ML20236N470

Proposed Tech Specs Reformatting Operability & Surveillance Requirements for Intermediate Range Channels & Revising Nominal Intermediate Range High Flux Trip Setpoint

ML20236N470
Person / Time
Site:	North Anna
Issue date:	11/09/1987
From:	VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:
Shared Package
ML20236N462	List: ML20236N460 ML20236N470
References
NUDOCS 8711160178
Download: ML20236N470 (27)

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I LIMITING SAFETY SYSTEM SETTINGS BASES The Power Range Negative Rate rip provides protection for control rod drop accidents. At high power, a rod drop accident could cause local flux j

peaking which could cause an unconservative local DNBR to exist. The Power Range Negative Rate Trip will prevent this from occurring by tripping the i

reactor. No credit is taken for operation of the Power Range Negative Rate Trip for those control rod drop accidents for which the DNBR's will be greater than the applicable design limit DNBR value for each fuel type.

Intermediate and Source Range, Nuclear Flux, f

The Source Range Nuclear Flux trip provides reactor core protection during l

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shutdown-(Modes 3, 4 and 5) when the reactor trip system breakers are in the closed position. The Source and Intermediate Range trips in addition to the Power Range trips provide core protection during reactor startup (Mode 2).

Reactor startup is prohibited unless the Source, Intermediate and Power Range trips are operable in accordance with Specificati g 3.3.1.1 The Source Range Channels will initiate a reactor trip at about 10 counts per second unless manually blocked when P-6 becomes active. The Intermediate Range Channels will initiate a reactor trip at a current level proportional to approximately

35. percent of RATED THERMAL POWER unless manually blocked when P-10 becomes active.

In the accident analyses, bounding transient results are based on i

reactivity excursions from an initially critical condition, where the source range trip.is assumed to be blocked.

Accidents initiated from a suberitical condition would produce less severe results since the source range trip would provide core protection at a lower power level. No credit was taken for operation of the trip associated with the Intermediate Range Channele in the j

accident analyses; however, their functional capability at the specified trip l

settings is required by this specification to enhance the overall reliability

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of the Reactor. Protection System.

Overtemperature AT

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The Overtemperature AT trip provides core protection to prevent DNB for all l

combinations of pressure, power, coolant temperature, and axial power distribution, provided that the transient is slow with respect to piping q

transient delays from the core to the temperature detectors (about 4 seconds),

and pressure is within the range between the High and Low Pressure reactor j

trips. This setpoint includes corrections for changes in density and heat d

capacity of water with temperature and dynamic compensation for piping delays j

from the core to the loop temperature detectors. With normal axial power

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distribution, this reactor trip limit is always below the core safety limit as shown in Figure 2.1-1.

If axial peaks are greater than design, as indicated by the difference between top and bottom power range nuclear detectors, the reactor trip is automatically reduced according to the notitions in Table 2.2-1.

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TABLE 3.3-1 (Continund)

TABLE NOTATION

• With the reactor trip system breakers in the closed position and the control rod-drive system capable of rod withdrawal.

• • The channel (s) associated with the protective functione derived from the out of service Reactor' Coolant Loop shall be placed in the tripped

. condition.

With the Reactor Trip Breaker open for surveillance testing in accordance with Specification Table 4.3-1 (item 21A).

'# The provisions of Specification 3.0.4 are not applicable.

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I High voltage to detector may be de-energized above P-6.

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1. 1. 1. Below the F-10 (Low Setpoint Power Range Neutron Flux Interlock) Setpoint' l

ACTION STATEMENTS j

l ACTION 1 -

. With the number of channels OPERABLE one less than required by i

I the Minimum Channels OPERABLE requirement, be in HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; however, one channel way be bypassed for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for surveillance testing per Specification 4.3.1.1.1 provided the other channel is operable, j

ACTION 2 -

With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and POWER OPERATION may proceed

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provided.the following conditions are satisfied:

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a.

The inoperable channel is placed in the tripped condition j

within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

-b.

The Minimum Channels OPERABLE requirement is met; however, the inoperable channel may be bypassed for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for surveillance testing of the redundant channel (s) per Specification 4.3.1.1.1.

c.

Either, THERMAL POWER is restricted to s 75% of RATED j

THERMAL and the Power Range, Neutron Flux trip setpoint is i

reduced to s 85% of RATED THERMAL POWER within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />; or, the QUADRANT POWER TILT RATIO is monitored at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

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d.

The QUADRANT POWER TILT RATIO shall be determined to be I

within the limit when above 75 percent of RATED THERMAL POWER with one Power Range Channel inoperable by using the moveable incore detectors to confirm that the normalized symmetric power distribution, obtained from 2 sets of 4

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symmetric thimble locations or a full-core flux map, is

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consistent with the indicated QUADRANT POWER TILT RATIO at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

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ACTION 3 -

With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement and with the THERMAL POWER level:

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' a '.. I Below P-6, ; restore ' the ' inoperable ' channel to !0PERABLE --

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iTHERMAL POWERiabove the P-10 setpoint.

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1 Tc. : LAbove the P-10 setpointi' POWER.0PERATION may continue.

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'. status prior to increasing:. THERMAL POWER. above' the.P-6 Setpoint.'

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JAbove.P-6,L. operation may-continue.

-ACTION 5 With'the number of channels OPERABLE?one less.than' required by

the' Minimum ~ Channels' OPERABLE' requirement, verify compliance j

with the SHUTDOWN MARGIN requirements of. Specification 3~.1.1.1-i

.or'3.lil'.2Was applicable, within'I hour and,at least-once per g

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.12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />s:thereafter.

JACTION 6 -

Not applicable..

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With the. number of OPERABLE channels'one-less than the Total.

h Number of Channels, STARTUP_and POWER: OPERATION may. proceed

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until-performance of thesnext required CHANNEL FUNCTIONAL TEST-provided:the inoperable channel is placed in.the. tripped-condition within I hour.-

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' ACTION 8' Not applicable t

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TABLE 4.3-1 (Continued) l l

NOTATION With the reacter trip ::~ tem breakers closed and the control rod drive syatem capable ot vod withdrawal.

Below P-10 (Low Setpoint Power Range Neutron Flux Interlock).

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Setpoint.

j (1) -

If not performed in previous 7 days.

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(2) -

Heat balance only, above 15% of RATED THERMAL POWER.

(3) -

Compare incore to excore axial offret above 15% of RATED THERMAL POWER. Adjust changal if absolute difference 2 3 j

percent.

Manual ESF functional input cbeck every 18 months.

(4) 1 Each train or logic channel shall be tested at least every 62 days on (5) a STAGGERED TEST BASIS.

(6) -

Neutron detectors may be excluded from CHANNEL CALIBRATION.

i (7). -

The CHANNEL FUNCTIONAL TEST shall independently verify the l

OPERABILITY of the undervoltage and shunt trip circuits for the q

manual reactor trip function.

The test shall also verify the operability of the Bypass Breaker Trip circuit (s).

(8) -

Local manual shunt trip prior to placing the bypass breaker into service.

1 Automatic undervoltage trip.

(9)

(10) -

The CHANNEL FLNCTIONAL TEST shall independently verify the OPERABILITY of the undervoltage and shunt trip attachments of the Reactor Trip Breakers.

(11) -

Monthly Surve111~ance in Modes 3*,

4*,

and 5* shall also include j

verification thet Permissives P-6 and P-10 art in their required

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state for existing plant conditions by observatio. if the permissive

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annunciator window.

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(12) -

Detector plateau curves shall be obtained and evalua-The l

provisions of specification 4.0.4 are not applicable

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LIMITING SAFETY SYSTEM SETTINGS BASES Intermediate and Source Range, Nuclear Flux

- The Source Range : Nuclear Flux trip -provides reactor core protection during shutdown' (Modes 3, 4 and 5) when the reactor trip system breakers are

_in the closed: position.

The Source and Intermediate Range trips in addition to the. Power. Range. trips provide core protection during reactor startup-(Mode 2).

Reactor startup is prohibited unless the Source, Intermediate and Power Range trips are operable in accordance with Specification.3.1.1. The Source

'{

Range. Channels will initiate a reactor trip at -about 10 counts:per second

)

unless manually blocked when P-6 becomes active.

The Intermediate Range Channels will. initiate a reactor trip at a current level proportional to approximately 35 percent' of RATED THERMAL POWER unless manually blocked when P-10 ' becomes active.

In.the accident analyses, bounding trancient results are based on reactivity excursions from an initially critical condition, where the source range trip is assumed to' be - blocked. _ Accidents initiated from a subcritical condition would produce less severe results since the source range trip would provide core protection at a lower power level.

No credit was taken for operation of trip associated with the Intermediate Range Channels in the. accident analyses; however, their functional capability at - the specified trip - settings is required by this specification to enhance the overall reliability of the Reactor Protection System.

Overtemperature Delta T The Overtemperature Delta T trip provides core protection to prevent DNB for all combinations of pressure, power, coolant temperature, and axial power distribution, provided that the transient is slow with respect to piping transit delays from the core to the temperature detectors (about 4 second),

and pressure is within the range between the High and Low Pressure reactor trips.

This setpoint includes corrections for changes in density and heat capacity of water with temperature and dynamic compensation for piping delays from the core to the loop temperature detectors.

With normal axial power distribution, this reactor trip limit is always below the core safety limit as shown in Figure 2.1-1.

If axial peaks are greater than design, as in<licated by the dif ferenca between top and bottom power range nuclear detectors, the reactor trip is automatically reduced according to the notations in Table 2.2-1.

l' Operation with a reactor coolant loop out of service below the 3 loop P-8 b

set point does not require reactor protection system set point modification

[

because the P-8 set point and associated trip will prevent DNB during 2 loop operation exclusive of the Overtemperature Delta T set point.

Two loop operation above the 3 loop P-8 set point is permissible af ter resetting the K1, K2 and K3 inputs to the Overtemperature Delta T channels and raising the P-8 set point to its 2 loop value.

In this mode of operation, the P-8 interlock and trip functions as a High Neutron Flux trip at the reduced power level.

NORTH ANNA - UNIT 2 B 2-4

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TABLE 3.3-1 (Continued)

TABLE NOTATION

• With the reactor trip system breakers in the closed position and the control' rod drive system capable of rod withdrawal.

• • The channel (s) associated with the protective functions derived from the out of service Reactor Coolant Loop shall be placed in the tripped condition, j

• • • With the Reactor Trip Breaker open for surveillance testing in accordance

'i with Specification Table 4.3-1 (item 21A).

1 The provisions of Specification 3.0.4 are not applicable.

1. C High voltage to detector may be de-energized above the P-6, (Block of Source Range Reactor Trip), setpoint.

.### -Below the P-10 (Low Setpoint Power Range Neutron Flux Interlock)

Setpoint.

ACTION STATEMENTS ACTION 1

- With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirements, be in HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; however, one channel may be bypassed for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for surveillance testing. per Specification 4.3.1.1.1 provided the other channel is OPERABLE.

With the number of OPERABLE channels one less than the Total ACTION 2 Number of Channels, STARTUP and POWER OPERATION may proceed provided the following conditions are satisfied:

a.

The inoperable channel is placed in the tripped condition within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

b.

The Minimum Channels OPERABLE requirement is met; however, the inoperable channel may be bypassed for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for surveillance testing of the redundant channel (s) per Specification 4.3.1.1.1.

c.

Either, THERMAL POWER is restricted to s 75% of RATED THERMAL POWER and the Power Range, Neutron Flux trip setpoint is reduced to % 85% of RATED THERMAL POWER within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />; or, the QUADRANT POWER TILT RATIO is monitored at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

1 d.

The QUADRANT POWER TILT RATIO shall be determined to be within the limit when above 75 percent of RATED THERMAL POWER with one Power Range Channel inoperable by using the movable incore detectors to confirm that the normalized symmetric power distribution, obtained from 2 sets of 4 l

symmetric thimble locations or a full-core flux map, is consistent with the indicated QUADRANT POWER TILT RATIO at l

least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

i NORTH ANNA - UNIT 2 3/4 3-5 Amendment No.

]

1 1

l TABLE 3.3-1 (Continued)

ACTION 3

- With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement and with the THERMAL POWER level:

a.

Below the P-6 (Block of Source Range Reactor Trip) setpoint, restore the inoperable channel to OPERABLE status prior to increasing THERMAL POWER above the P-6'Setpoint.

b.

Above the P-6 (Block of Source Range Reactor Trip) setpoint, but below the P-10 setpoint, restore the inoperable channel to OPERABLE status prior to increasing ' THERMAL POWER above the F-10 setpoint, c.

Above the P-10 setpoint, POWER OPERATION may continue.

ACTION 4 With the -number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement and with the THERMAL POWER level:

a.

Below the P-6 (Block of Source Range Reactor Trip) setpoint, restore the inoperable channel to OPERABLE status prior to increasing THERMAL POWER above the P-6 Setpoint.

b.

Above the P-6 (Block of Source Range Reactor Trip) setpoint, operation may continue.

With the nuuber of channels OPERABLE one less than required by ACTION 5 the Minimum Channels OPERABLE requirement, verify compliance with the SHUTDOWN MARGIN requirements of Specification 3.1.1.1 or 3.1.1.2,-as applicable, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter.

Not applicable.

ACTION 6 With the number of OPERABLE channels one less than the Total ACTION 7 Number of Channels, STARTUP and POWER OPERATION may proceed until performance of the next required CHANNEL FUNCTIONAL TEST provided the inoperable channel is placed in the tripped condition within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

Not applicable.

ACTION 8 NORTH ANNA - UNIT 2 3/4 3-6 Amendment No.

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TABLE 4.3-1 (Continued)

NOTATION With the reactor trip system breakers closed and the control rod drive systam capable of rod withdrawal.

Below P-10 (Low Setpoint Power Ranga Neutron Flux Interlock) Setpoint.

If not performed in previous 7 days.

(1)

(.2 ) -

Heat balante only, above 15% of RATED THERMAL POWER. Adjust channel if absolute difference > 2 percent.

(3) -

Compare incore to excore axial offset above 15% of RATED THERMAL POWER. Recalibrates if the absolute difference 2 3 percent.

Manual ESF functional input check every 18 months.

(4)

(5) -

Each train or logic channel shall be tested at least every 62 days on a STAGGERED TEST BASIS.

Neutron detectors may be excluded from CHANNEL CALIBRATION.

(6)

(7) -

Below the P-6, (Block of Source Range Reactor Trip), Setpoint.

(8) -

The CHANNEL FUNCTIONAL TEST shall independently verify the OPERABILITY of the undervoltage and shunt trip circuits for the Manual Reactor Trip Function.

The test shall also verify the OPERABILITY of the Bypass Breaker trip circuit (s).

(9) -

Local manual shunt trip prior to placing the bypass breaker into service.

(10) -

Automatic undervoltage trip.

(11) -

The CHANNEL FUNCTIONAL TEST shall independently verify the OPERABILITY of the undervoltage and shunt trip attachments of the Reactor Trip Breakers.

(12) -

Monthly Surveillance in Modes 3*, 4* and 5* shall also include verification that permissives P-6 and P-10 are in their required state for existing plant conditions by observation of the permissive annunciator window.

(13) -

Detector plateau curves shall be obtained and evaluated. The provisions of specification 4.0.4 are not applicable for entry into Mode 2 or 1.

I l

I NORTH ANNA - UNIT 2 3/4 3-14 Amendment No.

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Safety Evaluation

Proposed-Technical Specification:s Changes b

The Intermediate _ Range Nuclear Flux Trips provide backup reactor core protection during reactor startup.

The Intermediate Range channels are fed by two redundant gamma compensated ion chambers which are located external to the core in air, cooled ' instrument wells along a major core axis.

The IR circuitry provides ruonitoring of the flux level over an eight' decade range (10**-11'to 10**-3 amperes).

A reactor trip is gen-erated based on one out of two channels exceeding a current equivalent

.to 25% of rated thermal' power. The channels can be manually bypassed when permissive P-10 (2 of 4 power rango channels > 10 % of rated thermal power) is active.

No credit was taken for operation of the trips associated with the IR j

channels in the accident analyses. Redundant protecticn against reactiv-ity ' and power excursions during startup is provided by the following channels.

?

1. Power range neutron flux, low setpoint (2/4 channels = 25% of rated thermal power).

2. Power range neutron flux, high setpoint (2/4 channels = 109% of rated thermal power).

i

3. Power range neutron flux rate (2/4 channels = Impulse signal equivalent to 5% of rated power with a time constant =

l 2 seconds).

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Hender, the Technical Specifications changes submitted. In. Reference :1 t a.

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support the~ core power uprating, the source range flux trips-must be op-S?

'erable whenever.the. control rods are energized and capable:of withdrawal l

y duhing th'e shutdown modes!(modes 3, 4 and 5). ' Conversely,.the' power range

'l trips must,be' operable during modes 1 and 2.

Furthermore, during modes H

4 3-6?the Technical Specifications require that all potential boron di '

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' 'lution sources be locked,-~ sealed or otherwise ' secured closed except during;

' ' ) planned dilution or meheup activities. As a result, there is no operating condition under which the intermed'iate range trip provides sole overpower

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. protection.

It'is therefore a backup trip only, and.no credit is taken q

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,for the trip in the-safety analysis.

1 j

'Because the:IR flux detectors are located outside the core, the IR signal

.l

.has been shown historically to be.se~nsitive-to the core loading pattern q

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in use.

For example, the long-lived low leakage patterns currently in use at: North - Anna have a different IR detector response than the more t.raditional type of pattern used for the initial core loadings.

In ad-dition,-because the detectors do not cover as much of the full core length i

.~as the power range channols, the detector response is also sensitive to the core axial flux distribution.- As a result, such effects as varying j

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' core burnups or' control rod positions can also have a significant impact-j on the IR channel response. The variability in the channel. response has j

Jmade it difficult to maintain the channels in proper calibration.

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. References ' 2 ? and 3 provided / Licensee Event. Reports (LER's). de' scribing

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Linstances,in-which the: IR high' neutron flux trip setpoints were higher-

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thanLthe current equivalentEto'30%.o'f' Rated Therm'al Power (RTP)', which j

.n lis, tbe allowable 'valuetspecified in Section 2.2-1 of the Technical Spec-

'ifications..'The Reference 3 event, which occurred on Unit 2, was-the more l

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> severe violation,'in that.the effective trip setpoint'was determined to~

l be equivalent to approximately 55% of RTP, This condition existed shortly--

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after the Cycle 2/3 refueling.

J5.

Asa' result of these difficulties, Virginia Electric'and. Power Company 1

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1s,proposingajsetofchangestotheTechnicalSpecificationwhichgovern j

the' operability requirements and surveillance testing of the Intermediate R

'I Range t channels. : The proposed' changes are consistent with the current u.

Standard Tedhnical Specifications. In addition, the changes would result -

1 i

in the nominal Intermediate Range high flux trip setpoint being elevated from a current equivalent to 25% of rated thermal power to a current c

equivalent to 35% of rated thermal power.

The paragraphs below provide a detailed discussion and evaluation of the proposed changes.

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1. Revised Operability and Calibration Format

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. Attachments 1 and 2. provide revised copies of Tables 3.3-1 and 4.3-1 of the ~ Units 1 and 2 Technical Specifications, respectively. The proposed revisions provide channel operability and surveillance requirements for i

the Intermediate Range channels which are consistent with those in NUREG-0452, Revision 4,

" Standard -Technical Specifications for Westinghouse Pressurized Water Reactors" (Fall 1981). A point by by point discussion of the proposed revision follows.

A. In the APPLICABLE MODES column of Tabic 3.3-1, a footnote is added for Mode 1 to limit applicability to power levels below the P-10 (Low Setpoint Power Range Neutron Flux Interlock) setpoint.

Thus for power levels above 10%, the channels are not required to be OPERABLE. Under the existing specifications, a) a minimum of 2 IR channels are always required to be operable in Modes 1 and 2, with one channel to trip, and b) with less than l

2 channels an action statement is entered, in which operation

]

above 57, of RATED THERMAL POVER is allowed, but only if one of the two channels is operable.

Thus, under the proposed-changes 1

unneeded surveillance of the channels during periods when they

)

are providing no protection is eliminated.

I B. The requirement to have the IR channels operable whenever the reactor trip breakers are in the closed position and the rod drive system is capable of withdrawal has also been removed.

This

'g' requirement extended the IR channels down into modes 3-5 when the

{

trip breakers are closed. With the changes proposed in Reference 1, this extension is not required, since modes 3-5 are protected by the 1/2 Source Range trip logic whenever the trip breakers are closed.

C. Under the surveillance requirements for the Intermediate Range I

trips, a note has been added to the APPLICABLE modes column

.to limit applicability in Mode 1 to power levels below the P-10 interlock. The specified surveillance program continues to be applicable in Mode 2.

The requirement to perform channel checks whenever the trip breakers are closed in the other modes has been replaced by a note requiring monthly verification that Permissives P-6 and P-10 are in their required states for existing plant conditions by observation of the plant annunciator window.

Finally, the channel calibration requirement, to be performed 4

once/ refueling, has been annotated to require taking and evaluating detector p.'.ateau curves, consistent with the Standard Technical Specifications. The note also explains that the surveillance requirements for channel calibration need not be completed prior to enturing into the applicable modes.

Sucunarizing, the revised format requires operability and surveillance requirements for the IR trips for those conditions where the Intermediate 1

Range trips perform a backup function to the Power Range low setpoint trips, i.e. in Modes 1 and 2 below the P-10 interlock.

The requirement

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j t-verification (via permissive checks) that the power range low setpoint l

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and source range trips cannot be blocked below the-appropriate power

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l levels and therefore their ranges of protection are properly overlapped.

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2. Fevised Intermediate Range' Trip Setpoint I

Attachments 1 and 2 also provide revisions to Table 2.2-1 of the Units 1 I

i and 2 Technical Specifications, respectively.

As shown the IR setpoint has been raised from a nominal value of 25% of RATED THERMAL pcwcr to 35%.

Under the current nominal setpoint, the, variability in IR channel response with control rod position is such that in some conditions very little operating space exists between the P-10 permissive power level (which allows the operator to block the Intermediate Range trip function) and the power level at which the IR trip occurs.

As a result, there is a-potential for unneeded unit trips due to inadequate response time avail-able to the operator for blocking the trip during power ascension.

If the nominal setpoint is raised to 35%, the minimum amount of operating space between the interlock and a potential trip is essentially doubled, allowing for more time to block the trip when normal power ascension is occurring and its protective function is not required.

As shown in Attachments 1 and 2, the allowable IR trio setpoint has been -

raised to 40 % of Rated Thermal Power. The 5% margin between the nominal and allowable setpoint is retained to reflect the effects of instrument drift between surveillance tests.

This drift value is consistent with i

w-4 that allowed in the Standardized Technical Specifications and by the go-neric, Westinghouse setpoint methodology presented in Reference 4.

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3. Evaluation of Elevated IR Setpoint Yo11owing the standard Westinghouse setpoint methodology submitted in i

Reference 4, the difference between the nominal setpoint and the safety analysis value for a trip or actuation setpoint is developed from a sta-tistical combination of error terms which include process measurement i

accuracy, rack and sensor calibration accuracy, rack' and sensor drif t and

,j the effects of temperature and pressure variations on process instrumen-tation and the' associated protection system electronics.

Using the methodology of Reference 4, Virginia Electric and Power Company performed l

4 I

an e valuation of. the total channel error for the intermediate: range.

channels.

As might be expected, the major contribution to this ~ erro.r comes from the process measurement accuracy (PMA) term.

As applied to the IR channels, PMA includes the variability in the signal due to control j

rod shadowing and core burnup effects.

Variations with loading pattern t

were assumed to be calibrated out at the startup of each new fuel cycle.

j The PMA term.was developed based on an evaluation of the variation in radial power distributions with cycle burnup and rod position for several reload cycles for each unit.

Data for 9 different loading patterns were

)

.i examined, and a PMA term was selected which conservatively envelopes the data for all cycles.

Evaluation of the data resulted in a total statistical channel accuracy j

l of +/- 12% of rated thermal power for the IR channels, based on an assumed I

nominal setpoint of 35% of rated thermal power. Thus the highest expected setpoint for the proposed nominal value is 35% + 12% or about 47% of rated l

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thermal power.

This compare's to an expected upper limit of about 35% of rated thermal power for the current nominal setpoint.

It has'been previously demonstrated (see Reference 5) that raising. the effective flux trip setpoint from an assumed safety analysis value of 35%

of rated power to as much as 118% of rated power.has an insignificant.im-pact on low power reactivity excursion events such as rod withdrawal from subcritical and rod ejection. It is therefore concluded that the proposed revision to. the nominal and. allowed IR setpoints will not impair the ability'of. these trips to 'oack up the power range channels in protecting against react'ivity-and power excursions initiated froc }ow reactor power.

I The proposed changes do not involve a significant hazards consideration because operation of North Anna Units 1 and 2 in accordance with these

.would not:

r

1. involve a significant increase in the probability or con-sequences of an accident previously evaluated. There is no adverse impact on the safety analysis (since no credit is taken for the' trips in the existing analyses), and no degradation of the protection system redundancy or reliability.

This-latter conclusion is based on sensitivity studies which show that th'e effectiveness of the flux trip system in protect-ing against.the low power reactivity excursions examined in the j

i

-FSAR is not sensitive to realistic variations in the actual flux i

trip setpoint.

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create the probability of a new or different kind of accident from any accident previously identified, since the severity of

.the analyzed accidents is unchanged, and since only.a change l

to a setpoint and'the associated surveillance requirements'for

.the reactor protection system is involved.

3. involve a significant reduction in a margin of safety, since none'of thel safety analysis input or assumptions are changed, g,

nor are the probability nor the consequences of any previously analyzed accidents increased.

Therefore, pursuant to.1'O CFR'50.92 based on the above considerations it

[has been determined that this change does not involve a-significant safety hazards consideration.

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E REFERENCESJ

$1.' Letter;from W. L. Stewart (VP) to H. R.;Denton.(NRC), " Response to

~ Request for' Additional Information on Core Uprate", Serial No.

lj 85 -7 72A,1-February -_.6,1986.

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2. Letter from E. W..Harrel'1':(Vepco) to J.'P.:0'Reilly'(NRC),-Serid1'No.

up 1N-83-094', transoiltting LER 83-038/03L-06,LJuly 6, 1983.

+

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' 3.. ' Letter from E. W. : Harre11' (Vepco) to.J. P. O'Reilly '(NRC), Serial No.

N-83-095, transmitting'LER 83-038/03L-07, July.6,.-1983.

-_s J

--4! Westinghouse Report, i' Westinghouse Reactor Protection System /.-

Engineered Safety. Features' Actuation System Setpoint Methodology",

L(Westinghouse. Proprietary Class.2), transmitted by letter from

.C..M.

Stallings-(Vepco) to.H. R. Denton,(NRC), Serial No.'541,.

September 29,1978.

-5. " Letter :from W. L. ' Stewart (VP); to ' H. R.E Denton (NRC),'_ Serial No.

-523A,' February, 14,:1985. ':

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