ML20044D822

Proposed Tech Specs Table 3.3-2 Re Reactor Trip Sys Instrumentation Response Times

ML20044D822
Person / Time
Site:	Mcguire, McGuire
Issue date:	05/06/1993
From:	DUKE POWER CO.
To:
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ML20044D820	List: ML20044D819 ML20044D822
References
NUDOCS 9305200346
Download: ML20044D822 (15)

Text

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4 ATTACHMENT 2 PROPOSED TECHNICAL SPECIFICATION CHANGE l

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l 9305200346 930506 PDR ADOCK 05000369 P PDR l

2 TABLE 3.3-2 (Continued) Y.

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5 REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES A

c- FUNCTIONAL UNIT RESPONSE TIME 5

d 11. Low Reactor Coolant Flow

a. Single loop (Above P-8) 51.0 second R b. Two Loops (Above P-7 and below P-8) 11.0 second m

12. Steam Generator Water Level--Low-Low <2.0 '".,it 1), 3.5 (!Pi t2) seconds l

13. Undervoltage-Reactor Coolant Pumps <1.5 seconds

14. Underfrequency-Reactor Coolant Pumps <0.6 second w 15. Turbine Trip k

w a. Low Fluid Oil Pressor N.A.

A b. Turbine Stop Valve Liosure N.A.

O

16. Safety Injection Input from ESF N.A.

17. Reactor Trip System Interlocks N.A.

18. Reactor Trip Breakers N.A.

mm gg 19. Automatic Trip and Interlock Logic N.A.

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1. 9 ATTACHMENT 3 No Signi nt Hazards Analysis 10 CFR 50.91 requires that the following analysis be provided concerning whether the proposed amendment request involves a significant hazards consideration as defined in 10 CFR 50.92.

Standards for determination that an amendment does not involve a significant hazards consideration are if operation of the facility in accordance with the proposed amendment would not:

1) Involve a significant increase in the probability or consequences of an accident previously evaluated; or, 2) Create the possibility of a new or different kind of accident from any previously evaluated; or, 3) Involve a significant reduction in a margin of safety.

The proposed amendment is administrative in nature in that it only restores the time response for the Steam Generator Water Level--Low-Low for Unit 1 to the < 3.5 seconds. Since this had been approved in Amendment 43724, the Nuclear Regulatory Commission had previously concluded that the proposed change did ,

not involve a significant hazard. There have been no changes to any of the analyses that would change this conclusion.

Environmental Impact Analysis The proposed amendment has been reviewed against the criteria of 10 CFR 51.22 for environmental considerations. The proposed amendment does not involve a significant increase in the amounts, and no significant change in the types, of any ef fluent that may be released offsite and that there is no significant increase in individual or cumulative occupational exposure.

This is the same conclusion that was reached for the original approval of the change in Amendment 43/24. Therefore, the proposed amendment meets the criteria given in 10 CFR 51.22(c)(9) for a categorical exclusion from the requirement for an environmental impact statement.

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ATTACHMENT 4 PREVIOUSLY APPROVED TECHNICAL SPECIFICATIONS /SERs r

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i nhec 4 ,f UNITED STATES y 3ygjf[g g NUCLEAR REGULATORY COiMMISSION U, WASHING TON, D. C. 20555 s f[/dj

• % , , , , ,+*g SAFETY EVALUATION REPORT RELATED TO AMENDMENT NO. 42 TO FACILITY OPERATING LICENSE NPF-9 AND TO AMENDMENT NO. 23 TO FACILITY OPERATING LICENSE NPF-17 DUKE POWER COMPANY MCGUIRE NUCLEAR STATION, UNITS 1 AND 2 I. INTRODUCTION By letter dated November 16, 1984, Duke Power Company requested changes to Technical Specifications to reflect the transition to the use of optimized fuel assemblies (0FA). One of the requested changes was addressed by Amendment Nos. 39 and 20 to McGuire Nuclear Station, Units 1 and 2, Facility Operating Licenses NPF-9 and NPF-17, respectively. This evalua-tion addresses the remaining changes.

II. EVALUATION This submittal is closely related to previous submittals by Duke Power Company for the Unit 1 first reload and for the generic transition to 0FA loadings for Units 1 and 2 (enclosures to references 2 and 3).

The Unit 2, Cycle 2 reload is very similar to the Unit 1, Cycle 2 reload and change to 0FA fuel, and almost everything reviewed and approved for it and the associated Technical Specification changes as described by Amendment Nos. 32 and 13 to McGuire Nuclear Station, Units 1 and 2, Facility Operating Licenses NPF-9 and NPF-17, respectively, is directly applicable to Unit 2. Unit 2 is being reloaded with 60 new 0FA fuel assemblies as was Unit 1. The core parameters related to transient analyses, for the most part, will remain within tne range covered by the approved generic transition of 0FA analyses as did Unit 1. Where these parameters are changed, the transient events have been reexamined. There are a number of changes to Unit 2 relating to analysis methodology changes and operational parameter changes. These were covered in the Unit 1 and generic reviews. The following changes for Unit 2 were evaluated and found acceptable in the previous Unit 1 Amendment No. 32 and need not be further evaluated here.

1. Change to 0FA fuel; fuel mechanical destgn, nuclear design, thermal-hydraulic design

2. Change in axial power distribution control from constant axial offset control (CAOC) to relaxed axial offset control (RAOC) or base load operation

3. Change from standard thermal-hydraulic design methodology to improved l thermal-design procedure using WRB-1 1

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4. Change to allow a positive Moderator Temperature Coefficient over part of the operating power range ,

5. Change of shutdown margin in Modes 1, 2, 3 and 4 from 1.6 to 1.3 percent delta k 1

6. Change of F3g power dependent modifier from 0.2 to 0.3 7 Removal of rod bow related requirement for F AH*

l As with the Unit I reload, core nuclear parameters for Unit 2 reload fall within bounds used in analyses for the generic 0FA submittal analyses, and new transient and accident analyses are not required because of these parameters. As with Unit 1, the dropped rod events were reanalyzed for .

Unit 2, as required by the approved methodology, and were satisfactory.  :

Differences from Unit 1 Review The Unit 2 submittal (and review) differs from the Unit I submittal

• primarily in two areas, (1) a new loss of coolant accident (LOCA) analysis, and (2) a reevaluation of transients and accidents because of a core flow reduction relative to the generic 0FA and Unit I reload analyses.

LOCA L0re. for Unit I was reanalyzed using analysis applicable for transition l P.id full 0FA cores for McGuire 1 and 2 as discussed in the generic transi-tion 0FA report. However, the Unit 2 analysis used BART (WCAP-9561) for core reflood heat transfer calculations. BART is approved for use on non-UHI plants but had not been approved for a UHI plant such as McGuire Unit 2. This methodology and the analysis for Unit 2 have been reviewed and are acceptable. This analysis for Unit 2 met LOCA criteria using a power peaking factor, F , of 2.26, and this value has been incorporated in the Unit 2 TechnicalqSpecifications.

Reduced Core Flow The generic transition 0FA submittal assumed a Thermal Design Flow (TDF) of 386,000 gpm. For Unit 2, Cycle 2 the TDF will be 382,000 gpm. This is l a one percent reduction in core flow from the approved analysis. As a l result of this reduction, all relevant transients and accident analyses l from the generic report were reexamined and when necessary reanalyzed and departure from nucleate boiling (DNB) and non-DNB limits evaluated; and the protection system setpcints and time constants were reviewed and recalculated and changed where necessary.

The reexamination verified that the core DNB limits are unchanged from the generic 0FA report and Unit I reload values, and the DNR basis is met for all the relevant transients. The Technical Specification limits relating to DNB remain unchanged but the vessel exit boiling limits become more restrictive.

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Each event in which non-DNB limits are of interest was also reexamined.

The control rod withdrawal at zero power, loss of load, steamline break and locked rotor events were reexamined to verify that fuel and clad temperature and system pressure changes (which were all small) would remain within limits. For the steamline break this was determined via  ;

conclusions that the return to power was less severe. The loss of feedwater/ station blackout, rupture of main feedwater line, and limiting .

control rod eiection events were reanalyzed with reduced flow and found '

to fall within limits. The primary events for overtemperature and over-power 6T trip protection, control rod withdrawal at power and small steam- '

line breaks, were reanalyzed, using new setpoints and time constants and met DNB limits.

For the loss of feedwater/ station blackout and rupture of main feedwater line events the steam generator level low-low setpoint used a revised value in the reanalysis, and these values are in the new Technical Specifi-cations.

The new LOCA analysis used the reduced flow value. "

i Our review of this reexamination has concluded that a suitable examination of the effects of the decreased flow has been carried out and, with the l

related review of the Technical Specifications, appropriate core limits ,

will be maintained. ,

Technical Specifications A number of Technical Specification changes are proposed for the Unit 2, '

Cycle 2 reload operation. Most of these changes are the same as (or have only minor variations from) those for Unit 1, Cycle 2. This applies to ,

both specification changes and corresponding bases changes. The Unit'2 '

changes were presented both in Attachment 1 and Attachment 2A to the November 16, 1984, submittal. Attachment 1 also contained a few Technical Specification changes for both Units 1 and 2 that are primarily administra-tive changes. We have reviewed the proposed changes, line by line, and find them acceptable. The following list of changes (from Attachment 1) does not include further discussion where the change has already been discussed in the previous Unit 1 Amendment No. 32, 6

l Technical Specification Changes Section 2.1; Figure 2.1-Ib: The safety limits for DNB for Unit 2 have not further changed beyond the charges resulting from the use of 0FA fuel and corresponding changes in analyses methodology discussed in the Unit 1 SER, but the boiling limits are more restrictive because of the change in core flow. The bases for Section 2.1 have changed to reflect the 0FA related thermal-hydraulic methodology changes. These changes are acceptable.

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Section 2.2; Table 2.2-1: The change to a lower core flow, to the altered steam generator water leval low-low setpoint, and to the overpower and over temperature T setpo'.nts and time constants for Unit 2 are given in this table. These changes are approved as a result of the review of the analyses of the effects of the flow and setpoint changes on transients ,

and accidents and the changes from using 0FA fuel and the related '

methodology (as in Unit 1). The procedures and methodology for overpower, overtemperature T trip setpoint changes (Reference 4) are standard as used for all cycles of Westinghouse designed reactors approved by the staff and are acceptable.

Section 3/4.1.1.1 and Bases (and Bases for 3/4.1.2): The change for the shutdown margin in Modes 1, 2, 3 and 4 is from 1.6 to 1.3 percent ok as in Unit 1.

Section 3/4.1.1.3: This change is to allow a positive moderator tempera- l ture coefficient as in Unit 1.

Section 3/4.2.1 and Bases: The change is from CAOC to RAOC and Base Load Operation as in Unit 1 (including the supplemental review for Unit 1 Base Load).

i Section 3/4.2.2: The change in Unit 2 to an Fnof 2.26 is approved as a result of the approval of the LOCA analysis usYng this value. The change to Fgsurveillance (from Fxy surveillance) is as in Unit 1.  !

Section 3/4.2.3: The change in the F power factor from 0.2 to 0.3 and 0

the elimination of the rod bow factor $re as in Unit 1.

Table 3.3-2 and 3.3-4: The changes in time constants and setpoints are  !

the same as in Section 2.2, Table 2.2-1.

Section 3/4.3.3.2: This change in Unit 2 and for consistency, in Unit 1, reflects the elimination of F xy surveillance.

Section 3.5.1.1: The cold leg injection accumulator volume and pressure values are changed to those used in the LOCA analyses and are acceptable.

Section 6.9.1.9: This is a reporting requirement for W (zi values for RAOC as in Unit 1.

III. ENVIRONMENTAL CONSIDERATION These amendments involve a change in use of facility components located within the restricted area as defined in 10 CFR Part 20 and changes in I

survelliance requirements. The staff has determined that the amendments l

involve no significant increase in the amounts, and no sionificant change in the types, of any effluents that may be released offsite and that there is no significant increase in individual or cumulative occupational I

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l exposure. The Commission has previously issued a proposed finding that  ;

these amendments involve no significant hazards consideration, and there  ;

has been no public comment on such finding. Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR Section 51.22(c)(9). Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connec-tion with the issuance of these amendments.

t IV. CONCLUSION The Commission made a proposed determination that the amendments involve l no significant hazards consideration which was published in the Federal ,

l Register (49 FR 50802) on December 31, 1984, and consulted with the state  :

I of North Carolina. No public comments were received, and the state of North Carolina did not have any comments.

We have concluded, based on the considerations discussed above, that: I (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (2) such activities will be conducted in compliance with the Commission's regu-  ;

lations, and the issuance of these amendments will not be inimical to the f common defense and security or to the health and safety of the public.

V. REFERENCES i

1. Letter to H. R. Denton (NRC) from H. B. Tucker (Duke Power) "McGuire  :

Nuclear Station, Docket Nos. 50-369 and 50-370, McGuire 2/ Cycle 2 0FA l Reload", November 16, 1984.

2. Letter to E. G. Adensam (NRC) from H. B. Tucker (Duke Power) "McGuire Nuclear Station, Docket Nos. 50-369 and 50-370, McGuire 1/ Cycle 2 0FA Reload" December 12, 1983. .

3. Letter to E. G. Adensam (NRC) from H. B. Tucker (Duke Power) "McGuire l Nuclear Station, Docket Nos. 50-369 and 50-370", November 14, 1983.

I l 4 S. L. Ellenberger, et al., " Design Basis for the Thermal Overpower T l

and Thermal Overtemperature AT Trip Functions", WCAP-8745, March 1977. 1 Principal Contributors

Jon B. Hopkins, Licensing Branch No. 4, DL Howard J. Richings, Core Performance Branch, DSI l

l Dated: March 22, 1985 l

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TABLE 3.3-2 (Continued) \s n ~

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E2 REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES m

E FUNCTIONAL UNIT RESPONSE TIME ,

(( 12. Low Reactor Coolant Flow -

a. Single Loop (Abo'e v P-8) < l.0 second j[ b. Two Loops (Above P-7 and below P-8) 31.0second

• • <2.0(Unit 1),3.5(Unit 2) seconds

13. Steam Generator Water Level--Low-Low

14. Undervoltage-Reactor Coolant Pumps < 1.5 seconds

15. Underfrequency-Reactor Coolant Pumps < 0.6 second to 16. Turbine Trip 4

w a. Low Fluid Oil Pressure N.A. .

O b. Turbine Stop Valve Closure N.A.

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17. Safety Injection Input from ESF N.A.

18. Reactor Trip System Interlocks N.A.

19. Reactor Trip Breakers N.A.

32, 20. Automatic Trip and Interlock logic N.A.

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.g TABLE 3.3-2 (Continu d) .

5 REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES R .

h'FUNCTIONALUNIT RESPONSE TIME 5 i d 11. Low Reactor Coolant Flow ,

w

. Single Loop (Above P-8) a,

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b.

-Two Loops (Above P-7 and below P-8) 51.0 second-51.0 second

12. ' Steam Generator Water Level--Low-Low .52.0 (Unit 1), 3.5 (Unit 2) seconds

13. Undervoltage-Reactor Coolant Pumps <1.5 seconds-14.- Underfrequency-Reactor. Coolant' Pumps <0.6 second w 15. Turbine Trip

.h w a. Low Fluid Oil Pressure N. A.

4 b. Turbine Stop Valve Closure N.A. 1 o .

16. Safety Injection Input from ESF N. A.' i

' 17. Reactor Trip System Interlocks. N.A.

18. Reactor Trip Breakers N.A.

$ co ag 19. Automatic Trip and Interlock Logic N.A.

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A REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES e- FUNCTIONAL UNIT 5 RESPONSE TIME d 12. Low Reactor Coolant Flow w

= a. Single Loop (Above P-8) < l.0 second E b. Two Loops (Above P-7 and below P-8)

M 51.0second

13. Steam Generator Water Level--Low-Low f 3.5 seconds

14. Undervoltage-Reactor Coolant Pumps

< l.5 seconds

15. Underfrequency-Reactor Coolant Pumps < 0.6 second

16. Turbine Trip

a. Low Fluid Oil Pressure Y N. A.

b. Turbine Stop Valve Closure

$ N. A.

17. -Safety Injection Input from ESF N.A.

18. Reactor Trip System Interlocks N.A.

19. Reactor Trip Breakers N.A.

@@ 20. Automatic Trip and Interlock Logic N.A.

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(2) Changes to TS Table 2.2-1 I i l

The overtemperature delta-T (0 TDT) 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 function K values are used in the revised analysis performed by DPC, this change is acceptable.

The administrative change for both Units 1 and 2 to delete the reference to the  !

RTD Bypass System 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 TS Table 3.1-1 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 I 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) Changes to TS Table 3.3-4 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 eliminated.

The core cooling analysis for the steam line break event was reanalyzed assuming an uncompensated low steam line pressure setpoirt of 700 psig. The j l reanalysis indicated that DNB does not occur for this Condition IV event. 1l Therefore, the setpoint change is acceptable with respect to core protection. lI !

(6) Changes to TS Table 3.3-5 The response time for feedwater isolation is revised from 9 to 12 seconds, and l the response time for steam line isolation is revised from 7 to 10 seconds. l l The extended response times are assumed in the core cooling reanalysis for the  ! l steam line break event for the revised low steam line pressure setpoint. The l I

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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.js TABLE 3.3-2a C .

ll REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES ,

j! FUNCTIONAL UNIT RESPONSE TIME D! t

,, 1. Manual Reactor Trip N.A.

(( 2. Power Range, Neutron Flux 10.5 second (1)

3. Power Range, Neutron Flux, High Positive Rate N.A.

4. -Intermediate Range, Neutron Flux N.A. ,

5. Source Range, Neutron Flux .N.A.

)[ 6. Overtemperature AT 110.0_ seconds (1)(2) I

$5 7. Overpower AT 510.0 seconds (1)(2). l us

8. Pressurizer Pressure--Low . 12.0 seconds

9. Pressurizer Pressure--High '2.0 seconds ,

g g. 10. Pressurizer Water Level--High H.A. t S5 .

.E $ (1) Neutron detectors are exempt from response time testing. Response time of the neutron flux signal portion of the channel shall bi measured from detector output or input of first electronic component in channel.

Jijf (2) The i 10.0 second response time includes a 6.5 second delay for the RTDs mounted in thermowells.

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2 TABLE 3.3-2a (Continued) UNIT 1 -

REACTOR TRIP SYSTEM INSTRUMENTATION' RESPONSE TIMES '

A ,

e FUNCTIONAL UNIT RESPONSE TIME c

I g 11. Low Reactor. Coolant Flow

[ a. Single Loop (Above P-8) $1.0 second g b. Two Loops (Above P-7 and below P-8) 11.0 second

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• 12. Steam Generator Water Level--Low-Low 12.0 (Unit 1), 3.5 (Unit 2) seconds l

13. Undervoltage-Reactor Coolant Pumps <1.5 seconds

14. Under'requency-Reactor Coolant Pumps <0.6 second ,

, 15. Turbino Trip

• s

a. Low Fluid Oil. Pressure H.A.

y b. Turbine Stop Valve Closure N.A.

16. Safety Injection Input from ESF- N.A. l s

17. Reactor Trip System Interlocks - N.A.

l gg 18. ' Reactor Trip Breakers N.A.

$E 19. Aut6matic Trip and Interlock logic H.A. g E$

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