NL-09-1431, License Amendment Request for Technical Specification Table 3.3.1-1

License Amendment Request for Technical Specification Table 3.3.1-1

ML100340033
Person / Time
Site:	Vogtle
Issue date:	02/02/2010
From:	Ajluni M Southern Nuclear Operating Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NL-09-1431
Download: ML100340033 (18)

Text

Southern Nuclear Operating Compa!lV, Inc SOUTHERNA February 2, 2010 COMPANY Docket Nos.: 50-424 NL-09-1431 50-425 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Vogtle Electric Generating Plant License Amendment Request for Technical Specification Table 3.3.1-1 Ladies and Gentlemen:

In accordance with the provisions of 10 CFR 50.90, Southern Nuclear Operating Company (SNC) is submitting a request for an amendment to the Technical Specifications (TS) for Vogtle Electric Generating Plant (VEGP) Units 1 and 2.

The proposed change will add Surveillance Requirement (SR) 3.3.1.15 to VEGP TS Table 3.3.1-1, "Reactor Trip System Instrumentation," Function 3, "Power Range Neutron Flux High Positive Rate." SR 3.3.1.15 requires verification that the RTS RESPONSE TIME is within limits every 18 months on a STAGGERED TEST BASIS. Function 3 is the Power Range Neutron Flux High Positive Rate Trip (PFRT) function. This change is being proposed based on a reanalysis of the Rod Cluster Control Assembly Bank Withdrawal at Power event. provides a description and justification for the proposed change. provides the marked-up TS pages and TS Bases pages for the proposed change. Enclosure 3 provides the clean typed TS pages and TS Bases pages.

SNC requests approval of the proposed license amendments by October 31, 2010. The proposed change will be implemented 90 days from the date of issuance.

In accordance with 10 CFR 50.91, a copy of this amendment request is being provided to the designated Georgia State official.

U.S. Nuclear Regulatory Commission NL-09-1431 Page 2 This letter contains no NRC commitments.

If you have any questions, please advise.

Mr. M. J. Ajluni states he is the Manager - Nuclear Licensing of Southern Nuclear Operating Company, is authorized to execute this oath on behalf of Southern Nuclear Operating Company and to the best of his knowledge and belief, the facts set forth in this letter are true.

Respectfully submitted, SOUTHERN NUCLEAR OPERATING COMPANY M. J. Ajluni Manager - Nuclear Licensing Sworn to and subscribed before me this ~..J day of~~

_, 20.10.

C~~1r'it/~

My commission expires: '","k/~Wt<D /~ 0 / 3 MJAlDWM/lac

Enclosures:

1.

Basis for Proposed Change

2.

Technical Specifications and Bases Markup Pages

3.

Technical Specifications and Bases Clean Typed Pages cc: Southern Nuclear Operating Company Mr. J. T. Gasser, Executive Vice President Mr. T. E. Tynan, Vice President - Vogtle Ms. P. M. Marino, Vice President - Engineering RType: CVC7000 U. S. Nuclear Regulatory Commission Mr. L. A. Reyes, Regional Administrator Ms. D. N. Wright, NRR Project Manager - Vogtle Mr. M. Cain, Senior Resident Inspector - Vogtle Georgia Department of Public Health Mr. N. Holcomb, Commissioner - Department of Natural Resources

Vogtle Electric Generating Plant License Amendment Request for Technical Specification Table 3.3.1-1 Basis for Proposed Change

Vogtle Electric Generating Plant License Amendment Request for Technical Specification Table 3.3.1-1 Basis for Proposed Change Table of Contents 1.0 Summary Description 2.0 Detailed Description 3.0 Technical Evaluation 4.0 Regulatory Evaluation 4.1 Significant Hazards Consideration 4.2 Applicable Regulatory Requirements I Criteria 4.3 Precedent 4.4 Conclusions 5.0 Environmental Consideration 6.0 References Basis for Proposed Change 1.0 Summary Description In accordance with the provisions of 10 CFR 50.90, Southern Nuclear Operating Company (SNC) is proposing a change to the Vogtle Electric Generating Plant (VEGP) Unit 1 and Unit 2 Technical Specifications (TS). This proposed change would revise TS 3.3.1, "Reactor Trip System (RTS) Instrumentation" for VEGP Units 1 and 2 to add Surveillance Requirement (SR) 3.3.1.15 to Function 3, "Power Range Neutron Flux High Positive Rate" of TS Table 3.3.1-1, "Reactor Trip System Instrumentation."

2.0 Detailed Description TS 3.3.1 Table 3.3.1-1 Proposed Change The proposed change would revise TS 3.3.1 to add SR 3.3.1.15 to Function 3 of TS Table 3.3.1-1. SR 3.3.1.15 requires that RTS RESPONSE TIMES be verified to be within limits every 18 months on a STAGGERED TEST BASIS. Function 3 is the Power Range Neutron Flux High Positive Rate (PFRT) reactor trip function.

This change is being proposed based on a reanalysis of the Rod Control Cluster Control Assembly Bank Withdrawal at Power event.

3.0 Technical Evaluation SR 3.3.1.15 requires a verification that RTS RESPONSE TIMES are within limits every 18 months on a STAGGERED TEST BASIS, as defined in the VEGP TS.

As stated in the TS Bases, SR 3.3.1.15 verifies that the individual channel/train actuation response times are less than or equal to the maximum values assumed in the accident analysis. Response time testing acceptance criteria are included in FSAR, Chapter 16 (Reference 1). Individual component response times are not modeled in the analyses. The analyses model the overall or total elapsed time, from the point at which the parameter exceeds the trip setpoint value at the sensor to the point at which the equipment reaches the required functional state (Le., control and shutdown rods fully inserted in the reactor core).

VEGP FSAR Table 16.3-1 lists the associated response time as N.A. for item 3 (Power Range Neutron Flux High Positive Rate). As a result, the response time for the PFRT is not being verified.

RWAP RCS Overpressure Analyses for VEGP Nuclear Safety Advisory Letter (NSAL) 09-1 dated February 4,2009 (Reference 2), issued by Westinghouse discussed the potential for Reactor Coolant System (RCS) overpressurization as a result of a control rod bank withdrawal during power operation (RWAP) and the overall results of analyses crediting PFRT for this event. Westinghouse determined that the methodology used for the generic and plant-specific RWAP RCS overpressure analyses incorrectly assumed that a minimum initial power level creates the most limiting condition and did not fully address the full range of power operation. Previous analyses have assumed an initial power level of 10 percent of rated thermal power (RTP), minus calorimetric uncertainty. Further investigation has identified cases from higher initial power E1 - 2 Basis for Proposed Change levels that are more limiting. With the generic analysis key parameters and methodology, some cases exceeded the RCS overpressure limit for initial power levels in the range of 60 percent to 80 percent rated thermal power (RTP).

Westinghouse completed specific analyses for VEGP Units 1 and 2 which addressed the potential for RCS overpressure following a RWAP. The results for VEGP, using a conservative methodology, demonstrate that the RCS overpressure limit listed in VEGP TS 2.1.2 (i.e., 2735 psig) is not violated, assuming that credit is taken for a PFRT at or below 9 percent of RTP with a lag time constant of 2 seconds and a trip delay time of 0.65 second.

These analyses were based on the key parameters for the Vogtle units listed on Table A. The Vogtle analyses for a PFRT at 9 percent of RTP with 0.65 second trip time delay involved 680 individual LOFTRAN computer runs at various initial conditions and reactivity insertion rates. All cases modeled the high pressurizer pressure reactor trip, the high neutron 'flux reactor trip and the positive flux rate trip (PFRT) functions. Credit for all three trip functions is required in order to meet the RCS pressure acceptance criterion. Without credit for the PFRT, the maximum calculated RCS pressure would not have met the analysis acceptance criterion in some of the 680 cases analyzed.

These analyses calculated an acceptable result (i.e., RCS peak pressure not in excess of 2750 psia) with a Positive Flux Rate Trip (PFRT) at 9% of RTP, a lag time constant of 2.0 seconds, and a trip delay time of 0.65 second.

Table A: Key Reactor Trip Parameters Used for the VEGP Units 1 and 2 RWAP RCS Pressure Analysis 120 High Flux Trip SAL (%)

I 0.5 High Flux Trip Delav (sec) 2440 High Przr Pressure Trip Setpoint (Qsia}

2 High Przr Pressure Trip_[)~lay (sec;)

'9 Positive Flux Rate Trip SAL (%)

2 Positive Flux Rate Trip Time Constant (sec) 0.65 Positive Flux Rate Trip Delay (sec)

E1 - 3 Basis for Proposed Change Power Range Neutron Flux High Positive Rate Response Time Verification The PFRT response time of 0.65 second, is explicitly credited in the RWAP analysis and must be met. The PFRT response time is verified as per WCAP 14036-P-A, Revision 1 "Elimination of Periodic Protection Channel Response Time Tests" (Reference 3). This approach is based on having performed an initial baseline RTS response time test. Following the initial RTS response time test on this reactor trip function, periodic response time tests are not required consistent with WCAP-14036-P-A, Revision 1. since other required tests ensure that the response time continues to be met.

Since the PFRT response time is essentially the same as the Power Range Neutron Flux - High (PRNF - High) reactor trip function, the PFRT response time can be implicitly verified during the response time test for the PRNF-High reactor trip function as discussed below.

The PFRT circuitry is part of the Nuclear Instrumentation System (NIS). The circuit consists of the difference between the Power Range Nuclear Flux (PRNF) signal and that same signal with a first order lag. That is.

PFRT trip signal =Flux - FluX/(1 +rs),

where T is the time constant and s is the Laplacian operator.

The nominal trip setpoint (difference) for the PFRT reactor trip function contained in VEGP TS 3.3.1 is 5 percent of RTP with a time constant (r) ~ 2 seconds. Both the PRNF-High and PFRT reactor trip signals are sent from the NIS to the protection system via the same components and are processed from the same PRNF signals and identical bistables. There is no significant time delay for the PFRT added by the additional signal processing to form the difference between the flux signal and the lagged flux signal. Therefore, other than having different nominal trip setpoints, the PFRT trip signal sent to the protection system has a response time similar to the PRNF-High reactor trip function.

SNC has reviewed preoperational test data and confirmed that during preoperational testing the response time of the NIS and Solid State Protection System (SSPS) for the PFRT function was measured and recorded. The slowest response time measured for a channel was 0.100 second. Per VEGP FSAR Table 16.3-3A, the allocation time for processing is 0.220 second. Therefore, SNC meets the allocation times delineated in the VEGP FSAR. Per the Westinghouse analysis, the total function response time limit is < 0.65 second.

SNC has confirmed that the PRNF overall function response time is less than

<0.65 second. This confirmation is based upon the response time measured during preoperational testing of the NIS and SSPS processing time, and the most recent surveillance data for the other components that comprise the string.

Adding SR 3.3.1.15 to TS 3.3.1 Table 3.3.1-1 Function 3 would require verification of RTS response times for the PFRT.

E1 - 4 Basis for Proposed Change 4.0 Regulatory Evaluation 4.1 Significant Hazards Consideration Southern Nuclear Operating Company (SNC) has evaluated whether or not a significant hazards consideration is involved with the proposed changes by focusing on three standards set forth in 10 CFR 50.92(c) as discussed below:

1.

Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed change to Vogtle Electric Generating Plant (VEGP)

Technical Specification (TS) 3.3.1, "Reactor Trip System (RTS)

Instrumentation," Table 3.3.1-1, "Reactor Trip System Instrumentation" does not significantly increase the probability or consequences of an accident previously evaluated in the Update Final Safety Analysis Report (UFSAR). The overall protection system performance will remain within the bounds of the accident analysis since there are no hardware changes. The design of the Reactor Trip System (RTS) instrumentation, specifically the positive range neutron flux high positive rate trip (PFRT) function, will be unaffected. The reactor protection system will continue to function in a manner consistent with the plant design basis. All design, material, and construction standards that were applicable prior to the request are maintained.

The proposed change adds an additional surveillance requirement to assure that the PFRT is verified to be consistent with the safety analysis and licensing basis. In this specific case, a response time verification requirement will be added to the PFRT function.

The proposed changes will not modify any system interface. The proposed changes will not affect the probability of any event initiators. There will be no degradation in the performance of or an increase in the number of challenges imposed on safety-related equipment assumed to function during an accident situation.

There will be no change to normal plant operating parameters or accident mitigation performance. The proposed change will not alter any assumptions nor change any mitigation actions in the radiological consequences evaluations in the UFSAR.

The proposed change does not adversely affect accident initiators or precursors nor alter the design assumptions, conditions, or configuration of the facility or the manner in which the plant is operated and maintained. The proposed changes do not alter nor E1 - 5 Basis for Proposed Change prevent the ability of SSCs from performing their intended function to mitigate the consequences of an initiating event within the assumed acceptance limits. The proposed change is consistent with the safety analyses assumptions and resultant consequences.

The RCS overpressure limit listed in Specification 2.1.2 of the VEGP Technical Specifications (Le., 2735 psi g) is not violated.

2.

Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

There are no hardware changes nor are there any changes in the method by which any safety related plant system performs its safety function. This change will not affect the normal method of plant operation nor change any operating parameters. No performance requirements will be affected; however, the proposed change adds an additional surveillance requirement. The additional surveillance requirement is consistent with assumptions made in the safety analyses and licensing basis.

No new accident scenarios, transient precursors, failure mechanisms, or limiting single failures are introduced as a result of this change. There will be no adverse effect or challenges imposed on any safety-related system as a result of this change.

Therefore, the proposed change does not create the possibility of a new or different accident from any accident previously evaluated.

3.

Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No.

The proposed change does not affect the acceptance criteria for any analyzed event nor is there a change to any Safety Limits.

There will be no effect on the manner in which Safety Limits or Limiting Conditions of Operations are determined, nor will there be any effect on those plant systems necessary to assure the accomplishment of protection functions.

This change is consistent with the assumptions made in the safety analyses. The addition of an additional surveillance requirement increases the margin of safety by assuring that the associated safety analysis assumption on the PFRT response time is verified.

Therefore, the proposed change does not involve a significant reduction in the margin of safety.

E1 - 6 Basis for Proposed Change Based on the above, SNC concludes that the proposed amendment does not involve a significant hazards consideration under the standard set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

4.2 Applicable Regulatory Requirements I Criteria The following lists the regulatory requirements and plant - specific design bases related to the proposed changes.

• GOC-13 requires that instrumentation shall be provided to monitor variables and systems over their anticipated ranges for normal operation, for anticipated operational occurrences, and for accident conditions as appropriate to assure adequate safety, including those variables and systems that can affect the fission process, the integrity of the reactor core, the reactor coolant pressure boundary, and the containment and its associated systems.

• GOC-20 requires that the protection system(s) shall be designed (1) to initiate automatically the operation of appropriate systems including the reactivity control systems, to assure that specified acceptable fuel design limits are not exceeded as a result of anticipated operational occurrences and (2) to sense accident conditions and to initiate the operation of systems and components important to safety.

• GOC-21 requires that the protection system(s) shall be designed for high functional reliability and testability.

• GOC-22 through GOC-25 and GOC-29 require various design attributes for the protection system(s), including independence, safe failure modes, separation from control systems, requirements for reactivity control malfunctions, and protection against anticipated operational occurrences.

Regulatory Guide 1.22 describes an acceptable method for ensuring that the protection system is designed to permit periodic testing of its functioning during reactor operation.

10 CFR 50.55a paragraph (h), "Protection systems," states, in part, that "protective systems must meet the requirements stated in either IEEE Std. 279, "Criteria for Protection Systems for Nuclear Power Generating Stations," or in I EEE Std. 603-1991, "Criteria for Safety Systems for Nuclear Power Generating Stations."

10 CFR 50.36 paragraph (c)(1)(ii)(A), "Safety limits. limiting safety system settings, and limiting control settings" requires limiting E1 -7 Basis for Proposed Change safety system settings to be included in the Technical Specifications and to be "so chosen that automatic protective action will correct the abnormal situation before a safety limit is exceeded."

10 CFR 50.36 paragraph (c)(3), "Surveillance requirements,"

states "Surveillance requirements are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met.

4.3 Precedent None.

4.4 Conclusions In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

5.0 Environmental Consideration 10 CFR 51.22(c)(9) provides criteria for identification of licensing and regulatory actions eligible for categorical exclusion from performing an environmental assessment. A proposed amendment to an operating license for a facility requires no environmental assessment if operation of the facility in accordance with the proposed license amendment will not:

1.

Involve a significant hazards consideration,

2.

Result in a significant change in the types, or a significant increase in the amounts, of any effluents that may be released off site, or

3.

Result in a significant increase in individual or cumulative occupational radiation exposure.

SNC has evaluated the proposed changes and determines the changes do not involve (1) a significant hazards consideration, (2) a significant change in the types or significant increase in the amounts of any effluents that may be released off-site, or (3) a significant increase in the individual or cumulative occupational radiation exposure. Accordingly, the proposed change meets the eligibility criteria for categorical exclusion set forth in 10 E1 - 8 Basis for Proposed Change CFR 51.22(c)(9), and an environmental assessment of the proposed change is not required.

6.0 References

1. VEGP Updated Final Safety Analysis Report, Revision 16.

2. Westinghouse Nuclear Safety Advisory Letter (NSAL) 09-1, "Rod Withdrawal at Power Analysis for Reactor Control System Overpressure." February 4, 2009.

3. WCAP-14036-P-A, Revision 1 "Elimination of Periodic Protection Channel Response Time Tests."

E1 - 9

Vogtle Electric Generating Plant License Amendment Request for Technical Specification Table 3.3.1-1 Technical Specification and Bases Markup Pages

RTS Instrumentation 3.3.1 Table 3.3.1-1 (page 1 of 9)

Reactor Trip System Instrumentation APPLICABLE FUNCTION MODES OR OTHER SPECIFIED CONDITIONS REQUIRED CHANNELS CONDITIONS SURVEILLANCE REQUIREMENTS ALLOWABLE VALUE NOMINAL TRIP SETPOINT(n)

1.

Manual Reactor Trip 1,2 2

2 B

C SR 3.3.1.13 SR3.3.1.13 NA NA NA NA

2.

Power Range Neutron Flux

a.

High 1,2 4

o SR 3.3.1.1

<; 111.3% RTP 109% RTP SR 3.3.1.2 SR3.3.1.7 SR 3.3.1.11 SR 3.3.1.15

b.

Low 4

E SR 3.3.1.1

<;27.3% RTP 25% RTP SR 3.3.1.8 SR 3.3.1.11 SR 3.3.1.15

3.

Power Range 1,2 4

E SR 3.3.1.7

<; 6.3% RTP 5%RTP Neutron Flux High SR 3.3.1.11 with time with time Positive Rate constant constant SR3.3.1.15

2 sec

?: 2 sec r-------

4.

Intermediate 1 (b), 2(c) 2 F,G SR 3.3.1.1 41.9% RTP 25% RTP Range Neutron SR 3.3.1.8 Flux SR 3.3.1.11 2

H SR 3.3.1.1

<;41.9%RTP 25% RTP 2(d)

SR 3.3.1.8 SR 3.3.1.11 (a)

With Reactor Trip Breakers (RTBs) closed and Rod Control System capable of rod withdrawaL (b)

Below the P-10 (Power Range Neutron Flux) interlocks.

(c)

Above the P-6 (Intermediate Range Neutron Flux) interlocks.

(d)

Below the P-6 (Intermediate Range Neutron Flux) interlocks.

(n)

A channel is OPERABLE with an actual Trip Setpoint value outside its calibration tolerance band provided the Trip Setpoint value is conservative with respect to its associated Allowable Value and the channel is readjusted to within the established calibration tolerance band of the Nominal Trip Setpoint. A Trip Setpoint may be set more conservative than the Nominal Trip Setpoint as necessary in response to plant conditions.

Vogtle Units 1 and 2 3.3.1-14 Amendment No. ~ (Unit 1)

Amendment No..:J.OO (Unit 2)

BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY

3.

reactivity excursions...........-"

such as an inadvertent control rod withdrawal or RTS Instrumentation B 3.3.1

b.

Power Range Neutron Flux -

Low (continued) setpoint). This Function is automatically unblocked when three out of four power range channels are below the P-10 setpoint. Above the P-1 0 setpoint, positive reactivity additions are mitigated by the Power Range Neutron Flux -

High trip Function.

In MODE 3, 4, 5, or 6, the Power Range Neutron Flux Low trip Function does not have to be OPERABLE because the reactor is shut down and the NIS power range detectors cannot detect neutron levels in this range. Other RTS trip Functions and administrative controls provide protection against positive reactivity additions or power excursions in MODE 3, 4, 5, or 6.

Power Range Neutron Flux -

High Positive Rate The Power Range Neutron Flux -

High Positive Rate trip uses the same channels as discussed for Function 2 above.

e Power Range Neutron Flux -

High Positive Rate trip Fu ion ensures that protection is provided against rapid increa s in neutron flux that are characteristic of an RCCA drive rod using rupture and the accompanying ejection of the RCCA. Thi unction compliments the Power Range Neutron Flux -

High a Low Setpoint trip Functions to ensure that the criteria are met fo a rod ejection from the power range.

The LCO requires all four of the Power Range Neutron Flux High Positive Rate channels to be OPERABLE.

In MODE 1 or 2, when there is a potential to add a large amount of positive reactivity from a rod ejection accident (REA), the Power Range Neutron Flux High Positive Rate trip must be OPERABLE. In MODE 3,4, 5, or 6, the Power Range Neutron Flux -

High Positive Rate trip Function does not have to be Vogtle Units 1 and 2 B 3.3.1-13 IRevision No. 01

Vogtle Electric Generating Plant License Amendment Request for Technical Specification Table 3.3.1-1 Technical Specifications and Bases Clean Typed Pages

3.3.1 RTS Instrumentation Table 3.3.1*1 (page 1 of 9)

Reactor Trip System Instrumentation APPLICABLE MODES OR NOMINAL OTHER TRIP SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE SETPOINT(n)

FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS VALUE

1.

Manual Reactor 1,2 2

B SR 33.1.13 NA NA Trip 2

C SR 3.3.1.13 NA NA

2.

Power Range Neutron Flux

a.

High 1.2 4

D SR 3.3.1.1 s 111.3% RTP 109% RTP SR 3.3.1.2 SR3.3.1.7 SR 3.3.1.11 SR 3.3.1.15

b.

Low 4

E SR 3.3.1.1 s27.3% RTP 25% RTP SR 3.3.1.8 SR 3.3.1.11 SR 3.3.1.15

3.

Power Range 1,2 4

E SR 3.3.1.7 s 6.3% RTP 5% RTP Neutron Flux SR 3.3.1.11 with time with time High Positive SR 3.3.1.15 constant constant Rate

c. 2 sec 2: 2 sec

4.

Intermediate 1(b),2(c) 2 F,G SR 3.3.1.1 s41.9% RTP 25% RTP Range Neutron SR 3.3.1.8 Flux SR 3.3.1.11 2

H SR3.3.1.1 s41.9% RTP 25% RTP 2(d)

SR 3.3.1.8 SR 3.3.1.11 (continued)

(a)

With Reactor Trip Breakers (RTBs) closed and Rod Control System capable of rod withdrawal.

(b)

Below the P-10 (Power Range Neutron Flux) interlocks.

(c)

Above the P-6 (Intermediate Range Neutron Flux) interlocks.

(d)

Below the P-6 (Intermediate Range Neutron Flux) interlocks.

(n)

A channel is OPERABLE with an actual Trip Setpoint value outside its calibration tolerance band provided the Trip Setpoint value is conservative with respect to its associated Allowable Value and the channel is readjusted to within the established calibration tolerance band of the Nominal Trip Setpoint. A Trip Setpolnt may be set more conservative than the Nominal Trip Selpoint as necessary in response to plant conditions.

Vogtle Units 1 and 2 3.3,1-14 Amendment No.

(Unit 1)

Amendment No.

(Unit 2)

BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY RTS Instrumentation B 3.3.1

b.

Power Range Neutron Flux -

Low (continued) setpoint). This Function is automatically unblocked when three out of four power range channels are below the P-10 setpoint. Above the P-10 setpoint, positive reactivity additions are mitigated by the Power Range Neutron Flux -

High trip Function.

In MODE 3, 4, 5, or 6, the Power Range Neutron Flux Low trip Function does not have to be OPERABLE because the reactor is shut down and the NIS power range detectors cannot detect neutron levels in this range. Other RTS trip Functions and administrative controls provide protection against positive reactivity additions or power excursions in MODE 3, 4, 5, or 6.

3.

Power Range Neutron Flux -

High Positive Rate The Power Range Neutron Flux -

High Positive Rate trip uses the same channels as discussed for Function 2 above.

The Power Range Neutron Flux High Positive Rate trip Function ensures that protection is provided against rapid increases in neutron flux that are characteristic of an RCCA drive rod housing rupture and the accompanying ejection of the RCCA. This Function compliments the Power Range Neutron Flux -

High and Low Setpoint trip Functions to ensure that the criteria are met for reactivity excursions such as an inadvertent control rod withdrawal or a rod ejection from the power range.

The LCO requires all four of the Power Range Neutron Flux High Positive Rate channels to be OPERABLE.

In MODE 1 or 2, when there is a potential to add a large amount of positive reactivity from a rod ejection accident (REA), the Power Range Neutron Flux High Positive Rate trip must be OPERABLE. In MODE 3, 4, 5, or 6, the Power Range Neutron Flux -

High Positive Rate trip Function does not have to be Vogtle Units 1 and 2 B3.3.1-13