ML18022A712

Application for Amend to License NPF-63,revising Tech Specs in Support of Refueling & Operation W/Vantage 5 Fuel Design. W/Safety Evaluation,Sample Version of Core Operating Limits Rept Using Cycle 2 Info & Radiological Impact Assessment

ML18022A712
Person / Time
Site:	Harris
Issue date:	04/17/1989
From:	Cutter A CAROLINA POWER & LIGHT CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML18022A713	List: ML18022A712 ML18022A714
References
NLS-89-087, NLS-89-87, NUDOCS 8904250290
Download: ML18022A712 (71)

Text

Carolina Power 8 Ught Company P.O. Box 1551 ~ Raleigh. N.C. 27602 SERIAL: NLS-89"087 10CFR50.90 A. B CUTTER Vice President Nuclear Services Oepartment United States Nuclear Regulatory Commission ATTENTION: Document Control Desk Washington, DC 20555 SHEARON HARRIS NUCLEAR POWER PLANT DOCKET NO. 50-400/LICENSE NO. NPF-63 REQUEST FOR LICENSE AMENDMENT CYCLE 3 RELOAD SUBMITTAL Gentlemen:

In accordance with the Code of Ferteral Regulations, Title 10, Parts 50.90 and 2.101, Carolina Power 6 Light Company (CP&L) hereby requests a revision to the Technical Specifications for the Shearon Harris Nuclear Power Plant, Unit 1 (SHNPP). This proposed change revises numerous Technical Specifications in support of refueling and operation for SHNPP with VANTAGE 5 improved fuel design.

The proposed Technical Specification (TS) changes for SHNPP Cycle 3 primarily result from changes in four areas'1) change in fuel design'2) use of improved analytical methodologies', 3) incorporation of some cycle-specific TS parameters in a Core Operating Limits Report; and 4) Cycle 2 administrative corrections. The change in each of these areas is presented below:

1) Chan e in Fuel Desi n The Shearon Harris Nuclear Power Plant is currently operating in Cycle 2 with a Westinghouse 17 x 17 low-parasitic (LOPAR) fueled core. For subsequent cycles, it is planned to refuel and operate the Shearon Harris Plant with the Westinghouse VANTAGE 5 improved fuel design, except for the inclusion of a Debris Filter Bottom Nozzle (DFBN) instead of the VANTAGE 5 bottom nozzle. As a result, future core loadings would range from approximately a 67X LOPAR fuel and 33X VANTAGE 5 fuel in the Cycle 3 transition core to eventually an all VANTAGE 5 fueled core. The VANTAGE 5 fuel assembly is designed as a modification to the current LOPAR (standard fuel) and optimized fuel assembly (OFA) designs.

The VANTAGE 5 design features were conceptually packaged to be licensed as a single entity. This was accomplished via the NRC and approval of the "VANTAGE 5 Fuel Assembly Reference Core Report,"

WCAP-10444-P-A. The initial irradiation of a fuel region containing all the VANTAGE 5 design features occurred in the Callaway Plant during the last quarter of 1987. The Callaway VANTAGE 5 licensing submittal was made to the NRC on March 31, 1987 (USNRC-1470, Docket 8904250290 8904i7 PDR ADOCK 05000400 IxoOl PDC

'e Document Control Desk NLS-89-087 / page 2 No. 50-483). Several of the VANTAGE 5 design features, such as axial blankets, reconstitutable top nozzles, extended burnup modified fuel assemblies, and Integral Fuel Burnable Absorbers have been successfully licensed as individual design features and are currently in operating Westinghouse plants. Also, Shearon Harris is operating in reload Cycle 2 with LOPAR fueL containing several VANTAGE 5 design features. These include the reconstitutable top nozzle, axial blankets, and the extended burnup modified fuel assembly. The Debris Filter Bottom Nozzle (DFBN) introduced in Cycle 2, while not a VANTAGE 5 feature, will continue to be used.

A brief summary of the VANTAGE 5 design features and their major advantages compared to the LOPAR fuel design are given below.'nte ral Fuel Burnable Absorber (IFBA) The IFBA features a zirconium diboride coating on the fuel pellet surface on the central portion of the enriched U02 fuel stack. In a typical reload core, approximately one third of the fuel rods in the feed region are expected to include IFBAs. IFBAs provide power peaking and moderator temperature coefficient control.

Intermediate Flow Mixer (IFM) Three IFM grids Located between the four upper most Zircaloy grids provide increased DNB margin.

Increased margin permits an increase in the design basis F<H and F~.

Reconstitutable To Nozzle A mechanical disconnect feature facilitates the top nozzle removal.

Extended Burnu The VANTAGE 5 fuel design will be capable of achieving extended burnups. Changes in the design of both the top and bottom nozzles increase burnup margins by providing additional plenum space and room for fuel rod growth (used in Cycle 2). The basis for designing to extended burnup is contained in the approved Westinghouse Topical WCAP-10125-P-A (used in Cycle 2).

Blankets The axial blanket consists of a nominal six inches of natural UO> pellets at each end of the fuel stack to reduce neutron leakage an8 to improve uranium utilization (used in Cycle 2) ~

Fuel Pellet Diameter - The VANTAGE 5 fuel rod has the same clad wall thickness as the LOPAR fuel rod, but the VANTAGE 5 fueL rod diameter is reduced to optimize the water-to-uranium ratio, resulting in improved uranium utilization.

2) Use of Im roved Anal tical Methodolo ies The analysis of the LOPAR and VANTAGE 5 fuel is based on the NRC approved Improved Thermal Design Procedure (ITDP). The LOPAR fuel analysis uses the WRB-1 DNB correlation, while the VANTAGE 5 fuel utilizes the WRB-2 correlation. The existing analyses for Cycle 2 uses the W-3 correlation. The new DNB correlations take credit for the significant improvement in the accuracy of the critical heat flux predictions over previous DNB correlations. The W-3 DNB correlation continues to be used for some accidents which are beyond the ranges

Document Control Desk NLS-89-087 / page 3 of applicability of the new correlations. As a result of the new DNB correlations'mproved accuracy, confidence at a 95/95 level that the limiting power rod will not experience DNB is provided with a limiting DNBR value of 1.17 versus the existing 1.30.

The current design method employed in Cycle 2 to verify that the DNBR limit is met involves the initiation of the Chapter 15 transients at the simultaneous worst possible conditions for core power, temperature, pressure, and flow, in addition to conservative application of uncertainties on fuel fabrication parameters. The resulting DNBR versus time for the transient is then compared against the limit (i.e., 1.30). The ITDP methodology used for Cycle 3 recognizes that this approach is overly conservative and statistically combines the effects of variations in these parameters on DNBR. The resulting statistically determined "delta DNBR" is applied to the 1.17 limit (including additional plant-specific DNBR margin) to yield a limiting safety limit DNBR value to which the transient results can be compared. Note, that since the safety limit DNBR value already includes allowances for operational, nuclear, and thermal variations, the transients are initiated from nominal conditions for those parameters')

Core 0 eratin Limits Re ort 4

The proposed Technical Specification (TS) changes include the removal of the Rod Insertion Limits (RILs) and Relaxed Axial Offset Control (RAOC) operational limits from the Technical Specifications. The proposed TSs retain the LCO wording and surveillance require'ments but refer to the specific limits being supplied to the NRC per the proposed revisions to TS 6.9.1.6 in a Core Operating Limits Reports This is similar to the current TSs which transmit cycle-specific values to the NRC via the Radial Peaking Factor Report in TS 6.9.1.6. Removal of these is consistent with the recent guidance provided by the NRC in Generic Letter 88-16. Removal of the limits from the TSs will allow flexibility to change rod parking position for control rod wear without the delay and cost associated with a TS submittal. It also allows a cycle-specific RAOC curve to be implemented according to approved methodology without the time and expense associated with a TS change. A sample version of the Core Operating Limits Report using Cycle 2 information is enclosed as Attachment 5.

4) C cle 2 Administrative Corrections Based on Technical Specifications approved by the NRC for SHNPP Cycle 2 operation, two additional administrative corrections are incorporated. The first administrative correction involves the deletion of Surveillance Requirement 4.2.1.1.a.2 requiiing monitoring and logging of indicated Axial Flux Difference (AFD) for a 24-hour period after the automatic computer monitoring is returned to an operable status. The second administrative correction involves the deletion in Action 5 of Technical Specification 3 '.1 (Table 3.3-1)

Document Control Desk NLS-89-087 / page 4 of a reference to Technical Specification 3.1.1.1 which is no longer applicable. provides the safety evaluation for VANTAGE 5 fuel which includes the mechanical, nuclear, thermal and hydraulic, and accident evaluations. In addition, Section 7.0 of this- attachment provides a summary of the Technical Specifications sections being changed. Attachment 2 provides the marked-up pages to those Technical Specifications being revised. Attachments 3 and 4 forward the Non Loss-of-Coolant Accident (LOCA) and LOCA analyses, respectively. Attachment 5 is a sample version of the Core Operating Limits Report using Cycle 2 information. The significant hazards evaluation pursuant to 10CFR50.92 is included as Attachment 6. Lastly, the radiological impact assessment is forwarded as Attachment 7.

Carolina Power & Light Company requests approval of the proposed amendment by October 11, 1989 in order to support the upcoming SHNPP refueling outage currently scheduled to begin in November 1989. Please refer any questions regarding this submittal to Mr. John Eads at (919) 546-4165.

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Enclosures cc: Mr. R. A.~ Becker

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Mr. W. H. Bradford Mr. Dayne H. Brown Mr. S. D. Ebneter A. B. Cutter, having been first duly sworn, did depose and say that the information contained herein is true and correct to the best of his information, knowledge and belief; and the sources of his information are officers, employees, contractors, and agents, of Carolina Power 6 Light Company.

Notary ( al My commission expires: g 1" 99

ATTACHMENT 1 SAFETY EVALUATION FOR THE SHEARON HARRIS NUCLEAR POWER PLANT TRANSITION TO WESTINGHOUSE 17 x 17 VANTAGE 5 FUEL (277CRS/cn )

TABLE OF CONTENTS Section ~Pa e

1.0 INTRODUCTION

2.0

SUMMARY

AND CONCLUSIONS 3.0 MECHANICAL EVALUATION 4.0 NUCLEAR EVALUATION 5.0 THERMAL AND HYDRAULIC EVALUATION 12 6.0 ACCIDENT EVALUATION 17 7.0

SUMMARY

OF TECHNICAL SPECIFICATION CHANGES 27

8.0 REFERENCES

33 LIST OF TABLES Table No ~ Title ~Pa e 3.1

~ Comparison of 17 x 17 LOPAR and 17 x 17 VANTAGE 5 Fuel Assembly Design Parameters 5.1 Thermal and Hydraulic Design Parameters 14 7.1 Summary and Justification for Technical 30 Specification Changes LIST OF FIGURES

~Fi use No. Title ~Pa e 3.1 17 x 17 VANTAGE 5/LOPAR Fuel Assembly Comparison 10 (277CRS/lah)

e 1

INTRODUCTION 4

The Shearon Harris Nuclear Power Plant is currently operating in Cycle 2 with a Westinghouse 17 x 17 low-parasitic (LOPAR) fueled core. For subsequent cycles, it xs planned to refuel and operate the Shearon Harris Plant with the Westinghouse VANTAGE 5 improved fuel design (as defined in Reference 1, except for inclusion of a Debris Filter Bottom Nozzle (DFBN) instead of the VANTAGE 5 bottom nozzle). As a result, future core loadings would range from approximately 67K LOPAR fuel and 33X VANTAGE 5 fuel in the Cycle 3 transition core to eventually an all VANTAGE 5 fueled core. The VANTAGE 5 fuel assembly is designed as a modification to the current 17 x 17 LOPAR (standard fuel) and the optimized fuel assembly (OFA) designs, Reference 2.

The VANTAGE 5 design features were conceptually packaged to be licensed as a single entity. This was accomplished via the NRC review and approval of the "VANTAGE 5 Fuel Assembly Reference Core Report,"

WCAP-10444-P-A, Reference 1. The initial irradiati'on of a fuel region containing all the VANTAGE 5 design features occurred in the Callaway Plant during the last quarter of 1987. The Callaway VANTAGE 5 licensing submittal was made to the NRC on March 31, 1987 (USNRC-1470, Docket No. 50-483). Several of the VANTAGE 5 design features, such as axial blankets, reconstitutable top nozzles, extended burnup modified fuel assemblies, and Integral Fuel Burnable Absorbers have been successfully licensed as individual design features and are currently in operating Westinghouse plants. Also, Shearon Harris is operating in reload Cycle 2 with LOPAR fuel containing several VANTAGE 5 design features.

These include the reconstitutable top nozzle, axial blankets, and the extended burnup modified fuel assembly. The Debris Filter Bottom Nozzle (DFBN) introduced in Cycle 2 while not a VANTAGE 5 feature will continue to be used.

Descriptions and evaluations of the VANTAGE 5 design features are presented in Section 3.0 of this evaluation report. A brief summary of the VANTAGE 5 design features and their major advantages compared to the LOPAR fuel design are given below.

Inte ral Fuel Burnable Absorber (IFBA) The IFBA features a zirconium diboride coating on the fuel pellet surface on the central portion of the enriched U02 fuel stack. In a typical reload core, approximately one third of the fuel rods in the feed region are expected to include IFBAs. IFBAs provide power peaking and moderator temperature coefficient control.

Intermediate Flow Mixer (IFM) Three IFM grids located between the four upper most Zircaloy grids (Figure 3.1) provide increased DNB margin.

Increased margin permits an increase in the design basis F>H and F~.

Reconstitutable To Nozzle A mechanical disconnect feature facilitates the top nozzle removal. Changes in the design of both the top and bottom nozzles increase burnup margins by providing additional plenum space and room for fuel rod growth.

(277CRS/lah)

d' Extended Burnu The VANTAGE 5 fuel design will be capable of achieving extended burnups. The basis for designing to extended burnup is contained in the approved Westinghouse Topical WCAP-10125-P-A, Reference 3.

Blankets The axial blanket consists of a nominal six inches of natural UO2 pellets at each end of the fuel stack to reduce neutron leakage and to improve uranium utilization.

This report shalL serve as a reference safety evaluation/analysis report for the transition from the present Shearon Harris LOPAR fueled core to an all VANTAGE 5 fueled core. In this report, the differences between the VANTAGE 5 and LOPAR fuel assembly designs are examined and the effect of these differences on the cores during the transition to an all VANTAGE 5 core is evaluated. The transition and VANTAGE 5 core evaluation/analyses were performed at a core thermal power level of 2775 MWt with the following conservative assumptions made in the safety evaluations: a full power F<H of 1.62 for the LOPAR fuel and 1.65 for the VANTAGE 5 fuel, an increase in the maximum F~ to 2.45, 6X plant steam'enerator. tube plugging for the LOCA analysis'nd 0% for the non-LOCA analysis, and a positive moderator temperature coefficient (PMTC) of +5 pcm/'F from 0 to 70X power and then decreasing linearly to 0 pcm/'F between 70 to 100% power.

This report utilizes the standard reload design methods described in Reference 4 and will be used as a basic reference document in support of future Shearon Harris Reload Safety Evaluations (RSEs) for VANTAGE 5 fuel reloads. 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 changes needed. Attachments 2 and 3 contain the Technical Specification change pages and non-LOCA safety analyses results, respectively. Attachment 4 contains the large and small break LOCA safety analyses.

Consistent with the Westinghouse standard reload methodology, Reference 4, parameters are chosen to maximize the applicability of the safety evaluations for future cycles. The objective of subsequent cycle-specific RSEs will be to verify that applicable safety limits are satisfied based on the reference evaluation/analyses established in this report.

In order to demonstrate early performance of the VANTAGE 5 design product features in a commercial reactor, four 17 x 17 VANTAGE 5 demonstration assemblies were loaded into the V. C. Summer Unit 1 Cycle 2 core and began power 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 the VANTAGE 5

'ssumes a maximum of 6X of steam generator tubes in each generator are plugged.

(277CRS/I QI1 )

components. Likewise, the IFM grids on the VANTAGE 5 demonstration assemblies had no efEect on the adjacent fuel assemblies. All four demonstration assemblies were reinserted into the V. C. Summer core for a second cycle of irradiation. This cycle was completed during March of 1987, at which time the demonstration assemblies achieved an average burnup of about 30,000 MWD/MTU. The observed behavior of the four demonstration assemblies at the end of two cycles of irradiation was as good as that observed at the end of the Eirst cycle of irradiation. The four assemblies were reinserted Eor a third cycle of irradiation.

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 Eor two reactor cycLes. Turkey Point Unit 4 contains 112 fuel rods equally distributed in four demonstration assemblies. The IFBA coating performed well with no Loss of coating integrity or adherence. The IFM grid feature has been demonstrated at McGuire Unit 1. The demonstration assembly at McGuire has been irradiated for two reactor cycles and has demonstrated good mechanical integrity.

(277CRS/I ah)

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2.0

SUMMARY

AND CONCLUSIONS Consistent with the Westinghouse standard reload methodology for analyzing cycle-specific reloads, Reference 4, parameters were selected to conservatively bound the values for each subsequent reload cycle and to facilitate determination of the applicability of 10CFR50.59. The objective of subsequent cycle-specific reload safety evaluations will be to verify that applicable safety limits are satisfied for the actual core designs under consideration. The reference evaluation/analyses established in this report will serve as the basis for those cycle-specific evaluations. The mechanical, thermal and hydraulic, 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 Shearon Harris core transitions to an all VANTAGE 5 core are given in the following sections of this submittal'he transition design and safety evaluations consider the following conditions: 2775 MWt core thermal power, 592.7'F core average temperature, 2250 psia system pressure, 292,800 gpm RCS thermal design flow, "and 299,700 gpm minimum measured flow. These conditions are used in core design and safety evaluations to justify safe operation with the conservative assumptions noted in Section 1.0. The conditions summarized in the SER for the VANTAGE 5 reference core report, WCAP-10444, have been considered in the Shearon Harris plant-specific safety evaluations.

The results of evaluation/analysis described herein lead to the following conclusions:

1. The Westinghouse VANTAGE 5 reload fuel assemblies for the Shearon Harris Nuclear Plant are mechanically compatible with the current LOPAR fuel assemblies, control rods, secondary source rods, and reactor internals interfaces. The VANTAGE 5/LOPAR fuel assemblies satisfy the current design bases for the Shearon Harris reactor.

2. Changes in the nuclear characteristics due to the transition from LOPAR to VANTAGE 5 fuel will be within the range normally seen from cycle to cycle due to fuel management effects.

3. The reload VANTAGE 5 fuel assemblies are hydraulically compatible with previously irradiated LOPAR fuel assemblies.

4. The core design and safety analyses results documented in this report show the core's capability for operating safely for the rated Shearon Harris Plant design thermal power with an F<< of 1.62 for LOPAR fuel and 1.65 for VANTAGE 5 fuel and an F of 2.45, steam generator tube plugging levels up to 6X for LOCI analyses and OX for non-LOCA analysis, and a positive MTC of +5 pcm/'F from 0 to 70X power and then decreasing linearly to 0 pcm/'F at 100K power.

5. Plant operating limitations given in the Technical Specifications will be satisfied with the proposed changes noted in Section 7.0 of this report. A reference is established upon which to base Westinghouse reload safety evaluations for future reloads with (277CRS/cn )

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3.0 MECHANICAL EVALUATION This section evaluates the mechanical design and the compatibility of the 17 x 17 VANTAGE 5 fuel assembly with the current LOPAR fuel assemblies during the transition through mixed-fuel cores to an all VANTAGE 5 core. The VANTAGE 5 fuel assembly has been designed to be compatible with the LOPAR fuel assemblies, reactor internals interfaces, the fuel handling equipment, and the refueling equipment. The VANTAGE 5 design dimensions as shown on Figure 3.1 are essentially equivalent to the LOPAR design from an exterior assembly envelope and reactor internals interface standpoint. The design basis and design limits are essentially the same as those for the LOPAR designs. As such, compliance with the "Acceptance Criteria" of the Standard Review Plan (SRP, NUREG 0800) Section 4.2, Fuel System Design, was fully demonstrated.

The significant new mechanical features of the VANTAGE 5 design relative to the initial core/Cycle 1 LOPAR fuel design include the following'.

Integral Fuel Burnable Absorber (IFBA)

Intermediate Flow Mixer (IFM) Grids Reconstitutable Top Nozzle Slightly longer fuel rods Axial Blankets Replacement of six intermediate Inconel grids with Zircaloy grids Reduction in fuel rod, guide thimble, and instrumentation tube diameter Redesigned fuel rod bottom end plug to facilitate reconstitution capability Table 3.1 provides a comparison of the VANTAGE 5 and LOPAR fuel assembly design parameters.

The Debris Filter Bottom Nozzle (DFBN) introduced in the Region 4/Cycle 2 LOPAR design will continue to be utilized.

The new mechanical features of the VANTAGE 5 design relative to the current Region 4/Cycle 2 LOPAR fuel design in operation are the following:

Integral Fuel Burnable Absorbers (IFBA)

Intermediate Flow Mixer (IFM) grids Slightly longer fuel rods Replacement of six intermediate Inconel grids with Zircaloy grids Reduction in fuel rod, guide thimble, and instrumentation tube diameters Fuel Rod Performance The VANTAGE 5 fuel rod has the same clad wall thickness as the LOPAR fuel rod, but the VANTAGE 5 fuel rod diameter is reduced to optimize the water-to-uranium ratio. The VANTAGE 5 fuel rod length is larger to provide a longer plenum and bottom end plug. The bottom end plug has an (277CRS/ I 0I1 )

II internal-grip feature to facilitate rod loading on both designs and is Longer to provide a longer Lead-in for the removable top nozzle reconstitution feature.

Fuel rod performance for all fuel rod designs is shown to satisfy the NRC SRP fuel rod design bases on a region-by-region basis. The same bases are applicable to all fuel rod designs, including the Westinghouse LOPAR and VANTAGE 5 fuel designs, with the only difference being that the VANTAGE 5 fuel is designed to achieve a higher burnup consistent with WCAP-LOL25-P-A, Reference 3, and VANTAGE 5 fuel is designed to operate with a higher F<H limit. The design bases for Westinghouse VANTAGE 5 fuel are discussed in Reference 1.

There is no effect from a fuel rod design standpoint due to having fuel with more than one type of geometry simultaneously residing in the core during the transition cycles. The mechanical fuel rod design evaluation for each region incorporates all appropriate design features of the region, including any changes to the fuel rod or pellet geometry from that of previous fuel regions.

The IFBA coated fuel pellets are identical to the enriched uranium dioxide pellets except for the addition of a thin neutron absorber coating on the pellet cylindrical surface. Coated pellets occupy the central portion of the fuel column. The number and pattern of IFBA rods within an assembly may vary depending on specific application. The ends of the enriched coated pellets and enriched uncoated pellets are dished to allow for axial expansion at the pellet centerline and void volume for fission gas release. Analysis of IFBA rods includes any geometry changes necessary to model the presence of the burnable absorber, and conservatively models the gas release from the coating. An evaluation and test program for the IFBA design features are given in Section 2.5 of Reference 1.

Fuel performance evaluations are completed for each fuel region to demonstrate 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 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'n References 2, 5, 6, 7, and 8 to demonstrate that the SRP fuel rod design criteria (including the rod internal pressure design basis in Reference 9) will be satisfied.

Grid Assemblies The top and bottom Inconel (nonmixing vane) grids of the VANTAGE 5 fuel assemblies are nearly identical in design to the Inconel grids of the LOPAR fuel assemblies. The only differences are: 1) the grid spring and dimple heights have been modified to accommodate the reduced diameter fuel rod and 2) the grid spring force has been reduced in the top grid. The six intermediate (mixing vane) structuraL grids are made (277CRS/I ah )

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of Zircaloy material rather than the Inconel 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 midspan flow mixing in the hottest fuel assembly spans. Each IFM grid cell contains four dimples which are designed to prevent midspan channel closure in the spans containing IFMs 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 to be similar to current fuel designs to preclude grid hang-up and damage during fuel handling. Additionally, the grid envelope is smaller which further minimizes the potential for

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'damage and reduces calculated forces during seismic/LOCA events. A eoolable geometry 'is, therefore, assured at the IFM grid elevation as well as at the structural grid elevation.

Mechanical Com atibilit of Fuel Assemblies Based on the evaluation of the VANTAGE 5/LOPAR design differences and the results (Reference 1), it is concluded that the two designs are mechanically compatible with each other. The VANTAGE 5 fuel rod mechanical design bases remain unchanged from those used for the LOPAR fuel assemblies in the Shearon Harris Cycle 2 core.

Rod Bow It is predicted that the 17 x 17 VANTAGE 5 rod bow magnitudes, like those of the Westinghouse LOPAR fuel, will be within the bounds of existing 17 x 17 LOPAR assembly rod bow data. The current NRC approved methodology for comparing rod bow for two different fuel assembly designs is given in Reference 10.

Rod bow in fuel rods containing IFBAs is not expected to differ in magnitude or frequency from that currently observed in Westinghouse LOPAR fuel rods under similar operating conditions. No indications of abnormal rod bow have been observed during visual or dimensional inspections performed on the test IFBA rods. Rod growth measurements were also within predicted bounds.

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 LOPAR and VANTAGE 5 fuel assembly designs employing different grids, there is an unequal axial pressure distribution between the assemblies. Cross flow resulting from this unequal pressure distribution was evaluated to determine the induced rod vibration and subsequent wear. Hydraulic tests (Reference 1, Appendix A.l.4) were performed to verify that the (277CRS/lnh)

cross flows were negligible and also to check hydraulic compatibility of the LOPAR and VANTAGE 5 designs. The VANTAGE 5 fuel assembly was flow tested adjacent to a 17 x 17 OFA,, since vibration test results indicated that the cross flow 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.l.4, revealed that the VANTAGE 5 fuel assembly wear characteristic was similar to that of the 17 x 17 OFA when both sets of data were normalized to the test duration time. It was concluded that the VANTAGE 5 fuel rod wear would be less than the maximum wear depth established, Reference 12, for the 17 x 17 OFA at EOL.

8 (277CRS/lah)

TABLE 3.1 COMPARISON OF 17 x 17 LOPAR AND 17 x 17 VANTAGE 5 FUEL ASSEMBLY DESIGN PARAMETERS 17 x 17 17 x 17 Parameter LOPAR Desi n VANTAGE 5 Desi n Fuel Assembly Length, in. 159.765 , 159.915 159.975 Fuel Rod Length, in. 151.56 p 152.16 152.285 Assembly Envelope, in. 8.426 8.426 Compatible with Core Internals Yes Yes Fuel Rod Pitch, in. 0.496 0.496 Number of Fuel Rods/Assembly 264 264 Number of Guide Thimble Tubes/

Assembly 24 24 Number of Instrumentation Tubes/

Assembly Fuel Tube Material Zircaloy 4 Zircaloy 4 Fuel Rod Clad OD, in. 0.374 0.360 Fuel Rod Clad Thickness, in. 0.0225 0.0225 Fuel/Clad Gap, mil. 6.5 6.5 Fuel Pellet Diameter, in. 0.3225 0.3088 Fuel Pellet Length Enriched Fuel, in. 0.530(1), 0.387(2) 0.370 Unenriched Fuel, in. 0.545 0.500 Guide Thimble Material Zircaloy 4 Zircaloy 4 Guide Thimble OD, in. 0.482 0.474 (1) Cycle 1/initial core fuel (2) Cycle 2/Region 4 fuel (277CRS/ I ah )

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159.975 3.475 2.383 152.285

~ 8.426 ~ 8. 426 S 8.372 ~ 8.418 CI Cl 0

I I 53:60 122.31 IOI .76 bl .2I 2 25 I I 52 l33.47 112.92 92.37 71.82 51.27 30.72 5 84 17XI7 VANTAGE 5 FUEL ASSEMBLY 159 765 / 159 9152 3.670 / 3.475 2.738 / 2.389 151.56 / 152.16

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CI 1.32 2 30.26 5.6IL 153.60 !33.01 I 12.46 91.91 71.36 50.6 I 153.96 '33.371 I I 2.82 92.27 71 .72 51 . I 7 I

30.62 6. I 90 I 7X 7 RECONSTITUTABLE LOPAR FUEL ASSEMBLY I

SHEARON HARRIS UNIT I I-CYCLE .Figure 3.1 2I'-CYCLE I7XI7 VANTAGE S /'QPAR FUEL ASSEMBLY COMPARISON

4.0 NUCLEAR EVALUATION The evaluation of the transition and equilibrium cycle VANTAGE 5 cores presented in Reference 1, as well as the Shearon Harris, specific transition core evaluations, demonstrate that the impact of implementing VANTAGE 5 does not cause a significant change to the physics characteristics of the Shearon Harris cores beyond the normal range of variations seen from cycle to cycle.

The methods and core models used in the Shearon Harris reload transition core evaluations are described in =References 1, 4, 11, 13, and 14.

These licensed methods and models have been used for Shearon Harris and other previous Westinghouse reload designs using the OFA and VANTAGE 5 fuel. No changes to the nuclear design philosophy, methods, or models are necessary because, of the transition to VANTAGE 5 fuel.

From the nuclear design area, the following Shearon Harris Technical Specification changes are proposed:

1)

Increased F<H limits.'hese higher limits will allow loading pattern designs with reduced leakage which in turn will allow longer cycles.

2) Increased F~ limit. The increased F limit will provide greater flexibility with regard to accommodating the axially heterogeneous cores (blankets and short burnable absorbers).

3) It is proposed that the current Axial Flux Differ'ence (AFD) operating spaces for Relaxed Axial Offset Control (RAOC), Base Load Operations and Control Bank Insertion Limits be deleted from the Technical Specifications and instead be incorporated in the Core Operating Limits Report. This eliminates the potential necessity of Technical Specification amendments for future reload cycles, while providing adequate assurance that the correct operating limits will be followed.

Power distributions and peaking factors show slight changes as a result of the incorporation of reduced length IFBAs and increased peaking factor limits, in addition to the normal variations experienced with different loading patterns. The usual methods of enrichment variation and burnable absorber usage can be employed in the transition, and full VANTAGE 5 cores to ensure compliance with the peaking factor Technical Specifications.

The key safety parameters evaluated for the Shearon Harris reactor as it transitions to an all VANTAGE 5 core show little change relative to the range of parameters experienced for the all LOPAR fuel core. The changes in values of the key safety parameters are typical of the normal cycle-to-cycle variations experienced as loading patterns change. 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.

(277CRS/Iah)

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~.o THERMAL AND HYDRAULIC EVALUATION The analysis of the LOPAR and VANTAGE 5 fuel is based on the Improved Thermal Design Procedure (ITDP) described in Reference 15. The LOPAR fuel analysis uses the WRB-1 DNB correlation in Reference 17, while the VANTAGE 5 fuel utilizes the WRB-2 DNB correlation in Reference 1. These DNB correlations take credit for the significant improvement in the accuracy of the critical heat flux predictions over previous DNB correlations. The WRB-2 DNB correlations also takes credit for 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. In addition, the W-3 DNBR correlation is used where appropriate (e.g., accidents analyzed in Sections 15.2.1 and 15.4.2.1 of Attachment 3). Table 5.1 summarizes the pertinent thermal and hydraulic design parameters.

The design method employed to meet the DNB design basis is the 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 95 percent probability at 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 uncertainties. These DNBR uncertainties, combined with the DNBR limit, establish 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 DNBR values for the LOPAR fuel analysis are 1.35 for thimble cold wall cells (three fuel rods and a thimble tube) and 1.36 for typical cells (four fuel rods). The design DNBR values for the VANTAGE 5 fuel are 1.32 and 1.33 for thimble and typical cells, respectively.

, In addition to the above considerations, a plant-specific DNBR margin has'een considered in the analyses. In particular, safety analysis DNBR limits of 1.46 for thimble and 1.48 for typical cells for LOPAR fuel and 1.60 for both 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 follows. A fraction of the margin is utilized to accommodate the transition core penalty (12.5% for VANTAGE 5 fuel and none for LOPAR fuel) and the appropriate fuel rod bow DNBR penalty, Reference 10, which is less than 1.3%. The existing 7.5% margin in the LOPAR fuel and 16.9% margin in the VANTAGE 5 fuel between the design and safety analysis DNBR limits also includes a greater than 6.2% DNBR margin in the LOPAR fuel and a greater than 3.1% DNBR margin in the VANTAGE 5 fuel reserved for flexibility in the design.

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

The phenomena of fuel rod bowing, as described in Reference 10, must be accounted for in the DNBR safety analysis of Condition I and Condition II events for each plant application. Internal to the fuel 12 (277CRS/I Gh )

rod, the IFBA and fuel pellet designs are not expected to increase the propensity for fuel rods to bow. External to the VANTAGE 5 fuel rod, the Inconel nonmixing vane and Zircaloy mixing vane grids provide fuel rod support. Applicable generic credits for margin resulting from retained conservatism in the evaluation of DNBR are used to offset the effect of rod bow (Reference 18). The safety analysis for the Shearon Harris plant maintains sufficient margin between the safety analysis limit DNBR and the design limit DNBR to accommodate full flow and low flow DNBR penalties.

The Westinghouse transition core DNB methodology is given in References 2 and 19 and has been approved by the NRC via Reference 20.

Using this methodology, transition cores are analyzed as if they were full cores of one assembly type (full LOPAR or full VANTAGE 5), applying the applicable transition core penalties (References 21 through 23).

This penalty is included in the safety analysis limit DNBRs such that sufficient margin over the design limit DNBR exists to accommodate the transition core penalty and the appropriate rod bow DNBR penalty.

The fuel temperatures for use in safety analysis calculations for the VANTAGE 5 fuel are evaluated using the same methods as those used to evaluate the LOPAR fuel. Westinghouse uses the performance code described in Reference 5 to perform both design and licensing calculations. When the code is used to calculate fuel temperatures to be used as initial conditions in safety analyses, a conservative thermal safety model, Reference 6, is used.

13 (277CRS/Iah)

/t TABLE 5.1 THERMAL AND HYDRAULIC DESIGN PARAMETERS Design Thermal and H draulic Desi n Parameters Parameters (Using ITDP)

Reactor Core Heat Output, MW[ 2775 Reactor Core Heat Output, 10 BTU/Hr. 9469 Heat Generated in Fuel., X 97.4 Core Pressure, Nominal, psia 2280 Radial Power Distribution (LOPAR) 1 '6 [1+0 30(l P)]'c (v-5) 1.59 [1+0.35(1-P)]'K Limit DNBR for Design Transients Typical Flow Channel (LOPAR) 1.48 (v-5) 1.60 Thimble (Cold Wall) Flow Channel (LOPAR) 1.46 (v-5) 1.60 DNB Correlation -:< (LOPAR) WRB-1 (v-s) WRB-2 The 4% radial power uncertainty has been removed for statistical combination with other uncertainties in the ITDP analysis.

"--" See Section 5.0 for the use of the W-3 DNB correlation.

14 (277CRS/IQh)

TABLE 5.1 (Continued)

THERMAL AND HYDRAULIC DESIGN PARAMETERS Design Thermal H draulic Parameter Value Vessel Minimum Measured Flow Rate (including Bypass), 10 ibm/hr. 111.4 GPM 299,700 Vessel Thermal Design Flow Rate (including Bypass), 10 ibm/hr. 108.9 GPM 292,800 Core Flow Rate (excluding Bypass, based on Thermal Design Flow) 10 ibm/hr. 101.7 GPM 273,475 Fuel Assembly Flow Area for Heat Transfer, ft. 2 (LOPAR) 41.55 (V-5) 44.04 Cor~ Inlet Mass Velocity, 10 ibm/hr-ft (based on TDF) (LOPAR) 2.45 (V-5) 2.31 Assumes all LOPAR or VANTAGE 5 core.

15 (277CRS/cn )

TABLE 5.1 (Continued)

THERMAL AND HYDRAULIC DESIGN PARAMETERS Design Thermal H draulic Parameter Value (Based on Thermal Design Flow)

Nominal Vessel/Coze Inlet Temperature, 'F 557.4 Vessel Average Temperature, 'F 588.8 Core Average Temperature, 'F 592.7 Vessel Outlet Temperature, 'F 620.2 Average Temperature Rise in Vessel, 'F 62.8 Average Temperature Rise in Core, 'F 66.7 Heat Transfer Active Heat Transfer Surface Area, ft. (LOPAR) 48,598 (V-5) 46,779 Average Heat Flux, BTU/hr-ft (LOPAR) 189,820 (V-5) 197,200 Average Linear Power, kw/ft. 5.45 Peak Linear Power for Normal Operation, kw/ft. 13.40 Assumes all LOPAR or VANTAGE 5 core.

Based on 2.45 F~ peaking factor.

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ACCIDENT EVALUATION Non-LOCA Accidents This section addresses the impact of the VANTAGE 5 design features and modified safety analysis assumptions on the Shearon Harris non-LOCA accident analyses.

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

Fuel Rod Dimensions Integral Fuel Burnable Absorbers (IFBAs)

Intermediate Flow Mixer Grids (IFMs)

A brief description of each of these and its consideration in the safety analyses follows.'uel Rod Dimensions The VANTAGE 5 fuel rod dimensions which determine the safety analysis temperature versus linear power density relationship include rod diameter, pellet diameter, initial pellet-to-clad gap size, and stack height. The non-LOCA safety analysis fuel temperature and rod geometry assumptions consider this geometry change and bound both LOPAR (standard) and VANTAGE 5 fuel.

IFBAs IFBAs increase DNB margin by flattening the core power distribution. 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 effect of IFBAs on the reload safety analysis parameters is taken into account in the reload design process. The axial power distribution assumption in the safety analyses kinetics calculations has been determined to be applicable for evaluating the introduction of IFBAs in the Shearon Harris plant.

IFM Grids The IFM grid feature of the VANTAGE 5 fuel design increases DNB margin. The fuel safety analysis limit DNBR values contain significant DNBR margin (see Section 5.0). This DNB margin was set to ensure that the core thermal safety limits for the VANTAGE 5 fuel with an F H of 1.65 are acceptable. However, for the transition cycles, t e LOPAR fuel core limits with an F<H of 1.62 are more restrictive than the VANTAGE 5 fuel core limits. Thus, the most restrictive core limits correspond to the LOPAR fuel design. Any transition core penalty is accounted for with the available DNB margin.

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~JCql g1 1

The IFM grid feature of the VANTAGE 5 fuel design increases the core pressure drop. The control rod scram time to the dashpot is increased from 2.2 to 2.7 seconds. The increased drop time primarily effects 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.

Modified Safety Analysis Assumptions Listed below are the analysis assumptions which represent a departure from those currently used for Shearon Harris.

Increased Design Enthalpy Rise Hot Channel Factors (F>H) and FQ Increased F<H Part Power Multipliers Improved Thermal Design Procedure K(z) Third Line Segment Removal A brief description of each of these assumptions follows.

Increased Design Enthalpy Rise Peaking Factors (F>H) and FQ The F<H for the LOPAR and VANTAGE 5 fuel is 1.62 and 1.65 respectively. The non-LOCA calculations applicable for the VANTAGE 5 core have assumed a full power F<H of 1.65. This is a conservative safety analysis assumption for this report.

The design core limits for this report incorporate the increased F<H for both the LOPAR and VANTAGE 5 fuel.

The increase in the Technical Specification maximum LOCA FQ from 2.32 to 2.45 for both LOPAR and VANTAGE 5 fuel is conservatively accounted for in the non-LOCA transients.

Increased F<H Part Power Multipliers The F part power multipliers are 0.35 for VANTAGE 5 fuel and 0.30 for L "AR fuel. The revised values have been considered in the generation of the core limits for both fuel types. The Overtemperature Delta Temperature/Overpressure Delta Temperature (OTdT/OPAT) setpoints assumed in the non-LOCA analyses conservatively bound both sets of core limits. Furthermore, DNB-related events which do not trip on the OTAT/OPdT setpoints have conservatively accounted for the increase in the part power multipliers in the safety analyses performed for this report.

Im roved Thermal Desi n Procedure (ITDP)

The calculational method utilized to meet the DNB design basis is the ITDP which is discussed in Reference 15. Uncertainties in plant operating parameters are statistically treated such that there is at least a 95 percent probability at a 95 percent confidence level that the minimum DNBR will be greater than 1.17. Since the parameter 18 (277CRSicn )

uncertainties are considered in determining the design DNBR value, the plant safety analyses are performed using nominal input parameters without uncertainties.

The LOPAR fuel DNB analyses use the WRB-1 correlation, while the VANTAGE 5 fuel analyses use the WRB-2 correlation. The W-3 correlation was used in the analysis of the events in Sections 15.2.1 and 15.4.2.1 of Attachment 3.

K(z) Third Line Se ment Removal Elimination of the third line segment of the K(z) curve could impact the non-LOCA safety analyses assumptions for power distribution limits. However, nuclear design calculations have confirmed that the power distributions assumed in the safety analyses will be ensured with the adherence to the proposed K(z) curve. Therefore, this change to the Shearon Harris Technical Specifications is acceptable with respect to the non-LOCA safety analyses.

Non-LOCA Safety Evaluation Methodology The non-LOCA safety evaluation 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 reevaluation. This methodology is applicable to the evaluation of VANTAGE 5 transition and full cores.

Any required reevaluation identified by the reload methodology is one of two 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 made which conservatively "evaluates the magnitude of the effect and explains why the actual analysis of the event does not have to be repeated. Alternatively, should the deviation be large and/or expected to have a significant 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.

The key safety parameters are documented in Reference 4. Values of these safety parameters which bound both fuel types (LOPAR and 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. Reevaluation of the affected transients would take place and would be documented for the cycle-specific reload design, as per Reference 4.

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4*

6.1.4

~ ~ Conclusions Descriptions of the transients reanalyzed for this report, method of analysis, results, and conclusions are contained in Attachment 3.

The analytical procedures and computer codes used are identified in Section 15.1. Attachment 3 has been prepared conforming to the format of the Shearon Harris FSAR.

For each of the accidents reanalyzed, it was found that the appropriate safety criteria are met. In addition, an evaluation has been performed regarding the impact of VANTAGE 5 fuel and the modified safety analysis assumptions on the Steam Line Break Mass and Energy Release analysis inside containment. The results of this evaluation verify that the mass and energy releases previously calculated, are not adversely impacted by the transition to VANTAGE 5 fuel, or by the modified safety analysis assumptions.

6.2 LOCA Accidents 6.2.1 Large Break LOCA 6.2.1.1 Description of Analysis/Assumptions for 17 x 17 VANTAGE 5 Fuel The large break Loss-Of-Coolant Accident (LOCA) analysis for the Shearon Harris Nuclear Power Plant, applicable to a full core of VANTAGE 5 fuel assemblies, was performed to develop Shearon Harris specific peaking factor limits. This is consistent with the methodology employed in the Reference Core Report for 17 x 17 VANTAGE 5 fuel, Reference 1. The Westinghouse 1981 Evaluation Model with BASH, Reference 24, was utilized and a spectrum of cold leg breaks was analyzed for Shearon Harris. Other pertinent analysis assumptions include: a core thermal power of 2775 MWt, 6% uniform steam generator tube plugging, an F>H of 1.65 and fuel data based on the revised thermal model, Reference 6. The analysis results, tables, and figures are presented in Attachment 4.

VANTAGE 5 fuel features, as applied to the Shearon Harris Nuclear Power Plant, result in a fuel assembly that is more limiting than LOPAR fuel currently in the core with respect to large break LOCA ECCS performance, Reference l. As such, VANTAGE 5 fuel has been analyzed herein.

6.2.1.2 Method of Analysis The methods used in analyzing the Shearon Harris Nuclear Power Plant for VANTAGE 5 fuel, including computer codes used and assumptions, are described in detail in Attachment 4, Section 15.6.5.3 ' ~

6.2.1.3 Results

,The results of this analysis, including tabular and plotted results of the break spectrum analyzed, are provided in Attachment 4, which has been prepared using the NRC Standard Format and Content Guide,

'egulatory Guide 1.70, Revision 1, for accidents applicable to the Shearon Harris plant.

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Reference 24 states three restrictions related to the use of the 1981 Evaluation Model + BASH calculational model. The application of these restrictions to the plant-specific large break LOCA analysis was addressed with the following conclusions.

Shearon Harris is neither an Upper Head Injection (UHI) nor an Upper Plenum Injection (UPI) plant, so Restriction 1 does not apply.

Regarding Restriction 2, Reference 25 reports that a maximum safety injection assumption can potentially be limiting only for four loop, non-burst node limited analyses. It is noted that Shearon Harris specific LOCA analyses with minimum safeguards ECCS assumptions has resulted in a burst-node peak clad temperature. Any additional water added to the system will result in more rapid delivery of water to the burst-node elevation, earlier quench, and a lower peak clad temperature. In all previous applications of BASH to three loop plants modeling both minimum and maximum safety injection, the results have consistently demonstrated that minimum safety injection has been limiting (higher peak clad temperature). Based on these trends in analyses to date and noting the burst-node limited scenario exhibited here, it has been confirmed that the limiting large break scenario for the Shearon Harris plant is represented in the analysis reported in Attachment 4.

Generic sensitivity studies were performed by Westinghouse for a typical three loop plant using different power shapes. This sensitivity study (i.e., Reference 26) demonstrated that the chopped cosine was the most limiting power shape. A chopped cosine power shape was used in the large break LOCA analysis for the Shearon Harris plant, thus satisfying Restriction 3.

Conclusions The large break LOCA analysis performed for the Shearon Harris Power Plant has demonstrated that for breaks up to a double-ended severance of the reactor coolant piping, the Emergency Core Cooling System (ECCS) will. meet the acceptance criteria of Title 10 CFR Part 50, Section 46, that is:

1. The calculated peak cladding temperature wi:11 remain below the required 2200'F.

2. The amount of fuel cladding that reacts chemically with the water or steam does not exceed one percent of the total fuel rod cladding.

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

4. The core remains amenable to cooling during and after the LOCA.

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

21 (277CRS/I ah )

1* WW 5 Il lp T

f

The time sequence of events for all breaks analyzed is shown in Tables 15.6.5-1 and 15.6.5-4 of Attachment 4, Section 15.6.5.

The large break LOCA analysis for the Shearon Harris Power Plant, assuming a full core of VANTAGE 5 fuel (utilizing the 1981 evaluation model + BASH calculational model), resulted in a peak cladding temperature of 2106'F for the limiting Double-Ended Cold Leg Guillotine (DECLG) break at a total peaking factor of 2.45. The maximum local metal-water reaction was 8.025%, and the total core-wide metal-water reaction was less than 0.3% for all cases analyzed. The clad temperature transients turn around at a time when the core geometry was still amenable to cooling.

The impact of the transition core cycles are conservatively evaluated to be at most 50'F higher in calculated peak cladding temperature (PCT) which would yield a transition core PCT of 2156.0'F. The transition core penalty can be accommodated by the margin to the 10CFR50.46 limit of 2200'F.

'It can be determined from the results contained in Attachment 4, Section 15.6.5, that the ECCS analysis for the Shearon Harris Power Plant remain in compliance with the requirements of 10CFR50.46 including consideration for transition core configurations.

Small Break LOCA Description of Analysis/Assumptions for 17 x 17 VANTAGE 5 The small break loss-of-coolant accident (LOCA) was analyzed assuming a full core of VANTAGE 5 fuel to determine the peak clad temperature. This is consistent with the methodology employed in WCAP-10444-P-A, Reference 1, for 17 x 17 VANTAGE 5 fuel transition. The currently approved NOTRUMP Model Small Break ECCS Evaluation Model, Reference 27, was utilized for a spectrum of cold leg breaks. Attachment 4, 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 F>< of 1.65, total peaking factors corresponding to 2.5 at the core mid-plane, 6% uniform steam generator tube plugging, and the core thermal power level of 2775 MWt.

Sensitivity studies performed using the NOTRUMP small break evaluation model have demonstrated that VANTAGE 5 fuel is more limiting than OFA fuel in calculated ECCS performance. Similar studies using the WFLASH evaluation model, Reference 28, have previously shown that OFA fuel is more limiting than LOPAR fuel.

For the small break LOCA, the effect of the fuel difference is more pronounced during core uncovery periods and, therefore, shows up predominantly 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 more limiting of the two types of fuel residing in the core.

22 (277CRS/I ah)

41! vi Pi If l$ ad&

6.2.2.2

~ ~ ~ Method of Analysis The methods of analysis, including codes used and assumptions, are described in detail in Attachment 4, Section 15.6.5.3.1.

6.2.2.3 Results The results of this analysis, including tabular and plotted results of the break spectrum analyzed, are provided in Attachment 4.

6.2.2.4 Conclusions The small break VANTAGE 5 LOCA analysis for the Shearon Harris Nuclear Power Plant, utilizing the currently approved NOTRUMP evaluation model, resulted in a peak clad temperature of 1780'F for the 3.0-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 elevation and 2.388 at the peak power elevation. The maximum local metal-water reaction is 2.258 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 is still amenable to cooling.

Analyses presented in Attachment 4, Section 15.6.5 show that one centrifugal pump and one low head pump, together with the accumulators, provide sufficient core flooding to keep the calculated peak 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.

6.2.3 Transition Core Effects on LOCA When assessing the impact of transition cores on the LOCA analysis, it must be determined whether the transition core can have a greater calculated peak cladding temperature (PCT) than either a complete core of the LOPAR fuel 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.

Hydraulic resistance mismatch will exist only for a transition core and is the only unique difference between a complete core of either fuel type and the transition core.

6.2.3.1 Large Break LOCA The large break LOCA analysis was performed with a full core of VANTAGE 5 fuel and conservatively applies the blowdown results to transition cores'he VANTAGE 5 fuel differs hydraulically from the LOPAR assembly design it replaces. The difference in the total assembly hydraulic resistance between the two designs is approximately 10X higher for VANTAGE 5.

23 (277CRS/I ah )

An evaluation of hydraulic mismatch of approximately 10X showed an insignificant effect on blowdown cooling during a LOCA. The SATAN-VI computer code models the cross flows between the average core flow channel (N-1 fuel assemblies) and the hot assembly flow channel (one fuel 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 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 during blowdown for this evaluation. The results of this evaluation have shown that no peak cladding temperature penalty is observed during blowdown for the mixed core. Therefore, it is not necessary to perform a blowdown calculation for the VANTAGE 5 transition core configuration because the evaluation model blowdown calculation performed for the full core of VANTAGE 5, fuel is conservative and bounding.

'The other portion of the LOCA evaluation model impacted by the hydraulic resistance mismatch is the core reflood transient.

Analyses have been performed which accurately model mixed core cases during reflood. As expected, the increase in hydraulic resistance mismatch for the VANTAGE 5,fuel assembly with its intermediate flow mixing (IFM) grids was shown to produce a reduction in reflood steam flow rate for the VANTAGE 5 fuel assemblies during the transition period. This reduction in steam flow resulted in a peak clad temperature increase of up to 50'F for core axial elevations where PCTs can possibly occur. Therefore, the maximum PCT penalty possible for Westinghouse VANTAGE 5 fuel during transition cores is 50'F. Once a full core of VANTAGE 5 fuel is achieved, the large break LOCA analysis will apply without the cross flow penalty.

6.2.3.2 Small Break LOCA The NOTRUMP computer code, Reference 29, is used to model the core hydraulics during a small break event. Only one core flow channel is modeled in NOTRUMP since the core flow during a small break is relatively slow and this provides enough time to maintain flow equilibrium between fuel assemblies (i.e., no cross flow).

Therefore, hydraulic resistance mismatch is not a significant 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 bounding for all transition cycles.

6.2.4 Containment Integrity Mass and Energy Releases The effects that design changes to the reactor fuel can have on containment mass and energy releases, used to determine containment peak pressure, are dependent upon:

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I

1) The change in core fluid volume as a result of the new fuel design.

2) Increase or decrease in core stored energy.

3) Effect of the new fuel design on reflood flooding rates as a result of core flow area or hydraulic resistance changes.

The VANTAGE 5 fuel design utilizes a fuel rod of smaller diameter than the 17 x 17 LOPAR fuel presently installed in the Shearon Harris Nuclear Power Plant. This smaller fuel rod diameter leads to a reduction in core stored energy which is beneficial in reducing the mass and energy releases calculated for a hypothetical LOCA. The small VANTAGE 5 fuel rod will slightly increase core fluid volume and the use of intermediate flow mixing grids will increase hydraulic resistance, but these changes are offset by the reduction in core stored energy. Thus, the implementation of VANTAGE 5 fuel at the Shearon Harris Nuclear Power Plant will not result in an increase in the containment peak pressure reported in

the Shearon Harris FSAR or increase the off-site radiological consequences associated with high containment pressures resulting from a hypothetical LOCA. Based on this evaluation, a reanalysis of containment integrity mass and energy releases was deemed unnecessary for the implementation of VANTAGE 5 fuel at the Shearon Harris Nuclear Power Plant.

6.2.5 Steam Generator Tube Rupture 6.2.5.1 Introduction Design basis analyses of a steam generator tube rupture event at the Shearon Harris Nuclear Power Plant have been performed to assess the effect of the transition to a core with VANTAGE 5 fuel assemblies.

The analyses performed include a demonstration of margin to steam generator overfill in the event of a tube rupture and an analysis which demonstrates that the calculated offsite radiation doses are within the limits set forth in 10 CFR 100.

The analyses performed bound operation of the Shearon Harris Nuclear Power Plant at an NSSS power of 2785 MWt with a Standard Fuel/VANTAGE 5 fuel transition core, Standard fuel core, or VANTAGE 5 fuel core with up to 6X uniform steam generator tube plugging. Since the assumption that the initial. primary coolant is at the Standard Technical Specification limit will not change for Shearon Harris due to the proposed change in fuel, the parameters which impact the offsite radiation doses calculated for the FSAR SGTR analysis are primary to secondary break flow and the steam released from the ruptured steam generator to the atmosphere.

Therefore, the analyses to support the transition to VANTAGE 5 fuel assess the impact of the fuel change on primary to secondary break flow and steam released via the ruptured steam generator.

25 (277CRS/ I ah )

i'J V

Methodology The steam generator tube rupture analyses were performed for Shearon Harris using methodology and assumptions described in WCAP-11703 (Reference 31). Plant response to the event was modeled using the LOFTTR2 computer code with conservative assumptions of break size and location, condenser availability and initial secondary water mass in the ruptured steam generator. The analysis methodology includes the simulation of the operator actions for recovery from a steam generator tube rupture event based on the Shearon Harris Nuclear Power Plant Emergency Operating Procedures, which were developed from the Westinghouse Owners Group Emergency Response Guidelines.

Since the limiting single failure is different for the overfill analysis and the offsite radiation dose analysis, the two analyses were performed using different single failure assumptions. For the margin to overfill analysis, the single failure was assumed to be the failure of an intact steam generator Power Operated Relief Valve (PORV) to open on demand when cooldown of the Reactor Coolant System (RCS) is initiated. In the offsite radiation dose analysis, the ruptured steam generator PORV was assumed to fail open when the isolation of the ruptured steam generator is performed.

Results The LOFTTR2 analysis to determine the margin to overfill was performed for the time period from the steam generator tube rupture until the primary and secondary pressures are equaLized and break flow is terminated. The water volume in the secondary side of the ruptured steam generator was calculated as a function of time to demonstrate that overfill does not occur. The results of the analysis demonstrate that the transition to VANTAGE 5 fuel does not change the conclusion that there is margin to overfill calculated for "the Shearon Harris Nuclear Power Plant in the event of a tube rupture.

For the offsite radiation dose analysis, the primary to secondary break flow and the steam release to the atmosphere from both the ruptured and intact generators were calculated for use in determining the activity released to the atmosphere. The mass releases were calculated with the LOFTTR2 program from the initiation of the event until termination of the break flow. For the time period following break flow termination, steam releases from and feedwater flows to the ruptured and intact steam generators were determined from a mass and energy balance using the calculated RCS and steam generator conditions when primary to secondary tube leakage was terminated. The mass release information was used to calculate the radiation doses at the exclusion area boundary and low population zone assuming that the primary coolant activity is at the Standard Technical Specification limit prior to the accident. The results of the analysis to support the transition to VANTAGE 5 fuel show that the offsite doses for Shearon Harris are well within the

'llowable guidelines specified in the Standard Review Plan, NVREG-0800, Section 15.6.3, and 10 CFR 100.

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4* I 4

I 4

(

A more complete description of the Steam Generator Tube Rupture Offsite Radiation Dose analysis and results is provided in Attachment 4, which shows the recommended modifications to the Shearon Harris FSAR Chapter 15.6.3.

Blowdown Reactor Vessel and Loop Forces The forces created by a hypothetical break in the RCS piping are principally caused by the motion of the decompression wave through the RCS. The strength of the decompression wave is primarily a result of the assumed break opening time, break area, and RCS operating conditions of power, temperature, and pressure. The small changes in these parameters caused by a change in fuel from 17 x 17 LOPAR to VANTAGE 5 and assuming thimble plugs deleted will not significantly alter the decompression wave as calculated. The forces in the vicinity of the core are affected by the core flow area and volume. An increase in core flow area or volume will tend to more effectively- dissipate the decompression wave resulting in a

"'reduction of the forces acting on the reactor vessel internals.

VANTAGE 5 fuel, having a smaller rod diameter than LOPAR fuel, increases the core flow area and volume which is beneficial in reducing forces associated with a hypothesized LOCA. Forces acting on the RCS loop piping as a result of a hypothesized LOCA are not influenced by changes in fuel assembly design. Thus, the implementation of VANTAGE 5 fuel at the Shearon Harris Nuclear Power Plant will not result in an increase of the calculated consequences of a hypothesized LOCA on the reactor vessel internals or RCS loop piping. The current FSAR analysis for forces on the reactor internals and RCS piping resulting from a hypothesized LOCA is considered to be bounding to the application of VANTAGE 5 fuel at the Shearon Harris Nuclear Power Plant.

Post-LOCA Long-Term Core Cooling (ECCS Flows, Core Subcriticality, and Switchover of the ECCS to Hot Leg Recirculation)

The implementation of VANTAGE 5 fuel at the Shearon Harris Nuclear Power Plant does not impact the assumptions for decay heat, core reactivity, or boron concentration for sources of water residing in the containment sump Post-LOCA. Thus, these licensing requirements associated with LOCA are not significantly affected by the implementation of VANTAGE 5 fuel. Additionally, Westinghouse performs an independent check on core subcriticality for each fuel cycle operated at the Shearon Harris Plant.

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~.o

SUMMARY

OF TECHNICAL SPECIFICATION CHANGES i

The proposed changes t o the Shea ron Harr s Nuclear Power Plant Techn i ca 1 S pec i fi f ca t ion s are summarized in Table 7 . 1 . These changes re 1 ec t the f

impact o the design , analytical methodology, and safety analysis assumptions outlined in this document and are given in the pro po s ed Technical Specification change pages ( s ee Attachment 2 of this i i fi report) . A brief overvi ew of the s gn can t changes fo 1 ows . l 7.1 Core Safety Limits Core safety limits and associated bases for three-loop operation during Modes 1 and 2 are revised to reflect the impact of the transition to VANTAGE 5 fuel with:

1. The use of ITDP and the WRB-1 and WRB-2 DNB correlations.

2. An F<H of 1.62 for LOPAR fuel and 1.65 for VANTAGE 5 fuel (see Section 7.7).

3. An F<H part power multiplier 'of 0.30 for LOPAR fuel and 0.35 for VANTAGE 5 fuel.

limits correspond to those for the LOPAR fuel which are

'he proposed limiting during the transition period. Less limiting values will be possible with a full core of VANTAGE 5 fuel.

OPQT/OTAT Setpoints Revisions to the limiting safety system settings for the thermal overpower dT and overtemperature AT trip functions are proposed to maintain consistency with the non-LOCA accident analyses. These trip functions provide primary protection against departure from nucleate boiling and fuel centerline melting (excessive kw/ft) during postulated transients.

7.3 Rod Drop Time The VANTAGE 5 guide thimbles are identical to those in the LOPAR design except for a reduction in the guide thimble diameter and length above the dashpot. The reduction to the guide tube diameters is required due to the thicker Zircaloy grid straps and reduced cell size; whereas, the VANTAGE 5 thimble tube is shorter due to the reconstitutable top nozzle feature. To accommodate these changes, the scram time to the dashpot for accident analyses is increased from 2.2 seconds to 2.7 seconds for the transition to VANTAGE 5 fuel.

7.4 Core Operating Limits Report Current Technical Specifications contain certain operating limits which could change from cycle to cycle depending on the particular core design. The NRC has stated in Reference 30 that these limits can be removed from the Technical Specifications and placed in a cycle-specific report. CPSL proposes to .incorporate this concept for the control bank 28 (277CRS/lah)

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insertion limits and the relaxed axial offset control (RAOC) and base load axial flux difference (AFD) limits. To do this, the following changes are required to the Technical Specifications:

a. A defined formal report, the Core Operating Limits Report, is added to the Definitions section.

b. The affected Limiting Conditions for Operation and Surveillance sections are modified such that the current references to Technical Specification figures are removed and references to the Core Operating Limits Report are added.

c. The current Peaking Factor Limit Report section of the Administrative Controls is renamed to the Core Operating Limits Report and appropriate references are made to NRC approved methods used in generating the cycle-specific operating limits.

7.5 Heat Flux Hot Channel Factor F~

It is'proposed'to increase the F limit from 2.32 to 2.45 for greater flexibility and to accommodate tke axially heterogeneous aspects (blankets and short burnable absorbers) of the VANTAGE 5 core.

Furthermore, the K(z) curve, which defines the axial dependency of FOy is modified to remove the third line segment applicable to the top of the core.

The full power F limit value of 2.45 was selected to support a steam generator tube p)ugging level of up to 6% while still limiting large break LOCA peak clad temperature values to less than 2200'F, with transition core penalties included.

The axial power profile used to perform the small-break LOCA analysis was derived using the recently improved Westinghouse power shape methodology. Among the most notable aspects of this methodology are the use of a comprehensive data base and the elimination of the third line segment from the K(z) Technical Specification curve.

7.6 Nuclear Enthalpy Rise Hot Channel Factor The following F<H values (including uncertainties) are proposed for the VANTAGE 5 transition.

F<H

= 1.62 [1 + 0.30 (1-P)] for LOPAR fuel F<H

= 1.65 [1 + 0.35 (1-P)] for VANTAGE 5 fuel where P is the fraction of full power. These higher values allow increased fuel cycle design flexibility and lower leakage core loading patterns.

7.7 DNB Parameters The proposed limits on DNB related parameters (T and pressurizer pressure) assure that each are maintained within t5e normal steady state 29 (277CRS/lah)

+1

'8

'L l IV

envelope of operation assumed in the transient and accident analyses.

The proposed revisions are consistent with new accident analyses which utilize the ITDP for DNB evaluations.

7.8 Cycle 2 Corrections Based on changes approved by the NRC for Cycle 2 operations, two additional administrative corrections are incorporated.

Surveillance Requirement 4.2.1.1.a.2 requiring monitoring and logging of indicated Axial Flux Difference (AFD) for a 24-hour period after the automatic computer monitoring is returned to an operable status is being deleted. The purpose of this continued manual logging requirement was to ensure that AFD did not deviate outside the control band where penalty points were previously accumulated, since the computer was assumed not to have any history of AFD during the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> due to its inoperability. The 24-hour period was the time required to remove an accumulated penalty point. So after this period, the computer would have sufficient'istory to track cumulative penalty points. Cycle 2 Technical Specifications revised LCO 3.2.1 from constant axial offset control to relaxed axial offset control, and added a provision for baseload operations. The tracking of penalty points for AFD deviations was deleted. Therefore, the surveillance requirement for the 24-hour logging period following restoration of the computer no longer served any purpose and could have been deleted. The deletion of Surveillance Requirement 4.2.1.1.a.2 is administrative in nature.

b. Action 5 of Technical Specification 3.3.1 (Table 3.3-1) is applicable to the source range nuclear instrumentation during Modes 3, 4, and 5; no other item refers to Action 5. With one or both source range NIS inoperable, verification of adequate shutdown margin per the applicable specification is required to comply with Action 5. Prior to Cycle 2, shutdown margin was covered in Modes 3 and 4 by LCO 3.1.1.1 and in Mode 5 by LCO 3.1.1.2. Cycle 2 Technical Specification changes revised the mode applicability of LCO 3.1.1.1 to Modes 1 and 2 and LCO 3.1.1.2 to Modes 3, 4, and 5.

Therefore, since Action 5 applies only in Modes 3, 4, and 5, reference to LCO 3.1.1.1 for shutdown margin consideration is no longer applicable and could have been deleted in Cycle 2 Technical Specification changes. Although the existing wording in Action 5 states to comply with 3.1.1.1 or 3.1.1.2 "as applicable" and therefore Action 5 is technically still accurate, it is appropriate to remove the reference to LCO 3.1.1.1 to make the action statement completely consistent with other technical specifications. The deletion of reference to LCO 3.1.1.1 is administrative in nature.

30 (277CRS/cn )

TABLE 7.1

SUMMARY

AND JUSTIFICATION FOR TECHNICAL SPECIFICATION CHANGES

~Pa e ~Chan a Reason for Chan e Add a new definition for The Axial Flux Difference (AFD) the Core Operating Limits and Rod Insertion Limits (RIL)

Report. were transferred to the Core Operating Limits Report.

a 1V Delete Figure 3.1-2. The RIL were transferred to the Core Operating Limits Report.

Delete Figure 3.2-1. The AFD limits were transferred to the Core Operating Limits Report.

xlx Rename the Peaking Factor The RIL and AFD limits were Limits Report. transferred to the Core Operating Limits Report.

1-2 Add the Core Operating The RIL and AFD limits were Limits Report to the transferred to the Core Operating definitions. Limits Report.

Figure 2.1-1, Core Safety Safety limits changes are due to Limits increased F>H, DNB correlation changes and use of ITDP.

2-4, 5, 7, Reactor Trip Setpoint ITDP impacted the trip setpoints, 8, 9, 10 Parameters resulting in substantial increase in OTAT setpoint and a small decrease in OPAT setpoint.

B2-1 Core Safety Limit Bases The bases discussion was updated to reflect the new DNB correlations, ITDP methodology, and FAH change.

B2-4 DNBR limit from 1.30 to The DNBR limit was changed due to 1.17 the use of new DNB correlations.

B2-5 and Revise low flow trip ITDP impacted the trip setpoint ~

B2-6 setpoint from 91.7X to 90.5X.

3/4 2-2 Delete Surveillance Surveillance requirement to Requirement 4.2.1.l.a.2. monitor AFD for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> no longer applies based on NRC approval of Cycle 2 RAOC operation.

31 (277CRS/cn )

TABLE 7.1 (Continued)

SUMMARY

AND JUSTIFICATION FOR TECHNICAL SPECIFICATION CHANGES

~Pa e ~Chan a Reason for Chan e 3/4 3-7 Delete reference to Based on revision of TS, LCO 3.1.1.1. LCO 3.1.1.1 mode applicability to Modes 1 and 2 only, reference to this LCO is no longer applicable.

3/4 1-14, Add references to the Core RIL and AFD limits were 1-20, 1-21, Operating Limits Report and incorporated into the Core 1-22, 2-1, replace references to Operating Limits Report.

and 2-4 deleted figures.

3/4'-19 Rod drop time increase from The rod drop time was revised for 2.2 to 2.7 seconds. consistency with VANTAGE 5 fuel performance.

3/4 2-5 F~ limit increased from F~ and K(z) were revised for through 2.32 to 2.45, K(z) consistency with the new LOCA 3/4 2-8 (Figure 3.2-2), core analysis. The Core Operating operating. Limits Report was referenced for AFD and W(Z) limits.

3/4 2-9 RCS flow change from Resulted from ITDP.

292,800 gpm to 293,540 gpm.

3/4 2-9 Change in notation from Change in notation for N

FAHto FbH, change an consistency. Change in limit FAH limit. results from use of VANTAGE 5 fuel.

3/4 2-14 Tav < 592.6 changed to Resulted from discovery that Tav avg < 594.1. Pressurizer pressure transmitters used at site pressure > 2205 psig different than assumed in

~ changed to pressure > 2185. analysis with different uncertainty assumptions.

3/4 3-34 Low-Low T interlock ITDP impacted trip setpoints.

allowable value changed from 550.6 to 549.3'F.

B 3/4 2-1 DNBR Design Limit and F~ The DNBR design correlation limit Limit changed, and the F~ limit increased due to new LOCA analyses.

32 (277CRS/cn )

1 TABLE 7.1 (Continued)

SUMMARY

AND JUSTIFICATION FOR TECHNICAL SPECIFICATION CHANGES

~Pa e ~Chan a Reason for Chan e B 3/4 2-2 Update AFD bases to AFD limits were transferred to the reference the Core Core Operating Limits Report.

Operating Limits Report.

B 3/4 2-2a Correct page heading. Administrative correction.

B 3/4 2-4 Enthalpy Rise Hot Channel The enthalpy rise hot channel Factor, Rod Bow Generic factors were increased for both Margin, and Core Operating LOPAR and VANTAGE 5 fuel. Also, Limits Report reference the generic DNBR allowance for rod for W(Z) bow does not apply to the current DNB correlation and W(Z) has been transferred to the Core Operating Limits Report.

B 3/4 2-S RCS flow uncertainty This flow uncertainty is increased from 2a0 to 2.1X. calculated using the ITDP methodology.

B 3/4 2-6 DNB Limit and DNB The DNB parameters are calculated Parameters using the ITDP methodology, and old DNB correlation limit was replaced.

B 3/4 4-1 DNB Limit The old DNB correlation limit is replaced.

6-24 Replace the administrative The administrative control was control for the Peaking included in the Core Operating Factor Limit Report with Limits Report in accordance with the Core Operating Limits NRC Generic Letter 88-16.

Report.

6-24a Correct page heading. Administrative correction.

33 (277CRS/cn )

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8.0 REFERENCES

Davidson, S. L. and Kramer, W. R. (Ed.), "Reference Core Report VANTAGE 5 Fuel Assembly," WCAP-10444-P-A, September 1985.

2. Davidson, S. L. and Iorii, J. A., "Reference Core Report 17 x 17 Optimized Fuel Assembly," WCAP-9500-A, May 1982.

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

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

5 ~ Miller, J. V., "Improved Analytical Models Used in Westinghouse Fuel Rod Design Computations," WCAP-8720 (Proprietary),

October 1976.

6. Leech' ~ J p et. al -, "Revised PAD Code Thermal Safety Model,"

WCAP-8720-A2 (Proprietary), October 1982.

7. Weiner, R. A., et. al., "Improved Fuel Performance Models for Westinghouse Fuel Rod Design and Safety Evaluations,"

WCAP-10851-P-A, August 1988.

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

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

August 1978.

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

Miller, R. W., et. al., "Relaxation of Constant Axial Offset Control-F~ Surveillance Technical Specification," WCAP-10217-A, June 1983.

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

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, HE y Bomany LE HE p Sharp, D. R., "Improved Thermal Design Procedure," WCAP-8567, July 1975.

34 (277CRS/I ah )

g ~

Pg I

gL

)A

~ &g 4

P C

1 r

H II

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

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

18. Letter from C. Berlinger (NRC) to E. P. Rahe, Jr. (W),

Subject:

Request for Reduction in Fuel Assembly Burnup Limit for Calculation of Maximum Rod Bow Penalty, June 18, 1986.

19. 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.

20. Letter from C. 0. Thomas (NRC) to Rahe (W) Supplemental Acceptance No. 2 for Referencing Topical Report WCAP-9500, January 1983.

21. Letter from W. J. Johnson (Westinghouse) to M. W. Hodges (NRC),

NS-NRC-87-3208, dated October 2, 1987,

Subject:

"VANTAGE 5 DNB Transition Core Effects."

22. Letter from M. W. Hodges (NRC) to W. J. Johnson (W), NRC SER on VANTAGE 5 Transition Core Effects, dated February 24, 1988.

23. Schueren, P., McAtee, K. R., "Extension of Methodology for Calculating Transition Core DNBR Penalties," WCAP-11837, May 1988.

24. Kabadi, J. N., et. al., "The 1981 Version of the Westinghouse ECCS Evaluation Model Using the BASH Code," WCAP-10266-P-A, March 1987, (Westinghouse Proprietary).

25. Letter E. P. Rahe (W) to R. L. Tedesco (NRC), "Reporting of Max SI Issue," December 22, 1981, NS-EPR-2538.

26. Letter W. Johnson (W) to J. Lyons (NRC), "Submittal of WCAP-10266 Addendum 1, BASH Power Shape Sensitivity Studies," January 26, 1987, Revised June 22, 1987.

27. Lee, N. Rupprecht, S. D., Schwarz, W. R., Tauche, W. D.,

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

28. 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.

29. Meyer, P. E., "NOTRUMP, A Nodal Transient Small Break and General Network Code," WCAP-10079-P-A (Proprietary) and WCAP-10080-A (Nonproprietary) August 1985.

35 (277CRS/l ah )

30. Letter from D. M. Crutchfield (NRC) to All Power Reactor Licensees and Applicants, "Removal of Cycle-Specific Parameter Limits from Technical Specifications (Generic Letter 88-16)," October 4, 1988.

31. Holderbaum, D. F., Lewis, R. N., and Rubin, K., "LOFTTR2 Analysis for a Steam Generator Tube Rupture-Shearon Harris Nuclear Power Plant," WCAP-11703 (Proprietary)/WCAP-11704 (Nonproprietary),

January 1988.

36 (277CRS/I ah )

f