ML20059H795

Revised Resolution of Plant-Specific Differing Prof Opinion Issues Concerning McGuire Tech Specs

ML20059H795
Person / Time
Site:	Mcguire, McGuire
Issue date:	08/31/1990
From:	Desai K Office of Nuclear Reactor Regulation
To:
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ML20059H790	List: ML20059H786 ML20059H795
References
NUDOCS 9009190019
Download: ML20059H795 (79)

Text

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P ENCLOSURES 1. 2, & 3 t

RESOLUTION OF PLANT-SPECIFIC DF0 ,tSSUES CONCERNING i y

MCGUIRE TECHNICAL SPECIFICATIONS 8 t

by P

Kulin Desai Reactor Systems Branch Division of Systems Technology l

l APRIL 1990 REVISED AUGUST 1990 P -

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}- l DP0 CONCERNS ON MCGUIRE TECHNICAL SPECIFICATIONS i

ENCLOSURE I PLANT-SPECIFIC DP0 ISSUES RESOLVED BY TECHNICAL ~

SPECIFICATION AMENDMENT. i, e

ENCLOSURE-2 PLANT SPECIFIC DP0 ISSUES RESOLVED BY UPDATING i FSAR i  !

ENCLOSURE 3 PLANT-SPECIFIC DP0 ISSUES RESOLVED REQUIRING NO i LICENSEE ACTION i

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• l ENCLOSURE 1 DP0 CONCERNS ON MCGUIRE TECHNICAL SPECIFICATIONS ptANT SPECIFIC DP0 ISSUES RESOLVED BY TECHNICAL SPECIFICATION AMENDMEhT Question 6e Include response time in the definition of Table 3.3 4.. of the setpoint and provide appropriate Item 4d descriptors for the values in the TS.

(Reference 4) l Issue i Technical Specifications Table 3.3-4 specifies the Engineered Safety features Actuation System Instrumentation trip setpoints and allowable values for various functional units. 1 25 44 eddresses Negative Steam Line Pressure-Rate-High for Steam Line Isolation.

TS Values' descriptors are inconsistent in their format with rcspect to setpoint methodology values and inclusion of a negative sign is redundant to the setpoint definition.

Resolution The licensee changed the descriptor in the TS '

to make it consistent with the descriptor for the setpoint methodology values and eliminated a negative sign for better clarity.

These TS changes are administrative in nature.

The staff approved these changes in TS i

Amendment 102.(Unit 1)andTSAmendment84 I (Unit 2) respectively.

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. . l Questions 7d, 71 and 7k, Clarify the inconsitency between the TS Table 3.3 5, Item 2e values and FSAR values for these items..

Table 3.3-5, Item 3e Table 3.3-5, Item 4e I

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Issue '

I TS Table 3.3-5, lists the engineered safety j features response tine. Items 2e, 3e and 4e 1 indicate that response time is N.A." for the Containment Purge and Exhaust Isolation Systems for Containment Pressure-High, Pressurizer Pressure-Low-Low and Steam Line  !

Pressure-Low initiating signals.

FSAR offsite consequei ,es accident analyses -

took credit for the contaimnent purge and exhaust system isolation and assumed 4 seconds as response time in the analyses. FSAR Section '

9.5.12.3 indicates closure time for these valves is 3 second; and FSAR Section 7.3.1.2.6 indicates a 1 second response time for generating an engineering safety feature .

actuation signal.

_ Resolution The licensee proposed a TS change to make safety analysis values and TS values consistent by including 4 second response times for items 2e, 3e and 4e in TS' table

-f 3.3-5.

i The staff approved these changes in the TS Amendment #102 (Unit 1) and TS Amendment #84 (Unit 2) respectively.

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Question 71 Clarify the inconsistency between the safety Table 3.3-5, l analysis value and the.TS Value for steam line ltem 4h isolation response time. {

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FSAR feedwater system pipe break analysis sequence ~of events Table 15.2.3-1 indicates e that the low steam line pressure setpoint is-s

. reached in the ruptured steam generator in 420 seconds, and that all main steam line isolation valves would close in 427 seconds, a

Based on this information, the response time '

assumed in the safety analysis for steam line isolation is 7. seconds. The'TS allows steam line isolation time of 9 seconds.

Resolution The licensee propsed a TS change to make the allowed tieam line isolation response- time 7 seconds which is consistent with the FSAR.

This TS change was approved by the staff in the TS Amendment #29 (Unit 1) and TS Amendment i

1. 10 (Unit 2) respectively. I i

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Question 7n Clarify the inconsistency between the safety ,

Table 3.3-5, analysis value and the TS value for feedwater Item 6b -isolation re,sponse time. '

Issue '

Table 15.1.2-1 in the FSAR indicates that

• l following an excessive .feedwater flow event at full power, a High-High Steam Generator water '

level signal is generated in 27 seconds and  ;

feedwater isolation valves close in 36 i seconds. Consequently, the actLal feedwater isolation time is 9 seconds; however, the-TS [

lists 13 seconds for feedwater isolation.  !

Resolution The licensee proposed a TS change to make feedwater isolation response tine in the TS 9 seconds, which is consistent with the FSAR. This TS change was approved by the staff in the TS Amendment 102 (Unit #1) and 84 l (Unit #2) respectively, t

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Question 15 Clarify the inconsistency between the TS and FSAR '

TS 3/4.5.3 concerning the number of ECCS pumps operable when l the RCS tempereture is less than or equal to 300*F

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with respect to low temperature overpressure protection (LTOP).  :

Issue l

TS 3.5.3 presents ECCS subsystems - Tavg 4 350'F i during Mode 4 operation. The footnote states that a maximum of two ECCS pumps--one centrifugal  !

charging pump and one safety injection--pump shall i be operable whenever the temperature of one or more l

of the RCS cold legs is less than or equal to  ;

300'F.

The licensee performed the low temperature  ;

overpressureprotectionanalysis(FSAR5.2.2.3) .

assuming only one pump operation when the RCS temperature is less than or equal to 300'F. -

Resolution i l The footnote for TS 3.5.3 calls for two pumps to be operable, however, the plant procedures permit only -

the centrifugal pump to be lined-up for injection to the reactor vessel. The safety injection pump will be operable and may be run in the recir-culation mode; however, the safety injection pump flow path to the reactor vessel is normally blocked.

with closed valves not actuated on safety ^

injection. Thus, only centrifugal charging pump could inadvertently iaject during this mode which is consistent with the FSAR analysis. However, the licensee is in process to revise the footnote to make it consistent with the FSAk analysis.

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During the review process, the staff found that TS 3.4.9 concerning pressure and temperature limits for heatup ano cooldown curves had been revised such that the threshold for LTOPs protection shifted to 320'F from 300'F; but that the reference to this temperature threshold in the footnote to TS 3.5.3 had not been revised  ;

accordingly. 'This inconsistency was not i identified as a DP0 issue; but rather, found I incidentally during the review of the above DP0  !

issue. The staff has discussed this subject with

-the licensee and Darl Hood, the NRC Project n

Manager for McGuire. The licensee is in process i of revising the TS 3.5.3 to be consistent with 1

the TS 3.4.9.

Clarification of R. Licciardo's consnents dated June 19, 1990 l

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The DP0 reviewer raised the concern that the safety injection pump breakers -'

shou 19 be opened, locked and tagged to be consistent with the FSAR LTOP ana*ysis. 1 MtGuire's LTOP analysis is based on one centrifugal charging pump mass flow. I TS 3.5.3 defines the minimum number of ECCS pumps to be operable for 4 ten,9erature less than or equal to 350'F. Surveillance requirement, SR 4.5.3.2 l

spec.'fies that all pumps, except the minimum required operable pumps (which  !

means only one centrifugal charging pump for LTOP. considerations) shall be

' demonstrated inoperable by verifying that the motor circuit' breakers are secured in the open position or by verifying the discharge of each pump has been isolated from the RCS by at least two isolation valves (double isolation) '

with power removed from the valve operators at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> whenever the temperature of one or more of the RCS cold legs is less than or equal to 300' F. 'hus, there is an adequate protection provided for LTOP event. "

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However, there is an apparent inconsistency in the TS. The TS has a footnote that allows a maximum of one centrifugal charging pump and one SI pump to be operable whenever the temperature of one or more of the RCS cold legs is less than or equal to 300'F. This would invalidate the LTOP analysis. However, as ,

noted in our response, plant procedures only permit the charging pump to be lined up for injection.

We have discussed this subject matter with the licensee. The licensee has committed to eliminate this inconsistency as part of their planned threshold temperature TS change of their LTOP.

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ENCLOSURE 2 DP0 CONCERNS ON MCGUIRE TECHN!!AL SPECIFICATIONS PLANT-SPECIFIC DP0 ISSUES RESOLVED BY UPDATING FSAR Question 4a/4b Resolve the inconsistency between the TS response J TS Table 3.3-2, time value ofd5 2.0 secs with respect to the Items 9 and 10 valueforpressurizerpressure(Iowandhigh)on (Reference 4) page 7.2-14 of the FSAR.

Issue TS Table 3.3-2. itens 9 and 10 provide the maximum allowablepressurizerpressure(lowandhigh) reactor trip response time which are greater than the nominal value given in chapter 7 of the FSAR.

Resolution i

The licensee has updated page 7.2-15 in the FSAR to make reactor trip response time consistent with the TS for pressurizer pressure (low and high) trip functions.

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Question 4c Clarify whether the reactor is tripped due to TS Table 3.3-2, pressurizer pressure-low signal or pressurizer Item 17  !

pressure-low-low (ESFAS/safetyinjection) signal during an accidental depressurization of the main steam system; and if so, include' the appropriate response time in Table 3.3 2. Also, clarify q terminology'used in Note e for Table 7.2.1-4 in the FSAR. '

l Issue  !

A. TS Table 3.3-2, lists the reactor trip instrumentation response times. Item 17 in the table lists the input response time as "N. A." for pressurizerpressure-low-low-(safetyinjection).

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This would appear to be incorrect if this trip function is relied upon to mitigate the transient-associated with depressurization of the main steam system.

B. Note e for Table 7.2.1-4 in the FSAR makes rcference to a pressurizer low pressure-low level trip. This should be pressurizer pressure-low-low (safetyinjection).. >

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_ Resolution A.

During the transient associated with depressurization of the main steam system, the I reactor will trip at 1945 psig with the pressurizer pressure-low function during the transient. The pressurizer pressure-low-low (SI) setpoint is 1845 psig. Since this. trip function is not utilized to mitigate accidents other than LOCA, the TS will '

continue to list "N.A." in the TS Table 3.3-2. The actual response time of 2.0 seconds is listed ,

for this ESFAS function under item 3b of TS Table 3.3.5. Therefore, the present TS is correct an'd -

remains the same.

s B.

The licensee will revise the FSAR-Table 7.2.1-4, Note e for better terminology and clarity.  ;

Clarification to R. Licciardo's comments dated June19, 1990 '

i TS Tables 3.3-2 and 3.3-5 list reactor trip response times and engineered safety features response times respectively. Response times are provided for accidents and transients as appropriate based on the trip function which is taken credit for in the safety analysis. However, the other trip functions are always available and have their surveillance requirements to demonstrate their operability.

To eliminate duplicate surveillance testing requirements, trip functions response times are listed in either tables as appropriate.

This response is also applicable to Question 7b/79 and 7e comments.

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ENCLOSURE 3 DP0 CONCERNS ON MCGUIRE TECHNICAL SPECIFICATIONS

_ RESOLUTION OF PLANT SPECIFIC DP0 ISSUES RESOLVED REQUIRING NO LIC Question 1 Confirm the validity of McGuire Units 1/2 steam  ;

Table 2.2-1 generator instrumentation, setpoint and their (Reference 4) applicability. McGuire Unit I has D-2 steam generators and McGuire Unit 2 has D-3 SG.

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_ Issue Steam Gencrators D-2 and D-3 have a minor design difference et SG bottom plate. Both SGs have identical instrumentation hardware and setpoint. s Resolution The licensee performed a conservative safety analysis which is applicable to both units.

Instrumentation setpoints values are based on this analysis. Westinghouse RPS/ESFAS setpoint methodology is applicable to both units and approved by the staff.

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i Question la Verify that a time constant of > 2 seconds results l Table 2.2-1 in a slower response time which is less conservative. I Item 3 Issue I t

TS Table 2.2-1 represents reactors trip system-instrumentation trip setpoints including response j time. TS Table 2.2-1. Item 3 - concerns power l

range, neutron flux, high positive rate trip during -l a control rod ejection accident. 1 i

'esolution i

e An increased time constant results in a faster response and thus results in a shorter time from initiation of a transient to reactor trip.

The analysis assumes a time constant of 2 '

seconds. Therefore, the time constant of > 2 seconds is conservative. -

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Que;"on Ib (1) Verify that a time constant of ) 2 seconds result

, Table 'c.c *. in a slower response time which is less item 4 conservative.

(2) Resolve the inconsistency between setpoint methodology value and FSAR analysis value.

Issues TS Table 2.2-1 Item 4 specifies power range -

neutron flux, high negative rate during a control rod drop event. The reviewer questioned (1) the conservatism of the time constant used in processing the flux rate signal input to the RPS; and (2) the validity of statements in the setpoint methodology document which indicates that-the negative flux rate setpoint was not used in the safety analysis for McGuire.

Resolution (1)

An increased time constant results in a faster response and thus results in a shorter time from initiation of a transient to reactor trip.

Therefore, the time constant of > 2 seconds is conservative.

(2) As indicated in the FSAR the negative flux rate  !

trip setpoint was evaluated as part of the safety analysis for McGuire. The setpoint methodology document was indeed in error. The licensee has

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revised the setpoint methodology Table 3-4 to show a safety analysis limit of 6.9 % rated thermal power. TS trip setpoint and allowable values remain the same. 1 3

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o Question Ic Resolve the disparity between the setpoint

( TS Table 2.2-1 I methodology value and the FSAR safety analysis Item 9 value.

t Issue '

The setpoint methodology safety analysis value for pressurizer pressure-low is 1845 psig. While the ~

FSAR value for the same analysis is 1835 psig.

i Resolution The licensee has indentified the correct value to be 1835 psig. No change to the FSAR or TS was necessary.

Clarification to R. Licciardo's comments dated June 19, 1990 i

Set point methodology document is a reference document to demonstrate how the l

instrumentation drift and other uncertainties ore accounted for in setpoint l determination.

Unlike FSAR or Technical Specifications changes, it does not require an amendment.

Licensee keeps their files up to date with their revised calculations and can make changes in the setpoint methodology document without the staff's approval.

FSAR original analysis value of 1836 psig remains-the same.

This response also applies to Question Id.

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Question'Id-' Verify.:that the;FSAR safety analysis value assumed TS Table 2.2-1, in-the;feedwaterline'breakanalysisislonerthan 1 Item 13 the TS setpoint value. ~

Issue ,

a TS Table 2 ltem 13 lists- steam generator water levu-low-low recctor trip setpoint and i allowable' value? The reviewer questions whether the allowance for instrument error and '

uncertainties was applied in a conservative manner p

to ars .ve atLthe safety analysis value' listed in the setpoint methodology document.

Resolution The setpoint specified in the setpoint methodology.

document does suggest a non-conservative application of the allowance 1 for. channel error and drift. However, this value (i.e W STS + 10%) was not used in the McGuire TS. As' discussed below, j

the allowance for instrument error and other uncertainties has been properly applied for McGuire. j 1

The licensee performed the limiting'feedwater break analysis starting at full power and assuming a ~10w water level trip setpoint of 231 narrou range span.

The McGuire TS lihiit for the SG low-low-water level trip setpoint, t.t 100% rated thermal power is 40% of narrow range span which exceeds the M safety analysis value of-23% narrow range span.by more than-10%.

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' Clarification to R. Licciardo's comments t'ated June 19. 1990 -

-(1) The licensee's feedwater line rupture analysis assumes a SG water le N u .C

low reactor trip setpoint of 23% narro'w range span. The TS setpoint va oe is 40% of narrow range span. The staff has reviewed the derivation of the; TS value and has concluded that the effect of instrument error and channel

. drift has been appropriately added to the value used in the safety analyses, q

In addition, the reviewer raised a question related to the need to revise- i the setpoint methodology document to reflect these changes. The setpoint -l methodology document is a reference document which demonstrates how the instrumentation drift and other uncertainties are accounted for in setpoint determination. Unlike FSAR or TS changes, it does not require'an f

amendment. The licensee maintains separate calculation files to support their setpoint calculations used in the TS.

(2) Feedwater line rupture and . loss of normal feedwater events result in.

different containment environments. Instrumentation error is larger for the feedline rupture event due to the hostile environment which is created..

Therefore, the SG water level-low ~ low reactor trip.setpoint for the main feedwater line rupture is lowered to 23 percent-to account for the environmental effects.

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Question le Clarify whether pressurizer ' pressure - low signal '!

Table 2.2-1, orpressurizerpressure-low (safetyinjection)

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Item IBb signal trip the reactor during an accidental  !

.depressurization of the main steam system;from I, zero load.

Resolution  !

l An socidental .depressur12ation of the main steam system (inadvertent opening of a dump valve,- 1

-safety valve or relief valve)-.is initiated from hot- l shutdown conditions at zero power which is.the nost conservative' initial condition. Reactor is already tripped at the beginning of the transient i

.(hotshutdowncondition)._Thus,noexplicit ,

assumption is made regarding the cause of reactor -

trip for the- FSAR analysis. No credit is taken for the reactor trip on; pressurizer pressure when reactor _ power is below the'P-7' interlock.

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Question 2 Clarify why the existing minimum temperature for TS Page 3/4 1-6' criticalityL (Modes 1/2) is. 551*F which is -less than (TS3.1.1.4) j the programmed setpoint minimum value of 557'F for events from zero power.

Issue l The reviewer is concerned that translents orc accidents may be initiated at zero prwer conditions f rom a temperature lower than the progrcsmed -

setpoint minimum value of 557'F, i.e. the an wed minimum temperature for criticality of-551*F. '

o Resolution i

Accident evaluations for events. from zero' power are performed using the programmed setpoint minimum value.of 557'F. The difference between the hot zero power temperature and minimum temperature for criticality limit is required in order to allow for measurement of the moderator.

temperature coefficient. For most plants the

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minimum temperature for criticality is lower than. 1 hot zero powsr temperature.

The change in initial condition-from 557*F to 551*F for transients occuring at hot zero power would have a negligible impact on results and would be a less representative. input condition since the majority of time spent at hot zero power conditions is at a temperature of about 557'F.

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Clarification of R. Licciardos commen'ts dated June 19. 1990 The change in initial condition from:557 F to 551*F for transients occurring at hot zero power would have a negligible impact on results because the ,

moderator temperature coefficient (MTC) is not significantly affected in this ^

temperature band.

In addition, the analyzed input condition represents the expected plant operating condition at hot zero power temperature of aboutL i 557'F.

Therefore, it is not necessary to analyze hot zero power transients at 551*F.  !

This is a normal industry practice. >

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. l Question 3 Verify that,dUring shutdown in. Modes 3. 4 and 5

• TS Table 3.3-1, with reactor trip system breakers open, source item 6c.

range and neutron flux channel operability TS ,

requirements specify only one' channel operable while FSAR requires two channels to be operable.

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t Technical' Specifications require 2 source range- '

neutron flux channels be operable at all times except when in modes 3, 4 and 5 with the reactor; trip breakers.open. Reviewer suggested that assumptions of boron' dilution a'nalysis would require 2 operable channels at all times. .

Resolution The licensee has determined that boro dilution i

events during modes 1, 2 and 6 were analyzed for the.McGuire units. Consequently..the McGuire' '

- safety analysis does not provide a basis for requiring two operable source range channels- during modes 3, 4 and 5 of ~ operation. The. licensee has <

considered changing technical-specification 3.3.'1 to require two operable source range channels at. j all times during operation in mode 3, 4:and 5; but. -

Fes instead choosen to follow staff guidance in Generic Letter 85-05 to take action to assure that adequate protective measures to avoid boron dilution events are in place. "

Clarification of R. Licciardo's comments dated June 19, 1990 The staff reviewed the FSAR Section 15.4.6 concerning boron dilution events analyses.

We could not find the commitment.made by the licensee that two l

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e source range neutron-flux channels would be operable during'the modes 3, 4 and-5.

The licenne complies with the staff position requiring adequate operating procedures and administrative controls to prevent boron dilution events.

McGuire has both positive alarms and audible count rate meters to alert the opera ors to boron dilution events. Therefore, the plant complies with its ]

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e Question Sa- Clarify whether ap'plicable modes, . Modes .1 and 2 #

Table 3.3-3--

is appropriate or it should be modes 1 61d 3 #-:

Item 79- .under'P-11' interlock.

Issue TS Table 3.3-3 presents Endineered Safety Fea'tures .

Actuation System Instrumentation. Item 7g specifies -

-applicable modes and operability requirements for auto-start of motor driven auxiliary feedwater pumps-(motor-drivenJpumps)on.tripoffallmain feedwater pumps. The reviewer questioned whether  !

this feature could. be blocked during Mode.2 below

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the P-11 interlock because the threshhold for P-11 i could not be reached while in mode 2.

The # sign states that trip function may be blocked I in this mode below'the'P-11 (pressurizer pressure-j interlock setpoint) and which can occurionly in mode 3, therefore, the reviewer believes that i condition-should be on mode # 3.

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_ Resolution l

The statement that P-11 can only occur in mode 3  ;

is inaccurate. Mode 2 is defined as operation 1

with T,yg >/ 350'F, k,ff >, 0.99 and power.( 5% RTP.

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Therefore, subcritical operation with T,yg>. 350'F- i is in mode 2'if k,ff is not less than 0.99. '

Critical operation is restricted to T,yg b 551*F.3 q

but even then the pressure-temperature opera' ting limits permit pressures below 1955 psig. As a i

practical matter, pressure is maintained in the normaloperatingrange( 2235 psig') during mode 2.

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The defeat of auxiliary feedwater pump auto-start.

is accomplished by depressing a switch that is '

interlocked with the P-11 permissive. Thus, the 'i auto-start can only be defeated below a pressurizer

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pressure of 1955 psig. - However,, the same defeat switch will prevent auto-start on low-low steam generator water level (TS_ Table 3.3-3, Item 7c(1)'.

l Since this auto-start capability is required in-

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g Modes ~1, 2 and 3, blocking is not allowed in these' modes. The # is misleading _and will be eliminated by the licensee during the new STS development  ?

program.--

Clarification of R. Licciardo's comments dated June 19.1990 Auxiliaryfeedwatersystem(AFS)applicabilityinMode4isa~genericissue.

Our response is provided as a resolution of Generic -Issue No. 29A. (The new STS require operability of the one motor driven auxiliary feedwaterisystem~ pum in Mode 4 whenever a steam generator is relied on for heat removal. This is a change over the current STS which do not require AFS operability in Mode-.4).

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3 Question 6b Clarify TS items 7c(1) .and 7c(2) concerning the-

Table 3.3-4, Auxiliary Feedwater system-initi.ation. and the flow =

Items 7c(1)and(2) distribution following a feedwater-line break.

Issue TS Table 3.3-3 presents Engineered Safety. Features Actuation System Instrumentation. Items 7c(1)and' (2)discusstheauxiliary.feedwater: system initiation by the steam generator water level-low-low signal. Infomation 'in the table

. indicates that low-low level in one steam-generator is necessary to start the motor driven" pumps and low-low level in at least two steam-generators is= necessary to start the turbine driven pump., The reviewer questions whether the i level in the intact steam generator will be-low i I

enough during 'the feedline break incident to result in a start of the turbine driven AFW pump. j i

Resolution l

Ir, tta case of a'feedwater line breal. the auxiliary feedwater system is design!d to deliver

'450 GPM by either turbine driven purp'or two motor-driven pumps to thise intact:iteam generators while feeding one faulted generator.. l l

In the McGuire feedwater line break analysis, it J

was assumed that: (1)+theturbinedrivenpump t failed as the single failure consideration; (2) One j motor driven auxiliary feedwater pump supplies 110 gpm to an~ intact:SG (the remainder spills' out the l

I break in the faulted _ loop); and (3) the other-motor-driven pump supplies 170 gpm to each of the  !

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other two intact steam generator; thus maintaining

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450 gpm as. total flow to three intact system generators. These assumptions are consistent with' the design of-the:AFW system instrumentation and i

TS requirements ~ for that instrumentation. "

In the case of'a single failure of a motor' driven pump, it is assumed that the~ turbine driven-pump. i can etuate on low-low level in at least two steam generators. :The licensee has calculated that.

q during this accident condition, the mass invento_ry in the intact-steam generators is reduced significantly prior;to reactor' trip' on low-low-l 1evel.in the faulted loop. 'The shrinkage caused by-the bubble collapse from this reduced mass-a condition would cause low-low level to be reached in the other steam generators.-

Dus, in the case of a motor-driven pump single fai Wre consideration, the turbine-driven-pump can i

actuato on low-low level in two steam generators. j and would maintain 450 gpm flow-distribution  !

similar to the motor-driven pump to the intact-SGs.

Thus, with either motor-driven pump or turbine drivin pump single failure consideration, ,

the auxiliary.feedwater system can deliver the designed flow of 450 gpm.

Clarification of R. Licciardo's comments dated June19, 1990 Westinghouse has explicitly calculated the steam generator inventory for the j

case of a single failure of the motor driven pump. Actuation of turbine-driven i AFW pump on low-low steam generator has been demonstrated for this single failure case.

.TS 3.7.1.2, auxiliary feedwater system, requires each motor-driven and steam turbine driven pump be demonstrated operable once per 31 days by verifying adequate pump flow.

The flow distribution calculations are done by computerize analyses which utilize standard engineering techniques and conservative failure assumptions to minimize flows.

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. Question 6c- - Confirm the bases:for the setpoints and allowable L

Table 3.3-4, values as specified in the TS.

Item 9

! Issue ,

t TS Table 3.3-4, Item 9 presents ESFAS.

instrumentation trip setpoint and allowable value for 4KV Emergency Bus Undervoltage-Grid Degraded Voltage (Loss of Power). Reviewer. requested that bases for setpoints be confirmed.

Resolution i

The NRC staff issued a generic letter, dated . I August 12, 1976 requesting all-licensees to [

analyze their Class 1E: electrical distribution system to determine if the operability of safety related equipment could be-adversely affected by.

short term or long tecs degradation'of grid 1 system voltage. A supplemental generic letter issued' June 2, 1977 provided staff. positions pertaining to degraded grid voltage protection and:the selection of volt Je and time'setpoints. and appropriate technical specifications. The licensee's responses, including setpoints, were-reviewed by the staff and found acceptable as' discussed on Page 8-1 of Supplement 1 to the SER.

Clarification of R.-Licciardo's comments dated June 19. 1990 R. Licciardo raised the new issue and our response is as follows:

The undervoltage setpoints and the degraded voltage setpoints of the safety busses are provided to protect the equipment on the safety busses from degraded-

- voltage conditions and to ensure availability of the safety busses following i

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loss of offsite power. They are not designed to trip the reactor. The undervoltage trip on the reactor coolant pump provides an anticipatory trip of the reactor. The clearing of the loads from _the safety busses upon under- t voltage conditions, the startup of the diesel generators, and the subsequent  !

sequencing of the safety loads onto the safety busses do not necessarily result  !

in or require a reactor trip. Therefore, coordination between the setpoints for the undervoltage clearing of the safety busses and the setpoints for the ,

reactor cc:'snt pump undervoltage trip _ is not required nor desired. Nor is the undervoltage-orip function of the reactor coolant pump compromised if there is a concurrent loss of voltage on the safety busses since functioning of the -

reactor trip system does not depend on the AC safety busses in the short term.

This is because the reactor trip system is powered from the DC system and its associated inverters, i

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4-Question 7a and 7f Clarify the inconsistency between the TS response Table-3.3-5, Item 2a time values and the FSAR values used in the LOCA

. Table 3.3.-5, Item 3a analyses.  ;

Issue TS Table 3.3-5, lists engirc.ered safety features response time. Items 2a and 3a provide Safety Injection (ECCS)responsetimeof27-seconds (without.offsite power) due to containment.

.i pressure - high and pressurizer pressure-low-low a initiating signals.during LOCA analyses, i respectively. Reviewer questioned the response time betweenLitems 2a, 3a and 4a.

Resolution No LOCAs were analyzed for initial condition below P interlock.- Low head safety: injection pumps are required during the LOCA cases which results in a response time of 27 seconds (without oft..te <

power) for Items 2a and 3a t as shown in the table-below. Item 4a represents the main steamline break where the low head safety injection pumps are -

not expected to deliver flow.because of'the high RCS pressure.- Consequently, the response time.is shorter as indicated in the table below.

l Therefore, the additional 5 seconds delay for low head safety injection pumps to attain their - -

discharge pressure is not included in the safety analysis for steam line break.

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-TS Table 3.3-5 Initiating TS Response ,

j Item Signal

_ Time 2a. Safety Injection. Containment Pressure-High 27 seconds  !

(ECCS)-

l 3a. Safety Injection Pressurizer Pressure-Low-Low 27/12 seconds (without/with . -

off-site power)'

4a. Safety Injection' F W m Line Pressure-Low i 22/12 seconds i (ECCS)

-l Clarification of R.-Licciardblg, comments dated June 19, 1990' i

Safety injection flow rate to.the Reactor Coolant System as a function of the system pressure is used as.part of the input in the LOCA analyses. The Safety Injection (SI) system was assumed to be delivering to the RCS 25 seconds'after the low pressurizer pressure setpoilt was reached. The Technical-Specifica-tions permit a delay time of 27 seconds; however, the two seconds difference is more than offset by the following three factors:-

(1) These analyses-assume that no safety injection. flow reaches the reactor y

coolant system until the full 25 second delay has expired. Actually, '

i there will be some safety injection ' flow during the startup of the various safety injection. pumps, i l

(2) According to the Technical Specification requirements, the high head safety injection pumps are loaded onto the emergency buses within 13 seconds. i These pumps would therefore be providing their fuli ilow prior l

to 25 seconds. '

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(3) These analyses assume diesel generator startup upon the generation of a safety injection signal and take no credit for the start of the generator due to the loss-of offsite power, which is assumed to occur concurrently with the initiation of the LOCA. .

I The licensee has revised the LOCA analyses and will update the FSAR by September 1991.

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Question 7b and 79 Clarify the 2.0-seconds TS response time value Table 3.3-5. . Item 2b ' .versus the 1.0 seconds value on FSAR Page 7.3-8 Table 3.3-5,1 Item 3b .value, ine descriptor-(from SI) is incorrect and r:

should be deleted.

Issue-l p'  ;

TS Table 3.3-5, items 2b and .3b provide reactor I trip (from 51 signal) response time off 2 seconds for containment pressure-high and pressurizer -

pressure-low-low initiating signals respectively.

~

The lower value of'1.0 second on FSAR Page 7.3-8 is I the limit on the delay in receipt of S1 actuation-upon exceeding the high containment pressure setpoint..

Resolution The response time. listed in TS Table 3.3-5 is not related to 1.0 second limit _in FSAR page 7.3-8..-

The FSAR~value of 1.0 second is the time it takes to generate a safety injection signal.- The

{

description "(from SI)"'is correct in that the. i allowable delay for a reactor trip due to the.SI  !

actuation signal is 2 seconds. This value is '

independent of the setpoint and associated delay of  ;

the initiator of SI.

i 20

Clarification of R. Licciardo's comments dated June 19, 1990 The FSAR Table uses the word'" signal" to mean "setpoint" reached. TS Table 3.3-5 shows response times which is the time from the setpoint reached to full i

actuation of equipment. Thus, the values shown in the FSAR table are not 1

directly related to TS Table 3.3-5. '

Apparently, the DP0 reviewer has concluded that the reactor trip tines shown in the FSAR are based on reaching the $1 actuation signals. However, these  ;

are actually based on reaching the reactor trip signal on pressurizer pressure-low. Thus, these have no relationship to TS Table 3.3-5. Therefore, 4

no change to the TS is appropriate, j

Please also see our response in Question 4c.

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Question 7c and 7h Justify the' TS values used for containment Table 3.3-5, Item 2d isolation valves closure time for LOCA Table 3.3-5, item 3d analyses.

I Issue a

TS_ Table 3.3-5, Items 2d-and 3d. list containment isolation-phase "A" (2) response ~ times of 18 and-28 seconds for containment pressure-high and pressurizer pressure-low-low initiating signals for LOCA analysis with and without offsite power respectively. The reviewer questioned the-acceptability of the containment isolation response times. '

Resolution ,

The only isolation valves explicitly considered in the radiological consequences analysis of a LOCA-include the containment purge, exhaust and the '

process line isolation valves which connect >

containment to the environment. The containment purge and exhaust valves will close in 4 seconds.

7 The process lines with fluids will take longer time -  !

l to close.in comparison to the purge valves. The I process lines valves will close in about 18 seconds (with offsite power). However,' ANSI N271-1976/ANS 56.2,

" Containment Isolation Provisions for Fluid Systems" recommends that. _ in general,' closure times should be as low as reasonably attainable, based on manufacturers' recomended times and valve sizes,.

but generally not less than 15 seconds and in any case, no more than one minute. If these guidelines are met, releases through these process line valves-21

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- before closure need not be modeled in the dose calculation. Therefore, the TS ' containment- l

- . isolation valves closure time of-18 seconds is acce;_ '~~ '

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Question 7e ' Clarify the TS concerning auxiliary feedwater Table 3.3-5, system initiation' on Containment Pressure-High-

. Item 2f- in Modes 3 and 4.

' Issue-TS Table 3.3-5, Item 2f'provides auxiliary. feed-- ,

water system response time for. actuation from a containment pressure-high. initiating. signal as.  ;

"N.A."

I' Resolution FSAR accidents-analyses do not take any credit for >

actuation of the auxiliary feedwater system from a containment pressure-high signal. Consequently, N.A. has been entered for the response time'in table 3.3-5.- However, the TS Table 3.3-5, Note 5=

clarifiesLthat the response time for motor-driven auxiliary feedwater pumps on all: safety injection  :

signals shall be less thar, or equaloto 60 seconds.

Response time limit includes opening of valves:to establish safety _ injection path and attainment of-discharge pressure fo~r auxiliary feedwater pumps.

The AFW response time-as "N.A." is acceptable.

5 Clarification of R.- Licciardo's ' comments dated June 19, 1990 Please see our response in Question 4c.

Y 23 f

r Question 7j Clarify the-TS concerning auxiliary l feedwater

. Table 3.3-5, system under pressurizer-pressure-low-low-Item 3f . initiation signal.  :  !

Issue i TS Table 3.3-5, Item 3f provides auxiliary feed-- ,

water system: response time as "N. A." due to.

pressurizer- pressure-low-low initiating' signal.. _

The reviewer questioned the "N.A." entry for'this item.

Resolution

~

The main steamline depressurization event -

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]'

(inadvertent opening of a steam generator safety, reliefordumpvalve);assumesESFactuationon  !

pressurizer pressure-low-low initiating signal.. ,

For this event it is :onservatively assumed that. ,

auxiliary.feedwater is actuated at'the maximum flow ^

j rate at the initiation of-the eventL o.t accentuate

~

the cooldown. Any delay'in auxiliary feedwater actuation would be-beneficial and'therefore a response time requirement is not applicable: or 'l appropriate.

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• l Qhestion'7m Confirm that the TS containment spray 1 response Table'3.3.-5, time and FSAR analysis value are consistent.

I Resolution- '

'1

.I TS Table 3.3.-5, Item Sa 11sts' containment spray

)

-response time of f 45 seconds following a contain-ment pressure-high-high initiating signal. TS .

response time of 45' seconds.is consistent with'.the- , -

FSAR containment analysis actuation assumption as shown in FSAR Table 6.2.1-16.

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-QuestionL7o Confinn that the TS automatic switchover'to:

. Table 3.3-5, {

recirculation response time is consistent with the Item 12 FSAR assumption. tL Issue TS Table 3.3-5, Item 12 lists response time 6 60 ,

seconds for automatic switchover to recirculatier, resulting from a refueling water storage tank , t (RWST) level initiating s'ignal. The reviewer .

. ' questioned the basis for this value. ',

Resolution s

The containment sump valves are interlocked'with' the RWST isolation valves to the RHR pumps such that these isolation valves will~close when the contain-- .,

t ment sump valves reach their full'open position.

This automatic switchover provides an uninterrupted' flow of water to the RHR pumps.-

The auton.atic switchover to recirculation is initiated when the lev 91 setpoint is reached in the .;

RWST. The plant procedures.as delineated in FSAR-Table 6.3.2-3A/3B test'to ensure switchover delay

]

of 60 seconds which is consistent with the TS 1 response time.

Clarificaticn of R. Licciardo's: comments dated June 19, 1990 i

)

The FSAR ar.alysis documented in Table 6.3.2-3B is based on sequential operation of the sump valves (NI-1848 and NI-185A) and the RWST valves (ND-4B j andND-19A). As stated in Note 10 of'this table, the RWST valves (which close-1 after the sump valves have fully opened) finish closing no later than 10 I

26

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seccnds into Step 5_,

I Since this step is' begun at 110 seconds, the sequential operation of each. pair of valves therefore is assumed in the FSAR to require 120 seconds, reflecting the ellowable stroke time of 60 seconds per valve.

The valves actually stroke faster than this, allowing the 60~ seconds Technical Specification to be satisfied.'

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s-I Question 9' Justify TS _6c; ton requirement to rest' art an-idle Page.3/4L4-2 loop when.in Mode 3-with no reactor coolant loops

! TS 3.4.1.2 'i n operation; or ekplain how natural. circulation s.

is' accomplished with emergency procedures. '

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, TS 3.4.1.2, Action C states, "with no reactor:

coolant loop-in operation, suspend all operations involving a reduction in boron concentration of the RCStand immediately. initiate-corrective action to-return the required reactor coolant loop to ope ra tion. '- The reviewer questions the basis for these procedural'ac'.t ions and prepares alternate

. action which 'is to implement an E0P for natural

-- circulation.

. Resolution I i

For the condition of no reactor coolant loops in operation while in mode 3, the licensee will immediately initiate corrective-action to restart' the reactor coolant pumps to operation per the Abnormal Procedure, AP/1 and 2/A.5500/09," Plant Operations During Natural Circulation." -If -

restart of reactor coolant pumps is not successful, natural circulation-cooling is1 verified and maintained ~per this same procedure actions And

. their sequence are standard .in the industry and are '

acceptable to the staff. . It is to be'noted that E0Ps can only be entered following a reactor trip ,

or safety injection. l l

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4 Question lla. The operator aligns the Residual Heat . Removal TS Section. 3.4,5_ ~

. System at less than 400 psig and 350'F. The valves in the line=from the RWST are closed.

Resolution The " question is merely a statement of_ operator-action to align RHR. It remains true and requires no response.

LOCAs in lower modes of operation and loss.of RHR cooling in lower modes will be addressed generically in Question 5b.

Clarification of R. Licciardo's comments dated June 19, 1990 Dur response to Ouestions lla, 11b, and 11c is available:in Generic Issues resolution Item 5b.

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.4-Question 11b When-the sytem is in'the RHR-cooling modes, the

-TS 3.5 operator would place all safeguards: systems valves in the required positions for plant operation and r ,

place the safety injection,: centrifugal charging. ,

and residual heat removal pumps along with.SI-  ;

accumulator in ready.and then' manually. actuate SI.

I Resolution f

Thit " question" is a statement of operator action =

to align the ECCS for use from a shutdown condition. It remains true and requires no. l response.

' LOCAs in lower modes of operation. anJ. loss of RHR cooling in lower modes will be addressed generically in Question Sb.,

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Ouestion 12c The question 1: ..ot clearly stated.

TS 3.5 Resolution This " question

• is largely a quotation from the FSAR. The lest two paragraphs are statement {

j introducing a quotation fron the SER. This

{

questici! requires no' respoise. l LOCAs in lower modes of operation and loss of RHR cooling in lower rodes will be addressed-generically in-Question Sb.

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Question 12a Explain why FiAR value for nitrogen cover-pressure TS 3.5.1.1.d of cold leg accumlators should not be of higher value to account for channel error and drif t consideration.

. Issue FSAR safety analysis value is 400 psig for nitrogen cover-pressure et cold leg accumulators.

TS setpoint value is also 400 psig. How do we account for channel error and drift consideration?

Resolution Since the UHI system is removed, the licensee revised the value for nitrogen cover-pressure of cold leg accumlator to 585 psig in comparison to 400 psig with UHI accumlator. The alarm is set at 590 psig to account _for_ channel _ error and drift consideration.

In the near future, the licensee will consider the channel error and drift va' lues in the safety analysis when they revise the LOCA analyses to meet the SG tubes plugging requirement. The safety analysis' value will be 564 psig and the TS value will remain the serie, SBS psig.

Clarification of R. Lieciardo's comments dated June 19. 1990 The licensee has revised the LOCA analyses and will update the FSAR by September 1991.

This new analysis value will provide about 3 percent of margin to account for drif t and channel error which we find acceptable.

32 4

4 Question 12b Verify that the accumulators relief valves .

TS'4.5.1.1.1.d.1 setpoints are included in the Inservice Testing progran,.

Recolution The cold leg accumulators relief valves are not-required to perform a safety function either to l shutdown the reactor or-to mitigate the t consequences of an accident. Therefore, these  !

valves are not included in the IST program. '

However.'these valves are included in the licensee's preventive paintenance program at this time.

Clarification of R. Licciardo's comments _ dated June 19. 1990 The cold leg accumulator relief valves will be added into the IST program at the upcoming next 10 year inservice inspection interval assuming ANSI /ASME-0M-10 standard is incorporated into the 10 CFR 50. -!

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Question 13 Verify the water temperature value used in the TS 3.5.1.2.d safety analysis for UNI accumulator.

Verify that the accumulator relief valve setpoint is included in the Inservice Testing Program.

Issue (1) Should the accumulator water temperature value be in the technical specification?

(2) Should the accumulator relief valve setpoint be in the IST program.

Resolution (1) The safety analysis value related to UHI accumulator water temperature is assumed to be the upper bound value of 200'F. Since the UHI accumulator is not heated or located inside containment, there is no real mechanism for increasing temperatures.during operation.

Therefore, there is no need for TS or UHI accumulator water temperature.

(2) The UHI accumulator relief valve is not required l {

' to perform a 50fety function either to shutdown the reactor or to mitigate the consequences of an accident. Therefere. it is not in the IST program.

McGuire Units 1/2.are ice condenser plants with Opper Head Injection system. Experience has i

demonstrated that the- UHI system adds to the e

34 l-

I complexity of plant operation, requires additional maintenance'ano ,enerally reduces plant availability. 'he TS Amendment 57 (Unit 1) and 38 i (Unit 2) approved the removal of the UHI system for McGuire Units 1/2.

Clerification of R. Liccierdo's comments dated June 19, 1990 Upper Head Injection system is removed from the McGuire facility. The cotiments ,

are no longer applicable.

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Question 14 Verify the bases for the flow distributions in the TS 4.5.2.h {

ECCS system and how they meet minimum flow I conditions to intact loops during accident occurrences.

Resolution The ECCS flows assured in the LOCA analyses are the bases for the limits as specified in TS 4.S.2.h.  ;

Flow balance tests are performed during shutdown ,

to account for any change in the subsystem flow '

characteristics to ensure adequate flow for ECCS .

consideration. ECCS flow in.iected to the broken cold leg is assumed to spill in LOCA analyses.

The flow balance tests will place limits on the ,

branch lines to ensure that total. designated flow reaches the intact loops. '

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, i Question 17 FSAR page 9.2-13 states that "In the event of

. TS 3/4.7.5 solid layer of ice" forms on the Standby Nuclear i Service Water. Pond'(SNSWP), the operating train is manually aligned to SNSWP. Provide safety-related reason for this action.

i Resolution l

McGuire Units 1/2 have two sources for ultimate -

heat sink, the. primary source is a lake and the '

backup source is a pond. In the case of severe . -

prolonged cold weather, the operating train could be aligned manually from the control room to desolve the ice layer on the top of the pond.

Clarification of R. Licciardo's comments dated June 19, 1990 '

We have deleted the last two sentences. I Intake structure for the pond is 40 feet below'the top of the pond level which provides initial water source in the case of the ice layer on the top of the pond. r Discharge water is recirculated on the top of the pond which could also melt the ice.

Thus, the pond is available to satisfy all. design basis events upon the loss of the lake source.

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Question 18 Why TS are not applied to flow control velves TS 3/4.9.1 INV-171 A and INV-175 A?'

Resolution Surveillance Requirement 4.9.1.3 requires that valve #INV-250 shall be verified locked closed .

under administrative controls' at least once per 72 - [

hours during refueling operation. This valve is upstream of valves INY-171 A and INY-175 A and isolates the flow path to prevent the inadvertent j

dilution of the RCS boron concentration.

Therefore, INV-171 A and INV-175 A are not part of Tr.

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l REFERENCES '

t

1. Letter from Robert Licciardo to Brian Sheron, " Review of McGuire Technical Specifications," dated June. 11, 1984. {

2. Letter from Thomas Novak to H. B. Tucker, "Nequest for Comments on i McGuire Technical Specifications Concerns Resulting from Differing Professional Opinion," dated July 9, 1985.

3. Letter from H. Thompson to R. Bernero, " Disposition of Concerns Raised by R. Licciardo in his DP0 on the McGuire Technical Specifications," dated May 1985.

4. Letter from H. B. Tucker to Harold Denton, "NRC DP0 Concerns on McGuire Technical Specifications," dcted June 10, 1986.

i 5.

Memorandum from Thomas Murley to Robert Licciardo, " December 7, 1983 Differing Professional Opinion," dated December 29, 1989.

6. WCAP-8745-P-A, " Design Bases for the Thermal Overpower T and Thermal-Overtemperature T Trip Functions," dated March 1977.

7.

' NUREG-0964, " Technical Specifications McGuire Nuclear Station Unit Nos. 1 >

and 2." dated March 1983. '

8.

Letter from William Parker to Harold Denton, " Westinghouse Reactor Protection System / Engineered Safety Features Actuation System Setpoint '

Methodology, Duke Power Company, McGuire Unit 1," dated October 1981.

i 9.

Duke Power Wmpany, McGuire Nuclear Station Final Safety Analysis Report

- Volumes 5, 6, 7, 9,10 and 12.

! 10.

ANS-56.2, " Containment Isolation Provisions for Fluid Systems," 1976.  ;

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11. . - Generic" Litter 85-05, " Inadvertent Boron Dilution' Events."' January 85.. .i i

-12. Letter from. George Lear. to D. C. Switzeri " Millstone Nuclear.' Power

[

1 Station 'Jnits 1 and 2 " dated June 1977.  !

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ENCI4SURE 4 RESOLUTION OF GENERIC DPO ISSUES CONCERNING MCGUIRE TECHNICAL SPECIFICATIONS Technical Specifications Branch Division of Operational Events Assessment i

AUGUST 3, 1990

AVAIL /.ElLITY OF r. cps DEPAKTURE FR0f1 EUCLEATE BOILIlm (DNB) TS Concern 9A 1ssue The sssertion involvis.g Concern 9A consists cf the following three p6tts:

1. The LP0 asserts that the value for the reactor coolant systen, aver 6ge teraperature (Tav;)'given in the TS Table 3.2-1 is not consistent with the value giver, for Tavg in FSAR Figure 5.3.31 for the rated power conditions. Further-more, the LP0 asserts that the following should be provided in the TS:

t.) The setpoint and allowtble values of Tavg; b.) The related power level ascribed to Tavg; and c.) The reactor coolant system Tavg for the zero power cor.dition.

11. _The DP0 asserts that the values for pressurizer pressure in TS Table 3.2-1 are not consistent with the inforp.ation given in FSAR Tabic 15.1.2-2 and Table 4.1-1 of 0F0 reference 20. Also, the DP0 asserts that the setroint and allow-abic values of the pressuit:er pressure should be provided in the TS.

111. The 0P0 asserts that the pressurizer pressure should be provided in TS 2.1-1.and 3/4.4.3.

1.

I

!!esolutior.

ThevaluesofTayg)listedinTable3.0-1oftheTScreLimitingConditior$s of Operttion (LCOs cnd are derived fron. plant safety analyses. These limiting vtluts are esttilished in conjunction with liraiting values fcr other pr ncipali '

thtrraal-hydraulic pareraeters to ensure sufficient DNB margin. These liraits ensure transient, thtt the DNB safety limit will not be violated in the event of a plent FSAR Figure 5.3.3-1 is a plot of the expected Tsvg versus power level. The values of Tcvg ir it.c plot arc not derived from the plant' safety analyses.

They are estimates of the actuti values of Tavg that will exist when the plant is operated the way the licensee intencs. All the plotted Tavg salues are within the limits in TS Ttble 3.2-1.

a.) There is no instrumentation which monitors Tavg and generates a reactor I trip signal based on the values in Table 3.0-1. Therefore, setpoints and allowtbic value<, corresponding to the lir..its in Table 3.E-1 do not need to be specified in the TS. i i

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i b.)andc.) The Tav9 limits in TS Table 3.2-1 were derived by considerinc f plant transients initteted from all power-levels. Therefore,  :

they are boundirig values which are applicable at 'any power i level. !;o related power level needs to be ascribed to the '

Tevg values in TS Table 3.1-1, and no separate Tavg liuit needs  ;

to be specified in TS Table 3.2-1 for zero power operations.

Fesolution - 11.

The values of pressurizer pressure listed in Ta ble 3.2-1 of the TS are LCOs I and are derived from p1&nt safety analyses. Pressurizer pressure is another principal thermal-hydraulic parar:eter in the calculation of DNB. These l liraits ensure that the Dht safety lin.it will not be violated in the event of a plant trensient. ,

Since there tre no autor.atic reactor trips actuated based on the values in TS 3.2-1, there is no need to specify setpuints or allowable values. The '

instrurwr.tetion that would inititte a reactor trip based on these parar.eters is addressed in TS 3.3.1. '

The pressurizer pressure value in Table 4.1-1 of reference 20 of the DP0 is the nominal design pressure for the reactor coolcr.t system and re6ctor internals and is an expccted value for plant operation. It is an estimate of the actual ,

value of pressurizcr pressure that will exist when the plant is operated the "

l wty the licensee intends. The nominal vtlue is within the limits of TS Table

! 3.2-1. '

FSAR T6ble 15.1.E-F is part of the description of the plant safety analyses.  ;

These analyses include adjustments to account for stecdy state fluctuations and r.,eesurement er ror. The EPO suggests that the limits in TS Table 3.2-1  ;

should equal the reference 20 nominc1 value minus the adjustment specified -

in the safety analyses. This suggestion is not correct. The limits in the TS are derived b) making adjustments on safety analysis limiting values of the pressure - not nominal values. '

Pet 01ution - 111, in the new STS pressurizer pressure is included in the curves in Section 2.1.1(itiscIsoincludedintheSection2.1.1curvesofthecurrentSTS).

Specification 3/4.4.3 specifies the' operability of the pressurizer. The operability of the pressurizer is detern.4r.ed based on water volure and heater ,

captcity; therefore, pressurizer pressure does not need to be included in TS 3/4.4.3.

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REACTOR COOL /,hT LOOPS AIJC COOLAt;T CIRCl:LATIOi, Concern 10B Cuestions Lt. Eb,_8c, P,d, and Be

'IS 3/4.4.1 1ssue ,

The CPO asserts that four hetctor Coolcht System (RCS) loops should be required-to be OTERABLE in MOCE 3 (Hot Sttridby) to meet the assumptions of the safety analysis for a nurber of accider,t scentrics. Each of Questions Ce - 8d below  ;

discusses this conccrn for a different type of accider,t.

Outstion 04: OCCURREf;CEE WITH RAPIC RE/.CTIVITY 1hCREASE Perttihing to " Uncontrolled Rod Cluster Control Assenbly Bank Withdrawal from Sub-Critical Cor.dition."

This Technicti specification (TS) at the time of the CP0 submittel required l thtt two RCS locps be OPERAELE and one RCS loop be_in cperation in MOCE 3. The FSAR for McGuire (and other Westinghouse plants) assun.es four Reactor Coolant' Purps (RCPs) are running for this event. The DP0 asserts that any Technical Specification allowing operability of less than four RCS loops in 110DE 3 would not be in conformance with the FSAR and is non-conservative.

Question Eb: STEAli Lil:E BREAKS: OCCURREf;CES Pertaining to "flajor Rupture of a itain Stesm11ne." -

The licGuire FSAR statt.s that the resulting impact on shutdown margin for this ,

event during liODEE 3, 4, and 5 is in. proved over that of the design basis fcr.

f ro power, just critict1 end Tavg = EE7*F. The DP0 asserts, however thtt the design basis c6se nay not be the most limiting case. It states that .tI is conceiv6ble that tuo loop operction may be less conservative than either four RCPs contir.uing to opertte or four RCPs tripped on Safety Injection. The conclusion of the DP0 is that any Technical Specification ellowing operability l of less than four RCS loups in MOCE 3 would not be in conforn.ance with the l

FIF.R end is non-conservativt.

Question Oc: LOSS OF PRIMARY C00LAt:T: OCCURREt:CES Pertaining to "Small Creal LOCA (SBLOC/.)." '

The ItcGuire FSAR and PCAP C356 describe the SPLOCA as a design basis event when it oc:urs from the Rated Power (110DE 1) and Hot Stcr.dby (MODE 3) conditions.

The assertion centained in the DP0 is that "until further tvaluations are made, we niust conclu(e that the current Safety Analysis Limits of the SBLOCA event is four RCS pur.ps OPERABLE in 110DE 3 down to 4E5 psig/350'F" and that the operability of less than four RCS 1 cops in 110DE 3 would not be in conforn.ance  !

uith the current safety analysis limits and is riot conservative. The DP0 clso contains c sin.iltr assertion for the large brect LOCA scencrio.

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guption8d: OCCURREt:CES CAUT"!C A!! II:!TIAL lt: CREASE OF RCS TEMPERATURE The assertion centtined in tt 000 states that the increase of 'RCS ternperature events are of concern because of the potential influence of the positive

! noderator teo.perature ccef ficient. It discusses several events and states that all but one are licensing basis events from rated power. The conclusion of the LP0 is thtt these events are irnportant in MOEL 5 due to the positive moderator teniperature coefficient and states that operability of less than four RCS lecps in MODE 3 would not be in confornance with the safety analyses

• lin.its and is not conservative.

Question Be: AVAIL /.EILITY OF RCPs The CPO states that four RCS loops would be required in 110DE 3 to meet the requiren.er.ts of the licensing basis events frora zero power. In additicn it suggests thtt, in 1:0DE 4, c ninimum set of RCS purnps and loops be used to cool ar.d depressurire the plant down to effectively zero pressure in the steam generators before transferring the heat sink to the RHR system. This is to ensure control of Stean, Line Break ar.d LOCA ever.ts down to RCS conditions where kCS flows are nut necessary. The ptrt of this question addressing MODE 4 is addressed in Concern IEA.

Resolution in the new STS the LCO for RCS loops in H0DE 3 states:

[Two) RCS loops shall be OPERABLE, and

c. [Two) ECS loops shall le in operation when the reactor trip breakers are closed, or

b. One RCS loop shall be in operation wi.:n the reactor trip brtakers bre oper.

The nunbers in brackets indicate that each plant must supply the nunber of pun.ps which is required to nieet their s6fety analysis. For four loop Westing-

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hr.se pli.nts, analysis indicates thtt two 4s the approprdtte nur.ber of RCS loors. 1 At the tin.e of the DP0 submittal, this TS required that two RCS loops be OPER-ABLE and only one RCS loop be in operation it. MCCE 3. The FSAR for McGuire (and cther Westinghouse plents) assumed from two to four RCPs operating for m6ny of '

the accidents discussec above. Westinghouse acknowlecsed the discrepancy in c letter d6ted July 9, ISL4, from E. P. Rahe to D. Eisenhut. At that time, ktstinghouse reviewed thc S.'.fety antlyses for the accidents which are the it.ost litriting 6t zero pcrer for the reduced flow conditions of one RCP. These accidents are the steamline break, rod ejection, and control rod bank withdrawal f rora subcritical conditions. For the rod ejection and steamline break events, i Westinghouse deterrained that the inconsistency between the safety analysis and the Technical Specification would not impact the conclusions prestnted in the F St.R . The analyses showed that the applicable accident criteria were met with only ore RCS pump operating.

4 For the tant withdrtrti fron. subcritical ever.t. Westir.shouse perforr..ed calcu.

1ttdons which showed that the DtiBR design basis racy not be met when only one TCP is in operetion. Consequently, the Westinghouse STS were changed to '

require at least two F.C! loops in operation with the reactor trip breakers  ;

closed tc r..eet the safety analysis liraits for an inadvertent bank withdrawal from subtrit'ctl. ,

i for the SBLOCA en entlysis wts conducted by L'estinghouse assuming that all puraps were ir.4tit 11y operating followed by either all the pur..ps tripping or all the pumps continu4tig to operate. The general conclusion was that there was a smt11er peat clad temperature for the case of all the pumps operating when compared to the case of all the pumps tripped. This case forms the 1 bounding analysis since the retctor coolant punips are not automatically tripped during the SBLOCA and continue to operate of ter the SBLOCA. For ECCS analysis for Westir.ghouse four LOOP plants the most conservative results art obtcit.ed when tht: pCPs are assumed to be tripped at the initiation of a post-ulated LOCA. Tbc Cp0's assertion is unsubstantiated since the ECCS analysis demonstrated thet acceptebic fuel cladcitig temperatures resulted for the more i conservative scenario which resulted when the RCPs are assur.ied to be shut down.

Therefore, for the 14mitirig design ttsis events 6t zero power, the proposed new

!T! will ensure the safety analysis liraits are met. The other events described d

in the Dr0 6re not litu ting design Lasis events at zero power and are thereby 1

tcended by tFc liraitir.g events, i

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• 1 AVAILABILITY Of REACTOR COOLANT SYSTEM LOOPS II l'0DE 4 (HOT SHUTDOWN)

Co_ncern 101 Question 10 15 Fcge7 4 3 1%%UC The DP0 proposed two edditional requirenents for this specification. The first is thtt two RCS loops be Opertble whenever RHR loops are in operation, in order tu provide for the failure of a single motorized valve in the RHR/RCS suction line. The second is that surveillance requirements be added to i

requirt. t determination of the operability of the associated Auxiliary Feed- l wt.ter System and the Atmospheric Dun.p Valves. '

Encifically, the LP0 concerns the McCuire TSAR, which descrites a scentrio corrrrised of the f ailure of a single motorized valve in the RHR/RCS suction line concurrent with the loss of offsite rewer. For this scenario the DP0  :

esserts tht.t two RCS loops should be operabit whenever a p1 tnt has RPR oper-ating in 11 ode 4 Furthermore, the CPO asserts that the current specific 6tions ,

are not conservative because they lack operttility requirements for the i Auxiliery Teedwater Systens or Atmospheric Vahts in Mode 4 Resolu_ t_ ion '

The new West 4nghouse STS require two loops consisting of any combination of i RCS loop and RHR loops be Opertble and at lobst one 100p be in operttion in tiode 4. The Bases for this LCO stetes, "Any one loop in operation provides enough 11cw capacity to removt the decey bett from the core with forced cir. ,

culation. The second 1000, which is required to be OPERABLE, raets single fcilure criteria." Therefore, in order for a licensee to take credit for each loop, there c6nnct be a singit failure which could render both locrs inoperable. The McGuire design which is typical of Westinghouse plants -

includts a single RHR suction line which connects the reactor coolant loop to' the RHR punips. ,

This RHR suction line contains two n0torized valves in series. l The DP0 asserts that t single f ailure concurrent with the loss of offsite 1 power could cause one of these valves to fail close during Mode 4; thereby, '

e1 V ncting the ccre coolfi.g capabi141 ) of the RHR systen.. Thest valves art l opened and left open when core cooling via the RER is initiated in Mode 4 Since motorized valves fail in the "as is" position, these suction line valves remain open citer a single active it.ilure concurrent with a loss of offsite power in l' ode 4 resulting in the RHR system maintainir.g its full functional capability Therefore to require in the TS that 2 RCS loops be_ operable whenever the RHR loop (,s) are in operation is not necess6ry.

l 15 discustco in the resolution of concern 29A, the new Westinghouse STS require the operability of one rtotor driven AFS pump in 110de 4 when a steem generator

is relied on for heat removal. The new STS do not require operability of the j Atmospteric Durnp V61ves (ADys) in Mode 4  !

The preferrod frethod of removing hebt fron, tht' steam generttors in Mcco 4 is through the turbine bypass valves

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to tte condenser. If this p6th becomes unavailable, the hest load is low enough in Mode 4 thtt $G secondcry side sttaming would 16Le tine to reach a l bdgh enough pressure to necessitate venting. Several options (including the i opening of I.DVs) would be ava11tL1t to the operators during that tire to achieve i vthting or eldrainate the f.eed to vent. Ultimately, the safety valves would vent the pressure. The safety valve LCO does not require the safety valves to te operatie in liode 4; however, the TS definition of Operetility and the ASME ccce require crerability of the saftty valves when the steam gtnerator is optrable, The CPO also discussed concerns about tht depth of the Surveillance Require. )

ments (SR) and suggested that additional IR$ be added cr. the systems in this i LCO. Tht exdsting SRs are not intended to be complete tests of the s pcrformance; they are quick,- sin;ple, frequent chtcks (every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />)ystem to ensure thtt the equipo.cht is opertting properly. The more detailed testing is done in thepter 0 o1 the STS arid tre inservice test program. Therefore, there is no need to suppitment tbc oxisting 50s.  ;

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AVAILI.CILITY OF REACTOR COOLAHi SYSTEM LOOPS IN H0DE 5 (COLD SHUT 00L'G)

Concern _191, Question ft is 3.a.1 Issue The DP0 made the following assertions for the Cold Shutdown Mode of operation:

(1) If tht steam generttors are used for cooldng the Auxildary Feedwater

5) stem and Atr.ospheric Dunp Valves should be required to be operable.

(L) There is no btsis for ellowing the operating RHR pumps to be'de-energized for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

(3) The surveillance requiren,ents do not fully test all esttets of operability of the RHR and RCS Iceps.

L(solution The new Westinghcuse STS recuire one RCS loop or one RHR. loop to be operating and t itter one aeditional N:R loup to be OPERABLE or the secondary side water t or greater of the Low-Low level of at least two steam ger.erators to Trip Setpoint. A note in the Limitir.g Condition be [17]for Operation (LCO) 6110ws the RHR putt.p or the p,HR loop in operation to be de-energized for up to I hour provided: (1)nooptrationsare RCS boron concentration; and (2) coreperr.iitted thct wouloiscause outlet temperature reduction niaintained at of the least IC*f below saturation tenperature. The Surveillance Requireroents verify thtt at least or.e EHR or RCS loop is operating and that there is adequate tattr level in the SG.

In MOCE 5 with the RCS loops filleo, the objectives of this LCO are: (1)to remove decty heat genereted in the fuel; and (f) to prevent stratific6 tion of the soluble beric acid. In F0DE 5, an operating RHr. or RCS loop accomplishes I these functions. The other operable RHR loop or the two steam generators with l adequate secondary side wtter lesel 3rovide single failure protection. Under these corddtions of low Fett lead, t 3e bett sints in the two steam ger.erators provide adequate back-up cooling until a RCS or RHR loop can be put into operation. Also, under thesc low heat load conditions, operability of neither the Auxiliary feedwater Syster. nor the Attaospheric Dunp Valves is necessary.

In the new STS, the note 6llowing the operatir.g RHR or RCS pump to be de-entrgized for up to I hour 4s limited; it may only be eFercised once in an C hour period. This tint period is rieeded to perform surveillance testing.

t.s explained above, compensatory measures including close nonitoring of coolant teraperatures are required to exercise the 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> allowance. "The RHR or RCS loop would still be available to be restarted if coolant temperatures exceeded the surveillance limit dn the note. Cxperience in the use of this note has shown that plants do not experience heating or bcrit acid stratification prob ler.is.

2 The surveillances it this LCO do not include testing of alarms and design basis flow rates. The purpose of these surveillances is to provide quick, siniple, frequent checks (every 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />) to ensure that the equiptent is operating properly. The acre detailtd testing is done in chapter 5 of-the $1$

and the inservice testing program. Therefore, there is no need to augment the er.isting surveillt.nce requirements.

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, a STAtDEY I;UCLEAR SEf31CE WATER POLD (ULT 11 TATE FE/.T SINK)

Concern _20B 15 Section 3/4.7.4  ;

1ssue l

The PPO asserts that the applicability section of the Standby fluclear Service Water Pond (St:SWP) TS which includes Modes 1, T, 3, and 4 should also include Mod (s 5 and 6.

Lesolution The r.eed for oxrability of the Ultiraate Heat sink (UHS) in Modes 5 and 6 is  ;

addressed in tie ner STS through the definition of Operatility. UHf is i:equired as a support systin for other sy stetis such as OHR which are requirt-d by $15 tc be operable it liodes 5 and 6. In Modes 5 and 6 the heat load is low; therefore, the derr. ands on the UHS as a support system would Le well below the temperature anc solune recuireroents of the UHS LCO.

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REFUEL 1t$ OPERATI0t:S - LOW WATER LEVEL i

Concern 010 75Page3/I.911 Issue The DP0 asserts thtt both RHR locrs should be in operation in Mode 6 with less than 23 feet of water alcve the top of the reactor vessel fitnge. In support of this staternent the DFC pcstulates the loss of the operable RHR loop without operator ection; the CPC asserts thtt this scertardo would result.ir, boiling in E thinutes er.d rore uncovery in 100 indt .utes.

Resolution The new STS require that one RHR 1 cop be operating and the other RHR loop be operable under the icw water level conditions. The new STS.also require action to restore RHR coolfr.g if it is lost. The DP0 seems to express concern over a scenario where the ortrating RHR loop fails and the reactcr coolant heats up and uncovers the core before the operators become aware of the inoperable RHR  :

loop tred teke action to operete the other RHR loop. The operating RHR loop has an alarm for low PHR flow and cther instruments provide multiple, diverse  ;

indications of loss of RHR cooling to the operators. In addition, several operations personnel would be present in the. area of the reactor cavity. For these re6sor.s it is highly unlikely that a loss of RHR flow would go unnoticed trd uncorrected long enough to allow the core to become uncovered. The other EHR loop is required by STS to be opertble. Through the definition of Operabi141) the support systems necessary for cperation of the otFar RHR loop raust also be operable. Fir. ally, both offsite and energency' diesel generator power are required to be operable in !! ode 6 by STS 3.8.1.2. Therefore, an .

6ddit1ctt1 STS requirement tc hbve both RHR loups oper6 ting is not necessary. '

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AUXILIARY FEEDUATER SYSTEM Ccncern_ISt.

TT'Page 3/4.7-4 issue The E00 states that the TS should require op,erability of the steam driven i.uxiliary feedwater pur.:p in Mode 4. The DP0 also questions the derivation of the Steam tir.e - Pressure Low signal.

Fcsolution '

The new STS tccuire oper6bility of the one motor driven Auxiliary Feedwater Systtu (AFP) purp in Mode 4 wher.ever a steam generator is relied on for heat renovel. Once the plant is switched to 0.HR cooling, operability of the Auxiliery Fetdw6ter Syster. (AFS) is to longer required. This is a charge over the current STS which do not require I.FS operability in Mode 4 The currcrit STS assume th6t the p16 tit switches froin SG cooling to RHR cooling when a change fron. liode 3 to Mode 4 occurs. During the review of the riew STS, it wts found that son.e plants maintain coolirg vi6 the stear.. generatcrs into the upper ten.perature renge of Modt 4. Thest plants mainttin operability of the Auxiliary Feedwtter Systei. via administrative controls until cooling is switched to RHR.

The steani Line Pressure Low Signal used in the main stetni line break accident aralysis is derived f ror. ste6u line sensors downstresta of the stecr. generator flow restriction crifices. This results ir. a conservttive n;easure of ste6m generator pressure since the ste6m flow restrictors do not ccuse_ E significar.t pressure drop except during a doubled enced steam lir.e break. The blowdown phase of the double er.ded steam lirs bre6L lasts only a few seconds. Thc atturate pressure sensing in the steLm lines (the generation cf the stecu line pressure low signal) requires less than 2 seconds and steem line isolation requires less than 7 seconds. teriving this low pressure signti from sensors downstretta of the ste6i.. generator flow restriction orifices is conservative.

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l'.All: STEAli ISOLAT100 VALVES Concern _30A 7S tage 3/4.7-F issue The LP0 contains an assertion thct there is e conflict between TE Sectier.s 3.7.1.4, S.C.3 and TS Table 3.3-4 dealing with the applictbility r.ioces for operability of the Main Stccre Isolation Yalves.

F.esolution The l'ain Steara Isolation Velves (MSIV) have two accident mitigation functions.

Tirst, during a steam lir.e break the I:51Vs close to prevent blowdown fror.i r. ore than one steam generatur. This function is necessary in liodes 1, 2, and 3.

In lioct 4, the lower recctor coolcnt temperature reduces the consequences of the steamline blowdown such that itSIV closure is r.ot necessary. In the new STS, the LCOs which cddress this function in plant systeras and instrumentation ehtpters requirt PSIV opercbility in flodes 1, 2, end 3.

The second eccident mitigation function for tFt IISIVs is containment isolation, l This function is necessary in Modes 1, 2, 3, and 4 In the new STS, the LCOs which address this function ir, the contair.n.ent and instrun'entation chapters require 1151V opcrability in l' odes 1, 2, 3, and 4 a

STLI,h CENERATOP. POWER OPERATEC PELIEF YALVES (SCP0F.V)

Ccr.ct rn 31 A T!~fage 3/4.7.86 issue The 0F0 states thet the TS should include the SGPORVs since under the loss of offsite power cordition these vtives are riecessary for cooling down the plant by r.atural circulttion. Furtherraore, the bPO stetes that additional relieving captcity sFould be covered by TS since the reactor will opertte at power levels 65 high as I0t during the loss of offsite power condition.

F.esolution The loss of offsite power will cause the heactor Coolant Pun.ps (kCP) to trip sir.ce the only power sourcc for these punts is the off site grid.- At reactor power lovels greater than or equti to 10% the tripping of the RCP will it.itiate e reactor strani. At reattor power levels less tL nt10% the reactor would be n.anually strenc<ed by the operator. The power level for either scram is equiv-61ent to the initial cocay heat power level af ter a scrani. The required heat rcr-oval capacity is within the design liniits of r.stural circulttion.

Tte beses for the new STE stcte that the Atmospheric Cun.p Valves (ADYs) will be used to cool down the 91 ant for accidents which are accomp6nied by a loss of offsitt pcwer. Ther . .re, the 1.0Vs are part of the prinary success path for such acciderits trd cie required by the new STS in Modes 1, 2 and 3. PORVs are used to r..indniize the opening of the Main Stotn Safety Valves (MSSVs); the MSSys are part of the pritc.ary success path for events such as full power

J turbine trip without steari dun.p. Since the SGPORVs are not part of the pritnery success path, they do not nieet tht criteria for inclusior, in 15 pursucr.t to the Coni.dssion's Policy Stateront. Therefcre, operability of the SGPORVs is r.ot required by the ru !TS.

l COMP 01lElli C00L1llG WATER SYSTCM Concern 3EA TT fectior 2/4.7.3 1ssue The CTO states th6t the appliccbility of the Component Cooling Water Syster.

(CCWS) TS which includes l' odes 1, t, 3, and 4 should also include Modes E 6r.c 6.

Resolution The need for operability of the Component Cooling W6ter System (CCWS) in flodes 5 and C is addresstd in the riew STS through the definition of Orerability. CCWS is required as a support system for other systems such as NT. which arc lequired by STS to be operable in Modes E and C. Sirice the two trains of the CCWS are typicall) cross connected as in the McGuire Plant, one-train of CCES is adequbte to meet the support function for both RHR trains in Hoces 5 arid 6. Both trains of the CCWS are not required to be operable to provide sir.gle f ailure protectier, in !!cdes 5 ar.d 0 since the heat load is. low, arid there tre other methods which ctn be instituted by the operators to hardle the low heat load if the CCWS fails. I,1so, this allows licer sees to perform ric tt ssary ricir.tenance arid systern r,iodif ications.

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t SERVICE Kt.TER SYSTEM Concern 33A T5'fect ion 3/4.7.4 issue The DP0 states that the applicability section of the Service Water Systera (SWS)

TS rhich includes !! odes 1,'t, 3, and 4 should also include Modes 5 and 6.

i Resolption  ;

The need for operability of the Service Water System (Sn.'S) in Modes 5 and 6 is' addressed in the new STS through the definition of Operability. SWS is required as a support systen for ciher systems such as RHR which 6re required by STS to be operatie in Modes 5 and 6. Since the trains of the SWS are typically cross '

ccr.nected as in the McGuire Plant, ont train of SUS is adequate to neet the support fur.ction for both RHR trains in liodes 5 ar.d C. Both trains of the fi.'s are not required to be operable to provide single failure protection in Modes 5 ,

and 6 since the heat load is lov, and there are other methods which c6n be instituted by the operators to handle the low heat lotd if the SnlS fails.

Also, this allows licensees to perform necessary maintenance and system mcdifica t d or.s.

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LESIDUAL HEAT REMOVAL At:0 COOLANT ClkCULATI0li -

HIGh KATER LEVEL Concern 35A 15 3/4.9.E ,

l_ssue The Ett states thct the action statement sho'.sid require containmert isolation within 2.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> when no RHR loops are ope'able. c Also, the DP0 stetes that the TS shculd require operability of the cor.tainment sun.p and alternete cooling n,cthods in this Mode.

Lesolution in current SU , the action statement allows 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to isolate containment when t.o RHR loops are operat,le. In the nr.w $TS, the instruction to isolate containmtnt has been removed frca the action statenient. The new STS require the establ4shn.ent of alternate cooling methods whenever the RHR is uncytilable.-

The action statenent concentrates on the rnost important task of supplying core cooling M,d leaves the provision of containment isolation to the licensee's contingency procedures.

The alternate cooling r.,ethods do not need to be recuired to be operable in tiode E with the cavity ficoded. The Couraission's Interim Policy Stateinent on Technical Specif4 cation improvement states criteria for deciding which equipment and conditions should be included in TS. Under those criteria the primary success path system, El'R, is required by TS. The provision of alternato ecoling methods referred to in the first paragraph is left as the responsi-bility of the li er.uc's t contingency procedures, in Mode 6 with the cavity flooded, there is a lcrge volunt of water over the core er.d a low decay heat lohd. Underthcseconditdonstheoperatorhas(1) tite.nativecoolingn,ethods which can her.dh. the low decay heat load and (2) time to implement those alternatives.

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REFUElli;C OPERATIO!!! - LOW WATER LEVEL Conccro 3CA

- TI~Page 174.5-11 Issue

- The 0P0 states thtt the action staterient should require containcent ~ isolation ireediately when no kHR icops are operable. -

1:ciolutior2 I in the current STS, the action stater.cnt allows 4' hours to isolete contafraent-when no RilR loops'are oper6ble. In the new STS, the-instruction to isolate coutt.inment has Leen reinoved from the action stater.icnt. - As discussed in the resolution to concern 35A, the actiori statement concentrates on the most iciportant task of supplying core cooluz3 crid leaves the provisions of the containn. erit isolatior to the licer.see's contingency procedures.

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~ P.EACTOP, TF.IP liiSTRUliEt; TAT 10t! SETP0ll:T Concern 3CA Tabic 2.2.-l The assertion in the DP0 states that the Technical Specification nomer.fature

" Low Power Reactor Trips'Diock, P-7" Lis. incorrect and should be labeltd "High ~~!

Poster Reactor Trips Block".

Petolution 'l The nomenclature i's en acceptable description for this function without change; however, information describir.g the P-7' permissive and the P-10 and P-13 trips is discussed'in detail ~it, the new STS Bases under the title " Low Power. Reactor L Trips Block, _ P-7."  !

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REALTOP TRIP INSTRUMENTATION ELTPOINTS Concern 3B lible 1.2 ~1 13 sue The assertion in the DP0 is thr.t the LLsence of the permissivt P-7 [on P-10 tr.d P-33) introduces new eventt i to tvtluate for scfety. The 000.further ,

esserts that the imp 6ct of bloc.iing the Pressurizer Water Level-High trip below P-7 should be evnluated.

[esolution The new STS include in T6ble 3.3.1-1 the F-7 [on P-10 and P-13] interlock.

Stveral reattcr trips (ii.cluding Pressurizer Kater Level-High) are only required, wher operating above 101 power, the P-7 setpoint. The P-7 interlock en6bles and disables trips as rebctor power passes through the 10% power setpoint, Below 10f power, the RCS is cepable of sufficient natural circulation without j any RCP running to prevent DNB. i h.

The Pressurizer Water Level-High trip is a back-up signal for Presturizer-j Pressurt High trip and provides protection agcinst 'passir.g water through the pressurizer safety valves. A reactor trip is-actueted before the pressurizer is water sclid. Thest level channels provide ir.put to the pressurizer level control systEF and do not &ctuate thO safety Valves.

This trip must be operable in Mode I when there is a potential.for overfilling i

the pressurizer. This tr p is automatically enabled on increesing power by the F-7 interlock. On decrecsing power the absence of P-7, autor.:atically l blocks this trip. .Below the P-7 setpoint, trcnsients which could raise pressurizer water level will be slow and the operator will have sufficient time to evaluate ur.it conditions.and talt corrective actions. -

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o EtGINEEFEL SAFETY FEATur.ES ACTbi.'10N.

5)$1EM(ESFAS)INSTROP.EhTATION Question SB-Cor.cern 12B -

Ttble 3.3-3 Issue The assertion ir, the DP0 recomended tha't the sitff consider the consequences of not requiring autonatic switchover to recircul6 tion on RWST level for !! ode .

4 in addition to Mooes 1, 1, 6nd 3.

i F.esolution

.The r.ew STS and the current STS requirc the operab'lity of the switchover to containner.t sun.p or. RWST -level low' for Modes 1; E, 3, and 4.

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1' Concerr. 10A*

15 Iaoe 3/4.3_- Item 6c issue The assertior.s in the DP0 are as follov:s:

1) Penr range neutron flux trip (low and-high) setpoir.ts and interr..ediate range r.eutron ficy are used for events being initiated in'a Subtritical" conditior. cs described in FSAR (table 7.2.1-4); however, the TS does not require their operability. in Modes 3, 4, and 5.

2) Furtherrncie, the source range trip is required to be ope)6ble in Modes 3, 4, ~and 5, yet there is no technical specification for it.

Resolution

1) The Power Range, tieutron flux'-High Setpoint and Low Setpoint do not have to be operable in flodes 3, 4, and 5 l'uclear Instrumentation System (ills [power because thedetectors range retctor iscannot snutdown detectand the neutron-levels in the shutdown range. Other RTS functions and cdministr6tive controls provide protection against reactivity additions when in flodes 3, 4, and 5.  !

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The Intermediate RLnge Neutron flux trip dccs not have to be operable in licdes 3, 4, or 5 because the controls rods rnust be fully. inserted and '

only the shutdown rods may te withdrawn. The reactor'cannot be started up in this conditicr.. The core also has the rcquired Shutdown Margin to r.itigste the cor. sequences of a positive reactivity edditior accident and this margin is required to be onitored frequer.tly. In Mode 6, all rods tre fully ir.serted and the co e has an increased Shutdown Margin. Also, the i.15 interrnediate range cetectors cannot detect neutron _ levels in this range.

2) The new STS require the source range. neutron flux trip function to be ,

operable ir. Modcs 2, 3, 4, ard 5-with the reactor trip breakers closed '

and the red centrol system cepable of rod withdrawal. It is also i required to be opereble in hcde 3, 4 and E v:ith trip breakers open when.

the only function of the source range, monitor is indication. '

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0-P-11 It!TERLOCL Concern 20B TI Fige J/l.3-r issue i

The assertion in the DP0 is that the licensee needed to evaluate the consequcr4es of an event involving a itain Steam Line Breel below the P-11 interlock reactor trip such that the trip will not be initiated by the flegative Sttom Line ' essure Rate - High signal. This concern acknowledges the source range and '.ntern.ediate range nuclear flux trips under these (sn.all and intert.ediate size breaks) circumstances, on any returr. to power, as nut being .necessary because they are nct required in the safety analysis.

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Their current propcsed status precludes crediting their function capability '

and would leave or.ly the powcr range low setpoint trip to trip the reactor.

Furthernere, the resulting power levels of 35i as a safety analysis lic.it-would be unacceptablet without a substantial analysis of the event.

Resolut4cn The F-11 interlock permits a normal ur.it couldown and depressurization without actuation of safet ir.jection (SI) or main stean line isolation. With E/3 pressurizer pressu)re channels less than the P-11 setpoint, the operatur ctn-manually block the Pressurizer Pressure - Lcw and Stear. Line Pressure - Low SI signals and the Steam Line Pressure - Low Steam Line Isolation signal.' When the Stean. Lir.c Pressure ' Low Stean. Line Isolaticr. sigr.al is nar.ually blocked, the ri6i n steam isolation signal on Stear.; Line Pressure - Negative hate - High is enebled. This provides protectior, for a steam line break by closure of the j main sttan. isulttien valves ar.d initiatior cf a reactor trip. .

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.O ESFAS It!STRUMENTAT10t'

' Cor.cern 14/.

Table 2.3 4 1ssue-

.The DP0 asserts tht.t the operabdiity of the containment Phase B isoletion,on-a Ccntainment Pressure High-High sigt.al.should be required in flode,4. 'The=DPO-also asserts th6t a cor.tainment Phase B isolation is:necessary to' establish -

containr..ent integrity.

Resolution The-Containment Pressure lligh-High signal is initieted due to 6 large break j LOCA or st(tr: 'line breet and it actuctes conteir.raent s; ray and Phase B j containment isolation. Containment Pressure }ligh-High must be operable in Modes 1, I cr.d 3 when there is' sufficient energy in the primary and secondary ~~

sides to challer.se the contaiteent pressure High-1ligh setpoint. In Mode 4, 4 there'is insufficicnt energy in the prin.ary and seconctry. sides tc ch611enge the Containr.ient Pressure liigh-High set point. Therefore, operability- of the Containment Pressure High-High signal is not necessary.

Containraer.t P.ressure high actuates S1 and 51 actuates = containnent Phase A ,

isolation.. . Containment Phase A isolation isolates all lines into containraent' '  :

except those associated with the Engineered Safety features. The CCW System, 1 which is t)ficL11y.an Engineered Safety Features Systern cs in the McGuire. '

Plant, is not-isolated by the Phase A isolation. Containment- Phase A isolation i establishes containn.ent integrity tr.d allows the continued use of the Reactor i Coolant Pumps (RCPs) vib'ch rely cr. the CCW. The containnient Phase B isolation j is tctuated by Containn.er.t Pressure High-High and isolates the CCW. The high  !

pressure which caust.s the Contadrc:ent Pressure High-High signal indicates I accident cor.ditions for which RCP operation is not necessary.

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o-ESFAS lilSTitVEliTAT!0!; TF.lP. SETP01t;TS Coricern 1SA Table 3.3-4 issue b

The CP0 esserts thtt e new Functional Unit which is part of ESFAS should be inc1Litd in the 11, This new Functional Ur.it is " Closure of the Feedwater Isoittion, liain Feedwater, tr.c Bypass tiodulation Valves."

Resolution The new STS and the current STS include these valve closure functions under other functions in the ELFAS tables. The DP0 acknowledges this fact, but ,

asserts-that the function needs to be included as a separate functfun ir. the ESFAS tables. The 000 gives ne justification for including _this separate i function; therefore, no additional functional unit needs to be included in-the STS.

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