ML20044B407

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

ML20044B407
Person / Time
Site:	McGuire, Mcguire
Issue date:	04/30/1990
From:	Desai K Office of Nuclear Reactor Regulation
To:
Shared Package
ML20042H018	List: ML20042H017 ML20044B407
References
NUDOCS 9007230392
Download: ML20044B407 (43)

Text

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RESOLUTION OF PLANT-SPECIFIC DP0 ISSUES CONCERNING MCGUIRE TECHNICAL SPECIFICATIONS by Kulin Desai Reactor Systems Branch Division of Systems Technology APRIL 1990 9007230392 900607 DR ADOCK 05000369

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DP0 CONCERNS ON NCGUIRE TECHNICAL SPECIFICATIONS ENCLOSURE-1 PLANT-SPECIFIC DP0 ISSUES RESOLVED BY TECHNICAL SPECIFICATION AMENDMENT ENCLOSURE-2 PLANT-SPECIFIC DP0 ISSUES RESOLVED BY UPDATING FSAR ENCLOSURE-3 PLANT-SPECIFIC DP0 ISSUES REQUIRING NO LICENSEE ACTION 1

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ENCLOSURE 1 DP0 CONCERNS ON MCGUIRE TECHNICAL SPECIFICATIONS PLANT SPECIFIC DP0 ISSUES RESOLVED BY TECHNICAL SPECIFICATION' AMENDMENT Question 6a Include response time in the definition of

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Table 3.3-4 of the setpoint and provide appropriate j

Item 4d.

descriptors for the values in the TS.

(Reference 4)

Issue Technical Specifications Table 3.3 4 i

spec 1fies the Engineered 56fety Features i

Actuation System Instrumentation trip setpoints and allowable values for various l

t functional units.

Item 4d addresses Negative Steam Line Pressure-Rate-High for Steam Line Isolation.

TS Values' descriptors are inconsistent in their format with respect 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 Amendment 102 (Unit 1) and TS Amendment 84 (Unit ?) respectively.

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

l Table 3.3-5. Item 3e Table 3.3-5. Item 4e i

Issue TS Table 3.3-5. lists the engineered safety features response time.

Items Ee. 3e and 4e indicate that response time is "N.A." for the i

Contairment Purge and Exhaust Isolation Systems for Containment Pressure-High, Pressurizer Pressure-Low-Low and Steam Line l

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Pressure-Low initiating signals.

L FSAR offsite consequences accident analyses took credit for the contaimnent purge and l

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 seconds 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 3.3-5.

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, analysis value and the TS Value for steam line Item 4h isolation response time.

Issue FSAR feedwater system pipe break analysis sequence of events Table 15.2.3-1 indicates that the low steam line pressure setpoint is reached in the ruptured steam generator in 420 seconds, and that all rain steam line isolation valves would close in 427 seconds.

Based on this infonnation, 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 steam 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

1. 10 (Unit 2) respectively.

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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 response time.

Issue Table 15.1.2-1 in the FSAR indicates that following an excessive feedwater flow event at full power, a High-High Steam Generator water i

level signal is generated in 27 seconds and feedwater isolation valves close in 36 seconds. Consequently, the actual 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 isolatioa response time 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 (Unit #2) respectively.

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'Ouestion 15 Clarify the inconsistency between the TS and FSAR TS 3/4.5.3 concerning the number of ECCS pumps operable when the RCS temperature is less than or equal to 300'F with respect to low temperature overpressure protection (LTOP).

Issue TS 3.5.3 presents ECCS subsystems - Tavg ( 350'F during Mode 4 operation.

The footnote states that a maximum of two ECCS pumps--one centrifugal charging pump and one safety injection-pump shall be operable whenever the temperature of one or more of the RCS cold legs is less than or equal to 300'F.

The licensee performed the low temperature overpressure protection analysis (FSAR 5.2.2.3) assuming only one pump operation when the RCS temperature is less than or equal to 300'F.

Resolution The footnote for TS 3.5.3 calls for two pumps to be or,erable, 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 nonna11y blocked with closed valves not actuated on safety injection. Thus, only centrifugal charging pump could inadvertently inject 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 FSAR analysis.

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1 During the review process, the staff found that TS i

3.4.9 concerning pressure and temperature limits for heatup and cooldown curves had been revised I

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

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accordingly. This inconsistency was not d

i identified as a DP0 issue; but rather, found incidentally during the review of the above DP0 issue. The staff has discussed this subject with the licensee and Darl H00d, the NRC Profect

. Manager for McGuire. The licensee is in process l

of revising the TS 3.5.3 to be consistent with 4

the TS 3.4.9.

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ENCLOSURE 2 OP0 CONCERNS ON MCGUIRE TECHNICAL SPECIFICATIONS

_P_LANT-SPECIFIC DP0 ISSUES RESOLVED BY UPDATING FSAR i.

.Que'icion 4a/4b Resolve the_ inconsistency between'the TS response TS Table 3.3-2, time value of f; 2.0 secs with respect to the l:

Items 9 and 10 value for pressurizer pressure (low and high) on (Reference 4) page 7.2-14 of the FSAR.

Issue-V TS Table 3.3-2, items 9 and 10 provide the maximum allowable pressurizer pressure (low and high) reactor trip response time which are greater than the nominal value given in chapter 7 of the FSAR.

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-Resolution (1

The licensee has updated page 7.2-15 in the FSAR to make reactor trip response time consistent with i

the TS for pressurizer pressure (low and high) trip functions.

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.0uestion 4c-Clarify whether the' reactor'is tripped due to-iTS Table 3.3-2, pressurizer pressure-low signal or pressurizer w,

Item-17

. pressure-low-low (ESFAS/ safety injection) signal.

during an accidental.depressurization of the ma'in steam system; and if so, include the appropriate 1

response time in Table 3.3 2.

Also, clarify terminology used in Note e'for Table 7.2.1-4 in the FSAR.

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U issue A.

TS Table 3.3-2 11sts,the reactor trip instrumentation response times.. Item 17 in the table lists the input response time as "N.A." for pressurizer pressure-low-low-(safety injection).

1This knuld 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 reference to a pressurizer low pressure-low level trip.

This should be pressurizer pressure-low-low (safstyinjection).

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Resolution ih A.

During the transient associated with depressurization of the main' steam system, the-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 octual 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 and-remains the same.

B.

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

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T ENCLOSURE 3-1 k

DP0 CONCERNS ON MCGUIRE TECHNICAL SPECIFICATIONS

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. RESOLUTION OF PLANT-SPECIFIC DP0 ISSUES REQUIRING NO LICENSEE ACTION 4

Question. Confirm the validity of McGuire Units 1/2 steam Table l2,2-l' generator. instrumentation, setpoint and their 7

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' (Reference 4).

applicability. McGuire Unit I has D-2 steam-generators and McGuire Unit 2~has D-3 SG.

7 I,ssue Steam Generators D-2 and D-3 have a minor design difference at SG bottom plate. Both SGs have-V-

identical instrumentation hardware and setpoint.,

c Resolution U

The licensee perfomed a conservative safety analysis which is applicable to both units.

Instrumentation setpoints values are based on thist analysis. Festinghouse RPS/ESFAS-setpoint methodology is ' applicable to both units and approved by the staff.

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Question laL Verify that a time constant of y 2 seconds results-

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in a slower response time which is less conservative.

Item 3 Issue TS Table 2.2-1 represents reactors trip system

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instrumentation trip.setpoints including response time. TS Table 2.2-1 Item 3 - concerns power =

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range, neutron flux, high positive-rate trip during ~

a control rod ejection accident.

Resolution An increased time constant results in a faster' response and thus results in a shorter time from initi6 tion 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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' Question Ib.

(1)

Verify that.a-time constant of- > 2' seconds result Table 2.2 in a slower response time which is less Item 4 conservative.

(2)

Resolve the~ inconsistency between setpcist 6

methodology value and FSAR analysis value.

Issues y

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

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~ Question Ic Resolve the disparity between the_setpoint TS Table 2.2-1~,-

methodology value and the FSAR safety analysis

' Item 9.

value.

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.

Resolution The licensee has indentified the correct value to be 1835 psig.

No change to the FSAR or TS was necessary.

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. Question Id--

Verify that the FSAR safety analysis value assumed TS Table 2.2-1 in the feedwater line break analysis is lower then

.. Item 13 the TS setpoint value.

Issue TS Table.2.2-1, item 13 lists steam generator water level-low-low reactor trip setpoint and allowable value.- The reviewer questions whether-the allowance for instrument error and uncertainties was applied in a conservative manner to arrive at the safety analysis value listed in the setpoint methodology document, Resolution

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The setpoint specified in the setpoint methodology document does suggest'a non-conservative application of the allowance for channel error and 9

-drift.

However, this value (i.e W STS + 10%) was not used in the McGuire TS. As discussed below, the allowance for instrument error and other uncertainties has been properly applied for 2.

McGuire.

D The licensee perfomed the limiting feedwater break analysis starting at full power and assuming a low water level trip setpoint of 23% narrow range-span. The McGuire TS limit for the SG low-low water level trip setpoint, at 100% rated thermal-power is 40% of narrow range span which exceeds the' safety eq1ysis value of 23% narrow range span by more than 10%.

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- Question le Clarify whether pressurizer pressure _- low signal Table 2.2-1 z

or' pressurizer pressure - low (safety injection)-

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- Item 18b signal trip the reactor during an accidental depressurization of the main steam system from zero-load.

Resolution An accidental depressurization of the main steam fu system (inadvertentopeningofadumpvalve,_

safety valve or relief valve) is initiated from hot 5

shutdown conditions at zero power which is the most conservative initial condition.

Reactor is already. tripped at the beginning of the transient (hot shutdown condition). Thus, no explicit 4

assumption is made regarding the cause of reactor d,

trip for the FSAR analysis, h'o credit is taken for the reactor trip on pressurizer pressure'when reactor power is below the P-7 interlock.

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7 Question'2 Clarify why the existing minimum temperature for TS '/ age 3/41-6 criticality (Modes 1/2)is551*Fwhichislessthan

- (TS3.1.1.4) the prograuned setpoint minimum.value of 557'F for

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events from zero power.

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The reviewer is concerned that transients or.

1 accidents may be initiated at zero power conditions from a temperature lower than the programmed setpoint minimum value of 557'F, i.e. the allowed-minimum-temperature for criticality of 551*F.

Resolutior, Accident evaluations for events from zero power

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arc 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 plantsithe minimum temperature for criticality is lower than hot zero power 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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1 Ouestion 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 g

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

Y Issue b

. Technical Specifications require 2 source range neutron flux channels be operable at all times except when in modes.3, 4 and! S with the reactor trip breakers open.

Reviewer suggested that-assumptions of boron dilution analysis-would-require 2 operable channels at all times.

Resolution The licensee has determined that-boron dilution events during modes 1, 2 and 6 were analyzed.for the McGuire units.. Consequently, the McGuire safety analysis'does not provide a basis for

-e recuiring two operable source range channels during modes 3, 4 and 5 of operation.

Tne licensee has considered changing technical specification 3.3.1 to require two operable source range channels at all times during operation in mode 3. 4 and 5; but has 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.

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~ Question'Sa Clarify whether applicable modes, Modes 1 and 2-#-

Table 3.3-3 is appropriate or'it should be modes _1 and 3 #

' Item;79, under P-11 interlock.

Issue TS Table 3.3-3 presents Engine'ered Safety Features --

Actuation System Instrumentation.

Item 79 specifies y,

_ applicable modes and operability requirements for auto-start of motor driven' auxiliary feedwater pumps (motor-driven pumps) on trip of all main p

feedwater pumps. The reviewer questioned whether this feature could be blocked during Mode 2 below the P-11 interlock because the threshhold for P >

could not be reached while in mode 2.

The # sign states that trip function ~may be blocked.

in this mode below the P-11 (pressurizer pressure interlock setpoint) and which can occur only -in rode 3, therefore. the reviewer believes that condition should be on mode'# 3.

Resolution The statement that P-11 can only occur in mode 3 is inaccurate. Mode 2 is defined as operation with T,yg & 350*F, k,ff D 0.99 and power f 5% RTP.-

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

Critical operation is restricted to T,yg ) 551*F, but even then the pressure-temperature operating limits permit pressures below 1955 psig. As a practical matter, pressure is maintained in the nonnal operating range (

2235 psig) during mode 2.

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.The defeat of auxiliary feedwater pump auto-st' art

.fs-accomplished by depressing a switch that -is

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interlocked with the P-11' permissive. Thus, the-auto-start can only be~ defeated below a' pressurizer 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, Jtem 7c(1).

Since this auto-start capabil ty is required in-Modes 1, 2 and 3, blocking is not allowed in these-modes. The # is misleading ai d will. be eliminated by'the licensee during the new STS development program.

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~ Question 6b Clarify TS items 7c(1) and 7c(2) concerning the Table 3.3-4..

' Auxiliary Feedwater system initiation 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) discuss-the auxiliary feedwater system initiation by the steam generator water level-low-low signal. -Information 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 i:, necessary to start the turbine driven pump. The reviewer questions-whether the level in the intact steam generator will be low enough during the feedline break inciaent to result in a start of the turbine driven AFW pump.

Resolution 1

In the case of a feedwater line break, the auxiliary feedwater system is des 1

',<>d to deliver 9

450 GPM by either turbine driven pump or two motor-driven pumps.to three intact steam generators.

while feeding one faulted generator.

I, the McGuire feedwater line break analysis, it was assumed that:

(1) the turbine driven pump failed as the single failure consideration; (2) One motor driven auxiliary feedwater pump supplies 110 gpm to an intact SG (the remainder spills out the break in the faulted loop); and (3) the other motor-driven pump supplies 170 gpm to each of the ether two intact steam generator; thus maintaining

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l450 gpm as-total flow to three intact system

-_ generators. These (;sumptions are consistent with the design of the AFW system instrumentation and-TS requirements for that instrumentation, s

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-In the case of a single failure of a motor driven-pump,.it is assumed that the turbine driven pump.

can actuate on low-low level in at least two steam generators. The licensee has calculated that during this accident condition, the mass inventory

'in the intact steam generators is reduced

- significantly prior to reactor trip on low-low level in the faulted loop. The shrinkage caused.by the bubble collapse from this reduced mass condition would cause low-low level to be reached in the other steam generators.

Thus, in the case of a motor-driven pump. single failure consideration, the turbine-driven pump can

[ actuate on low-low level in two steam generators and would maintain 450 gpm ficw distributicn 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.

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-Questionf6c'

' Confirm the bases for the setpoints-and allowable

Table 3.3-4;-

values as specified in the TS.

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Item 9-Issue v.

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.

1 Resolution The NRC staff issued a ' generic letter, dated August 12, 1976 requesting all licensees to analyze their Class IE electrical distribution system to determine if the operability of safety related equipment could be adversely affected by short term or long term degradation of grid system 9

voltage. ll supplemental genericiletter issued June 2, 1977 provided staff positions pertaining-to degraded grid voltage protection and the selection of voltage and time setpoints, and gc_

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 I to the SER.

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

~ Tabli 3.3.-5, item 3a

. analyses..

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hsue TS Table-3.3-5,-lists engineered safety features response time.

Items 2a and 3a provide Safety 4

Injection (ECCS) response time of 27 seconds (without offsite power) due to containment pressur' - high and pressurizer pressure-low-low -

e initiating signals during LOCA analyses, respectively.

Reviewer questioned the response.

tine between items 2a 3a and 4a.

Resolution No'LOCAs were analyzed for initial condition below P-11 interlock.

Low head safety injection pumps are required during the LOCA cases which results in a response time 'of 27 seconds -(without offsite power) for Items 2a and 3a as shown-in the table.

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Item 4a. represents the main steamline

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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.

Therefore, the additional 5 seconds delay for 1r,w.

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 Initiating:

TS Response Item Signal-Time _

2a. Safety Injection Containment Pressure-High 27 seconds

'(ECCS):

m 3a. Safety Injection Pressurizer Pressure-Low-Low 27/12 seconds (without/with off-site power)

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4a. Safety injection Steam Line Pressure-Low 22/12 seconds (ECCS)

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16 Question 7b and 79 Clarify the'2,0 seconds TS response. time value r

s Table'3.3-5, Item 2b

.versus the l.0 seconds value on FSAR Page 7,3-8 1 Table 3.3-5, Item 3b value. The descriptor (from SI) is incorrectLand should be deleted.

Issue a

TS Table-3.3-5, items Pb and 3b provide reactor.

trip (from SI signal) response time of 4 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 ts the limit on the delay in receipt of Si 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-cf 1.0 second is the tine it takes to generate e safety injection signal. The description "(from SI)" is' correct in that the 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.

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(17-Question 7cland 7h-Justify the TS values used for containment isolation.

' Table 3.3-5, item 2d-valves closure time for LOCA analyses.

= Table ~3.3-5, item 3d-issue 1

f TS Table 3.3-5, Items 2d and.3d list containment isolation-phase "A" I2) 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.

The process lines with fluids will take longer time to close in corpparison to the purge valves.

The process lines valves will close in about 18.. seconds (with offsite power). However, ANSI N271-1976/ANS 56.2, "Containw at Isolation Provisions for Fluid Systems" recommends that, in general. closure times should be as low as reasonably attainable, based on manufacturers' reconsnended times and valve sizes, but generally not less than 15 seconds and in any m

case, no more than one minute.

If these guidelines are met, releases through these process line valves before closure need not be modeled in the dose calculation.

Therefore, the TS containment isolation valves closure time of 18 seconds is acceptable.

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Question 7e; Clarify the TS concerning auxiliary feedwater-

,iTable'3.3-5, system initiation on Containment Pressure-High item 2f in Nodes 3 and 4.

a Issue I

TS Table 3.3'-5, item 2f provides auxiliary feed-water system response time for actuation from a~

containaent pressure-high initiating signal'as; "N.A."

Resolution FSAR accidents analyses do not take any credit for actuation of the' auxiliary feedwater_ system from a containrent pressure-high signal. Consequently..

N.A. has been entered for the response' time in x

table 3.3-5.

However, the.TS Table 3.3-5, Note.5 clarifies that tha response time for motor-driven auxiliary feedwater_ pumps on all safety ' injection signals shall be less than or equal to 60l seconds.

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Response tine limit. includes opening of valves to establish. safety >1njection_ path and attainment of discharge pressure-_for auxiliary,feedwater pumps.

The AFW response time _as "N.A." is acceptable.

Xu 19 Question 7j Clarify the TS' concerning auxiliary feedwater Table 3.3-5, system under pressurizer-pressure-low-low Item 3f.

> initiation signal.

Issue' 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 steam 11ne depressurization event (inadvertent opening-of a steam generator' safety, relief or dump valve) assumes ESF actuation on pressurizer pressure-low-low initiating signal.

For this event it is conservatively assurred that auxiliary feedwater is actuated at the maximum flow rate at the initiation of the event to accentuate the cooldown. Any delay in auxiliary feedwater actuation would.be-beneficial and therefore a response time requirement is not applicable or appropriate.

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4 Question-7m Confirm that the TS containment spray response' i

Table 3;3.-5.-

time'and FSAR analysis value are ' consistent.-

w Resolution o

TS Table 3.3-5,-Item'5a lists containment: spray

response time of dG 45' seconds following a-contain-;

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ment ~ ptcssure-high-high initiating signal.1TS j

response _ time of.45 seconds is consistent with the J

FSAR.conta1nmeni :nalysis actuation assumption as.

shown in FSAR Table 6.2.1-16.

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21-Question 70L

. Confirm that' the TS automatic switchover to '

't 1 Table 3.3-5,

~ recirculation response time is consistent with the

!! tem 12-FSAR assumption.

Issue 1

TS Table 3.3-5. Item 12 lists response time 4F 60 seconds for automatic switchover to recirculation resulting from a refueling water storage tank (RWST) level initiating signal. The reviewer a

questioned the basis for this value.

Resolution The containment sump valves are interlocked with-the RWST isolation valves to the RHR pumpsLsuch that these isolation valves will close when the contain-ment sump valves reach their full open position.

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

The automatic switchover'to recirculation is initiated when the level setpoint-is reached in the.

RWST.

The plant procedures as delineated in FSAR Teble 6.3.2-3A/3B test to ensure switchover delay of 60- seconds which is consistent with the TS response. tine.

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^N QusstionL9l Justify TS action requirement.to restart an idle Page 3/4 4. loop when in Mode 3 with no reactor coolant-loops-TS 3.4.1.2 in operation; or explain how natural circulation

. is accomplished with emergency procedures.

Issue.

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

coolant loop in operation, suspend all operations involving a-reduction'in boron concentration of the RCS and imediately initiate corrective action to return the required reactor coolant looptto operation." The reviewer questions the basis'for these procedural actions and prepares alternate action which is to implement an E0P for natural circulation.

Resolution 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 is 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 ercered following a reactor trip or safety injectio..

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4,,U g3 e Ji Question lla The operator aligns. the Residual Heat Removal-DTS Section 3.4.5 System at less than 400 psig and 350'F._ Tim l

valves in the.line from the RWST are closed.

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.The " question" is merely a statement of operator

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- ce actiore to align' RHR.

It remains true and requires n

no response.

LOCAs in lower modes of operation and loss of RHR cooling in lower modes will be addressed L

generically in Question Sb.

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-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 place the safety injection, centrifugal charging, and residual heat removal pumps along with 51 accumulator in ready and then manually actuate SI'.-

Resolution This " question" is a statement of operator action to align the ECCS for use from a shutdown-condition.

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 Ouestion 5b.

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' The question is-not clearly stated.

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Resolution j

This " Question" is largely a quotation from the i

FSAR. The last two' paragraphs are statement introductig a quotation frem the SER.-'This question requires no response.

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

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26 Question !?a Explain why TSAR value for nitrogen cover-pressure TS 3.5.3.3.d of cold leg accumlators should not be of higher value to account for channel error and drift consideration.

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

TS setpoint value is also 400 psig.

How do we account for channel error and drift consideration?

j Resolution Since the VH1 system is removed, the licensee p

revised the value for nitrogen cover-pressure of cold leg accumlator to 585 psig in comparison to l

a00 psig with UHI accumlator.

The alarm is set at 590 psig to account for channel error and drift j

considera tion, e

L In the near future the licensee will-consider the channel error ar.d drift values in the safety analysis when l

they reviz,e the LOCA analyses to meet the SG tubes pluggino requirement..The safety enalysis value will be 564 psig and the TS value will remain the same 585 psig.

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. c :s.. o Question !?b Verify that the accumulators relief valves TS 4.5.1.1.1.d.1 setpoints are included in the Inservice Testing program.

Resolution The cold leg accumulators rel!ef valves are not required to perform a safety function either to shutdown the reactor or to mitigote the consequences of an accident.

Therefore, these valves are not included in the IST program.

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

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5 o4 28 Ouestion 13 Verify the water temperature value used in the j

TS 3.5.1.2.d safety analysis for UHI accumulator.

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Yerify that the accumulator relief valve setpoint is included in the Inservice Testing Program, i

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

,(2) Should the accumulator relief valve setpoint be in j

the IST program.

Resolution (1) The safety analysis value related to UHI l

accumulator water temperature is assumed to be the upper bound value of 100'F. Since the UHI accumulator is not heated or located inside containment, there is no real mechanism for t

I increasing temperatures during operation.

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

(2) The UHI accumulator relief valve is not required to perform a safety function either to shutdown i

the reactor or to mitigate the consequences of an accident.

Therefore, it is not in the IST

program, u

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McGuire Units 1/2 are ice condenser plants with 3:

Upper Head Injection system. Experience has demonstrated that the UHI system adds to the l

complexity of plant operation, requires additionel l.

maintenance and generally reduces plant L

availability.

The TS Amendment 57 (Unit 1) and 3P L

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! Unit 2) approved the removal of the UHI system for McGuire Units 1/P.

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29 Ouestion 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.

f Resolution The ECCS flw' assumed in the LOCA analyses are the bases for the limits as specified in TS 4.5.2.h.

Flow balance tests are perfomed du *ing shutdown to account for any change in the subsystem flow L

characteristics to ensure adequate flow for ECCS e

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consideration.

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

The flow balance tests will place limits on the l

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 i

TS 3/4.7.5' solid layer of ice" forms on the Standby Nuclear Service Water Pond (SNSWP), the operating train is manually aligned to SNSWP.

Provide safety-related l

reason for this action.

Resnlution 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.

In ten years of operation, the licensee never experienced this kind of situation or any l'

cperating problems.

Therefore, the licensee deleted this action and description from the FSAR and does not require any TS surveillance for this system.

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Question 18 Why TS are not applied to flow control valves TS 3/4.9.1 INV-171 A and INV-175 A7

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Resolution i

Surveillance Requirement 4.9.1.3 requires that valve #INV-250 shall be verifled locked closed i

urder administrative controls at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> during refueling operation. This valve is

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upstream of valves INV-171 A and INV-175 A and i

isolates the flow path to prevent the inadvertent dilution of the RCS boron concentration.

Therefore INV-17. A and INV-175 A bre not part of TS.

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_ REFERENCES l

l 1.

Letter from Robert Licciardo to Brian Sheron, " Review of McGuire l

Technical Specifications," dated June 11, 1984.

2.

Letter from Thomas Novak to H. B. Tucker, " Request for Comments on McGuire Technical Specifications Concerns Resulting from Differing Professional Opinion," deted July 9,1085.

3.

Letter from H. Thompson to R. Bernero, " Disposition of Concerns Raised by i

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," dated June 10, 1986.

5 Memorandum from Thomas Murley to Robert Licciardo " December 7, 1983 j

Differing Professional Opinion " dated December 29, 1989.

6.

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

l 7.

NUREG-0964, " Technical Specifications McGuire Nuclear Station Unit Nos. I and 2 " dated March 1903, 8.

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

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

9.

Duke Power Company, McGuire Nuclear Station Final Safety Analysis Report l

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

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

11. Generic Letter 85-05. " Inadvertent Boron Dilution Events," January 85.

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12. Letter from George Lear to D. C. Switzer, " Millstone Nuclear Power

_ Station Units 1 and 2." dated June 1977.

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