE 5 fuel design increases DNB margin. The fuel safety analysis limit DNBR values contain at least 18% DNB margin (see Section5'.0).

This DNB margin was set to ensure that the core thermal safety limits for the VANTAGE 5 fuel with an FaH of 1.65 are the same as or are slightly less restrictive than the OFA core thermal safety limits which include an FAH of 1.55 and the effects of the increased bypass flow due to VANTAGE 5 fuel. The Callaway 0FA transition core report, Reference 3, 22 l

nw e-mw

\\

O A

- ([

concluded that the OFA fuel core limits were more restrictive than the LOPAR fuel core limits. Thus the most restrictive core limits correspond to the OFA l

fuel design. Any transition core penalty is accounted for with generic DNB margin.

The IFN grid feature of the VANTAGE 5 fuel design increases the core pressure drop. The control rod scram time to the dashpot is increased from 2.4 to 2.7 I

i seconds. The increased drop time primarily affects the fast reactivity transients. These accidents were reanalyzed for this report. The revised safety analysis assumption was incorporated in all the reanalyzed events requiring this parameter and the remaining transients have been evaluated.

Core flow areas and loss coefficients were preserved in the design of the reconstitutable top nozzle.

As such, no parameters important to non-LOCA safety analyses are impacted.

Fuel Enrichment The VANTAGE 5 fuel design increased fuel enrichment is conservatively bounded 4

by the maximum safety analysis assumption of 5.0 w/o.

Extended Burnuo Fuel Assembly Design WCAP-10125-P-A, " Extended Burnup Evaluation of Westinghouse Fuel", Reference 12, evaluates the impact of extended burnup on the design and operation of Westinghouse fuel. The major effect of the extended burnup rod design is on power shaping between fresh and burned assemblies.

i 6.1.1 Increased Design Enthalpy Rise Peaking Factor (FAH)

The FAH for the OFA fuel during the transition cycles, consistent with the current licensing basis, is 1.55. The non-LOCA transients reanalyzed for the VANTAGE 5 fuel have assumed a full power FAH of 1.65.

This is a conservative safety analysis assumption for this report.

Safety analyses 7

which incorporate explicit modeling of this peaking factor were reanalyzed.

They are:

23 37547 6-87032?

r o

Locked rotor (rods-in-DNB calculation) y o

Partial Loss of Forced Reactor Coolant Flow o

Complete Loss of Forced Reactor Coolant Flow Startup of An Inactive Reactor Coolant Loop At An Incorrect Temperature o

The design core limits for this report incorporate the increased FAH for the VANTAGE 5 fuel.

6.1.2 Positive Moderator Temperature Coefficient A positive moderator temperature coefficient (PMTC) of +5 pcm/ degree F from 0%

to 70% power and decreasing linearly to O pcm/ degree F at 100% power was incorporated into the safety analyses performed for this report.

Transients sensitive to a positive or near zero moderator temperature coefficient were reanalyzed. All other transients were evaluated. The reanalyzed transients

~

- include:

o Control Rod Assembly Withdrawal From Subcritical l

o Control Rod Assembly Withdrawal At Power

~

~

o Loss of Reactor Coolant Flow (Partial and Complete) o Locked Rotor o

Loss of External Load / Turbine Trip o

RCS Depressurization a

Loss of Normal Feedwater o

Feedline Break o

Control Rod Ejection o

Loss of Non-Emergency AC Power In general, the analyses presented here are based on a +5 pcm/ degree F moderator temperature coefficient, which is assumed to remain constant for variations in temperature.

Exceptions are rod ejection and rod withdrawal from suberitical which are based on a MTC of +5 pcm/ degree F at zero power nominal average temperature and which, due to moderator temperature feedback modeled in the TWINKLE diffusion-theory code, becomes less positive for higher temperatures. A summary of the transient modeling assumption for each transient is provided in Section 15.0 of Attachment 5, Appendix A.

f 24 375 # S-870327

O

()

The non-LOCA safety analyses not listed above include those resulting in excessive heat removal from the reactor coolant system for which a large negative moderator temperature coefficient is more limiting, and these for which heatup effects following reactor trip are not sensitive to the moderator temperature-coefficient. These transients were evaluated to remain applicable for PMTC.

The baron dilution transient was reanalyzed to incorporate the increases in RCS boron concentration expected with PMTC.

Incorporation of the described level of PMTC into the safety analyses is, in all cases, a conservative assumption for this report.

The Chapter 15 Baron Oilution analysis has incorporated baron concentrations intended to remain applicable to future Callaway fuel cycles in which PMTC may be incorporated.

The assumptions are conservative with respect to Cycle 3 specific boron concentrations and are also beyond the RCS baron concentration range specified in the currently applicable Callaway Technical Specifications.

Consistent with the Westinghouse Reload Methodology, analyses are performed

(

specific to Cycle 3 to verify the adequacy of the boron concentration in the RWST and accumulator tanks. Since PMTC has not been incorporated into the Callaway Cycle 3 fuel design, changes to the RWST and accumulator boron concentrations are not required and the existing related technical specifications remain applicable.

6.1.3 Increased Steam Generator Tube Plugging The Callaway 0FA transition core report, Reference 3, provided safety analysis justification to support up to 10% plant total steam generator tube plugging -

not to exceed 10% in any single steam generator. All non-LOCA safety analyses reanalyzed for this report have incorporated any necessary changes to model 15%

plant total steam generator tube plugging.

No conclusions are made regarding the impact of this assumption on those transients not requiring explicit reanalysis for this report. The licensing basis supported by this report continues to be 10%.

J 25 3754F S-87C327

o 6.1.4 Increase in Fg from 2.32 to 2.50 is The increase in the Technical Specification maximum Fg conservatively accounted for in the non-LOCA transients reanalyzed for this report and in those transients evaluated to remain applicable for this report.

The reanalyzed transients include locked rotor and rod ejection.

6.1.5 Non-LOCA Safety Evaluation Methodology The non-LOCA safety 6 valuation process is described in References 1 and 2.

The process determines if a core configuration is bounded by existing safety analyses in order to confirm that applicable safety criteria are satisfied.

The methodology systematically identifies parameter changes on a cycle-by-cycle basis which may invalidate existing safety analysis assumptions and identifies the transients which require re-evaluation. This methodology is applicable to the evaluation of VANTAGE 5 transition and full cores.

Any required re-evaluation identified by the reload methodology is one of two j

types.

If the identified parameter is only slightly out of bounds, or the transient is relatively insensitive to that parameter, a simple evaluation may be i

made which conservatively evaluates the magnitude of the effect and explains why l

the actual analysis of the event does not have to be repeated. Alternatively, should the deviation be large and/or expected to have a significantly or not easily quantifiable effect on the transients, reanalyses are required.

The reanalysis approach will typically utilize the analytical methods which have been used in previous submittals to the NRC. These methods are'those which have been presented in FSARs, subsequent submittals to the NRC for a specific plant, reference SARs, or report submittals for NRC approval.

l The key safety parameters are documented in Reference 4.

Values of these safety parameters which bound all three fuel types (LOPAR, OFA, VANTAGE 5) were assumed in the safety analyses.

For subsequent fuel reloads, the key safety parameters will be evaluated to determine if violations of these bounding values exist.

Re-evaluation of the affected transients would take piace and would be documented for the cycle specific reload design, as per Reference 4.

26 375 # s-470327

)

O 477s

,g).

6.1.6 Conclusions

~

Descriptions of the transients reanalyzed for this report, method of analysis, results, and conclusions are contained in Attachment 5, Appendix A.

The analytical procedures and computer codes used were established in previous analyses for both 0FA and LOPAR fuel cores or are otherwise identified and described in this report. Appendix A has been prepared using the NRC Standard Format and Content Guide, Regulatory Guide 1.70, for accidents applicable to the Callaway Plant.

For each of the accidents reanalyzed, it was found that the appropriate safety criteria are met. For each of the accidents not reanalyzed, evaluations were performed which determined that the existing conclusions remain applicable for the proposed changes to the plant.

These transients, descriptions, results, and conclusions are documented in the Callaway FSAR and the Callaway 0FA transition core report, Reference 3.

6.2 LOCA Accidents 6.2.1 Large Break LOCA 6.2.1.1 Description of Analysis / Assumptions for 17x17 VANTAGE 5 Fuel The large break loss-of-coolant (LOCA) accident analysis for the Callaway Plant, applicable to a full core of VANTAGE 5 fuel assemblies, was performed to develop the Callaway Plant specific peaking factor limits. This is consistent with the methodology employed in the Reference Core Report for 17x17 VANTAGE 5, Reference 1.

The BASH model, References 20 and 25, was utilized for a spectrum of cold leg breaks. Other pertinent analysis assumptions include:

the core thermal power of i

3565 MWt, 15% uniform (i.e., in all steam generators) steam generator tube of 2.5 and fuel data based on the new fuel plugging, an FaH of 1.65, an Fg thermal model, Reference 7.

The analysis results, tables, and figures are presented in Attachment 5, Appendix B.

J

,r 27 375eF 6-47C227

VANTAGE 5 fuel features, as applied to the Callaway Plant, result in a fuel assembly that is more limiting than either a LOPAR or 0FA fuel assembly with respect to large break LOCA ECCS performance, Reference 1.

As such, VANTAGri 5 fuel has been analyzed herein.

The fuel assembly hydraulic resistance mismatch between 17x17 LOPAR and VANTAGE 5 assemblies is greater than the mismatch between 17x17 0FA and VANTAGE 5 assemblies. Evaluation of the transition core effect on calculated peak clad temperature (PCT) for 17x17 LOPAR to VANTAGE 5 assembly transition core is, therefore, bounding and valid for 17x17 0FA to VANTAGE 5 transition core cases.

Also, the evaluation for the 17x17 LOPAR to VANTAGE 5 transition is bounding and valid for transition cores containing all three fuel types; LOPAR, OFA, and VANTAGE 5, Reference 1.

When assessing the impact of transition cores on the large break LOCA analysis, it must be determined whether the transition core can have a greater calculated 1

~ peak clad temperature (PCT) than either a complete core of the LOPAR assembly design or a complete core of the VANTAGE 5 design.

For a given peaking factor, the only mechanism available to cause a transition core to have a greater calculated PCT than a full core of either fuel is the possibility of flow redistribution due to fuel assembly hydraulic resistance mismatch.

This

- hydraulic resistance mismatch may exist only for transition cores and is the only unique difference between a complete core of either fuel type and the transition Core.

2 The LOCA analysis performed assumes a full core of VANTAGE 5 and conserva-tively applies to transition cores. The VANTAGE 5 design differs hydraulically from the LOPAR assembly design it replaces. The difference in the total assembly hydraulic resistance between the two designs is approximately 10%.

i An evaluation of a hydraulic mismatch of approximately 10% showed an insignificant effect on blowdown cooling during a LOCA. The SATAN-VI code models I

the crossflows between the average core flow channel (N-1 fuel assemblies) and the hot assembly flow channel (one flow assembly) during blowdown.

To better understand the transition core large break LOCA blowdown transient phenomena, conservative blowdown fuel clad heatup calculations have been performed to 28 37$47 6-870327

=

t h~

determine the clad temperature effect on the new fuel design for mixed core configurations. The effect was determined by reducing the axial flow in the hot assembly at the appropriate elevations to simulate the effects of the transition core hydraulic resistance mismatch.

In addition, the Westinghouse blowdown evaluation model was modified to account for grid heat transfer enhancement l

during blowdown for this evaluation. The results of this analysis have shown that no peak clad temperature penalty is observed during blowdown for the mixed core. Therefore, it is not necessary to perform a blowdown calculation for VANTAGE 5 transition core configurations because the evaluation model blowdown calculation performed for' the full VANTAGE 5 core is conservative and bounding.

Since the overall resistance of the two types of fuel is essentially the same, only the crossflows during core reflood due to the smaller rod size and different grid designs need to be evaluated. The LOCA analysis uses the BASH computer code to calculate the reflood transient, Reference 20, which utilizes the BART code, Reference 21. A detailed description of the BASH code is given in Attachment 5, Appendix B.

Fuel assembly design specific analyses have been performed with a

- ('

version of the BART computer code, Reference 21, which accurately models mixed core cases during reflood.

Westinghouse transition core designs, including a

~

specific 17x17 LOPAR to VANTAGE 5 transition core case, were analyzed. For this case, BART modeled both fuel assembly types and predicted the reduction in axial flow at the appropriate elevations. As expected, the increase in hydraulic resistance for the VANTAGE 5 assembly was shown to produce a reduction in reflood i

i steam flow rate for the VANTAGE 5 fuel at mixing vane grid elevations during the transition core period. This reduction in steam flow rate is partially offset by the fuel grid heat transfer enhancement predicted by the BART model during reflood. The various fuel assembly specific transition core analyses performed resulted in peak clad temperature increases of up to 50*F for core axial elevations that bound the location of the PCT. Therefore, the maximum PCT penalty possible for VANTAGE 5 during transition cores is 50*F (Reference 1).

j Once a full core of the VANTAGE 5 fuel is achieved, the large break LOCA analysis will apply without the crossflow penalty.

j M

i I

I M E$*$I$)II v-y 9

y w --

---w--

w ywi.

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a 6.2.1.2 Methods of. Analysis v

The methods of analysis including codes used and assumptions are described in detail in Attachment 5, Appendi.x B, Section 15.6.5.

6.2.1.3 Results The results of this analysis, including tabular and plotted results of the break spectrum analyzed, are provided in Attachment 5, Appendix B, Section 15.6.5 which' has been prepared using the NRC Standard Format and Content Guide, Regulatory i

Guide 1.70, for accidents applicable to the Callaway Plant.

Reference 25 stated three restrictions to the 1981 evaluation model using BASH.

The application of these restrictions to the plant specific LOCA analysis was addressed with the following conclusions:

The Callaway Plant is neither an Upper Head Injection (UHI) or Upper Plenum Injection (UPI) plant, so restriction 1 does act apply.

For the plant specific LOCA analysis, both minimum and maximum safeguards analyses were performed to address restriction 2.

The CD = 0.6 DECLG with minimum safeguards was identified as the most limiting break.

Sensitivity studies were performed for a typical 4-loop plant using different power shapes demonstrating that the chopped cosine is the limiting power shape, Reference 26. A chopped cosine power shape was used in the large break LOCA analysis for the Callaway Plant, so this sensitivity study was used to address restriction 3.

6.2.1.4 Conclusions For breaks up to and including the double-ended severance of a reactor coolant pipe, the emergency core cooling system will meet the acceptance criteria as presented in 10CFR50.46.

That is:

1.

The calculated peak fuel element clad temperature is below the requirement of 2200*F.

30 375W S-470327

O The amount of fuel element cladding that reacts chemically with water or 2.

steam does not exceed 1 percent of the total amount of Zircaloy in the reactor.

3.

The localized cladding oxidation limit of 17 percant is not exceeded during or after quenching.

4.

Tne core remains amenable to cooling during and after.the break.

5.

The core temperature is reduced and decay heat is removed for an extended period of time. This is required to remove the heat from the long-lived radioactivity remaining in the core.

In order to demonstrate that criterion 5 is met, a specific evaluation of the core subcriticality is made for each reload cycle.

For Cycle 3 this evaluation demonstrates that there is sufficient baron available in the ECCS fluid to maintain the core suberitical for long term core cooling.

For 2

(

subsequent cycles, this will be re-evaluated for the specific core design.

The time sequence of events for all breaks analyzed is shown in Table 15.6-6 of Attachment 5, Appendix B, Section 15.6.5.

The large break VANTAGE 5 LOCA analysis for the Callaway Plant, utilizing the j

BASH model, resulted in a peak clad temperature of 2004*F for the limiting break case at a peaking factor of 2.50. The maximum local metal-water reaction is 5.17 percent, and the total metal-water reaction is less than 0.3 percent for all cases analyzed. The clad temperature transients turn around j

at a time when the core geometry is still amenable to cooling.

tm t 5 Fe c

w ch d nt gn 10CFR50.46 limits.

l It can be seen from the results contained in Attachment 5, Appendix B, Section 15.6.5 that this ECCS analysis for the Callaway Plant remains in compliance with the requirements of 10CFR50.46.

31 37548 6-870327 i

6.2.2 Small Break LOCA 6.2.2.1 Description of Analysis / Assumptions for 17x17 VANTAGE 5 Including Transition. Impact The small break loss-of-coolant accident (LOCA) was analyzed assuming a full core of VANTAGE 5 fuel to determine the peak clad temperature (PCT). This is consistent with the methodology employed in WCAP-10444-P-A, Reference 1, for 1

the 17x17 VANTAGE 5 transition. The currently approved NOTRUMP Model Small Break ECCS Evaluation Model, Reference 22, was utilized for a spectrum of cold leg breaks. Attachment 5, Appendix B, Section 15.6.5, includes a full description of the analysis and assumptions utilized for the Westinghouse VANTAGE 5 ECCS LOCA analysis.

Pertinent assumptions include an FAH of 1.65, I

total peaking factors corresponding to 2.5 at the core mid plane, 15% steam generator tube plugging, and the core thermal power level of 3565 MWt.

Sensitivity studies performed using the NOTRUMP small break evaluation model i

have demonstrated that VANTAGE 5 fuel is more limiting than 0FA fuel in calculated ECCS performance. Similar studies using the WFLASH evaluation model, Reference 23, have previously shown that OFA fuel is more limiting than LOPAR fuel. Far the small b m k LOCA, the effect of the fuel difference is most pronounced during core uncovery periocs and, therefore shows up pre-dominantly in the LOCTA-IV calculation in the evaluation model analysis.

Consequently, the previous conclusion drawn from the WFLASH studies, regarding the fuel difference, may be extended to this NOTRUMP analysis.

Thus, only VANTAGE 5 fuel was analyzed, since it is the most limiting of the three types of fuel residing in the core.

When assessing the LOCA impact of transition cores, it must be determined whether the transition core can have a greater calculated peak clad tempera-j ture (PCT) than either a complete core of the reference fuel design or a l

complete core of the improved fuel design.

For a given peaking factor, the only mechanism available to cause a transition core to have a greater calculated PCT than a full core of either fuel is the possibility of flow redistribution due to fuel assembly hydraulic resistance mismatch. This 1

32 375476-870327 i

AI hydraulic mismatch might exist only for transition cores and is the only.

'd unique difference between a complete core of either fuel type and a transition core.

~

The NOTRUMP computer code, Reference 24, is used to model the core h during a small break event.

Only one core flow channel is modeled in the NOTRUMP code, since the core flow during a small break is relatively slow providing enough time to maintain flow equilibrium between fuel assemblies

~"

(i.e., no cross flow). Therefore, hydraulic resistance mismatch is not a factor for small break.

Thus it is not necessary to perform a small break evaluation for transition cores, and it is sufficient to reference the small break LOCA for the complete core of the VANTAGE 5 fuel design as boun all transition cycles.

6.2.2.2 Method of Analysis The methods of analysis including codes used and assumptions.are describ

(

detail in Attachment 5, Appendix B, Section 15.6.5.

l 6.2.2.3 Results I

i The results of this analysis including tabular and plotted results of the break spectrun, analyzed are provided in Attachment 5, Appendix B, Section 15.6.5 using the NRC Regulatory Guide 1.70 format.

6.2.2.4 Conclusions The small break VANTAGE 5 LOCA analysis for the Callaway Plant utiliz

\\

currently approved NOTRUMP Evaluation Model resulted in a peak clad temperature of 1528'F for the 4-inch diameter cold leg break.

The analysis assumed the limiting small break power shape consistent with a LOCA F envelope of 2.50 at core midplane eleve. tion and 2.31 at the top of the core.

0 The maximum local metal-water reaction is 0.46 percent and the total core metal water reaction is less than 0.3 percent for all cases analyzed.The clad temperature transients turn around at a time when the core geometry i k

still amenable to cooling.

s m o now 33

[

Nixed core hydraulic resistance mismatch is not a significant factor for small 3

break LOCA analysis.

Therefore, it is not necessary to perform any additional small break evaluations for transition coros and it is sufficient to re.ference the small break LOCA applicable to the compiate core of the VANTAGE 5 design.

l Analyses presented in Attachment 5, Appendix 8, Section 15.6.5 show that one l

centrifugal charging pump and one safetyinjection pump, together with the l

accumulators, provide sufficient core flooding to keep the calculated peak l

clad temperature well below the required limits of 10CFR50.46.

It can also be seen that the ECCS analysis remains in compliance with all other requirements of 10CFR50.46. Adequate protection is therefore afforded by the ECCS in the event of a small break LOCA.

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(f 7.0

SUMMARY

OF TECHNICAL SPECIFICATION CHANGES Table 7.1 presents a list of the Technical Specification changes.

The changes noted in Table 7.1 are given in the proposed Technical Specification page changes (see Attachroat 2 of this report).

In addition to these Technical Specification changes for VANTAGE 5 fuel, Union Electric requests a change to Technical Specification 3/4.5.1 to facilitate operation. This Technical Specification deals with the accumulator water level measurement band. An evaluation was performed for Callaway using the new accumulator water volumes to support a change to Technical Specificat. ion 3/4.5.1. It was determined that a plus or minus 61 gallon change in accumulator water volume would have an insignificant effect on the safety

. analyses results.

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37$d'8-470327

. _ - - _ ~ -

TABLE 7.1 SUte4ARY AND JUSTIFICATION FOR cat _LAWAY TECHNICAL SPECIFICATION CHANGES FOR VANTAGE 5 FUEL PAGE SECTION DESCHIPTION OF CHANGE JUSTIFICATION 2-2 Figure Revised Reactor Core Safety Limits The new limits result from an increase 2.1-1 in the reactor coolant core bypass flow due to the VANTAGE 5 fuel design.

2-7, Table 2.2-1 Revised value for Delta T The change is due to the ITDP g.

2-9 calculations performed to reflect the VANTAGE 5 fuel design.

1

~

.2-4 Table 2.2-1 values for Z.S. and the Allowable Value The change in setpoints is due to the 2-8 revised for OT Delta T and OP Delta T.

ITDP calculations performed to 2-10 Revised the value for Z and the allowaole reflect the VANTAGE S fuel design, value for Reactor Coolant Flow.

8 2-1 2.1.1 Added WRS-2 correlation and design and This change reflects the DNS Correlation Basis Safety Analysis DNOR limits for the used for the VANTAGE 5 fuel.

VANTAGE 5 fuel.

3/4 1-19 3.1.3.4 Revised rod drop time to <2.7 seconds.

This change is a result of an increase in the core hydraulic resistance due to the VANTAGE 5 fuel design.

3/4 2-1 3.2.1 Added limit to the LCO for restricted AFD This change reflects the 3/4 2-2 operation. Revised Action Statements operational requirements for the accordingly.

peaking factor limits.

3/4 2-6 4.2.2.2 Revised the surveillance from a F This change was made for 3/4 2-7 4.2.2.3 surveillance to a F surveillance 7Y operational flexibility.

o Added a F surveillance for restricted o

AFD operafton.

I 37547 6 870327 l

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1. ;)

t TABLE 7.1 (Continued)

SUGARGARY AND JUSTIFICATION FOR CALt AWAY TECHNICAL SPECIFICATION CHANGES FOR VANTAE 5 FUEL PAGE SECTION DESCRIPTION OF CilANCE JUSTIFICATION 3/4 2-14 Table 3.2-1 Revised indicated DNB parameter for This change reflects the use of lower T

ITDP uncertainty values, versus the

avg, uncertainty values used in the non-ITDP analyses.

3/4 5-1 3.5.1 Revised the range for the RCS accumulators.

This change facilitates plant operation and does not affect the safety analyses.

This change was made because the F [,

B 3/4 2-1 3/4.2 Deleted definition of F,y(z) e surveillance was replaced with a F Basis surveillance.

8 3/4 2-1 3/4.2.1 Added discussion on normal operation and This discussion supports the change Basis restricted AFD modes of operation.

made in Specification 3.2.1.

w N

8 3/4 2-4 3/4.2.2 Revised discussion on design and These changes support the VANTAE 5 3/4.2.3 Safety Analysis DNBR limits to include fuel design and the F surveillance.

g Basis those for the VANTAGE 5 fuel design.

Replaced discussion on the F surveillance with discussion 7n the F surveillance.

g I

B 3/4 2-5 3/4.2.5 Revised indicated and analytical These changes reflect the use of lower Basis limits in the basis.

ITDP uncertainty values versus i

the uncertainty values used in the non-ITDP analyses.

6-21 6.9.1.9 Revised the discusF100 on the This change supports the Fg Peaking Factor Limit Report surveillance.

3754F G 870327

8.0 REFERENCES

1. Davidson, S.L. and Kramer, W.R.; (Ed.) " Reference Core Report VANTAGE 5 4

Fuel Assembly," WCAP-10444-P-A, September 1985.

2. Davidson, S.L.; forii, J.

A., " Reference Core Report - 17x17 Optimized Fuel Assembly," WCAP-9500-A, May 1982.

3. Union Electric Callaway Plant OFA Transition Core Submittal to NRC, l

ULNRC-1207, November 15, 1985.

4. Davidson, S. L.'(Ed.), et al., " Westinghouse Reload Safety Evaluation Methodology, "WCAP-9272-P-A, July 1985.

5. Letter from E. P. Rahe (Westinghouse) to L. E. Phillips (NRC) dated April 12,1984,NS-EPR-2893,

Subject:

Fuel Handling Load Criteria (6g vs 4g).

6. Miller, J. V., " Improved Analytical Models Used in Westinghouse Fuel Rod Design Ccmputations," WCAP-8720 (Proprietary), October 1976.

7. Leech, W. J., et al., " Revised PAD Code Thermal Safety Model" WCAP-8720-A2 l

(Proprietary), October,1982.

8. George, R. A., (et al.), " Revised Clad Flattening Model," WCAP-8377 (Proprietary) and WCAP-8381 (Non-Proprietary), July 1974.

4

9. Risher, D.

H., (et al.), " Safety Analysis for the Revised Fuel Rod Internal Pressure Design Basis," WCAP-8963-P-A (Proprietary), August 1978.

1

10. Davidson, S.

L., Iorii, J. A. (Eds.), " Verification Testing and Analyses of the 17x17 Optimized Fuel Assembly," WCAP-9401-P-A, August 1981.

I

11. Skaritka, J.,.(Ed.), " Fuel Rod Bow Evaluation, WCAP-8691, Revision 1 (Proprietary), July 1979.

38 n w -vox:

'Y P'-"

~-,y y.-

12. Davidson, S. L. (Ed.) et al., " Extended Burnup Evaluation of Westinghouse Fuel," WCAP-10125-P-A, December 1985.

13. Camden, T. M., et al., "PALADON-Westinghouse Nodal Computer Code,"

WCAP-9485-P-A, December 1979 and Supplement 1, September 1981.

14. Davidson, S. L. (Ed.), et al., "ANC:

Westinghouse Advanced Nodal Computer Code," WCAP-10965-P-A, September 1986..

15. Chelemer, H., Boman, L. H., Sharp, D. R., " Improved Thermal Design Procedure," WCAP-8567, July 1975.

16.

Letter from NRC to Westinghouse from Stolz to Eicheidinger, SER on WCAP-7956, 8054, 8567 and 8762 dated April 1978.

17.

Letter from E. P. Rahe (W) to Miller (NRC) dated March 19, 1982, NS-EPR-2573, WCAP-9500 and WCAPS 9401/9402 NRC SER Mixed Core Compatibility Items.

18.

Letter from C. O. Thomas (NRC) to Rahe (W) - Supplemental Acceptance No. 2 for Referencing Topical Report WCAP-9500, January 1983.

19. Motley, F.

E., et al., "New Westinghouse Correlation WRS-1, for Predicting Critical Heat Flux in Rod Bundles with Mixing Vane grids," WCAP-8762-P-A and WCAP-8763-A, July 1984.

20.

Kabadi, J. N., et. al., "The 1981 Version of the Westinghobse ECCS Evaluation Model Using the BASH Code," WCAP-10266, Revision 2, i

August 1986,(WestinghouseProprietary).

21.

Young, M., et. al., "BART-U.:

A Computer Code for the Best Estimate Analyzed Reflood Transients," WCAP-9561-P-A, 1984-(Westinghouse Proprietary).

l arso e-s7asca

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22. Lee, N., Rupprecht, S. D., Schwartz, W. R., Tauche, W. D.,

" Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code," WCAP-10054-P-A (Proprietary) and WCAP-10081-A (Non-Proprietary) August 1985.

23. Esposito, V. J.,

Kesavan, K., and Maul, B. J.; "W-FLASH-A Fortran-IV Computer Program for Simulation of Transients in a Multi-Loop PWR,"

WCAP-8200 (Proprietary), July 1973.

24. Meyer, P. E., "NOTRUMP, A Nodal Transient Small Break and General Network Code," WCAP-10079-P-A (Proprietary) and WCAP-10080-A (Non-Proprietary),

August 1985.

25. Letter C. E. Rossi (NRC) to E. P. Rahe, Jr. (W), " Acceptance for Referencing of Licensing Topical Report WCAP-10266," November 1986.

26. Letter W. Johnson E) to J. Lyons (NRC), " Submittal of WCAP-10266 Addendum 1, BASH Power Shape Sensitivity Studies," January 1987.40-37w e-amc2