ND-17-1559, Request for License Amendment: Technical Specification Changes to Support Control Rod Testing in Cold Shutdown with Reactor Coolant Pumps Not in Operation (LAR-17-034)

Request for License Amendment: Technical Specification Changes to Support Control Rod Testing in Cold Shutdown with Reactor Coolant Pumps Not in Operation (LAR-17-034)

ML17265A822
Person / Time
Site:	Vogtle
Issue date:	09/22/2017
From:	Aughtman A Southern Nuclear Operating Co
To:	Document Control Desk, Office of New Reactors
References
LAR-17-034, ND-17-1559
Download: ML17265A822 (29)

Text

Southern Nuclear Operating Company, Inc.

42 Inverness Center Parkway Birmingham, AL 35242 September 22, 2017 Docket Nos.: 52-025 ND-17-1559 52-026 10 CFR 50.90 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Southern Nuclear Operating Company Vogtle Electric Generating Plant Units 3 and 4 Request for License Amendment:

Technical Specification Changes to Support Control Rod Testing in Cold Shutdown With Reactor Coolant Pumps Not in Operation (LAR-17-034)

Ladies and Gentlemen:

Pursuant to 10 CFR 52.98(c) and in accordance with 10 CFR 50.90, Southern Nuclear Operating Company (SNC) requests an amendment to the combined licenses (COLs) for Vogtle Electric Generating Plant (VEGP) Units 3 and 4 (License Numbers NPF-91 and NPF-92, respectively). The requested amendment proposes changes to COL Appendix A, Technical Specifications (TS).

The requested amendment proposes changes to add new TS 3.1.10, Rod Withdrawal Test Exception - MODE 5, and modify TS Limiting Condition for Operation (LCO) 3.0.7, to allow rod movement and rod drop time testing under cold conditions (MODE 5). Additionally, the LCO TS 3.4.8, Minimum Reactor Coolant System (RCS) Flow, Applicability is revised to reflect its safety analysis basis. provides the description, technical evaluation, regulatory evaluation (including the Significant Hazards Consideration Determination) and environmental considerations for the proposed changes. identifies the requested changes and provides markups depicting the requested changes to the VEGP Units 3 and 4 licensing basis documents. provides conforming TS Bases changes for information only.

This letter contains no regulatory commitments. This letter has been reviewed and determined not to contain security related information.

SNC requests NRC staff approval of the license amendment by March 30, 2018. Approval by this date will allow sufficient time to implement licensing basis changes necessary to support procedure development in relation to conducting the necessary operator training to support plant operations. SNC expects to implement this proposed amendment within 30 days of approval of the requested changes.

U.S. Nuclear Regulatory Commission ND-17-1559 Page 2 of 4 In accordance with 10 CFR 50.91, SNC is notifying the State of Georgia of this license amendment request by transmitting a copy of this letter and its enclosures to the designated State Official.

Should you have any questions, please contact Mr. Wesley Sparkman at (205) 992-5061.

I declare under penalty of perjury that the foregoing is true and correct. Executed on the 22"d of September 2017.

Respectfully submitted, Amy G. Aughtman Nuclear Development Licensing Director Southern Nuclear Operating Company Enclosures 1) Vogtle Electric Generating Plant (VEGP) Units 3 and 4 - Request for License Amendment Regarding Technical Specification Changes to Support Control Rod Testing in Cold Shutdown With Reactor Coolant Pumps Not in Operation (LAR-17-034)

2) Vogtle Electric Generating Plant (VEGP) Units 3 and 4 - Proposed Changes to Licensing Basis Documents (LAR-17-034)

3) Vogtle Electric Generating Plant (VEGP) Units 3 and 4- Conforming Technical Specification Bases Changes (For Information Only) (LAR-17-034)

U.S. Nuclear Regulatory Commission ND-17-1559 Page 3 of 4 cc:

Southern Nuclear Operating Company / Georgia Power Company Mr. S. E. Kuczynski (w/o enclosures)

Mr. M. D. Rauckhorst Mr. D. G. Bost (w/o enclosures)

Mr. M. D. Meier (w/o enclosures)

Mr. D. H. Jones (w/o enclosures)

Mr. D. L. McKinney (w/o enclosures)

Mr. T. W. Yelverton (w/o enclosures)

Mr. B. H. Whitley Mr. J. J. Hutto Mr. C. R. Pierce Ms. A. G. Aughtman Mr. D. L. Fulton Mr. M. J. Yox Mr. E. W. Rasmussen Mr. J. Tupik Mr. W. A. Sparkman Ms. A. C. Chamberlain Mr. M. K. Washington Ms. A. L. Pugh Mr. J. D. Williams Document Services RTYPE: VND.LI.L00 File AR.01.02.06 Nuclear Regulatory Commission Mr. W. Jones (w/o enclosures)

Ms. J. Dixon-Herrity Mr. C. Patel Ms. J. M. Heisserer Mr. B. Kemker Mr. G. Khouri Ms. S. Temple Ms. V. Ordaz Mr. T.E. Chandler Ms. P. Braxton Mr. T. Brimfield Mr. C. J. Even Mr. A. Lerch State of Georgia Mr. R. Dunn

U.S. Nuclear Regulatory Commission ND-17-1559 Page 4 of 4 Oglethorpe Power Corporation Mr. M. W. Price Mr. K. T. Haynes Ms. A. Whaley Municipal Electric Authority of Georgia Mr. J. E. Fuller Mr. S. M. Jackson Dalton Utilities Mr. T. Bundros Westinghouse Electric Company, LLC Mr. R. Easterling (w/o enclosures)

Mr. G. Koucheravy (w/o enclosures)

Mr. P. A. Russ Mr. M. L. Clyde Ms. L. Iler Mr. D. Hawkins Mr. J. Coward Other Mr. S. W. Kline, Bechtel Power Corporation Ms. L. A. Matis, Tetra Tech NUS, Inc.

Dr. W. R. Jacobs, Jr., Ph.D., GDS Associates, Inc.

Mr. S. Roetger, Georgia Public Service Commission Ms. S. W. Kernizan, Georgia Public Service Commission Mr. K. C. Greene, Troutman Sanders Mr. S. Blanton, Balch Bingham Mr. R. Grumbir, APOG NDDocumentinBox@duke-energy.com, Duke Energy Mr. S. Franzone, Florida Power & Light

Southern Nuclear Operating Company ND-17-1559 Enclosure 1 Vogtle Electric Generating Plant (VEGP) Units 3 and 4 Request for License Amendment Regarding Technical Specification Changes to Support Control Rod Testing in Cold Shutdown With Reactor Coolant Pumps Not in Operation (LAR-17-034)

(This Enclosure consists of 16 pages, including this cover page)

ND-17-1559 Request for License Amendment Regarding Technical Specification Changes to Support Control Rod Testing in Cold Shutdown With Reactor Coolant Pumps Not in Operation (LAR-17-034)

Table of Contents

1.

SUMMARY

DESCRIPTION

2. DETAILED DESCRIPTION

3. TECHNICAL EVALUATION

4. REGULATORY EVALUATION 4.1. Applicable Regulatory Requirements/Criteria 4.2. Precedent 4.3. Significant Hazards Consideration Determination 4.4. Conclusions

5. ENVIRONMENTAL CONSIDERATIONS

6. REFERENCES Page 2 of 16

ND-17-1559 Request for License Amendment Regarding Technical Specification Changes to Support Control Rod Testing in Cold Shutdown With Reactor Coolant Pumps Not in Operation (LAR-17-034)

Pursuant to 10 CFR 52.98(c) and in accordance with 10 CFR 50.90, Southern Nuclear Operating Company (SNC) hereby requests an amendment to Combined License (COL) Nos.

NPF-91 and NPF-92 for Vogtle Electric Generating Plant (VEGP) Units 3 and 4, respectively.

1.

SUMMARY

DESCRIPTION The requested amendment proposes changes to COL Appendix A, Technical Specifications (TS). The proposed changes add new TS 3.1.10, Rod Withdrawal Test Exception - MODE 5, and modify TS Limiting Condition for Operation (LCO) 3.0.7, to allow rod movement and rod drop time testing under cold conditions (MODE 5). Additionally, the LCO Applicability of TS 3.4.8, Minimum Reactor Coolant System (RCS) Flow, is revised to reflect its safety analysis basis.

New TS 3.1.10 provides the necessary exception to TS 3.4.4, RCS Loops, which otherwise requires full RCS flow in MODE 5 whenever the Plant Control System is capable of rod withdrawal or one or more rods are not fully inserted. A reference to TS 3.1.10 is added to TS LCO 3.0.7 to include the new test exception. The LCO Applicability of TS 3.4.8, Minimum RCS Flow, is revised to align with the purpose of that TS with respect to the initial conditions credited for the analysis of an inadvertent boron dilution event.

2. DETAILED DESCRIPTION This amendment request adds a new test exception, TS 3.1.10, to allow rod movement and rod drop time testing in MODE. The MODE 5 requirement for four RCPs in operation with variable speed control bypassed is imposed by TS LCO 3.4.4, RCS Loops, when the Plant Control System (PLS) is capable of rod withdrawal or one or more rods are not fully inserted. Whenever the PLS is capable of rod withdrawal or one or more rods are not fully inserted, there is the possibility of an inadvertent rod withdrawal from subcritical, resulting in a power excursion in the area of the withdrawn rod. Reactor coolant flow is needed to prevent a departure from nucleate boiling (DNB). The required RCP flow is an initial condition assumed in the analysis of the uncontrolled rod cluster control assembly bank withdrawal from subcritical event. New TS 3.1.10 provides the necessary exception to TS 3.4.4, RCS Loops, which otherwise requires full RCS flow in MODE 5 whenever the PLS is capable of rod withdrawal or one or more rods are not fully inserted. A reference to TS 3.1.10 is also added to TS LCO 3.0.7 to include this new test exception. New TS 3.1.10 reflects the appropriate controls and initial conditions to allow rod movement and rod drop time testing under cold conditions (MODE 5) with LCO 3.4.4 not met.

The LCO Applicability of TS 3.4.8, Minimum RCS Flow, is revised to be applicable regardless of the status of the plant control system capability of rod withdrawal or rod insertion status. The proposed change to TS 3.4.8 will align the revised LCO Applicability (i.e., MODES 3, 4, and 5 with unborated water sources not isolated from the RCS) with the initial conditions credited for the inadvertent boron dilution event, which assumes a minimum mixing flow through the reactor core. Furthermore, with the proposed addition of TS 3.1.10 test exception allowing rod movement in MODE 5, the existing Applicability for TS 3.4.8 (which currently includes "with Plant Control System incapable of rod withdrawal [and] all rods fully inserted") would be exited Page 3 of 16

ND-17-1559 Request for License Amendment Regarding Technical Specification Changes to Support Control Rod Testing in Cold Shutdown With Reactor Coolant Pumps Not in Operation (LAR-17-034) and the requirement for minimum mixing required for unborated water sources being unisolated during MODE 5 would be removed.

Licensing Basis Change Descriptions TS LCO 3.0.7 is revised to add a reference to new test exception TS 3.1.10.

New TS 3.1.10, Rod Withdrawal test Exception - MODE 5, is added to require the following LCO:

During the performance of rod movement and rod drop time testing, the requirements of LCO 3.4.4, RCS Loops, may be suspended provided boron concentration of the reactor coolant system is greater than the all rods out (ARO) boron concentration that provides keff < 0.99.

New TS 3.1.10 has the following LCO Applicability:

MODE 5 with LCO 3.4.4 not met.

New TS 3.1.10 has one Condition. Condition A states:

Requirements of the LCO not met.

New Required Action A.1 reads:

Initiate action to fully insert all rods.

The Required Action A.1 Completion Time is immediately.

New Required Action A.2 reads:

Place the Plant Control System in a condition incapable of rod withdrawal.

The Required Action A.2 Completion Time is 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

New Surveillance Requirement (SR) 3.1.10.1, with a Frequency of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, is added for TS 3.1.10. SR 3.1.10.1 reads:

Verify boron concentration of the reactor coolant system is greater than the ARO boron concentration providing keff < 0.99.

The LCO Applicability for TS 3.4.8, Minimum RCS Flow, currently reads:

MODES 3, 4, and 5 with Plant Control System incapable of rod withdrawal, all rods fully inserted, and unborated water sources not isolated from the RCS.

The LCO Applicability for TS 3.4.8, Minimum RCS Flow, is revised to read:

MODES 3, 4, and 5 with unborated water sources not isolated from the RCS.

Conforming TS Bases changes are provided for information only in Enclosure 3.

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ND-17-1559 Request for License Amendment Regarding Technical Specification Changes to Support Control Rod Testing in Cold Shutdown With Reactor Coolant Pumps Not in Operation (LAR-17-034)

3. TECHNICAL EVALUATION Reactor Coolant System (RCS) Overview The primary function of the RCS is removal of the heat generated in the fuel due to the fission process, and transfer of this heat, via the steam generators (SGs) to the secondary plant. The reactor coolant is circulated through two loops connected in parallel to the reactor vessel, each containing a SG, two reactor coolant pumps (RCPs), and appropriate flow and temperature instrumentation for both control and protection. The SGs provide the heat sink to the secondary coolant. The RCPs circulate the primary coolant through the reactor vessel and SGs at a sufficient rate to ensure proper heat transfer and prevent fuel damage. This forced circulation of the reactor coolant ensures mixing of the coolant for proper boration and chemistry control.

The RCPs must be started using the variable speed controller with the Plant Control System (PLS) incapable of rod withdrawal and all rods fully inserted. The controller shall be bypassed prior to making the PLS capable of rod withdrawal or withdrawing one or more rods.

Chemical and Volume Control System (CVS) Overview One of the primary functions of the CVS is to maintain the reactor coolant chemistry conditions by controlling the concentration of boron in the coolant for plant startups, normal dilution to compensate for fuel depletion, shutdown boration, and isolation of excessive makeup. In the dilute mode of operation, unborated demineralized water may be supplied directly to the reactor coolant system. Another of the primary functions of the CVS is to maintain the reactor coolant inventory by providing water makeup for RCS LEAKAGE, shrinkage of the reactor coolant during cooldowns, and RCS boron concentration changes. In the automatic makeup mode of operation, the pressurizer water level starts and stops CVS makeup to the RCS. Although the CVS is not considered a safety-related system, certain functions of the system are safety-related. The appropriate components have been classified and designed as safety-related. The safety-related functions provided by the CVS include containment isolation of CVS lines penetrating containment, termination of inadvertent boron dilution, and preservation of the RCS pressure boundary, including isolation of CVS letdown from the RCS.

One of the initial assumptions in the analysis of an inadvertent boron dilution event is the assumption that the increase in core reactivity, created by the dilution event, can be detected by the source range neutron flux instrumentation. The source range neutron flux instrumentation will then supply a flux doubling signal to the CVS demineralized water isolation valves and the CVS makeup line isolation valves causing these valves to close and terminate the boron dilution event.

Protection and Safety Monitoring System (PMS) Overview The PMS senses plant conditions, generates the signals to trip the reactor and actuate engineered safety features (ESF), and provides the equipment necessary to monitor plant safety-related functions during and following designated events.

As discussed in Updated Final Safety Analysis Report (UFSAR) Subsection 7.1.1, the PMS provides detection of off-nominal conditions and actuation of appropriate safety-related Page 5 of 16

ND-17-1559 Request for License Amendment Regarding Technical Specification Changes to Support Control Rod Testing in Cold Shutdown With Reactor Coolant Pumps Not in Operation (LAR-17-034) functions necessary to achieve and maintain the plant in a safe shutdown condition. In addition, the PMS provides the equipment necessary to monitor the plant safety-related functions during and following an accident as discussed in NRC Regulatory Guide 1.97.

The PMS performs the reactor trip system (RTS) functions discussed in UFSAR Section 7.2, the ESF actuation functions discussed in UFSAR Section 7.3, and the post-accident monitoring functions discussed in UFSAR Section 7.5.

Source Range Neutron Flux Instrumentation Overview In MODE 3, 4, or 5 with the reactor shutdown, the Source Range Neutron Flux trip function must be OPERABLE if the PLS is capable of rod withdrawal or one or more rods are not fully inserted to provide core protection against an inadvertent rod withdrawal accident during a subcritical condition. The source range detectors are also required to be OPERABLE to provide protection for events like an inadvertent boron dilution. These functions are discussed in UFSAR Sections 7.2 and 7.3, and OPERABILITY requirements are specified in TS 3.3.2, Reactor Trip System (RTS) Source Range Instrumentation, and TS 3.3.8, Engineered Safety Feature Actuation System (ESFAS) Instrumentation, Function 17, Source Range Neutron Flux Doubling.

Plant Control System (PLS) Overview The PLS integrates the control of reactor power, reactor cooling, and multiple reactor support systems required for normal and transient conditions. The primary means for PLS to regulate reactor power and power distribution is to position clusters of control rods in the reactor core using the rod control system. The control rods are moved into and out of the reactor core by means of electromagnetic jacking mechanisms, called control rod drive mechanisms, located on the reactor vessel head.

New TS 3.1.10 and MODE 5 Testing Requirements In MODE 3, 4, or 5 with the PLS capable of rod withdrawal or one or more rods not fully inserted, TS 3.4.4, RCS Loops, requires the RCPs to be OPERABLE and in operation with variable speed control bypassed. New TS 3.1.10 allows a test exception from the requirements of TS 3.4.4, RCS Loops, for rod movement and rod drop time testing under cold (MODE 5) conditions since TS 3.4.4 would otherwise require full RCS flow in MODE 5 whenever the PLS is capable of rod withdrawal or one or more rods are not fully inserted. New TS 3.1.10 is needed to reflect the appropriate controls and initial conditions to allow rod movement and rod drop time testing under cold conditions (MODE 5) with LCO 3.4.4 not met.

Although the initial testing of the Rod Control System, Rod Position Indication, Control Rod Drive Mechanisms (CRDMs), and Rod Drop Times is performed on one bank at a time during the initial startup test program, retesting of rods that fail to meet the rod drop time test acceptance criteria may be performed on an individual rod cluster control assembly (RCCA), on a group of RCCAs, on a bank of RCCAs, or on all RCCAs at the same time. Therefore, prior to making the PLS capable of rod withdrawal for testing, the RCS will be borated to a concentration sufficient to assure that the reactor core remains subcritical with all shutdown, control, and gray rods fully withdrawn.

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ND-17-1559 Request for License Amendment Regarding Technical Specification Changes to Support Control Rod Testing in Cold Shutdown With Reactor Coolant Pumps Not in Operation (LAR-17-034)

Protection from potential reactivity events during shutdown conditions is provided by adequate shutdown margin (SDM) to ensure the reactor core remains subcritical at all times. As such, prevention of criticality is the acceptance criterion for exercising test exception LCO 3.1.10. The LCO 3.1.10 required boration will assure that the core remains subcritical at all times during any planned rod withdrawals or any inadvertent rod withdrawal errors. As such, there is no challenge to DNB or other fuel acceptance criteria.

A reference to new TS 3.1.10 is also added to TS LCO 3.0.7 to include this new test exception.

In MODE 5, the typical tests that require rod withdrawal include:

a. Cold Rod Drop Time Testing. UFSAR Subsection 14.2.10.1.14 requires a determination of rod drop times for each RCCA under both cold no-flow and hot full-flow conditions during initial startup testing. In addition, future post-refueling outage startup activities may warrant performance of cold rod drop time testing in MODE 5.

Performing cold rod drop time testing following reactor vessel assembly would provide an early opportunity to discover any significant anomalies in rod drop performance that might indicate that the rod drop time test of SR 3.1.4.3 would not be met when performed.

b. Digital Rod Position Indication (DRPI) System and Bank Demand Position Indication System Surveillance Testing. UFSAR Subsection 14.2.10.1.12 requires testing of the rod position indication system at no-load operating temperature and pressure (MODE 3) with at least one RCP operating during initial startup testing. Rod position indication is also tested under cold conditions during initial startup testing and is needed to support the performance of rod drop time testing. In addition, future post-refueling outage startup activities may warrant performance of rod position indication system testing in MODE 5. Performing cold rod position indication testing following reactor vessel assembly would provide assurance that the system is operating properly.

c. Rod Control System Testing. UFSAR Subsection 14.2.10.1.11 requires testing of the rod control system at no-load operating temperature and pressure during initial startup testing. In addition, future post-refueling outage startup activities may warrant performance of rod control system testing in MODE 5. The rod control portion of the PLS is also tested under cold conditions following reactor vessel assembly and is needed in order to withdraw the RCCAs for rod drop time testing.

d. Control Rod Drive Mechanism (CRDM) Testing. UFSAR Subsection 14.2.10.1.13 requires testing of the CRDMs under both cold shutdown and hot standby conditions.

In addition, future post-refueling outage startup activities may warrant performance of CRDM testing in MODE 5. The CRDMs are also required in order to withdraw RCCAs and to release them for rod drop time testing following reactor vessel assembly.

LCO Note 3 in TS 3.4.4, RCS Loops, allows an exception for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> (per 8-hour period) to perform necessary testing such as the verification of the RCP coastdown performance and validation of the rod drop times during cold conditions, both with and without flow. It is estimated that it will take 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to complete the rod movement and rod drop time testing at Page 7 of 16

ND-17-1559 Request for License Amendment Regarding Technical Specification Changes to Support Control Rod Testing in Cold Shutdown With Reactor Coolant Pumps Not in Operation (LAR-17-034)

MODE 5 conditions. New LCO 3.1.10 allows an exception to TS 3.4.4 with respect to the full RCS flow required by that LCO and also provides sufficient time to perform the rod movement and rod drop time testing.

Subcritical boron concentrations must be met assuming all control, shutdown, and gray rods are fully withdrawn. The boration requirements provide sufficient reactivity margin to assure that the reactor core remains subcritical and acceptable fuel design limits will not be exceeded for normal shutdown and anticipated operational occurrences (AOOs).

If an unexpected reactivity transient were to occur during the MODE 5 rod movement and rod drop time testing, defense in depth is afforded by the Source Range Neutron Flux - High Setpoint Reactor Trip Function (TS 3.3.2) and the Manual Reactor Trip Function (TS 3.3.5).

Both Functions are required to be OPERABLE in MODE 5 when the Plant Control System is capable of rod withdrawal or one or more rods are not fully inserted.

TS 3.4.8 LCO Applicability Change The purpose of TS 3.4.8 is to establish the credited RCS mixing when unborated water sources are not isolated from the RCS. The PLS capability for rod withdrawal or rod position is not necessary for LCO Applicability; therefore, changes to TS 3.4.8 LCO Applicability are proposed.

The inadvertent boron dilution event is analyzed in MODES 3, 4, and 5 with an assumption of sufficient RCS flow to promote adequate mixing. This is assured by TS 3.4.8, Minimum RCS Flow, which is applicable when unborated water sources are not isolated from the RCS.

An initial condition in the analysis of an inadvertent boron dilution event in MODE 3, 4, or 5 is the assumption of a minimum mixing flow (3,000 gpm) in the RCS. In this scenario, unborated water is inadvertently introduced into the RCS, is uniformly mixed with the primary coolant, and flows to the core. The increase in reactivity is detected by the source range instrumentation which provides a signal to terminate the inadvertent dilution before the available SDM is lost.

Function 17 of TS Table 3.3.8-1, Source Range Neutron Flux Doubling, provides this protection.

If there is inadequate mixing in the RCS, the unborated water may stratify in the primary system, and there will be no indication by the source range instrumentation that a dilution event is in progress. When primary flow is increased, such that unborated water enters the active core region, dilution may progress to the point that mitigation by the source range instrumentation is too late to prevent the loss of SDM.

Summary Conclusions The proposed changes do not adversely impact any functions associated with containing, controlling, channeling, monitoring, or processing radioactive or nonradioactive materials, nor do they diminish the functionality of any design or operational features that are credited with controlling the release of effluents during plant operation. The types and quantities of expected plant effluents are not changed. No effluent release path is associated with the proposed changes to the TS. Therefore, neither radioactive nor nonradioactive material effluents are affected by this proposed change.

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ND-17-1559 Request for License Amendment Regarding Technical Specification Changes to Support Control Rod Testing in Cold Shutdown With Reactor Coolant Pumps Not in Operation (LAR-17-034)

These proposed changes do not adversely impact radiologically controlled zones. Plant radiation zones, radiation controls established to satisfy 10 CFR Part 20 requirements, and expected amounts and types of radioactive materials are not affected by the proposed changes.

Therefore, individual and cumulative radiation exposures are not affected by this change.

The change activity has no adverse impact on the emergency plan or the physical security plan implementation, because there are no changes to physical access to credited equipment inside the Nuclear Island (including containment or the auxiliary building) and no adverse impact to plant personnels ability to respond to any plant operations or security event.

4. REGULATORY EVALUATION 4.1 Applicable Regulatory Requirements/Criteria 10 CFR 52.98(c) requires NRC approval for any modification to, addition to, or deletion from the terms and conditions of a Combined License (COL). This activity involves a change to COL Appendix A, Technical Specifications (TS); therefore, this activity requires NRC approval prior to making the plant-specific changes in this license amendment request.

10 CFR 52, Appendix D, VIII.C.6 states that after issuance of a license, Changes to the plant-specific TS (Technical Specifications) will be treated as license amendments under 10 CFR 50.90. 10 CFR 50.90 addresses the applications for amendments of licenses, construction permits, and early site permits. As discussed above, a change to COL Appendix A is requested, and thus a license amendment request (LAR) (as supplied herein) is required.

Section 182a of the Atomic Energy Act requires applicants for nuclear power plant operating licenses to include TS as part of the license. The TSs ensure the operational capability of structures, systems, and components that are required to protect the health and safety of the public. The NRCs requirements related to the content of the TSs are contained in Section 50.36 of Title 10 of the Code of Federal Regulations (10 CFR 50.36) which requires that the TSs include items in the following specific categories: (1) safety limits, limiting safety systems settings, and limiting control settings; (2) limiting conditions for operation (LCOs); (3) surveillance requirements per 10 CFR 50.36(c)(3); (4) design features; and (5) administrative controls.

This amendment application is related to the second category above since a new test exception TS 3.1.10 is added to allow for rod movement and rod drop testing in MODE 5. New TS 3.1.10 provides the necessary exception to TS 3.4.4, RCS Loops, which otherwise requires full RCS flow in MODE 5 whenever the Plant Control System is capable of rod withdrawal or one or more rods are not fully inserted. A reference to TS 3.1.10 is added to TS LCO 3.0.7 to include the new test exception. The LCO Applicability of TS 3.4.8, Minimum RCS Flow, is revised to align with the purpose of that TS with respect to the initial conditions credited for the analysis of an inadvertent boron dilution event.

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ND-17-1559 Request for License Amendment Regarding Technical Specification Changes to Support Control Rod Testing in Cold Shutdown With Reactor Coolant Pumps Not in Operation (LAR-17-034)

The following regulatory requirements apply to the Technical Specifications revised by the proposed amendment request:

10 CFR 50, Appendix A, General Design Criterion (GDC) 10 requires that the reactor core and associated coolant, control, and protection systems shall be designed with appropriate margin to assure that specified acceptable fuel design limits are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences.

The reactor core and associated coolant, control, and protection systems are designed to the following criteria:

No fuel damage occurs during normal core operation and operational transients (Condition I) or during transient conditions arising from occurrences of moderate frequency (Condition II). For normal operation, the plant is designed to accommodate a fuel defect level of up to 0.25 percent. Fuel damage, as used here, is defined as penetration of the fission product barrier, that is, the fuel rod cladding. The small number of clad defects that may occur are within the capability of the plant cleanup system and are consistent with the plant design bases.

The reactor can be returned to a safe shutdown state following a Condition III event, with only a small fraction of the fuel rods damaged, although sufficient fuel damage might occur to preclude the immediate resumption of operation.

The core remains intact with acceptable heat transfer geometry following transients arising from occurrences of limiting faults (Condition IV).

The reactor protection system is designed to actuate a reactor trip whenever necessary to prevent exceeding the fuel design limits. The core design, together with the process and decay heat removal systems, provide this capability under expected conditions of normal operation, with appropriate margins for uncertainties and anticipated transient situations. This includes the effects of the loss of reactor coolant flow, trip of the turbine generator, loss of normal feedwater, and loss of both normal and preferred power sources.

The proposed amendment does not involve a design change to any plant system. The compliance discussion in UFSAR Subsection 3.1.2 is unaffected.

10 CFR 50, Appendix A, GDC 25 requires that the protection system shall be designed to assure that specified acceptable fuel design limits are not exceeded for any single malfunction of the reactivity control systems, such as accidental withdrawal (not ejection or dropout) of the control rods.

The protection system is designed to limit reactivity transients so that the fuel design limits are not exceeded. Reactor shutdown by control rod insertion is independent of the normal control functions since the trip breakers interrupt power to the rod mechanisms regardless of existing control signals. Thus, in the postulated accidental withdrawal of a control rod or control rod bank (assumed to be initiated by a control malfunction), neutron flux, temperature, pressure, Page 10 of 16

ND-17-1559 Request for License Amendment Regarding Technical Specification Changes to Support Control Rod Testing in Cold Shutdown With Reactor Coolant Pumps Not in Operation (LAR-17-034) level, and flow signals would be generated independently. Any of these signals (trip demands) would operate the breakers to trip the reactor.

The AP1000 is designed to automatically terminate an inadvertent boron dilution event during manual or automatic operation at power, and also during startup and shutdown conditions.

The proposed amendment does not involve a design change to any plant system. The compliance discussion in UFSAR subsection 3.1.3 is unaffected.

10 CFR 50, Appendix A, GDC 26 requires that two independent reactivity control systems of different design principles shall be provided. One of the systems shall use control rods, preferably including a positive means for inserting the rods, and shall be capable of reliably controlling reactivity changes to assure that under conditions of normal operation, including anticipated operational occurrences, and with appropriate margin for malfunctions such as stuck rods, specified acceptable fuel design limits are not exceeded. The second reactivity control system shall be capable of reliably controlling the rate of reactivity changes resulting from planned, normal power changes (including xenon burnout) to assure that the acceptable fuel design limits are not exceeded. One of the systems shall be capable of holding the reactor core subcritical under cold conditions.

Two reactivity control systems are provided. These are rod cluster control assemblies and gray rod assemblies, and chemical shim (boric acid). The rod cluster control and gray rod assemblies are inserted into the core by the force of gravity.

During operation, the shutdown rod banks are fully withdrawn. The control rod system automatically maintains a programmed average reactor temperature compensating for reactivity effects associated with scheduled and transient load changes.

The shutdown and control rod banks are designed to provide reactivity margin to shut down the reactor during normal operating conditions and during anticipated operational occurrences, without exceeding specified fuel design limits. The safety analyses assume the most restrictive time in the core operating cycle and that the most reactive control rod cluster assembly is in the fully withdrawn position.

The safety-related passive systems provide the required boration to establish and maintain safe shutdown condition for the reactor core.

The proposed amendment does not involve a design change to any plant system. The compliance discussion in UFSAR Subsection 3.1.3 is unaffected.

10 CFR 50, Appendix A, GDC 27 requires that the reactivity control systems shall be designed to have a combined capability, in conjunction with poison addition by the emergency core cooling system, of reliably controlling reactivity changes to assure that under postulated accident conditions and with appropriate margin for stuck rods the capability to cool the core is maintained.

The plant is provided with the means of making and holding the core subcritical under any anticipated conditions and with appropriate margin for contingencies. Combined use of the control rod and the chemical shim control system permits the necessary shutdown margin to be Page 11 of 16

ND-17-1559 Request for License Amendment Regarding Technical Specification Changes to Support Control Rod Testing in Cold Shutdown With Reactor Coolant Pumps Not in Operation (LAR-17-034) maintained during long-term xenon decay and plant cooldown. The single highest worth control rod assembly is assumed to be stuck in the fully withdrawn position for this determination.

The proposed amendment does not involve a design change to any plant system. The compliance discussion in UFSAR Subsection 3.1.3 is unaffected.

10 CFR 50, Appendix A, GDC 28 requires that the reactivity control systems shall be designed with appropriate limits on the potential amount and rate of reactivity increase to assure that the effects of postulated reactivity accidents can neither (l) result in damage to the reactor coolant pressure boundary greater than limited local yielding nor (2) sufficiently disturb the core, its support structures, or other reactor pressure vessel internals to impair significantly the capability to cool the core. These postulated reactivity accidents shall include consideration of rod ejection (unless prevented by positive means), rod dropout, steam line rupture, changes in reactor coolant temperature and pressure, and cold water addition.

The maximum reactivity worth of the control rods and the maximum rates of reactivity increase employing control rods and boron removal are limited by design and operating procedures.

The appropriate reactivity addition rate for the withdrawal of control rods and the dilution rate of the boric acid in the reactor coolant system are specified in the precautions, limitations, and setpoint document and the control system setpoint study. Technical Specifications explicitly specify control rod bank alignment and insertion limits in addition to shutdown margin reactivity requirements.

The control rod reactivity addition rate is determined by the allowable rod control system withdrawal speed, in conjunction with the control rod worth, which varies throughout the operating cycle. The capability to change boron concentration is determined by the various plant systems that provide makeup to the reactor coolant system.

Core cooling capability following events such as rod ejection and steam line breaks is provided by keeping the reactor coolant pressure boundary stresses within faulted condition limits, as specified by applicable ASME codes. Structural deformations are also checked and limited to values that do not jeopardize the operation of needed safety-related features.

The proposed amendment does not involve a design change to any plant system. The compliance discussion in UFSAR Subsection 3.1.3 is unaffected.

10 CFR 50, Appendix A, GDC 29 requires that the protection and reactivity control systems shall be designed to assure an extremely high probability of accomplishing their safety functions in the event of anticipated operational occurrences.

The protection and reactivity control systems have an extremely high probability of performing their required safety-related functions in the event of anticipated operational occurrences. High quality equipment, diversity, and redundancy, support this probability. Loss of power to the protection system results in a reactor trip. Defense in depth is designed into AP1000 to reduce challenges to the protection and reactivity control systems.

The proposed amendment does not involve a design change to any plant system. The compliance discussion in UFSAR Subsection 3.1.3 is unaffected.

Page 12 of 16

ND-17-1559 Request for License Amendment Regarding Technical Specification Changes to Support Control Rod Testing in Cold Shutdown With Reactor Coolant Pumps Not in Operation (LAR-17-034) 4.2 Precedent No precedent is identified.

4.3 Significant Hazards Consideration Determination Southern Nuclear Operating Company (SNC) is requesting an amendment to Combined License (COL) Nos. NPF-91 and NPF-92 for Vogtle Electric Generating Plant (VEGP) Units 3 and 4, respectively. The requested amendment proposes to amend COL Appendix A, Technical Specifications (TS) by adding a new test exception TS 3.1.10, "Rod Withdrawal Test Exception - MODE 5," to allow for control rod movement and control rod drop testing in MODE 5. New TS 3.1.10 provides the necessary exception to TS 3.4.4, RCS Loops, which otherwise requires full RCS flow in MODE 5 whenever the Plant Control System is capable of rod withdrawal or one or more rods are not fully inserted. A reference to TS 3.1.10 is also added to TS LCO 3.0.7 to include the addition of this new test exception. The LCO Applicability of TS 3.4.8, Minimum RCS Flow, is revised to align with the purpose of that TS with respect to the initial conditions credited for the analysis of an inadvertent boron dilution event.

An evaluation to determine whether or not a significant hazards consideration is involved with the proposed amendment was completed by focusing on the three standards set forth in 10 CFR 50.92, Issuance of amendment, as discussed below:

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

Response: No.

There are no design changes associated with the proposed amendment. All design, material, and construction standards that were applicable prior to this amendment request will continue to be applicable.

The Plant Control System (PLS), Reactor Coolant System (RCS), Chemical and Volume Control System (CVS), and Protection and Safety Monitoring System (PMS) will continue to function in a manner consistent with the existing plant design basis.

There will be no changes to the PLS, RCS, CVS, or PMS operating limits.

The proposed amendment will not affect accident initiators or precursors or alter the design, conditions, and configuration of the facility, or the manner in which the plant is operated and maintained, with respect to such initiators or precursors.

The proposed amendment will preclude reactor core criticality during the use of new TS 3.1.10. The proposed amendment will not alter the ability of structures, systems, and components (SSCs) to perform their specified safety functions.

Accident analysis acceptance criteria will continue to be met with the proposed changes. The proposed changes will not affect the source term, containment isolation, or radiological release assumptions used in evaluating the radiological consequences of any accident previously evaluated. The proposed changes will not Page 13 of 16

ND-17-1559 Request for License Amendment Regarding Technical Specification Changes to Support Control Rod Testing in Cold Shutdown With Reactor Coolant Pumps Not in Operation (LAR-17-034) alter any assumptions or change any mitigation actions in the radiological consequence evaluations in the Updated Final Safety Analysis Report (UFSAR).

The applicable radiological dose acceptance criteria will continue to be met.

The proposed amendment adds a new test exception TS 3.1.10, revises TS LCO 3.0.7 to reference the new TS 3.1.10, and modifies the LCO Applicability of TS 3.4.8 to be consistent with the purpose of that TS as an initial condition of the inadvertent boron dilution analyses, but does not physically alter any safety-related systems.

Therefore, the proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No.

With respect to any new or different kind of accident, there are no proposed design changes nor are there any changes in the method by which any safety-related plant SSC performs its specified safety function. The proposed change will not affect the normal method of plant operation or change any operating parameters. No equipment performance requirements will be affected. The proposed change will not alter any assumptions made in the safety analyses.

The proposed amendment adds a new test exception TS 3.1.10, revises TS LCO 3.0.7 to reference the new TS 3.1.10, and modifies the LCO Applicability of TS 3.4.8 to be consistent with the purpose of that TS as an initial condition of the inadvertent boron dilution analyses. The proposed change does not involve a physical modification of the plant.

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

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

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

Response: No.

There will be no effect on those plant systems necessary to effect the accomplishment of protection functions. No instrument setpoints or system response times are affected. None of the acceptance criteria for any accident analysis will be Page 14 of 16

ND-17-1559 Request for License Amendment Regarding Technical Specification Changes to Support Control Rod Testing in Cold Shutdown With Reactor Coolant Pumps Not in Operation (LAR-17-034) changed. The proposed amendment will have no impact on the radiological consequences of a design basis accident.

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

Based on the above, it is concluded that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

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. Pursuant to 10 CFR 50.92(c), the requested change does not involve a Significant Hazards Consideration

5. ENVIRONMENTAL CONSIDERATIONS The requested amendment proposes to amend COL Appendix A, Technical Specifications (TS).

The proposed changes revise COL Appendix A, plant-specific Technical Specifications (TS) by adding a new test exception TS 3.1.10 to allow for rod movement and rod drop testing in MODE 5. New TS 3.1.10 provides the necessary exception to TS 3.4.4, RCS Loops, which otherwise requires full RCS flow in MODE 5 whenever the Plant Control System is capable of rod withdrawal or one or more rods are not fully inserted. A reference to TS 3.1.10 is also added to TS LCO 3.0.7 to include the addition of this new test exception. The LCO Applicability of TS 3.4.8, Minimum RCS Flow, is revised to align with the purpose of that TS with respect to the initial conditions credited for the analysis of an inadvertent boron dilution event.

The details of the proposed changes are provided in Sections 2 and 3 of this license amendment request.

This review has determined that the proposed changes require an amendment to the COL.

However, a review of the anticipated construction and operational effects of the requested amendment has determined that the requested amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9), in that:

(i) There is no significant hazards consideration.

As documented in Section 4.3, Significant Hazards Consideration Determination, of this license amendment request, an evaluation was completed to determine whether or not a significant hazards consideration is involved by focusing on the three standards set forth in 10 CFR 50.92, Issuance of amendment. The Significant Hazards Consideration determined that (1) the proposed amendment does not involve a significant increase in the probability or consequences of an accident Page 15 of 16

ND-17-1559 Request for License Amendment Regarding Technical Specification Changes to Support Control Rod Testing in Cold Shutdown With Reactor Coolant Pumps Not in Operation (LAR-17-034) previously evaluated; (2) the proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated; and (3) the proposed amendment does not involve a significant reduction in a margin of safety.

Therefore, it is concluded that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and accordingly, a finding of no significant hazards consideration is justified.

(ii) There is no significant change in the types or significant increase in the amounts of any effluents that may be released offsite.

The changes are unrelated to any aspects of plant construction or operation that would introduce any changes to effluent types (e.g., effluents containing chemicals or biocides, sanitary system effluents, and other effluents) or affect any plant radiological or non-radiological effluent release quantities. Furthermore, the proposed amendment does not diminish the functionality of any design or operational features that are credited with controlling the release of effluents during plant operation.

Therefore, it is concluded that the proposed amendment does not involve a significant change in the types or a significant increase in the amounts of any effluents that may be released offsite.

(iii) There is no significant increase in individual or cumulative occupational radiation exposure.

The proposed amendment does not affect plant radiation zones (addressed in UFSAR Section 12.3), and controls under 10 CFR 20 preclude a significant increase in occupational radiation exposure. Therefore, the proposed amendment does not involve a significant increase in individual or cumulative occupational radiation exposure.

Based on the above review of the proposed amendment, it has been determined that anticipated construction and operational effects of the proposed amendment do not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in the individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).

Therefore, pursuant to 10 CFR 51.22(b), an environmental impact statement or environmental assessment of the proposed amendment is not required.

6. REFERENCES None Page 16 of 16

Southern Nuclear Operating Company ND-17-1559 Enclosure 2 Vogtle Electric Generating Plant (VEGP) Units 3 and 4 Proposed Changes to Licensing Basis Documents (LAR-17-034)

Insertions Denoted by Blue Underline and Deletions by Red Strikethrough Omitted text is identified by three asterisks ( * * * )

(This Enclosure consists of 3 pages, including this cover page)

ND-17-1559 Proposed Changes to Licensing Basis Documents (LAR-17-034)

Revise COL Appendix A, Technical Specification 3.0.7, as follows:

LCO 3.0.7 Test Exception LCOs 3.1.8 and 3.1.10 allows specified Technical Specification (TS) requirements to be changed to permit performance of special tests and operations. Unless otherwise specified, all other TS requirements remain unchanged. Compliance with Test Exception LCOs is optional. When a Test Exception LCO is desired to be met but is not met, the ACTIONS of the Test Exception LCO shall be met. When a Test Exception LCO is not desired to be met, entry into a MODE or other specified condition in the Applicability shall be made in accordance with the other applicable Specifications.

Revise COL Appendix A, Technical Specification to include new LCO 3.1.10, as follows:

3.1.10 Rod Withdrawal Test Exception - MODE 5 LCO 3.1.10 During the performance of rod movement and rod drop time testing, the requirements of LCO 3.4.4, RCS Loops, may be suspended provided boron concentration of the reactor coolant system is greater than the all rods out (ARO) boron concentration that provides keff < 0.99.

APPLICABILITY: MODE 5 with LCO 3.4.4 not met.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Requirements of the A.1 Initiate action to fully insert Immediately LCO not met. all rods.

AND A.2 Place the Plant Control 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> System in a condition incapable of rod withdrawal.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.10.1 Verify boron concentration of the reactor coolant 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> system is greater than the ARO boron concentration providing keff < 0.99.

Page 2 of 3

ND-17-1559 Proposed Changes to Licensing Basis Documents (LAR-17-034)

Revise COL Appendix A, Technical Specification 3.4.8, as follows:

APPLICABILITY: MODES 3, 4, and 5 with Plant Control System incapable of rod withdrawal, all rods fully inserted, and unborated water sources not isolated from the RCS.

Page 3 of 3

Southern Nuclear Operating Company ND-17-1559 Enclosure 3 Vogtle Electric Generating Plant (VEGP) Units 3 and 4 Conforming Technical Specification Bases Changes (For Information Only)

(LAR-17-034)

Insertions Denoted by Blue Underline and Deletions by Red Strikethrough Omitted text is identified by three asterisks ( * * * )

(This Enclosure consists of 6 pages, including this cover page)

ND-17-1559 Conforming Technical Specification Bases Changes (For Information Only) (LAR-17-034)

Technical Specification Bases, Section B 3.0.7 LCO 3.0.7 There are certain special tests and operations required to be performed at various times over the life of the unit. These special tests and operations are necessary to demonstrate select unit performance characteristics, to perform special maintenance activities, and to perform special evolutions. Test Exception LCOs 3.1.8 and 3.1.10 allows specified Technical Specification (TS) requirements to be changed to permit performance of these special tests and operations, which otherwise could not be performed if required * *

• Technical Specification Bases, Section B 3.1.10 B 3.1.10 Rod Withdrawal Test Exception - MODE 5 BASES BACKGROUND In MODE 5, there are exceptions required to allow performance of tests that require rod withdrawal and movement. The typical tests that require rod withdrawal in MODE 5 are:

a. Cold Rod Drop Time Testing. Successful performance of cold rod drop time testing in MODE 5 following reactor vessel (RV) assembly provides an early opportunity to discover any significant anomalies in rod drop performance that might indicate that the rod drop time test of SR 3.1.4.3 in LCO 3.1.4, Rod Group Alignment Limits, would not be met when performed;

b. DRPI System and Bank Demand Position Indication System Surveillance Testing. SR 3.1.7.1 specifies verification that each DRPI agrees within 12 steps of the group demand position for the full indicated range of rod travel, prior to criticality after each removal of the reactor head. These indication systems are needed to support Cold Rod Drop time testing;

c. Rod Control System Testing. Successful testing of the Rod Control System in MODE 5 is necessary to support the performance of Cold Rod Drop time testing following RV assembly; and

d. Control Rod Drive Mechanism (CRDM) Testing. Successful testing of the CRDMs in MODE 5 is necessary to support the performance of Cold Rod Drop time testing following RV assembly.

Page 2 of 6

ND-17-1559 Conforming Technical Specification Bases Changes (For Information Only) (LAR-17-034)

B 3.1.10 Rod Withdrawal Test Exception - MODE 5 (continued)

The MODE 5 Rod Withdrawal Test Exception permits relaxation of LCO 3.4.4 to allow performance of tests that require rod withdrawal and movement. Compensating requirements are applied to provide adequate assurance of subcriticality during the testing.

Preliminary rod drop time testing, Digital Rod Position Indication (DRPI)

System and Bank Demand Position Indication System surveillance testing, Rod Control System testing, and Control Rod Drive Mechanism (CRDM) testing may be performed prior to plant heatup following refueling.

APPLICABLE The tests described above require operating the plant with less than the SAFETY RCS flow required by LCO 3.4.4, RCS Loops. However, this exception ANALYSES is acceptable under the subcritical boron concentration limitations imposed by the LCO. Protection from potential reactivity events during shutdown conditions is provided by adequate boration to ensure the reactor core remains subcritical at all times. As such, prevention of criticality is the acceptance criterion for exercising this test exception.

The required boration will assure that the core remains subcritical at all times during any planned rod withdrawals or any inadvertent rod withdrawal errors. As such, there is no challenge to DNB or other fuel acceptance criteria.

As described in LCO 3.0.7, compliance with Test Exception LCOs is optional, and therefore no criteria of 10 CFR 50.36(c)(2)(ii) apply. Test exception LCOs provide flexibility to perform certain operations by appropriately modifying requirements of other LCOs. A discussion of the criteria satisfied for the other LCOs is provided in their respective Bases.

LCO This LCO allows a MODE 5 exception to the LCO 3.4.4 requirement that 100% reactor coolant system flow is required whenever the Plant Control System is capable of rod withdrawal or one or more rods are not fully inserted. An exception to this requirement is needed to permit tests involving rod withdrawal.

Subcritical boron concentrations must be met during the use of this test exception LCO, assuming that all shutdown, control, and gray rods are fully withdrawn. The boration limitation specified precludes reactor criticality.

Page 3 of 6

ND-17-1559 Conforming Technical Specification Bases Changes (For Information Only) (LAR-17-034)

B 3.1.10 Rod Withdrawal Test Exception - MODE 5 (continued)

APPLICABILITY This LCO is applicable in MODE 5 with LCO 3.4.4, RCS Loops, not met.

This allows testing that requires rod withdrawal. During testing, rod withdrawal is enabled with the Reactor Trip Breakers (RTBs) closed and the Plant Control System capable of rod withdrawal.

Rod withdrawal under this LCO may only be enabled in MODE 5 to perform testing.

ACTIONS A.1 and A.2 If the LCO is not met, the reactor must be brought to a MODE in which the LCO does not apply. To achieve this status, action is initiated immediately to fully insert all rods and, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, the PLS is placed in a condition incapable of rod withdrawal. This precludes rod withdrawal events which are unanalyzed. The Completion Times for Required Actions A.1 and A.2 are reasonable, based on operating experience, to reach the specified condition in an orderly manner and without challenging plant systems SURVEILLANCE SR 3.1.10.1 REQUIREMENTS This SR requires verification that boron concentration of the reactor coolant system is greater than the all-rods-out boron concentration providing a keff < 0.99. Assuming all rods are fully withdrawn ensures that the reactor will remain subcritical for all planned tests.

REFERENCES None Page 4 of 6

ND-17-1559 Conforming Technical Specification Bases Changes (For Information Only) (LAR-17-034)

Technical Specification Bases, Section B 3.4.4 APPLICABLE SAFETY ANALYSES (continued)

Therefore, in MODE 3, 4 or 5 with the PLS capable of rod withdrawal or one or more rods not fully inserted, accidental control rod withdrawal from subcritical is postulated and requires the RCPs to be OPERABLE and in operation to ensure that the accident analysis limits are met.

In MODES 3, 4 and 5 with the PLS incapable of rod withdrawal and all rods fully inserted, RCS circulation is considered in the determination of the time available for mitigation of the accidental boron dilution event.

This is addressed in LCO 3.4.8, Minimum RCS Flow.

RCS Loops satisfy Criteria 2 and 3 of 10 CFR 50.36(c)(2)(ii).

LCO (continued)

This test is generally performed in MODE 3 during the initial startup testing program, and as such should only be performed once. If, however, changes are made to the RCS that would cause a change to the flow characteristics of the RCS, the input values of the coastdown curve may need to be revalidated by conducting the test again.

Another test performed during the startup testing program is the validation of the rod drop times during cold conditions, both with and without flow.

The nNo-flow tests testing may be performed in MODE 3, 4, or 5, and requires that the pumps be stopped for a short period of time. The Note 3 permits removing all RCPs from operation in order to perform this testing and validate the assumed analysis values. As with the validation of the pump coastdown curve, this test should only be performed once, unless the flow characteristics of the RCS are changed. The 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> time period specified is adequate to perform the desired tests and experience has shown that boron stratification is not a problem during this short period with no forced flow.

Page 5 of 6

ND-17-1559 Conforming Technical Specification Bases Changes (For Information Only) (LAR-17-034)

Technical Specification Bases, Section B 3.4.8 LCO (continued)

Note 1 permits all RCPS to be removed from operation for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> per 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> period. The purpose of the Note is to permit tests that are designed to validate various accident analysis values. * *

• Another test performed during the startup testing program is the validation of the rod drop times during cold conditions, both with and without flow.

The nNo-flow tests testing may be performed in MODE 3, 4, or 5, and requires that the pumps be stopped for a short period of time. The Note permits removing all RCPs from operation in order to perform this testing and validate the assumed analysis values. As with the validation of the pump coastdown curve, this test should only be performed once, unless the flow characteristics of the RCS are changed. The 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> time period specified is adequate to perform the desired tests and experience has shown that boron stratification is not a problem during this short period with no forced flow.

APPLICABILITY Minimum RCS flow is required in MODES 3, 4, and 5 with the Plant Control System incapable of rod withdrawal, all rods inserted, and unborated water sources not isolated from the RCS because an inadvertent BDE is considered possible in these MODES.

Page 6 of 6