ML20086H561

Safety Evaluation Supporting Amends 128 & 110 to Licenses NPF-9 & NPF-17,respectively

ML20086H561
Person / Time
Site:	McGuire, Mcguire
Issue date:	11/27/1991
From:	Office of Nuclear Reactor Regulation
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NUDOCS 9112090220
Download: ML20086H561 (14)

Text

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UNITED STATES E]-

( ",h NUCLEAR REGULATORY COMMISSION

E W ASHINGTON, D. C. 20555 7;

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SAFETY EVALUATION BY THE OFFICE OF HUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO.128TO FACILITY OPERAT1HG LICENSE NPF-9 AND AMENDMENT NO.110TO FACILITY OPERATING LICENSE IlPF-17 DUKE POWER COMPANY ltCGUIRE NUCLEAR STATION, UNITS 1 AND 2 DOCKET NOS. 50-369 AND 50-370

1.0 INTRODUCTION

By letter dated June 26, 1991, as supplemented September 16, 1991 and November 7, 1991, the Duke Power Company (licensee or DPC) submitted a request for changes to the ficGuire Nuclear Station, Units 1 and 2, Technical Specifications (TS). The requested changes are to support the McGuire Unit 1 Cycle 8 reload with B&W fuel, in addition, DPC performed the reload analysis with DPC methodology and in the process revised several input assumptions in the Chapter 15 accident analyses. The September 16, 1991 and November 7, 1991, letters provided clarifying information that did not change the initial proposed no significant hazards consideration determination.

This amendment is in response to a change in the fuel design and in the supporting analytical methodology for the McGuire Units.

For both Unit 1 and for Unit 2, during fuel cycles 1 through 7 the fuel was supplied by Westinghouse, (with the exception of demonstration fuel assemblies) and the supporting analyses were based principally on Westinghouse methodology. Beginning with Cycle B, fuel manufactured by the B&W Fuel Company (BWFC) will be utili;:ed. For Unit l's Cycle 8, which will begin in December 1991, this will result in a mixed core of BWFC and Westinghouse fuel. The analytical methodology for this cycle has been developed by the BWFC and Duke Power Company (DPC). This methodology has been reviewed and approved by the NRC staff in response to a series of BWFC and DPC topical reports as referenced in the application and in this evaluation.

Unit 2 continues at this time to operate with Westinghouse fuel based on Westinghouse developed analytical methodology. DPC plans to transition Unit 2 to BWFC fuel based on BWFC and DPC analytical methodology beginning with Cycle 8 which is scheduled to begin in early 1992. Therefore, to accommodate this transition, it is necessary to reflect TS limits for each unit based on its supporting analytical methodology. This is done by including where necessary (primarily in TS Sections 2.0 Safety Limits and Limiting Safety System Settings, 3/4.2 Power Distribution Limits, 3/4.3 Instrumentation and the BASES) separate TS pages for each unit marked to show their unit applicability.

The DPC has also utilized this amendment to include changes which are equally applicable to Units 1 and 2 and therefore do not require separate TS pages for each unit. These include:

(a) deletion of obsolete references in Tables 2.2-la, 2.2-1b, 3.3-2a and 3.3-2b regarding the now removed resistance temperature ol t roo02N a t lin r{ DR ADOD: OS00 s cs pop l

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- - detector (RTD) byp6ss system, (b) a nouenclature change in event titles and deletion of a non-applicable event in Table 3.1-1, (c) a change to TS 3.4.1.2 to require 3 operable reactor coolant pumps in Mode 3, (d) a change to TS 3.5.1.1 to the required average cold leg accumulator boron concentration and a change in nomenclature, (e) a change to TS 4.5.2 in the charging, safety injection and RHR pump head, and (f) a change to TS 6.9.1.9 to reflect the addition of references for the Urit 1 Core Operating Limits Report. These changes are addressed in the following safety evaluation.

Several_ changes proposed by DPC are not included in this amendment. These include the changes to the tolerance ranges for the pressurizer safety valves for Units I and 2 (TS 3.4.2.1 and 3 d.2.2) and for the main steam safety valves for Unit 1 (TS 3.7.1.1).

These proposed changes will be addressed separately as the NRC staff completes its review of them.

2.0 EVALUATION The DPC submittal contains Technical Specifications (TS) changes, e:hanges to the Core Operating Limits Report (COLR), markups of the appropriate FSAR Chapters, and design information relative to Cycle 8 reload. The McGuire Unit 1 plant operated in Cycle 7 with 190 Westinghouse 17x17 Optimized Fuel Assemblies (0FAs), and 3 Mark-BW 17x17 demonstration assemblies. This reload is the first time that a complete batch of B&W Fuel Company's Mark-BW 17x17 fuel design will be used at McGuire. A complete reload batch is operating at Catawba 1 in its current cycle. The use of Mark-BW fuel design in Catawba and NcGuire plants has been previously approved by the NRC via the topical reports l

BAW-10173-P-A,-Revision 2 and BAW-10174-A Revision 1.

The McGuire Cycle 8 reload is the first time DPC has performed reload safety analysis for its Westinghouse Units.

instead of relying on the B&W Fuel Company for reload design and services, DPC has developed its own reload methodology to support the use of Mark-BW fuel design in Catawba nd McGuire plants. The methods and analytical models used by DPC for McGuire Unit 1 Cycle 8 fuel assembly mechanical design, nuclear design, thermal-hydraulic analyses, and non-LOCA safety analysis have been approved by the NRC.

2.1 Fuel System Design The Mark-BW 17x17 fuel assembly design is similar in desigrr 'a the Westinghouse Standard 17x17 fuel assembly design. The unique features wr the Mark-BW 17x17 design include the Zircaloy intermediate spacer grids, the spacer grid restraint system, and the use of Zircaloy grids with standard lattice design.

The mechanical analyses and thermal performance for the Mark-BW 17x17 design i

were performed by DPC with the methodology describe:1 in the approved topical report DPC-NE-2001-P-A, Revision 1 and therefore, are acceptable.

i P.2 Nuclear Design The core physics parameters for Cycle 8 were generated by DPC with the PD007 and EPRT-N0DE-P computer codes using the methodology described in the approved topical report DPC-NE-2010-A. The Reactor Protection System limits and Operational limits for the core were verified through analysis of the Cycle 8 nuclear design using the methodology described in the approved topical report DPC-NE-2011-P-A.

2.3 Thermal-Hydraulic Design The thermal-hydraulic analyses supporting Cycle 8 operation were performed by DPC with the VIPRE-01 computer code and their approved statistical core design (SCD) methodology. The statistical core design methodology is a technique that statistically combines uncertainties associated with the core st6tepo4t parameters, code /model, and CHF correlation to determine a statistical DNBR limit (SDL). The SDL for use with the BWCt1V CHF correlation in VIPRE-01 is determined to be 1.40.

To provide design flexibility, a 10.7% margin is added to the SDL which yields a design DNBR limit (DDL) of 1.55 for the generic Mark-CW and ficCuire Unit 1 Cycle 8 analyses.

Reactor core safety limits for Cycle 8 were generated utilizing BWCllV CHF correlation and SCD methodology with a 1.55 DDL, for a full core of Mark-BW assemblies and a radial enthalpy rise hot channel factor of 1.50.

The hydraulic compatibility of the Mark-BW and the OFA assemblies had been addressed in the approved topical report BAW-10173-P-A Revision 2.

The results of the' hydraulic compatibility test indicated that the total pressure drop across the Mark-BW fuel is 2.4% lower than total pressure drop across the OFA fuel.

The approach taken by the licensee to address the transition core penalty is similar to that used in the topical report BAW-10173-P-A, Revision 2.

The licensee determined a generic transition core penalty by modelling a conservative core configuration witn ene 0FA assembly as tne hot assembly located in a Mark-BW core.

Bounding pwer shapes during normal and accident conditions were analyzed, yielding a maimum DNBR penalty of 3.8% for 0FA fuel.

The licensee address the transition core penalty for OcA fuel by applying the 3.8% DNBR penalty against the 10.7% generic margin inc.uded in the design DNBR limit.

2.4 Accident Analyses 2.4.1 Non-LOCA Analysis The effects of this reload on each Non-LOCA event previously analyzed in the

-FSAR have been evaluated. The following transients were reanalyzed to account for the differences in the core physics parameters of the Mark-BW fuel and the changes in TSs:

(1) Steam System Piping Failure;

(?) Feedwater System Pipe Break; (3) Partial Loss of Forced Reactor Coolant Flow; (4) Complete Loss of Forced Reactor Coolant Flow; (5

Locked Rotor; (6

Uncontrolled Bank Withdrawal from Subcritical; (7

Uncontrolled Bank Withdrawal et Power; (8) Dropped RCCA/RCCA Bank; (9) Single RCCA Withdrawal; (10) RCCA Ejection; (11)TurbineTrip.

The methods and results for analyses of the Steam System Piping Failure, Rod Ejection, and Dropped RCCA/RCCA Bank transients are documented in DPC topical report DPC-NE-3001 and follow-up correspondence from DPC which have been reviewed and found acceptable by the NRC staff.

These analyses show that all acceptance criteria for these transients continue to be met and previous conclusions in the FSAR regarding these events remain valid for Cycle 8 operation.

In addition, a bounding approach has been used to perform the analyses making them new reference analyses to be used in future reload evaluations.

The analysis of transients other than the three discussed above have been performed with RETRAN-02 NSSS transient analysis models and VIPRE-01 tbcrmal-hydraulic subchannel analysis model described in DPC topical report DPC-NE-3000.

These models have been reviewed and approved for use by the staff with certain limitations. The basis for choosing the initial conditions and assumptions used in the analysis of each transient are provided in DPC topical report DPC-NE-3002 and found acceptable by the staff with certain limitations. The specific analyses of the transients for Cycle 8 reload core conditions and operating limits were presented to the staff in a meeting with the licensee on October 7-8, 1991, and documented in follow-up submittals dated October 16, 1991 and November 5, 1991. The results of these analyses have been revicwed by the staff and found to be entirely consistent with results of reference analyses performed for a similar reload core configuration by the Babcock and Wilcox Fuel Company. These references analyses have been reviewed by the staff previously ano % nd acceptable for referencing for either McGuire or its sister plant Catawba. The results of NcSuire Cycle 8 analyses performed by DPC show that acceptance criteria for Departure from Nucleate Boiling (DNB), Peak Clad Average Temperature Fuel Centerline Temperature, Hot Leg Boiling, Peak Reactor Coolant System Pressure and Peak Secondary System Pressure for all these transients continue to be met and previous conclusions in the FSAR regarding these transients remain valid for Cycle 8 operation.

2.4.2 LOCA Analysis The LOCA analysis for McGuire Unit I transition cores with mixed Mark-BW and 0FA assemblies and future cores with all Mark-BW fuel has been reviewed previously by the NRC and found acceptable.

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1 2.5 Technical Specification fTSI Chan g The major portion of the TS changes stem from changes in safety analysis methodology, especially the power distribution and operaticnal control methodology.

Additional changes were made to reflect the existing and revised safety analyses input assumptions, providt operational flexibility or reduce the potential for a spurious trip, and correct existing errors or ncn-conservatisms in existing TSs. All the TS changes have been reviewed with the exception of those associated with increastd tolerances on the pressurizer safety valve and main steam line safety valve lift setpoints. The changes in safety valve lift setpoints will be reviewed in 1992.

2.5.1 Chances due to Revised Power Distribution and Operational Control FeWodology The primary cause for most of the significant changes to the McGuire 1 Cycle 8 technical specifications is the new core power distribution related core operating limit methodology, developed by DPC for use in McGuire and' Catawba reload analysis and cycle operation control. These revisions to the TS involve both format and value changes to the Limiting Conditions of Operation and surveillance requirements. The power distribution TS (3/4.2) are the specifications primarily affected by the methodology changes with the principal changes being to TS 3/4.2.2 and 3/4.2.3 which involve F and F limits and surveillance. The methodology changes af fecting TS 3.2 (aNd are 0

presented (primarily) in the staff approved DFC topical report DPC-NE-2011-P-A with additional related information in topical isports DPC-NE-3001-P-A and DPC-NF-2010-P-A.

The operational methodology used by the DPC is similar to that used by the B&W Fuel Company (BWFC) (described in BAW-10163P-A) for analysis, control and surveillance of BWFC reloads in Pestinghouse reactors, and used by BWFC for DPC for the Catawba 1 Cycle 6 reload. The use of the EWFC methodology and the resulting TS changes was reviewed by the staff and approved in Amendment No. 88 to the Catawba operating license. Since the Catawba 1 operational methodology is similar to that developed for McGuire 1, the TS changes required to implement the methodologies are also similar for both reactors. The changes proposed for the McGuire TS relating to methodology changes, affecting primarily TS 2.1 (Safety Limits) and TS 3/4.2.1 through 3/4.2.5 (Power Distribution Limits), are essentially same as the changes approved for Catawba 1.

Approximately the same revisSns, deletions and additions are made, and very similar justifications are tresented for the changes. There are only a few differences in terminology, form.t, formulation and numerical values of parameters and limits as needed to accomaodate the differences in cycle parameters for the two reactors and small dffferences in details of the methodologies between the DPC and BWFC versions of the operational methods.

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I The methodology adopted by DPC for McGuire 1 for power distribution operational limits control and surveillance has been approved by NRC review and is applicable to McGuire 1.

The proposed TS flow directly f rom the approved methodology and are directly parallel to the appruved TS for Catawba 1.

The relationship of the McGuire 1 TS changes to the methodology has been explained by DPC in the TS justification section of the DPC submittal and the staff review has found the justifications to be acceptable. Deviations between the two sets of TS have been e/amined by the staff in this review and found to be reasonable and acceptable. This review has concluded that this methodology change and the expression of that change in the revised TS for ticGuire 1 Cycle 8, is also acceptable.

The following changes have been proposed for the McGuire 1 TS due to changes to the power distribution / operational control methodology.

The administrative changes to distinguish the revised Power Distribution TS for Unit 1 from the TS for Unit 2 are acceptable.

TS Table 2.2 The OTDT f (delta-1) limits were changed and an axial l

imbalance penalty, f (delta-T), is applied to the OPDT reactor trip to produce p

a reactor trip on high AFDs for credible overpower events protected by the OPDT trip function. These changes to OTDT f (delta-1) and OPDT f (delta-1) are part p

of the approved methodology provided in)DPC-NC-2011-P-A and are acceptable.

TS 3/4.2.1 Axial Flux Difference - The references to the current Westinghouse methodology features such as RAOC and baseload operation are removed and replaced with an envelope of allowed axial flux difference values at various powers, following the DPC methodology. The axic' flux difference setpoint envelope is provided in the Core Operating Limit Report. These changes are in accord with the approved methodology and are acceptable.

TS 3/4.2.2 Heat Flux Hot Char.nel Factor; and TS 3/4.2.3 Nuclear Enthalpy Rise Hot Channel Factor - The Westinghouse methodology is removed from both of these IS and the DPC methodology is inserted. These two sets of changes constitute the primary changes to allow adoption of the new nethodology. The changes are similar to those approved for Catawba 1.

The details and basis of the changes were described in the justification secticn of the DPC McGuire 1 submittal and the approved methodology on which the changes were based is provided in DPC-ME-2011-P-A. The review has indicated that the TS changes have been appropriately described and justified and are in accord with the methodology and are therefore acceptable.

TS 3/4.2.4 Ouadrant power Tilt - The TS was changed to indicate that the required reduction in Ifiermal power below rated power for a quadrant power tilt begins, in the DPC methodology, when tilt exceeds 102 percent, rather than when exceeding 100 percent as with the Westinghouse methodology. This is in accord with the approved methodology and is acceptable.

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-- TS 3/4.2.5 DNB Parameters - The reactor coolant flow rate limit is moved from T5 3.2.3 to tMs T5, and is incorporated in a new Figure 3.2-1, where it is combined with the power level to provide permitted, restricted or prohibited operating regions. -This figure defines trade-offs in power and flow and has been verified by a number of thermal evaluations.. It provides comparable margins to those provided in the previous Wtstinghouse design TS 3.2.3, which is now revised. The change'is in accord with the approved methodology and is acceptable.

- The Bases for the safety limits and power distribution TS which have been changed for McGuire.1 have also been revised to reflect the new methodology.

These revisions present the changes and reasons for the changes in a satisfactory manner and are acceptable.

2.5.2 COLR TS Change The licensee's-proposed changes to the TS are in accordance with the guidance provided by Generic Letter 88-16 and are addressed below.

(1) The approved cycle-specific core operating limits in Amendment No.105 (Unit 1), were revised to reflect the use of the new approved reload analysis methodology as follows:-

(a) Specification 3/4.2.1

-The. axial flux difference limit for this specification and for this surveillance requirement is specified in-the COLR. Due to the use of

.the-new approved reload analysis methodology the target band and the base load operation.are no. longer applicable and are deleted.

(b) Specification 3.2.2 and Surveillance Requirement 4.2.2 The heat' flux hot channel factor (F ). limit.at rated thermal power, the normalized Fn limit as a functi0n of core height-K(z) for both l

-Mark-BW and 0FA-Yuel, F xyz limits for the operational and PS design n

peaking and adjustment to K value from OT delta-T KSLOPE for this specificationandforthislurveillance'requirementarespecifiedin

-the COLR.

-(c) Specification 3.2.3 and Surveillance Requirement 4.2.3 rise hot channel-factor limit- (F' Thenuclear'enthalpfLC0)andforthissurveillance(SNv=).MARP)for thermal this specification power reduction (RRH) and reduction in OT delta-T Ky set point-(TRH) are specified in the COLR.

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.- 4 The bases of affected specifications have been modified by the licensee to include appropriate reference to the COLR.

Based on our review, we conclude that the changes to these baser are acceptable.

(2) Specification 6.9.1.9 Core Operating Limits Report of the Administrative Controls section of the TS for Units 1 and 2 is revised to include currently proposed TS changes and to add additional NRC approved methodologies to support the values of cycle-specific parameter limits that are applicable for the current fuel cycle. The approved methodologies are the following:

(a) WCAP-9272-P-A, " Westinghouse Reload Safety Evaluation Methodology,"

July 1985 (W Proprietary).

(Methodology for Specifications 3.1.1.3 - Moderator Temperature Coefficient, 3.1.3.5 - Shutdown Bank Insertion Limits, 3.1.3.6. -

Control Bank Insertion Limits, 3.2.1 - Axial Flux Difference 3.2.2 -

Heat Flux Hot channel Factor, and 3.2.3 - Nuclear E-thalpy Rise Hot Channel Factor.)

(b) :WCAP-10216-P-A, " Relaxation of Constant Axial Offset Control F0 Surveillance Technical Specification," June 1983 (W Proprietary).

(Methodology for Specifications 3.2.1 - Axial Flux Difference (Relaxed Axial-Offset control) and 3.2.2 - Heat-Flux Hot Channel Factor (W(Z)SurveillanceRequirementsforF Methodology.)

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-(c) WCAP-10266-P-A Rev. 2, "The 1981 Version of Westinghouse Evaluation-Model Using BASH Code," March 1987, (W Proprietary).

(Methodology-for Specification 3.2.2 - Heat Flux Hot Channel Factor.)

(d)- BAW-10168P, Rev.1, "B&W Loss-of-Coolant Accident Evaluation Model -

for Recirculating Steam Generator Plants," September 1989 (B&W-Proprietary).

(Methodology for Specification 3.2.2 - Heat Flux Hot Channel Factor.)

(eE: UFC-NE-2011P, " Duke Power Company Nuclear Design' Methodology for Core Operating Limits of Wettinghouse Reactors," March 1990 (DPC Proprietary).-

-(Methodology for Specification 3.1.3.5 - Shutdown Rod insertion Limits, 3.1.3.6 - Control Bank Insertion Limits 3.2.1 - Axial Flux-Difference, 3.2.2 - Heat Flux Hot Channel Factor, and 3.2..V - Nuclear-Enthalpy Rise Hot Channel Factor.)

(f) DPC-NE-3001P, " Multidimensional Reactor Transients and Safety Analysis Physics Parameter Methodology," Harch 1991 (DPC Proprietary).

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-8 (liethodology for $pecif tation 3.1.1.3 - Moderator Temperature Coefficient, 3.1.3.5 Shutdown Rod Insertion limits, 3.1.3.6 -

Control Bank Insertion 61mits, 3.2.1 - Axial Flux Difference, 3.2.2 -

Heat Flux Hot Channel Factor, and 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor.)

(g) DPC-tie-2010P, " Duke Power Company McGuire fluclear Station Catawba Nuclear Station fluclear Physics Nethodology for Reload Design " April 1984 (DFC Froprietary).

(flethodology for Specification 3.1.1.3 - tioderator Temperature Coefficient.)

(h) DPC-NE-3002, "FSAR Chapter 15 System Transient Analysis Methodology,"

August 1991.

(llethodology used in the system thermal-hydraulic analyses which determine the core operating limits.)

(1) OPC-NE-3000 Rev.1, " Thermal-Hydraulic Analysis Methodology," May 1989.

(Modelingusedinthesystemthermal-hydraulicanalyses.)

This specification continues to require that all changes in cycle-specific parameter limits be documented in the COLR before each reload cycle or remaining part of.a reload cycle and submitted upon issuance to NRC.

Based on our rey;ew, the NRC staff concludes that the modifications proposed by the licensee are lit accordance with the the NRC guidance in Generic Letter 88-16 on modifying cycle-specific parameter limits in TS. Because. plant operation continues to be limited in accordance with the values of cycle-specific parameter limits that are established using NRC approved methodologies, the NRC staff concluoes that this change does not have an adverse impact on plant safety. Accordingly, the staff finds that the proposed changes are acceptable.

2.5.3 Other TS Changes (1)' Core Safety Limits (Figure 2.1-1)

Figure 2.1-1 of the TS is revised to reflect the use of BWCHV CHF correlation arf DPC's statistical core design methodology with a 1.55 desich DNBR limit.

Tr4 revised core safety limits are based on a full core of Mark-BW assemblies.

The licensee addresses the transition core penalty for OFA fuel by applying a 3.8% DNBR penalty against the 10.7% generic margin included in t'1e design DNBR limit. -Since the BWCHV CHF correlation and the SCD methodology are approved we conclude-that the revised core safety limits are acceptable for Unit li these changes are not applicable at this U me to Unit 2.

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.g-(2) Charges to TS-Table 2.2-1 The w etemperature delta-T (OTDT) and overpower delta-T (0PDT) trip function K values on TS Table 2.2-1 are revised to reflect the use of the approved BWCMV CHF correlation and the approved statistical core design methodology with a 1.55 thermal design DNBR limit. The revised overtemperature and overpower trip functior V

• es are used in the revised analysis performed by DPC, this change 3

• able.

The adm.

r a e change for both Units 1 and 2 to delete the reference to the RTD Bypasa

.m reflects the removal of that systems. This change is acceptable.

(3) Changes to TS Table 3.1-1 for Units 1 and 2 TS Table 3.1-1 is revised to include all accident analyses that would require reevaluation in the event that one full-length Rod Cluster Control Assembly is inoperable. Deletion of large break LOCA analysis from 15 Table 3.11 is acceptable since LBLOCA analysis does not take credit for any control rod insertion. This change is acceptable.

(4) Changes to Delete power Range Neutron flux Negative Rate Trip The changes to delete the power Range Neutron Flux Negative Rate trip function from TS Tables 2.2-1, 3.3-1, 3.3-2, and 4.3-1 is acceptable for Unit 1 since no credit is taken for this trip function in control rod drop accidents or any other FSAR licensing basis accidents; this change is not applicable at present to Unit 2.

(5)ChangestoTSTable3.3-4 The low steam line pressure setpoint for safety injection and main steam line isolation is revised from 585 psig to 775 psig, the allowable value is revised

-to 755 psig. The dynamic compensation of steam pressure signal is eliminattd.

The core cooling analysis for the steam line break event was reanalyzed assuming an uncompensated low steam line pressure setpoint of 700 psig. The re6nalysis indicated that DNB does not occur for this Condition IV event.

Therefore, the setpint change is acceptable with respect tn core protection.

(6) Changes to TS Teble 3.3-5 The response time for feedwater isolation is revised from 9 to 12 seconds, and the response time for steam line isolation is revised from 7 to 10 seconds.

The extended response times are assumed in the core cooling reanalysis for the steam line break event for the revised *iow steam line pressure setpoint.

The reanalysis indicated that DNB does not occur for this Condition IV event.

Therefore, the response time change is acceptable with respect to core protection.

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. (7) Changes to TS 3.4.1.2 for Units 1 and 2 The number of operable loops required in Hode 3 operation is revised from two to three to reflect the revised analysis for the uncontrolled bank withdrawal from suberitical nr low power startup condition event. This change is acceptable.

(8) Changes to TS 3.5.1.1 for Units 1 and 2 The required average accumulator boron concentration in ACTIONS c.? and c.3 is revised from 1500 to 1800 ppm, and the basis for averaging is revised to all four accumulators instead of the limiting three.

An increased average accumulator boron concentration will ensure long term subcriticality following a LOCA. This change is at.ceptable.

Changing the basis for the volume weighted average to four accumulators instead of the limiting three is acceptable since all the accumulators will be emptied into the containment sump in the long term following a LOCA.

(9) Changes to TS 4.7.1.4 The permissible stroke time for main steam line isolation valves to close is revised from 5 to 3 seconds. The longer isolation time has been assumed in the main steam line creak core cooling analysis. The results of the analysis show the consequences of the accident with respect to core protection to be acceptable; and therefore, the longer stroke time is considered acceptable.

(10) Changes to TS 4.5.2 (f a h) for Units 1 and 2 The ECCS pump performance requirements are revised.

The required developed head and delivered flow specifications for centrifugal charging pump, safety injection pump, and residual heat removal pump were revised to provide test margin. Since the ECCS pump performance at the revised TSs values meets all acceptance criteria in both the current FSAR analyses, and in the revised analysis performed by DPC, this change is acceptable.

2.5.4 containment Response for TS 4.7.14. TS Table 3.3-4. TS Table 3.3-5, The proposed TS changes to increase the MSIV stroke time from 5 to 8 seconds and to increase the response time for steam line isolation from 7 to 10 seconds would delay the completion of MSIV isolation following a MSLB, and therefore, could extend the period of blowdown. The additional steam released into containment would increase the peak containment temperature. The staff concern is whether the equipment needed for accident mitigation is qualified to the elevated containment temperature. To resolve the concern, the licensee and the staff held several telecons which are summarized in the licensee's letter of November 5, 1991, f

_ The licensee indicated that while the delay of MSlV isolation would increase the blowdown mass, it would also reduce the primary coolant temperaturo, thereby reducing the enthalpy of the steam in the faulted steam generator.

This would in turn, reduce the specific energy of the steam released to the containment. The licensee estimated that the net effect of the additional 3-second blowdown from intact steam generators and the decrease in the specific energy of the blowdown from the faulted steam generator would be to slightly increase the peak containment temperature by less than 1"F for the limiting design basis HSLB of 0.6 f te break.

Based on the peak containment terperature of 326*F (shown in Figure 6.2.1-16 of the McGuire FSAR) and the equipment qualification temperature of 340*F (stated in the-response in November 5, 1991, letter), there is sufficient margin to acconrnodate the slight increase of containment temperature.

Therefore, the staff concludes the TS changes to increase HSIV stroke time and steam line isolation response time are acceptable.

Based on the licensee's assessment, the proposed TS changes to remove the lead / lag dynamic tompensation on the steam line pressure signal and to increase the low steam line pressure setpoint for main steam isolation from 585 psig to 775 psig would result in later main steam isolation for large steam line breaks and earlier isolation for small breaks. Therefore, for small steam line breaks, there is no adverse impact to centainment temperature.

The steam line isolation is actuated by low steam line pressure or high-high containment pressure.

For large breaks, the removal of the lead / lag compensation results in a delayed isolation on low steam line pressure even with the increased setpoint of 775 psig. However, for the breaks analyzed in the FSAR, the main steam isolation occurs on high-high containment pressure at essentially the same time as isolation on lead /1cg compensated steam line pressure. Therefore, the mass and energy release data reported in the FSAR are still valid, and the containment temperature is not affected.

Feedwater isolation in a MSLB is actuated by, among other signals, high containment pressure and low steam line pressure. The combined effect of longer feedwater isolation response time (from 9 to 12 seconds) and removal of lead / lag compensation of~the steam line pressure signal is to delay feedwater isolation following a HSLB. However, this delay does not have an adverse impact on the steam blowdown because isolation on high containment pressure occurs before the time assumed in the FSAR for isolation on low steam line pressure. The licensee's assessment for a spectrum of breaks indicated that for all, except the 1.4 f te large breaks, feedwater isolation is completed in a shorter period of time. The 1.4 ft8 break is not a limiting break.

For the limiting break of 0.6 f t',

the high containment pressure signal occurs at ?.5 seconds and the isolation is completed in 14.5 seconds, in the current FSAR case, the low steam line pressure signal occurs at 8.2 seconds and the isolation is completed in 15.2 seconds.

Therefore, there is no adverse impact on the mass and energy releases resulting from the longer feedwater isolation time and the removal of the dynamic compensation of steam line pressure.

_. _ Based on the above evaluation, the staff concurs with the licensee that the above proposed TS changes have insignificant adverse impact on peak contaircent temperature, ard the equipment quelification temperature of 340*F adequately bounds the containment temperature profile. Therefore, the staff concludes that the above proposed TS changes are acceptable.

2.5.5 Radiological Consequences The NRC staff reviewed the FSAR markups provided as part of the h1C8 TS submittal to asts.rtain whether the MIC8 changes affect our assumptions and parar'eters used to assess the radiological consequences of the Chapter 15 accident analyses. We find that our assumptions and conclusions regarding radiological consequences stated in NUREG-0822 dated March 1978 and Supplement 4 dated January 1981 remain unaltered.

Furthermore, the FSAR markups are consistent with the applicable assumptions and parameters used to support Amendment Hos. 122/104 and 118/100 to the TS for McGuire Nuclear Station, Units 1 and 2, respectively. The radiological consequences from the postulated Chapter 15 accident analyses continue to meet the regulatory criteria previously applied to the ikGuire station and are, therefore, acceptable.

2.5.6 Conclusion We have reviewed the licensee's submittal in support of the McGuire Unit 1 Cycle 8 operation with C&W Fuel Company's Mark-BW fuel and other changes applicable to both Units 1 ano 2 as noted above. We have concluded that McGuire Unit 1 Cycle 8 operation, Mark-BW fuel assently mechanical design, nuclear design, thermal-hydraulic analyses and accident analysis for McGuire Unit 1 Cycle 8 are acceptable. Technical specification changes proposed in the submittal including the other duel unit changes have been *sily approved with the exception of the propnsed increase in the tolerances for the pressurizer safety valve and main steam line safety vhlve lift setpoints. The review of these proposed changes will be completed in 1992.

3.0 STATE CONSULTATION

in accordance with the Commission's regulations, the North Carolina State official was notified of the proposed issuance of the amendments.

The State official had no coments.

4.0 ENVIRONMENTAL CONSIDERATION

The amendments change requirenents with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Fart 20.

The NRC staff has determined that the amendments involve no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and tha! there is no significant increase in individual or cumulative occupational radiation exposure. The Comission has previously issued a proposed finding that the amenduents involve no significant hazards consideration, and there has been no public corrnent on such finding (56 FR 47233). Accordingly, the amendments meet the eligibility criteria

M 14 -

fo" categorical exclusion set forth in 10 CFR 51.22(c)(9).

Pursuant to 10 CFR 51.??(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amer.dments.

5.0 C0_NCLUS10H The Commission has concluded, based on the considerations discussed above, that:

(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activitit.s will be conducted in compliance with the Connission's regulations, and (3) the issuance of the amendments will not be inirnical to the connon defense and security or to the health and safety of the public.

Principal Contributors:

P. Huang, HRR/SRXB T. Huang, NRR/SRXB C. Y. Li, NRR/SPLB k

Date:

November 27, 1991 t

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