ML19332G119

Proposed Tech Specs Re Reactor Trip Sys Instrumentation & ESFAS

ML19332G119
Person / Time
Site:	Summer
Issue date:	12/11/1989
From:	SOUTH CAROLINA ELECTRIC & GAS CO.
To:
Shared Package
ML19332G115	List: ML19332G114 ML19332G119
References
NUDOCS 8912200184
Download: ML19332G119 (15)

Text

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Attachment 1 to Document' Control Desk Letter December 11, 1989~ _

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Page 1 of 5 .

4m vm u, 04 C . TABLE 2.2-1

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• REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS

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@$ g -Total xa .-.

Functional Unit Allowance (TA) . Z_ S, Trip 5etpoint om [ 1. Manual Reactor Trip Allowable Value

, as Not Applicable NA NA NA NA

om 2. Power Range, Neutron Flux .

R High Setpoint 7.5 4.56' O $109% of RTP $111.2% of RTP N$ tow 5etpoint 8.3 4.56 0

" $25% of RTP 127.2% of RTP

3. Power. Range, Neutron Flux g

! 1. 6 0.5 0 High Positive Rate

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. <5% of RTP with <6.3% of RTP with i i tine constant i time constant

>2 seconds >2 seconds

4. Power Range, Neutron Flux I 1.6 0.5 0

{ , High Negative Rate <5% of RTP with 4 i time constant <6.3% of RIP with i time constant 32 seconds 32 seconds

5. Intermediate Range, 1 17.0 8.4 0 Neutron Flux <25% of RTP

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<31% of RIP

6. Source Range, Neutron Flux 17.0 10

_0 1105 cps $1.4 x 105 cps

7. Overtemperature AT

9. 8 1 See note 1 See note 2 1.2** .1

8. Overpower AT I 9G r, 5.2 4-9 i 9.

See note 3 See note 4 N Pressurizer Pressure-Low 3.1 0.71 i

' 1.5. 31870 psig 11859 psig i

ka 10. Pressurizer Pressure-High 3.1 0.71 1. 5 12380 psig 12391 psig.

f  % 11. Pressurizer Water Level-High 5. 0  !

! 2 L 1.5 $92% of instrument span $93.8% of instrument i P 12. /, W ' ,6 Loss of Flow J.'s ,4 % .

ar4> -.4-6 >90% of loop .\ .,

189-f% of loop i design flow

• design flow

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"* " Loop cWign flow = 96.500 gum RIP - RATED THERMAL POWER span for Delta-T (RTDs) and 1.2% for Pressurizer Pressure.

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Attachment 1 to Document Control Desk Letter .

December 11, 1989 .O'.

Page 2 of 5 -

5E TABLE 2.2-1 (Continued) '

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o REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS '

NOTATION (Continued)

) NOTE 1: (Continued) and f power range nuclear ton chambers; with gains to be selected based response during plant startup tests such that:

(i) for q t ~ 9b between - 24 percent and + 4 percent f i (al) = 0 where qt and qb are percent RATED THERMAL POWER in the top and bottom halves of the core respectively, +q 55 and q

( b total THERMAL POWER in percent of RATED THERMAL POWER.

(ii) .

ry for each percent that the magnitude of'qt -

gbexceeds -24 percent, the AT trip setpoint L

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shall be automatically reduced by 2.27 percent of its value at RATED THERMAL POWER.

(iii)

for each percent thatLthe. magnitude of qt -

qbexceeds +4 percent, the AT trip setpoint i

i shall be automatically reduced by 2.13 percent of its value at RATED THERMAL POWER.

NOTE 2:

j 2. channel's maximum trip setpoint shall not exceed its computed trip point by more than percent AT Span.

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' NOTE 3: OVERPOWER AT g

~,

AT i AT, [K. - Ks f3['f5 3 T - Ks [T - T")]

g Where: AT = ,

E as defined in Note 1 g AT, =

c.

as defined in Note 1 2 K4 1.0875

? 1 i Ks >

,s 0.02/*F temperature for increasing average temperature and 0 for decreasing average C'

3I'f5 3

=

The function generated by the rate-lag controller for' T,, dynamic compensation

- V 1  %? W at--* * ' i y ""

"r. _w  % .p

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. Attachment 1 to-Document Control. Desk Letter. .. .

December 11,.1989

• Page 3 of 5 .

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!ABLE 2.2-1 (Continued)

. REACIOR TRIP SYSTEM INSTRUMENTATION TRIP.'SETPOINTS' E

NOTATION (Continued)

[ HOTE 3 (continued)

=

T3 Time constant utilized in the rate-lag controller for 1

,13 1 10 secs.

Ks 1 i =

0.00156/*F for i > T" and Ks = 0 for. I _< I" as defined in Note 1 1" <

587.4*F Reference T ,g at RATED THEllMAL POWER S =

as defined in Note 1 NOTE 4:

.Feercent channel's maximum AT Span. trip setpoint shall not exceed its' computed trip an point by more th 4

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Attachment p-to Document Control Desk . Letter:

December ~ 1Y, 1989 ' , ~

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4 TA8LE 3.3-4 (Continued)

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ENGINEERED' SAFETY FEATURE ACTUATION SYSTEN INSTRUMENTATION c-

$ Total l y Functional Unit i

Allowance (TA) _Z Trip Setpoint

_S Allowable Value

{ " 4. STEAN LINE ISOLATION

a. -Manu I NA NA NA~ NA

b. Automatic Actuation Logic NA MA NA NA' NA and Actuation Relays NA c.

Reactor Building Pressure- 3.0 0.71 High 2 .

1.5 16.35 16.61

d. Steam Flow in Two Steamlines- 20.0 Nigh, Cofncident with 13.16 1.5/ < a function 1.5 < a function defined t' ~ defined as as follows: .A ap

• follows: A AP corresponding to 44%-

Y corresponding of full steam flow y to 40% of full between UK and 20%

steam flow load and then a ap

! between 0% and increasing linearly.

20% load and to a Ap corre-then a op sponding to 114.0%

increasing of full steam I 11 pearly to a flow at full load.

Ap correspond-l ing to 110% of

• l full steam flow l

. av! ,8 t_ f sSLg log ;S T ,,- Low-Low 4.0 l PB.1 .;

>593*F [fMh4*F i e. Steam 11ne Pressure - Low .

) 20.0 10.71 1.5 1675 psig i >635 psig (1) Time constants utilized in lead lag controller for steamline pressure low are as follows: -

ti >_ 50 secs. In < 5 secs.

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, , . . - _ _.._,1 - ,_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - - _ _ =____:______.________------._ -_=-.

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' Attachment I to Document Control. Desk Letter _- .' .

December 11,'1989 Pcge 5 of 5 a-

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I TABLE 3.3-4 (Continued)

E 3 , ENGINEERED SAFETY FEATURE ACTUATION SYSTEP INSTRUNENTATION TRIP SETPOINTS c Total

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• Functional Unit Allowance (TA) Z 5 Trip Setpoint Alloweble Value

" 9. ENGINEERED SAFETY FEATUR;i ACTUATION SYSTEM IFTERLOCKS ~

i INTERLOCNS

a. Presserizer Pressure, P-11 3.1 .71 1.5 1985 psig 11974 psig &

<I ' P8 . .

T! .8 SS a., _,

b. T,,g Low-Low. P-12

.rrs. 4 as 4.0 M 1z2 l 464*F 1550-4*F & <555.f F w c. Reactor Trip, P-4 NA NA k NA NA E Y

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ATTACHMENT 2 '

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F DESCRIPTION OF AMENDMENT REQUEST SAFETY EVALUATION .

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Attachment 2 to Document Control Desk letter December 11, 1989 Page 1 of 4 i I

i Description of amendment request

• i The following are the current values as required by Technical Specifications that are affected by this request:

Specification 2.2.1 " Reactor Trip System Instrumentation Setpoints" Table 2.2-1:

Item 7 - Overtemp AT: Z=7.29, S=1.9,forDelta-T(RTDs)

Item 8 - Overpower AT: Z=2.26, S=1.9 Item 12 - Loss of flow: Z=1.0, S=1.5, Allowable Valuea 89.2% of loop design flow. '

Note 2 - applies to Item 7 and states that "The channel's maximum trip setpoint shall not exceed its computed trip setpoint by more than 2.0% AT span."

Note 4 - applies to Item 8 and states that "The channel's maximum trip setpoint shall not exceed its computed trip setpoint by more than 2.0% AT span."

Specification 3.3.2"EngineeredSafetyFeatureActuationSystem(ESFAS)

Instrumentation" Table 3.3-4 Item 4d - Tavg - Low Low: 2 1.12 S=1.2, Trip setpointa 5530F, Allowable Valuea 550.6.

Item 9b - Tavg - Low Low, P-12: Z 1.12 S=1.2 Trip setpointa 5530F, Allowable Valuea 550.6 and s555.4. '

Allowable Values for the trip setpoints have been specified in Tables 2.2-1 and 3.3-4. These values accommodate the instrument drift assumed to occur between operational tests and-the accuracy to which setpoints can be measured and calibrated. Operation with setpoints less conservative than the Trip Setpoint but within the Allowable Value is acceptable since an allowance has

' been made in the safety analysis to accommodate this error. An optional provision exist for determining the OPERABILITY of a channel when its trip setpoint is found to exceed the Allowable Value. The methodology of this option utilizes the "as measured" deviation from the specified calibration point for rack and sensor components in conjunction with a statistical

! combination of the other uncertainties of the instrumentation to measure the process variable and the uncertainties in calibrating the instrumentation.

In Equation 2.2-1, Z + R + S s TA, the interactive effects of the errors in the rack and the sensor, and the "as measured" values of the errors are considered. "Z," as specified in Tables 2.2-1 and 3.3-4, in percent span, is the statistical summation of errors assumed in the analysis excluding those l

F Attachment 2 to Document Control Desk letter l December 11, 1989 Page 2 of 4 i

associated with the sensor and rack drift and the accuracy of their measurement. "TA," or Total Allowance, is the difference, in percent span, '

between the trip setpoint and the value used in the analysis for actuation.

"R." or Rack Error, is the "as measured" deviation, in percent span, for the affected channel from the specified trip setpoint. "S." or Sensor Error, is either the "as measured" deviation of the sensor from its calibration point or the value specified in Tables 2.2-1 and 3.3-4, in percent span, from the analysis assumptions.

Specification 2.2.1 Trip Setpoint Limits ensures that the reactor core and reactor coolant system are prevented from exceeding their safety limits during normal operation and design basis anticipated operational occurrences and to assi;t the ESFAS system in mitigating the consequences of accidents.

Specification 3.3.2 Trip Setpoint Limits are consistent with the assumptions used in the accirient analyses and act as components of the facility design to provide protection and mitigation of accident and transient conditions.

The methodology used to derive the trip setpoints is based upon combining all of the uncertainties in the channels. Inherent to the determination of the trip setpoints are the magnitudes of these channels uncertainties. Sensors and other instrumentation utilized in these channels are expected to be capable of operating within the allowances of these uncertainty magnitudes.

Rack drift in excess of the Allowable Value exhibits the behavior that the rack has not met its allowance.

This request supports a modification (MRF 21315) that would remove the existing RTD bypass manifolds and replace the temperature instrumentation with fast response RTD's located in thermowells installed in the reactor ,

coolant loop piping. Enclosure 1 is WCAP-12189 (proprietary class 2) and '

WCAP-12190-(non proprietary) which extensively evaluates the effects of this modification on V. C. Summer Nuclear Station. The values related to the new fast response RTDs were used to perform new uncertainties calculations for the control and protection functions affected by the modification. Tables 3.1-1 through 3.1-9 of Enclosure 1 notes the values used and the results obtained from these calculations. As a result of this analysis, certain values in the Technical Specifications are affected. "S" and "Z," as '

described above, for Overtemperature Delta-T, Overpower Delta-T and Low Flow and the trip setpoints and allowable values with respect to Steam Line Isolation Low Low Tavg and P12 Low Low Tavg limits need to be changed to reflect the properties of the fast response RTDs.

Attachment 2 to Document Control Desk Letter  ;

December 11, 1989 )

Page 3 of 4 The results of the evaluation, as r.cted in section 3.3 of Enclosure 1 necessitates changing Technical Specifications to the values indicated below:

Specification 2.2.1 " Reactor Trip System Instrumentation Setpoints" ,

Table 2.2-1 Item 7 - Overtemperature AT: Z=7.21 S=1.6 for Delta T (RTDs)

Item 8 - Overpower AT: Z=1.96, S=1.6 '

Item 12 - Loss of Flow: Z=1.48, S=0.6 Allowable value288.9% of loop design flow.

Note 2 - Applies to Item 7 and is changed to "The channels maximum trip setpoint shall not exceed its computed trip point by more than 2.2 percent AT span."

Note 4 - Applies to Item 8 and is changed to "The channels maximum trip setpoint shall not exceed its computed trip point by more than 2.4 percent AT span."

Specification 3.3.2"EngineeredSafetyFeatureActuationSystem(ESFAS)

Instrumentation" Table 3.3-4 ,

Item 4d - Tavg-Low Low: Z= 71, S=0.8, Trip setpoint 25520F, Allowable Valuem 548.4.

Item 9b - Tavg-Low Low. P-12: Z=.71, S=0.8 Trip setpointa 5520F, Allowable Valuem 548.4of and s555.60F, t Safety Evaluation:

! i The following is an evaluation to determine if safety has been compromised by changing the values, as indicated above, in Technical Specifications when evaluated against the FSAR Accident Analysis and the VANTAGE 5 Reload Transition Safety Report (RTSR). The sensor uncertainties of the fast response RTDs along with system uncertainties associated with increasing the number of hot leg RTDs from one to three were calculated and determined to be bounded by the allowed margin of the analyses as explained in section 4.2 Enclosure 1. Sections 4.3 and 4.4 of Enclosure 1 supports the conclusion that the changes in Technical Specifications, required as a result of the modification, remains consistent with both the FSAR Accident Analysis and the L VANTAGE 5 RTSR assumptions. The above results are derived via methodology l-.

consistent with Westinghouse Setpoint Methodology for Protection Systems, (WCAP-11170) which was previously submitted to the commission. The changes in plant hardware will adhere to applicable ASME and IEEE codes, Regulatory Guidelines and NRC General Design Criteria to ensure the new components continue to function in a consistent manner towards safety.

i Attachment 2 to Document Control Desk Letter December 11, 1989 )

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lt should be noted that Enclosure 1 addresses a change in the response time limit for overtemperature Delta-T and overpower Delta-T. This change was ,

applicable to the assumptions made in the Vantage 5 RTSR. These assumptions '

have been incorporated since Enclosure 1 was published; therefore, the response tihes have been approved by the NRC and ir,corporated in Technical Specifications.

Finally, the modification eliminates certain operating obstacles associated with the RTD bypass manifolds such as leakage associated with valves and flanges, potential valve failures and radiation exposure during Reactor Building maintenance. i In conclusion the method of utilizing fast-response RTDs installed in the .

reactor coolant loop piping as a means for RCS temperature indication has  :'

undergone extensive analyses, evaluation and testing as described in this report. The incorporation of this system into the V. C. Summer Nuclear Station design meets all safety, licensing and control requirements necessary '

for safe operation of this facility. The fast response RTDs installed in the reactor coolant loop piping adequately replace the present hot and cold leg temperature measurement system and enhances ALARA efforts as well as improve plant reliability.

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ATTACHMENT 3 DESCRIPTION OF AMENDMENT REQUEST l NO SIGNIFICANT HAZARDS DETERMINATION l

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l Attachment 3 to Document Control Desk Letter December 11, 1989 Page 1 of 4 Description of amendmentrequest:

The'following are the current values as required by Technical Specifications that are affected by this request:

Specification 2.2.1 Reactor Trip System Instrumentation Setpoints" Table 2.2-1:

Item 7 - Overtemp AT: Z=7.29 S=1.9,forDelta-T(RTDs)

Item 8 - Overpower AT: Z=2.26, S=1.9 Item 12 - Loss of flow: Z=1.0, S=1.5. Allowable Valuea 89.2% of loop design flow.

Note 2 - applies to Item 7 and states that "The channel's maximum trip setpoint shall not exceed its computed trip setpoint by more than 2.0% AT span."

Note 4 - applies to Item 8 and states that "The channel's maximum trip setpoint shall not exceed its computed trip setpoint by more than 2.0% AT span."

Specification 3.3.2"EngineeredSafetyFeatureActuationSystem(ESFAS)

Instrumentation" Table 3.3-4 Item 4d - Tavg - Low Low: Z=1.12 S=1.2 Trip setpointa 5530F ,

Allowable valuea 550.6.

Item 9b - Tavg - Low Low, P-12: Z=1.12 S=1.2, Trip setpointa 5530F -

Allowable Valuem 550.6 and s555.4.

Allowable Values for the trip setpoints have been specified in Tables 2.2-1 and 3.3-4. These values accommodate the instrument drift assumed to occur between operational tests and the accuracy to which setpoints can be measured and calibrated. Operation with setpoints less conservative than the Trip Setpoint but within the Allowable Value is acceptable since an allowance has been made in the safety analysis to accommodate this error. An optional provision exist for determining the OPERABILITY of a channel when its trip setpoint is found to exceed the Allowable Value. The methodology of this option utilizes the "as measured" deviation from the specified calibration point for rack and sensor components in conjunction with a statistical combination of the other uncertainties of the instrumentation to measure the process variable and the uncertainties in calibrating the instrumentation.

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. Attachment 3 to Document Control Desk letter o December 11, 1989

- Page 2 of 4 In Equation 2.2-1, Z + R + S s TA, the interactive effects of the errors in the rack and the sensor, and the "as measured" values of the errors are ,

considered. "Zt " as specified in Tables 2.2-1 and 3.3-4, in percent span, is the statistical summation of errors assumed in the analysis excluding those associated with the sensor and rack drift and the accuracy of their measurement. "TA " or Total Allowance, is the difference, in percent span, between the trip setpoint and the value used in the analysis for actuation.

"R," or Rack Error, is the "as measured" deviation, in percent span, for the affected channel from the specified trip setpoint. "S." or Sensor Error, is either the "as measured" deviation of the sensor from its calibration point or the value specified in Tables 2.2-1 and 3.3-4, in percent span, from the analysis assumptions.

Specification 2.2.1 Trip Setpoint Limits ensures that the reactor core and reactor coolant system are prevented from exceeding their safety limits during normal operation and design basis anticipated operational occurrences and to assist the ESFAS system in mitigating the consequences of accidents.

Specification 3.3.2 Trip Setpoint Limits are consistent with the assumptions used in the accident analyses and act as components of the facility design to provide protection and mitigation of accident and transient conditions.

The methodology used to derive the trip setpoints is based upon combining all of the uncertainties in the channels. Inherent to the determination of the trip setpoints are the magnitudes of these channels uncertainties. Sensors and other instrumentation utilized in these channels are expected to be capable of operating within the allowances of these uncertainty magnitudes. l Rack drift in excess of the Allowable Value exhibits the behavior that the i rack has not met its allowance.

This request supports a modification (MRF 21315) that would remove the existing RTD bypass manifolds and replace the temperature instrumentation with fast response RTD's located in thermowells installed in the reactor coolant loop piping. Enclosure 1 is WCAP-12189 (proprietary class 2) and WCAP-12190 (non proprietary) which extensively evaluates the effects of this modification on V. C. Summer Nuclear Station. The values related to the new fast response RTDs were used to perform new uncertainties calculations for the control and protection functions affected by the modification. Tables 3.1-1 through 3.1-9 of Enclosure 1 notes the values used and the results obtained from these calculations. As a result of this analysis, certain values in the Technical Specifications are affected. "S" and "Z " as described above, for Overtemperature Delta-T, Overpower Delta-T and Low Flow and the trip setpoints and allowable values with respect to Steam Line Isolation Low Low Tavg and P12 Low Low Tavg limits need to be changed to reflect the properties of the fast response RTDs.

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.- Attachment 3 to Document Control Desk Letter o December 11, 1989 ,

Page 3 of 4 The results of the evaluation, as noted in section 3.3 of Enclosure 1 necessitates changing Technical Specifications to the values indicated below: }

Specification 2.2.1 " Reactor Trip System Instrumentation Setpoints" {

Table 2.2-1 Item 7 - Overtemperature AT: Z=7.21, $=1.6 for Delta T (RTDs)

Item 8 - Overpower AT: Z-1.96, S=1.6 Item 12 - Loss of Flow: Z=1.48, $=0.6 Allowable value288.9% of loop design flow. >

t Note 2 - Applies to Item 7 and is changed to "The channels maximum trip setpoint shall not exceed its computed trip point by more than .

2.2 percent AT span." i Note 4 - Applies to Item 8 and is changed to "The channels maximum trip setpoint shall not exceed its computed trip point by more than 2.4 percent AT span."

Specification 3.3.2"EngineeredSafetyFeatureActuationSystem(ESFAS)

Instrumentation" Table 3.3-4 Item 4d - Tavg-Low Low: Z=.71 S=0.8 Trip setpoint 25520F, Allowable Valuea 548.4.

Item 9b - lavg-Low Low. P-12: Z=.71 S=0.8 Trip setpointa 5520F,  ;

Allowable Valuea 548.4of and s555.60F, Basis for proposed no significant hazards considera tion:

SCE&G has evaluated the proposed changes against the Significant Hazards Criteria of 10CFR50.92. The results of SCE&G's evaluation demonstrate that the changes do not involve any significant hazards as described below,

a. The probability or consequences of an accident previously evaluated is not significantly increased. The Loss-Of-Coolant-Accident (LOCA) and non-LOCA accident analyses were reviewed verifying that the variations in uncertainty associated with certain reactor trip functions, reflected in the Technical Specifications changes, do not invalidate the current Reload Transition Safety Report (RTSR) analyses of record. Therefore, the design basis conclusions are still met. Additionally, it was determined that sufficient allowance exists in the current RTSR assumptions such that the total temperature measurement uncertainty and protection system response time for the new RTDs do not impact the RTSR results. With respect to a LOCA, conservative nominal input values were assumed in the analysis and not plant specific input values; the*efore,

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. Attachment 3 to Document Control Desk Letter

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-' Page 4 of 4 i slight variation in uncertainties do not affect the RTSR results. These .

conclusions are based upon calculations supporting the revised Tables 2.2-1 and 3.3-4 provided in the attached WCAP-12189. The new values were obtained using methodology consistent with the Westinghouse Setpoint Methodology for Protection Systems WCAP-11770, which was ,

previously submitted to the Commission. ,

b. The possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis reports is not created.

At V. C. Summer Nuclear Station, the three ilot leg RTDs and one cold leg RTD will utilize the existing penetrations into the RCS piping from the bypass system with only slight modifications. Caps and welds sealing the crossover leg bypass return piping nozzle, as well as the modification and welding for the existing penetrations, will be qualified in accordance with the ASME code and are consistent with current plant designs. Consideration has been given to plant response in the remote possibility that a thermowell would be ejected from its boss. It has been concluded that the affect of this flow area is insignificant on the results of the large break LOCA analyses and bounded by the results for the small break LOCA analyses.

The function of the delta-T/Tavg protection channels is not changed becausa of the bypass elimination. The newly installed fast response RTDs perform the same function in both Thot and Tcold applications. The three Thot signals are electronically averaged, with the capability to manually add an electronic bias to a two-RTD average should one RTD fail. These measured temperature values will still serve as input to two-out-of-three voting logic for protection functions. Spare RTDs are installed and can be manually activated should the on-line RTD fail.

The basis for the instrumentation and control design meets the criteria of applicable IEEE standards, regulatory guides and general design criteria which satisfy electrical separation, seismic and environmental qualification and single failure criteria. Therefore, the possibility of a new or different kind of accident does not exist.

c. The margin of safety as defined in the basis of the Technical Specifications is not significantly reduced by the affect of the change of the response time and setpoint uncertainties. The investigation of the affect of these variables on non-LOCA and LOCA transients has verified that plant operation will be maintained within the bounds of safe, analyzed conditions as defined in the RTSR with the revised Technical Specifications. Conclusions presented in the RTSR remain valid. The specific analyses and supporting calculations supporting l these conclusions are provided in the attached WCAP-12189. As such, no reduction in the margin of safety between the RTSR acceptance limit and the ultimate safety limit (such as departure from nucleate boiling i ratio) has taken place for operation with the new RTD system.

Therefore, based on the above considerations it has been determined that no significant hazard considerations exist.

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