ML21053A342

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Proposed License Amendment Request to Clarify Shutdown Bank Technical Specification Requirements and Add Alternative Control Rod Position Monitoring Requirements
ML21053A342
Person / Time
Site: Millstone Dominion icon.png
Issue date: 02/22/2021
From: Mark D. Sartain
Dominion Energy Nuclear Connecticut
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
21-005
Download: ML21053A342 (29)
v • d • e


Text

Dom inion Energy Nuclea r Con necticut, Inc.

5000 Domin ion Bou levard, Glen Allen, VA 23060 ~ Dominion Dom in io n En ergy.com ~ Energy February 22, 2021 U.S. Nuclear Regulatory Commission Serial No. 21 -005 Attention: Document Control Desk NRA/SS: RO Washington, DC 20555 Docket No. 50-423 License No. NPF-49 DOMINION ENERGY NUCLEAR CONNECTICUT, INC.

MILLSTONE POWER STATION UNIT 3 PROPOSED LICENSE AMENDMENT REQUEST TO CLARIFY SHUTDOWN BANK TECHNICAL SPECIFICATION REQUIREMENTS AND ADD AL TERNA TE CONTROL ROD POSITION MONITORING REQUIREMENTS Pursuant to 10 CFR 50.90, Dominion Energy Nuclear Connecticut, Inc. (DENC) is submitting a license amendment request (LAR) to revise the Technical Specifications (TS) for Millstone Power Station Unit 3 (MPS3). The LAR proposes revisions to TS 3.1.3.2 to provide an alternative monitoring option for the condition where a maximum of one digital rod position indicator per bank is inoperable. Specifically, as an alternative to determining the position of the non-indicating rod(s) indirectly by the movable incore detectors at a frequency of once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, the change would allow rod position verification to be performed based on the occurrence of rod movement or power level change. This revision is consistent with Technical Specification Task Force Traveler 547, Revision 1 (ADAMS Accession No. ML15365A610), and would provide alternate TS Actions to allow the position of the rod to be monitored by a means other than movable incore detectors. The LAR proposes revisions to TS 3.1.3.5 to replace shutdown "rods" with shutdown "banks," which is consistent with wording in the Standard Technical Specifications for Westinghouse Plants as provided in NUREG-1431, Revision 4 (ADAMS Accession No. ML 121 OOA222) . The LAR also includes administrative changes to revise the title of TS 3.1 .3.6, to reflect that the requirements apply to control "banks," and modify TS 6.9.1.6.a and TS 6.9.1.6.b to cite the revised titles of TSs 3.1.3.5 and 3.1.3.6. provides DENC's description and assessment of the proposed change.

Attachments 2 and 3 provide the marked-up TS and TS Bases pages, respectively. The marked-up TS Bases pages are provided for information only. The changes to the affected TS Bases pages will be incorporated in accordance with the TS Bases Control Program after this amendment is approved.

The proposed amendment does not involve a Significant Hazards Consideration under the standards set forth in 10 CFR 50.92. The basis for this determination is included in . DENC has also determined that operation with the proposed change will not result in any significant increase in the amount of effluents that may be released offsite, or any significant increase in individual or cumulative occupational radiation exposure. Therefore, the proposed amendment is eligible for categorical exclusion from an environmental assessment as set forth in 10 CFR 51 .22(c)(9). Pursuant to 10 CFR

Serial No.21-005 Docket No. 50-423 Page 2 of 3 51.22(b), no environmental impact statement or environmental assessment is needed in connection with approval of the proposed change.

The proposed amendment has been reviewed and approved by the station's Facility Safety Review Committee.

DENG requests approval of this LAR by February 28, 2022, with a 90-day implementation period.

In accordance with 10 CFR 50.91(b), a copy of this LAR is being provided to the State of Connecticut.

If you have any questions or require additional information, please contact Mr. Shayan Sinha at (804) 273-4687.

  • Sincerely, Mark D. Sartain Vice President - Nuclear Engineering and Fleet Support COMMONWEALTH OF VIRGINIA )

)

COUNTY OF HENRICO )

The foregoing document was acknowledged before me, in and for the County and Commonwealth aforesaid, today by Mark D. Sartain, who is Vice President - Nuclear Engineering and Fleet Support of Dominion Energy Nuclear Connecticut, Inc. He has affirmed before me that he is duly authorized to execute and file the foregoing document on behalf of that company, and that the statements in the document are true to the best of his knowledge and belief.

SLY Acknowledged before me this 22,,J day of f.eh,,uwy ,2021. blic of Virginia l/~J.-+Y------

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My Commission Expires: 653 I A'l~

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Attachments:

1. Description and Assessment of Proposed Change
2. Marked-up Technical Specification Pages
3. Marked-up Technical Specification Bases Pages (for information only)

Commitments made in this letter: None

Serial No. 21 -005 Docket No. 50-423 Page 3 of 3 cc: U.S. Nuclear Regulatory Commission Region I 2100 Renaissance Blvd, Suite 100 King of Prussia, PA 19406-2713 Richard V. Guzman Senior Project Manager U.S. Nuclear Regulatory Commission One White Flint North, Mail Stop 08 C2 11555 Rockville Pike Rockville, MD 20852-2738 NRC Senior Resident Inspector Millstone Power Station Director, Radiation Division Department of Energy and Environmental Protection 79 Elm Street Hartford, CT 06106-5127

Serial No.21-005 Docket No. 50-423 Attachment 1 DESCRIPTION AND ASSESSMENT OF PROPOSED CHANGE Dominion Energy Nuclear Connecticut, Inc.

Millstone Power Station Unit 3

Serial No.21-005 Docket No. 50-423 Attachment 1, Page 1 of 15 1.0

SUMMARY

DESCRIPTION Pursuant to 10 CFR 50.90, Dominion Energy Nuclear Connecticut, Inc. (DENC) is submitting a license amendment request (LAR) to revise the Technical Specifications (TS) for Millstone Power Station Unit 3 (MPS3) . The LAR proposes revisions to TS 3.1.3.2 to provide an alternative monitoring option for the condition where a maximum of one digital rod position indicator (DRPI) per bank is inoperable. Specifically, as an alternative to determining the position of the non-indicating rod(s) indirectly by the movable incore detectors at a frequency of once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, the change would allow rod position verification to be performed based on the occurrence of rod movement or power level change. This revision is consistent with Technical Specification Task Force (TSTF)

Traveler 547, Revision 1 (Reference 1), and would provide alternate TS Actions to allow the position of the rod to be monitored by a means other than movable incore detectors.

The LAR also proposes revisions to TS 3.1.3.5 to replace shutdown "rods" with shutdown "banks," which is consistent with wording in the Standard Technical Specifications (STS) for Westinghouse Plants, as provided in NUREG-1431, Revision 4 (Reference 3) . The LAR also includes administrative changes to revise the title of TS 3.1.3.6 to reflect that the requirements apply to control "banks," and modify TS 6.9.1.6.a and TS 6.9.1.6.b to cite the revised titles of TSs 3.1.3.5 and 3.1.3.6. This attachment provides DENC's description and assessment of the proposed change .

2.0 DETAILED DESCRIPTION 2.1 System Design and Operation The rod cluster control assemblies (RCCA), or rods, are moved by their control rod drive mechanisms (CROM). Each CROM moves its RCCA one step (approximately 5/8 inch) at a time, but at varying rates depending on the signal output from the rod control system. The RCCAs are divided among control banks and shutdown banks. Each bank may be further subdivided into two groups to provide for precise reactivity control. If a bank of RCCAs consists of two groups, the groups are moved in a staggered fashion but always within one step of each other. MPS3 has four control banks and five shutdown banks.

The shutdown banks are maintained either in the fully inserted or fully withdrawn position. The control banks are moved in an overlap pattern, using the following withdrawal sequence: when control bank A reaches a predetermined height in the core , control bank B begins to move out with control bank A. Control bank A stops at the position of maximum withdrawal, and control bank B continues to move out. When control bank B reaches a predetermined height, control bank C begins to move out with control bank B. This sequence continues until control banks A, B, and C are at the fully withdrawn position, and control bank D is approximately

Serial No.21-005 Docket No. 50-423 Attachment 1, Page 2 of 15 halfway withdrawn. The insertion sequence is the opposite of the withdrawal sequence.

The control banks are used for precise reactivity control of the reactor. The positions of the control banks are normally automatically controlled (only in the insertion direction) by the rod control system but can also be manually controlled. The control banks must be maintained above insertion limits and are typically near the fully withdrawn position during normal full power operations.

The rod insertion limits of the shutdown and control rods are initial assumptions in all safety analyses that assume rod insertion upon reactor trip. The insertion limits ensure sufficient shutdown margin (SOM) is available when required for a reactor shutdown . The sequence and overlap limits on the control rods govern the withdrawal sequence and overlap of the control rod banks to ensure consistent reactivity changes due to rod movement. The alignment limits govern the position of individual rods with respect to each other to maintain a consistent power distribution across the reactor core. The shutdown and control bank insertion and alignment limits, axial flux difference (AFD), and quadrant power tilt ratio (QPTR) are process variables that are used to monitor and control the three-dimensional power distribution of the reactor core. Additionally, the control bank insertion limits control the reactivity that could be added in the event of a rod ejection accident. The TS requirements on rod alignment ensure that the assumptions in the safety analyses will remain valid. Mechanical or electrical failures may cause a rod to become inoperable (i.e., not trippable),

unable to be moved, or to become misaligned from its group. The requirements on rod operability ensure that on a reactor trip, the assumed reactivity will be inserted. Rod operability requirements (i.e., trippability) are not dependent upon the alignment requirements, which ensure that the rods and banks maintain the correct power distribution and rod alignment. The rod operability requirement is satisfied if the rod will fully insert in the required rod drop time assumed in the safety analyses. Rod control malfunctions that result in the inability to move a rod (e.g., rod lift coil failures), but that do not impact trippability, do not result in rod inoperability.

The associated Limiting Condition for Operation (LCO) require both rod operability (i.e., trippability) and rod alignment, and provide appropriate required Actions when the LCO is not met.

The axial position of shutdown rods and control rods is indicated by two separate and independent systems, which are the bank demand position indication system (commonly called group step counters) and the DRPI system. The bank demand position indication system counts the pulses from the rod control system that moves the respective group of rods. There is one step counter for each group of rods. Individual rods in a group all receive the same signal to move, and therefore should all be at the same

Serial No.21-005 Docket No. 50-423 Attachment 1, Page 3 of 15 position indicated by the group step counter for that group. The bank demand position indication system is considered relatively precise (+/- 1 step or+/- 5/8 inch). If a rod does not move one step for each demand pulse, the step counter will still count the pulse but incorrectly reflect the position of the rod.

The DRPI system provides a more accurate indication of actual rod position, but at a lower precision than the step counters. DRPI measures the actual position of each full-length rod using a detector that consists of discrete coils mounted concentrically with the rod drive pressure housing. The coils are located axially along the pressure housing and magnetically sense the entry and presence of the rod drive shaft through its centerline. For each detector, the coils are interlaced into two data channels, and are connected to the containment electronics (Data A and B) by separate multi-conductor cables. By employing two separate channels of information, the DRPI system can continue to function (at reduced accuracy) if one channel fails.

2.2 Current Technical Specification Requirement MPS3 LCO 3.1.3.2 requires the DRPI system and the Demand Position Indication system to be OPERABLE and capable of determining the control rod positions within +/-12 steps, in MODES 1 and 2. The Actions for TS 3.1.3.2 must be performed if a maximum of one DRPI per bank is inoperable. Per TS 3.1.3.2 Action a.1, the position of the nonindicating rod(s) must be determined indirectly by the movable incore detectors at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and immediately after any motion of the nonindicating rod which exceeds 24 steps in one direction since the last determination of the rod's position. As an alternative, per TS 3.1.3.2 Action a.2, THERMAL POWER can be reduced to less than 50% of RATED THERMAL POWER within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

MPS3 LCO 3.1.3.5 requires all shutdown rods to be limited in physical insertion as specified in the CORE OPERATING LIMITS REPORT (COLR),

in MODES 1 and 2 with Kett greater than or equal to 1. The Actions for TS 3.1.3.5 must be performed within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> if a maximum of one shutdown rod

  • is inserted beyond the insertion limits specified in the COLR (except for surveillance testing pursuant to TS 4.1.3.1.2). Per TS 3.1.3.5 Action a., the rod must be restored to within the limit specified in the COLR. As an alternative, per TS 3.1.3.5 Action b., the rod may be declared inoperable with TS 3.1.3.1 applied. The Surveillance Requirements (SRs) state that each shutdown rod shall be determined to be within the insertion limits specified in the COLR within 15 minutes prior to withdrawal of any rods in Control Bank A, B, C, or D during an approach to reactor criticality (SR 4.1.3.5.a), and at the frequency specified in the Surveillance Frequency Control Program (SR 4.1.3.5.b).

Serial No.21-005 Docket No. 50-423 Attachment 1, Page 4 of 15 MPS3 LCO 3.1.3.6 requires all control banks to be limited in physical insertion as specified in the COLR, in MODES 1 and 2.

MPS3 TS 6.9.1.6, Core Operating Limits Report, requires core operating limits to be established for each reload cycle and contains references to the approved analytical methods used to determine the core operating limits.

TS 6.9.1.6.a includes the list of specifications that require core operating limits to be established and documented in the CORE OPERATING LIMITS REPORT before each reload cycle or any remaining part of a reload cycle.

TS 6.9.1.6.b provides a list of documents that define the methods used to determine the core operating limits for MPS3.

2.3 Reason for the Proposed Change Change 1: Provide an Alternative to Frequent Verification of Rod Position Using the Movable lncore Detectors (TS 3.1.3.2)

If one or more DRPI are inoperable, MPS3 TS 3.1.3.2, Action a.1, requires verification of the position of the associated rods using the movable incore detector system once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. The proposed change revises TS 3.1.3.2 to provide an alternative to using the moveable incore detectors every 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> (approximately 90 times per month) by utilizing a different monitoring method based on occurrence of rod movement or power level change. This change will reduce the wear on the movable incore detector system. Wear of the movable incore detector system does not pose a reduction in the margin of safety, but excessive wear could result in a loss of functionality of the system. This could lead to the inability to complete SRs and a plant shutdown.

Change 2: Clarification of Shutdown Bank TS Requirements (TS 3.1.3.5)

MPS3 TS 3.1.3.5 is currently applicable to "shutdown rods," but TS 3.1.3.1 also addresses individual shutdown rod positions and deviations. The proposed change from shutdown "rods" to shutdown "banks" clarifies operability requirements for these rods and banks.

Change 3: Administrative Change to TSs for Control Banks (TS 3.1.3.6)

The proposed change revises the title for MPS3 TS 3.1.3.6 from "Control Rod Insertion Limits," to "Control Bank Insertion Limits." This change provides consistency with the current LCO requirements, Actions and SRs in TS 3.1.3.6.

Serial No.21-005 Docket No. 50-423 Attachment 1, Page 5 of 15 2.4 Description of Proposed Change The following is a detailed description of the proposed MPS3 TS changes (added text is shown below in bold type, deleted text is shown in strikethrough):

  • Newly proposed TS 3.1.3.2, Action a.2 is added as follows:

Verify the position of the nonindicating rod(s) indirectly using the movable incore detectors within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and once per 31 days thereafter, and within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> after discovery of each unintended rod movement, and within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> after each movement of the nonindicating rod(s) greater than 12 steps, and prior to THERMAL POWER exceeding 50% RATED THERMAL POWER, and within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> after reaching RA TED THERMAL POWER, or

  • Existing TS 3.1.3.2, Action a.2 is renumbered as Action a.3.
  • The title of TS 3.1.3.5 is revised to "SHUTDOWN RGQ. BANK INSERTION LIMITS."

AU-Each shutdown bank roa-s shall be limited in physical insertion as specified in the CORE OPERATING LIMITS REPORT (COLR).

With a maximum of one shutdown bank f9G inserted beyond the insertion limits specified in the COLR except for surveillance testing pursuant to Specification 4.1.3.1.2, 1.vithin 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> either:

a. Restore the bank f9G to within the limit specified in the COLR within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, or
b. Declare the rod to be inoperable and apply Specification 3.1.3.1 Be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

Each shutdown bank f9G shall be determined to be within the insertion limits specified in the COLR:

  • The title of TS 3.1 .3.6 is revised to "CONTROL RGQ. BANK INSERTION LIMITS."

Serial No.21-005 Docket No. 50-423 Attachment 1, Page 6 of 15

  • TS 6.9.1.6.a is revised as follows, with respect to specifications that require core operating limits to be established:

6.9 .1.6.a Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT before each reload cycle or any remaining part of a reload cycle for the following:

1. Reactor Core Safety Limit for Specification 2.1.1.
2. Overtemperature b. T and Overpower b. T setpoint parameters for Specification 2.2.1.
3. SHUTDOWN MARGIN for Specifications 3/4.1.1.1.1, 3/4 .1.1.1.2, and 3/4.1.1.2.
4. Moderator Temperature Coefficient BOL and EOL limits and 300 ppm surveillance limit for Specification 3/4.1.1.3.
5. Shutdown Rea Bank Insertion Limits for Specification 3/4.1.3.5.
6. Control Rea Bank Insertion Limits for Specification 3/4.1.3.6.
  • TS 6.9.1.6.b is revised as follows, with respect to specifications that utilize the methodology from Reference 1 (WCAP-9272-P-A) to establish core operating limits:
1. WCAP-9272-P-A, "WESTINGHOUSE RELOAD SAFETY EVALUATION METHODOLOGY," (W Proprietary). Methodology for Specifications:
  • 2.1.1 Reactor Core Safety Limits

Loops Not Filled

  • 3.1.1.3 Moderator Temperature Coefficient
  • 3.1.3.5 Shutdown Red-Bank Insertion Limits
  • 3.1.3 .6 Control Rea Bank Insertion Limits
  • TS 6.9.1.6.b is revised as follows, with respect to specifications that utilize the methodology from Reference 20 (VEP-FRD-42-A) to establish core operating limits:

Serial No.21-005 Docket No. 50-423 Attachment 1, Page 7 of 15

20. VEP-FRD-42-A, "Reload Nuclear Design Methodology."

Methodology for Specifications:

  • 2.1.1 Reactor Core Safety Limits

Loops Not Filled

  • 3.1 .1.3 Moderator Temperature Coefficient
  • 3.1.3.5 Shutdown Roo Bank Insertion Limits
  • 3.1 .3.6 Control Roo Bank Insertion Limits Attachments 2 and 3 provide the marked-up TS and TS Bases pages, respectively. The marked-up TS Bases pages are provided for information only. The TS Bases revision also replaces a reference to MPS3 TS 3.1.3.6 with MPS FSAR 4.3.2.2.6, which describes the sequencing of control banks with overlapping banks. The changes to the affected TS Bases page will be incorporated in accordance with the TS Bases Control Program when this amendment is approved .

3.0 TECHNICAL EVALUATION

Change 1: Provide an Alternative to Frequent Verification of Rod Position Using the Movable lncore Detectors (TS 3.1 .3.2)

MPS3 TS 3.1.3.2, proposed TS Action a.2 is consistent with TS 3.1.7, proposed Action A.2.1 in TSTF-547. The proposed action continues to use the movable incore detector system to monitor the position of the rod with the inoperable DRPI. The initial position of the rod is determined within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and every 31 days of full power operation thereafter.

The 8-hour initial completion time is the same as existing TS Action a.1 and the 31-day period coincides with the typical frequency of power distribution . SRs that utilize the movable incore detector system. If there is unintended movement of a rod or if a rod with an inoperable DRPI is moved more than 12 steps, the movable incore detectors are used to verify the rod position within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> . If there are changes in core power, which could result in changes in rod position, the rod position must be verified before exceeding 50%

Rated Thermal Power (RTP) and within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of reaching full power. This confirms the position of the rod with an inoperable DRPI to ensure that power distribution requirements are not violated and to establish a starting point for the proposed alternate monitoring actions.

An unintended rod movement is defined as the release of the rod's stationary gripper when no action was demanded , either manually or automatically, from the rod control

Serial No.21-005 Docket No. 50-423 Attachment 1, Page 8 of 15 system or a rod motion in a direction other than the direction demanded by the rod control system. Verifying that no unintended rod movement has occurred is performed by monitoring the rod control system stationary gripper coil current for indications of rod movement.

The ability to immediately detect a rod drop or misalignment is not directly provided by the movable incore detectors used in current TS Action a.1, or by the alternate monitoring method proposed in Actions a.2 and a.3. However, should there be a drop of a rod, it will typically be detectable by the excore power range detectors. Additionally, a negative reactivity insertion corresponding to the reactivity worth of the dropped rod may cause a change in core parameters, such AFD and QPTR.

Note that the proposed TS Actions provide an alternative to the existing rod position indication requirements. The rod group alignment limits and the bank insertion limits of LCO 3.1.3.1, LCO 3.1.3.5, and LCO 3.1.3.6 continue to require the rods to be operable and within the insertion limits.

The NRC staff provided a review of this change in the Safety Evaluation (Reference 3) for TSTF-547, Revision 1. In this Safety Evaluation, the NRC concluded that if the rod position indication is failed for an individual rod, its position is determined indirectly by use of the moveable incore detectors. The NRC staff determined that this change, which verifies rod position using the movable incore detectors based on the occurrence of events requiring rod motion, rather than determining position on a specified frequency, is acceptable because events requiring rod motion of the shutdown banks and control banks A, B, and C are relatively infrequent during steady state operation. Events involving significant movement of rods in control bank D are also relatively infrequent. The indirect determination of rod position is required after significant changes in power level or following substantial rod motion.

The NRC staff concluded that the addition of an alternative monitoring scheme to indirectly determine the position of rods associated with an inoperable DRPI is acceptable. TS 3.1.3.2, as modified, continues to specify the minimum performance level of equipment needed for safe operation of the facility as an LCO and continues to specify the appropriate remedial measures if the LCO is not met. The NRC staff found that the requirements of 10 CFR 50.36(c)(2) continue to be met because the minimum performa11ce level of equipment needed for safe operation of the facility is contained in the LCO and the appropriate remedial measures are specified if the LCO is not met.

It should be noted that DENG proposes to omit the requirement to "Restore inoperable

[D]RPI to OPERABLE status, "with a Completion Time of "Prior to entering MODE 2 from MODE 3," in proposed TS Action a.2, which was Proposed Action A.2.2 for TS 3.1.7 in TSTF-547 (which utilized the STS format). Previous submittals by other Licensees have concluded that this required Action was included in TSTF-547 in error. Because STS 3.1.7 Actions A.1 and A.3 permit continued operation for an unlimited period of time in the Applicability of STS 3.1. 7 in TSTF-547, STS LCO 3.0.4.a may be used to enter Mode 2 from Mode 3. As Actions A.1, A.2, and A.3 are joined by a logical OR, a licensee may

Serial No.21-005 Docket No. 50-423 Attachment 1, Page 9 of 15 choose to follow Action A.2 (which includes A.2.1 and A.2.2) after entering Mode 2.

TSTF-547 did not add a Note requiring the Action to be followed as an "otherwise stated" allowance in LCO 3.0.2, so Action A.2.2 does not apply in Mode 3 and is not restrictive after Mode 2 is entered. For all these reasons, proposed Action A.2 .2 in TSTF-547 is moot. Further, the requirement is not needed to protect plant safety. The NRC staffs Safety Evaluation for TSTF-547 (Reference 2) noted that the monitoring method in Action A.2.1 is more appropriate than the existing method in Action A.1. Therefore, its use should not be restricted. This variation has been previously approved in Salem Unit 1 &

2 Amendments 330/311 (Reference 7).

Change 2: Clarification of Shutdown Bank TS Requirements (TS 3.1.3.5)

TS 3.1.3.5 states "All shutdown rods shall be limited in physical insertion as specified in the COLR." Additionally, the title of TS 3.1.3.5 is "SHUTDOWN ROD INSERTION LIMIT."

This proposed change modifies the TS LCO, Actions, and Surveillances Requirements by referring to each bank instead of all rods.

The purpose of TS 3.1.3.5 is to ensure that sufficient negative reactivity is available to shut down the reactor and to maintain SOM. The proposed change is acceptable because TS 3.1.3.5 will continue to ensure that sufficient negative reactivity is available to shut down the reactor and to maintain the SOM. The proposed change is also appropriate because TS 3.1.3.1 continues to address individual shutdown rod positions and deviations. The proposed change is consistent with TS 3.1.3.6, which addresses control rod banks rather than single control rods (see administrative title correction in Change 3),

and TS 3.1.5, "Shutdown Bank Insertion Limits," in the STS for Westinghouse Plants, as provided in NUREG-1431, Revision 4 (Reference 3).

A revision to TS 3.1.3.5, Action b. is proposed to "Be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />." The TS 3.1.3.5 Actions require that within one hour, either the rod be restored to within the insertion limit specified in the COLR (Action a.) or declared inoperable and Specification 3.1 .3.1 applied (Action b.). With the change from shutdown "rods" to shutdown "banks," Action b. is no longer appropriate because TS 3.1.3.1 provides operability requirements for individual shutdown rods. The proposed change is consistent with LCO 3.0.3, and TS 3.1.5, Action B.1 in the STS for Westinghouse Plants (Reference 3).

Administrative changes are also made to TS 6.9.1.6 .a .and TS 6.9.1.6.b to reflect the revis.ed title of TS 3.1 .3.5 in the list of specifications that' utilize the methodologies from Reference 1 (WCAP-9272-P-A) and Reference 20 (VEP-FRD-42-A) to establish core operating limits.

Change 3: Administrative Change to TS for Control Banks (TS 3.1.3.6)

The current title for MPS3 TS 3.1.3.6 is "Control Rod Insertion Limits." However, the LCO states that the "The control banks (emphasis added) shall be limited in physical insertion as specified in the CORE OPERATING LIMITS REPORT (COLR)." Similarly, the Actions

Serial No.21-005 Docket No. 50-423 Attachment 1, Page 10 of 15 and SRs in TS 3.1.3.6 also apply to control "banks," rather than control "rods." Therefore, the proposed change revises the title for MPS3 TS 3.1.3.6 to "Control Bank Insertion Limits." This change is considered administrative in nature to provide consistency with the current LCO requirements, Actions and SRs in TS 3.1.3.6. Additional administrative changes are also made to TS 6.9.1.6.a and TS 6.9.1.6.b to reflect the revised title of TS 3.1.3.6 in the list of specifications that utilize the methodologies from Reference 1 (WCAP-9272-P-A) and Reference 20 (VEP-FRD-42-A) to establish core operating limits.

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements and Criteria

  • Section 182a of the Atomic Energy Act requires applicants for nuclear power plant operating licenses to include TS as part of the license. The TS ensure the operational capability of structures, systems, and components that are required to protect the health and safety of the public.
  • 10 CFR 50.90 requires NRC approval for any modification to, addition to, or deletion from the plant TSs. Therefore, this activity requires NRC approval prior to making the proposed plant-specific changes included in this license amendment request.
  • 10 CFR 50.36 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 third and fourth categories above, since changes to LCOs and surveillance requirements are proposed .
  • 10 CFR 50, Appendix A, GDC 26, "Reactivity control system redundancy and capability," states that control rods, preferably including a positive means for inserting the rods, 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.

Serial No.21-005 Docket No. 50-423 Attachment 1, Page 11 of 15

  • 10 CFR 50, Appendix A, GDC 28, "Reactivity limits," states 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 (1) 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.

4.2 Precedents Change 1 (Alternate Control Rod Position Monitoring) is consistent with a TS revision contained in TSTF-547, Revision 1 (Reference 1), which has been approved by the NRC (Reference 2). The change has also been approved for plants such as Prairie Island Nuclear Generating Plant, Units 1 and 2 (Reference 4), and Salem Nuclear Generating Station, Units 1 and 2 (Reference 7).

Change 2 (Clarification of Shutdown Bank TS Requirements) is consistent with a TS revision contained in the Westinghouse ITS (Reference 3), which has been approved by the NRC for plants such as Sequoyah Nuclear Plant, Units 1 and 2 (Reference 5), and Seabrook Station, Unit 1 (Reference 6).

4.3 No Significant Hazards Consideration Dominion Energy Nuclear Connecticut, Inc. (DENC) is submitting a license amendment request (LAR) to revise the Technical Specifications (TS) for Millstone Power Station Unit 3 (MPS3). The LAR proposes revisions to TS 3.1.3.2 to provide an alternative monitoring option for the condition where a maximum of one digital rod position indicator per bank is inoperable.

Specifically, as an alternative to determining the position of the non-indicating rod(s) indirectly by the movable incore detectors at a frequency of once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, the change would allow rod position verification to be performed based on the occurrence of rod movement or power level changes. This revision is consistent with Technical Specification Task Force Traveler 547, Revision 1 (ADAMS Accession No. ML15365A610), and would provide alternate TS Actions to allow the position of the rod to be monitored by a means other than movable incore detectors. The LAR proposes revisions to TS 3.1.3 .5 to replace shutdown "rods" with shutdown "banks," which is consistent with wording in the Standard Technical Specifications for Westinghouse Plants, as provided in NUREG-1431, Revision 4 (ADAMS Accession No. ML12100A222). The LAR also includes administrative changes to revise the title of TS 3.1 .3.6, to reflect that the requirements apply to control "banks," and modify TS 6.9.1.6.a and TS 6.9.1.6.b to cite the revised titles for TSs 3.1.3.5 and 3.1.3.6.

Serial No.21-005 Docket No. 50-423 Attachment 1, Page 12 of 15 OENC has evaluated whether a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

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

Response: No Control and shutdown rods are assumed to insert into the core to shut down the reactor in evaluated accidents. Rod insertion limits ensure that adequate negative reactivity is available to provide the assumed shutdown margin (SOM). Rod alignment and overlap limits maintain an appropriate power distribution and reactivity insertion profile.

Control and shutdown rods are initiators to several accidents previously evaluated, such as rod ejection. The proposed change does not change the limiting conditions for operation (LCO) for the rods or make any technical changes to the Surveillance Requirements (SRs) governing the rods .

Therefore, the proposed change has no significant effect on the probability of any accident previously evaluated.

Revising the TS Actions to provide an alternative to frequent use of the moveable incore detector system to verify the position of rods with inoperable rod position indicator does not change the requirement for the rods to be aligned and within the insertion limits . Therefore, the assumptions used in any accidents previously evaluated are unchanged and there is no significant increase in the consequences.

The proposed change from shutdown "rods" to shutdown "banks" in TS 3.1.3.5 clarifies operability requirements for these rods and banks. The purpose of TS 3.1.3.5 is to ensure that sufficient negative reactivity is available to shut down the. reactor and to maintain SOM. The proposed change will continue to ensure that sufficient negative reactivity is available to shut down the reactor and to maintain the SOM. The proposed change is also appropriate because TS 3.1.3.1 continues to address individual shutdown rod positions and deviations. TS 3.1.3 .5, Action b. is being revised to provide an appropriate action for a shutdown bank inserted beyond insertion limits, and the proposed action is consistent with LCO 3.0.3.

The proposed change to revise the title of TS 3.1.3.6 is administrative in nature to provide consistency with the current LCO requirements, Actions and SRs.

Serial No.21-005 Docket No. 50-423 Attachment 1, Page 13 of 15 Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated .

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

Response: No The proposed changes do not involve a physical alteration of the plant (i.e.,

no new or different type of equipment will be installed). The changes do not alter assumptions made in the safety analyses. The proposed changes do not make any technical changes to LCOs for the rods or SRs governing the rods. The proposed change to actions maintains or improves safety when equipment is inoperable and does not introduce new failure modes.

Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any previously evaluated.

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

Response: No The proposed change to allow an alternative method of verifying rod position in TS 3.1 .3.2 has no effect on the safety margin as actual rod position is not affected. The proposed change from shutdown "rods" to shutdown "banks" does not change the requirements of TS 3.1.3.5, which will continue to ensure that sufficient negative reactivity is available to shut down the reactor and to maintain the SOM . The proposed change to revise the title of TS 3.1 .3.6 is administrative in nature to provide consistency with the current LCO requirements, Actions and SRs. The proposed changes do not affect the margin of safety 'as the changes do not affect the ability of the rods to perform their specified safety function .

Therefore, the proposed changes do not involve a significant reduction in a margin of safety.

Based on the above information, DENC concludes that the proposed changes do not involve a significant hazards consideration, under standards set forth in 10 CFR 50.92(c), "Issuance of Amendment," and accordingly, a finding of "no significant hazards consideration" is justified.

Serial No.21-005 Docket No. 50-423 Attachment 1, Page 14 of 15 4.4 Conclusion Based on the above discussions, (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

S A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement.

However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or a significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or .

cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9).

Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

6.0 REFERENCES

1. TSTF Comments on Draft Model Safety Evaluation of Traveler TSTF-54 7, Revision 0, "Clarification of Rod Position Requirements," and Transmittal of TSTF-547, Revision 1, December 31, 2015 (ADAMS Accession No. ML15365A610).
2. Final Safety Evaluation (SE) of TSTF-547, Rev. 1, "Clarification of Rod Position Requirements," March 4, 2016 (ADAMS Accession No. ML15328A350).
3. NUREG-1431, Vol 1,1Rev 4, "Standard Technical Specifications: Westinghouse Plants- Specifications,:' April 30, 2012 (ADAMS Accession No. ML12100A222).
4. Letter from USNRC to Mr. Christopher P. Domingos (Northern States Power Company), "Prairie Island Nuclear Generating Plant, Units 1 and 2 - Issuance of Amendment Nos. 233 and 221 Re: Adoption of Technical Specifications Task Force (TSTF) Traveler TSTF-547, "Clarification of Rod Position Requirements" (EPID L-2019-LLA-0295)," November 18, 2020 (ADAMS Accession Number M L20283A342) .

Serial No.21-005 Docket No. 50-423 Attachment 1, Page 15 of 15

5. Letter from USNRC to Mr. Joseph W. Shea (Tennessee Valley Authority),

"Sequoyah Nuclear Plant, Units 1 and 2 - Issuance of Amendments for the Conversion to the Improved Technical Specifications with Beyond Scope Issues (TAC Nos. MF3128 and MF3129)," September 30, 2015 (ADAMS Accession Number ML15238B460).

6. Letter from USN RC to Mr. Don Moul (Florida Power & Light Company), "Seabrook Station, Unit No. 1 - Issuance of Amendment No. 162 Re: Revisions to Technical Specifications Associated with Movable Control Rods (EPID L-2018-LLA-0272),"

November 18, 2019 (ADAMS Accession Number ML19224A563).

7. Letter from USNRC to Mr. Eric Carr (PSEG Nuclear LLC), "Salem Nuclear Generating Station, Unit Nos. 1 and 2 - Issuance of Amendment Nos. 330 and 311 Re: Revise Technical Specifications to Adopt TSTF-547 (EPID L-2019-LLA-0018)," November 18, 2019 (ADAMS Accession Number ML19275D694).

Serial No.21-005 Docket No. 50-423 Attachment 2 MARKED-UP TECHNICAL SPECIFICATIONS PAGES Dominion Energy Nuclear Connecticut, Inc.

Millstone Power Station -Unit 3

Serial No.21-005 Docket No. 50-423 Attachment 2, Page 1 of 6 Febl'tlftfY 25, 291 *I REACTIVITY CONTROL SYSTEMS POSITION INDICATION SYSTEMS - OPERATING LIMITING CONDITION FOR OPERATION 3.1.3.2 TI1e Digital Rod Position Indication System and the Demand Position Indication System shall be OPERABLE and capable of determining the control rod positions within +/- 12 steps.

2. Verify the position of th e nonindlcating rod(s) Indirectly using the movable lncore detectors APPLICABILITY: MODES 1 and 2. within 8 hou rs and once per 31 days thereafter. and within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> afte r discovery of each unintended rod movement , an d within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> efler each movemen t of the nonindiceUng ACTION: rod(s) greater th an 12 steps. and prior to THERMAL POWER exceeding 50% RATED THER MAL POWER, and wl hin 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> after reach ing RATED THERMAL POWER , or
a. of one digital rod position indicator per bank inoperable:
1. De rmine the position of the nonindicating rod(s) indirectly by the ovable incore detectors at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and inimediately after any motion of the noni.ndicating rod which exceeds 24 steps in one direction since the last detem1ination of the rod 's position, or Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
b. With a maximum of one demand position indicator per bank inoperable:
1. Verify that all digital rod position indicators for the affected bank are OPERABLE and that the most withdrawn rod and the least withdrawn rod of the bank are within a maximum of 12 steps of each other at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, or
2. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.1.3.2. l Each digital rod position indicator shall be detemtined to be OPERABLE by verifying that the Demand Position Indication System and the Digital Rod Position Indication System agree within 12 steps at the frequency specified in the Surveillance Frequency Control Program except ,r during time intervals when the rod position deviation monitor is inoperable, then compare the Demand Position Indication System and the Digital Rod Position Indication System at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

4.1.3.2.2 Each of the above required digital rod position indicator(s) shall be determined to be OPERABLE by verifying that the digital rod position indicators agree with the demand position indicators within 12 steps when exercised over the foll-range of rod travel at the frequency L specified in the Surveillance Frequency Control Program. f MILLSTONE - UNIT 3 3/4 1-23 Amendment No. W, 60, ~ , ~ . 229,

~

Serial No.21-005 Docket No. 50-423 Attachment 2, Page 2 of 6 Febftlary 25, 2014 REACTIVITY C SHUTDOWN ~ INSERTION LIMITS LIMITING CONDITION FOR OPERATION 3.1.3.5

~~

AH shutdown f6$ shall be limited in physical insertion as specified in the CORE OPERATING LIMITS REPORT (COLR).

APPLICABILITY: MODES l* and2* **.

ACTION:

b.

in at least HOT STA NDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> 4.1.3.5 Each shutdown ree shall be detennined to be within the insertion limits specified in theCOLR:

a. Within 15 minutes prior to withdrawal of any rods in Control Bank A, B, C, or D during an approach to reactor criticality, and
b. At the frequency specified in the Surveillance Frequency Control Progran1.
  • See Special Test Exceptions Specifications 3.10.2 and 3.10.3.
    • With Keff greater than or equal to 1.

MILLSTONE - UNIT 3 3/4 1-26 Amendment No. 6G, 229-, ~

Serial No.21-005 Docket No. 50-423 Attachment 2, Page 3 of 6 Febfl:lftl) 25, 2014 LIMITING CONDITION FOR OPERATION 3.1.3.6 The control banks shall be limited in physical insertion as specified in the CORE OPERATING LIMITS REPORT (COLR).

APPLICABILITY: MODES 1* and 2* ** .

ACTION:

With the control banks inserted beyond the insertion limits specified in the COLR, except for surveillance testing pursuant to Specification 4.1.3.1.2:

a. Restore the contrnl banks to within the limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, or
b. Reduce THERMAL POWER within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to less than or equal to that fraction of RATED THERMAL POWER which is allowed by the bank position using the insertion limits specified in the COLR, or
c. Be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.1.3.6 The position of each control bank shall be determined to be within the insertion limits at the frequency specified in the Surveillance Frequency Control Program except during ti.me ,r intervals when the rod insertion limit monitor is inoperable, then verify the individual rod positions at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

  • See Special Test Exceptions Specifications 3.10.2 and 3.10.3.
    • With 3/4ff greater than or equal to 1.

MILLSTONE - UNIT 3 3/4 1-27 Amendment No. W, W, ~, ~

R:eiss1::1etl by NRG Lettef tltNetl 8epte1nhef 27, 2006

Serial No.21-005 Docket No. 50-423 Attachment 2, Page 4 of 6

-FOR INFORMATION ONLY-July 28, 2016 ADMINISTRATIVE CONTROLS MONTHLY OPERATING REPORTS 6.9.1.5 Deleted CORE OPERATING LIMITS REPORT 6.9. l.6. a Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT before each reload cycle or any remaining part of a reload cycle for the following:

l. Reactor Core Safety Limit for Specification 2.1.1.
2. Overtemperature Lff and Oveipower tff setpoint parameters for Specification 2.2.1.
3. SHUTDOWN MARGIN for Specifications 3/4.1.1.1.1, 3/4.1.1.1.2, and 3/4.1.1.2.
4. Moderator Temperature Coefficient BOL and EOL limits and 300 ppm surveillance limit for Specification 3/4.1.1.3.

Insertion

7. AXIAL FLUX DIFFERENCE Limits for Specification 3/4.2.1.1 .
8. Heat Flux Hot Chatmel Factor Limits for Specification 3/4.2.2.1.
9. RCS Total Flow Rate, Nuclear Enthalpy Rise Hot Chatuiel Factor, and Power Factor Multiplier for Specification 3/4.2.3.1.
10. DNB Parameters for Specification 3/4.2.5.
11. Shutdown Margin Monitor minimum count rate for Specification 3/4.3.5.
12. Boron Concentration for Specification 3/4.9. l. l.

MILLSTONE - UNIT 3 6-19a Amendment No.~, ?rt-, 69, 86, S&,

US, ~ . ~.'&.1/4, 268

Serial No.21-005 Docket No. 50-423 Attachment 2, Page 5 of 6

.htly 28, 2016 ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT (Cont.)

6.9.1.6.b 111e analytical methods used to detennine the core operating limits in Specification 6.9.1.6.a shall be those previously reviewed and approved by the NRC and identified below. 111e CORE OPERATING LIMITS REPORT will contain the complete identification for each of the TS referenced topical reports used to prepare the CORE OPERATING LIMITS REPORT (i.e.,

report number, title, revision, date, and any supplements).

I. WCAP-9272-P-A, "WESTINGHOUSE RELOAD SAFETY EVALUATION METHODOLOGY," r:J:!_ Proprietary). Methodology for Specifications:

  • 2.1.1 Reactor Core Safety Limits

.1.1.2

  • 3.1.3.6
  • 3.2.1.1 AXIAL FLUX DIFFERENCE
  • 3.2.2.1 Heat Flux Hot Channel Factor
  • 3.2.3.1 RCS Total Flow Rate, Nuclear Enthalpy Rise Hot Chatmel Factor
  • 3.9.1.1 REFUELING Boron Concentration
  • 3.2.5 DNB Parameters
2. Deleted
3. Deleted
4. WCAP-10216-P-A-RlA, "RELAXATION OF CONSTANT AXIAL OFFSET CONTROLFQ SURVEILLANCE TECHNICAL SPECIFICATION,"

r:J:!_ Proprietary). (Methodology for Specifications 3.2.1.1--AXIAL FLUX DIFFERENCE and 3.2.2.1--Heat Flux Hot Channel Factor)

5. WCAP-12945-P-A, "CODE QUALIFICATION DOCUMENT FOR BEST ESTIMATE LOCAANALYSIS," (W Proprietary). (Methodology for Specification 3.2.2.1--Heat Flux Hot Chatmel Factor.) ,r
6. WCAP-16009-P-A, "REALISTIC LARGE-BREAK LOCA EVALUATION METHODOLOGY USING THE AUTOMATED STATISTICAL TREATMENT OF UNCERTAINTY METHOD (ASTRUM)," (W Proprietary). (Methodology for Specification 3.2.2.1--Heat Flux Hot Channel Factor.) --r MILLSTONE - UNIT 3 6-20 Amendment No. ~ . ?r/-, 6G, 69, s-1-,

~ . +:7G, ~ . ~ . ~ . ,!-42., ~. ~

Serial No.21-005 Docket No. 50-423 Attachment 2, Page 6 of 6 Ally 28, 2916 ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT (Cont.)

19. WCAP-12610-P-A & CENPD-404-P-A, Addendum 1-A, "Optimized ZIRLOn.1,"

<Y:f.. Proprietary). (Methodology for Specification 3.2.2.1 - Heat Flux Hot Channel ,+--

Factor.)

20. VEP-FRD-42-A, "Reload Nuclear Design Methodology." Methodology for Specifications:
  • 2.1.1 Reactor Core Safety Limits

.1.2

  • 3.1.3.6
  • 3.2.2.1 Heat Flux Hot Channel Factor
  • 3.2.3.1 Nuclear Enthalpy Rise Hot Channel Factor
  • 3.9.1.1 REFUELING Boron Concentration
21. VEP-NE-1-A, "Relaxed Power Distribution Control Methodology and Associated FQ Surveillance Technical Specifications." Methodology for Specifications:
  • 3.2.1.1 AXIAL FLUX DIFFERENCE
  • 3.2.2.1 Heat Flux Hot Channel Factor
22. VEP-NE-2-A, "Statistical DNBR Evaluation Methodology." Methodology for Specifications:
  • 3.2.3.1 RCS Flow Rate, Nuclear Enthalpy Rise Hot Channel Factor
  • 3.2.5 DNB Paramete!}--
23. DOM-NAF-2-P-A, "Reactor Core Thenual-Hydraulics Using the VIPRE-D Computer Code," including Appendix C, "Qualification of the Westinghouse WRB-2M CHF Correlation in the Dominion VIPRE-D Computer Code," and Appendix D, "Qualification of the ABB-NV and WLOP CHF Correlations in the Dominion VIPRE-D Computer Code." Methodology for Specifications:
  • 3.2.3.1 RCS Flow Rate, Nuclear Enthalpy Rise Hot Channel .Factor
  • 3.2.5 DNB Parameters MILLSTONE - UNIT 3 6-20b Amendment No. 268

Serial No.21-005 Docket No. 50-423 Attachment 3 MARKED-UP TECHNICAL SPECIFICATIONS BASES PAGES (FOR INFORMATION ONLY)

Dominion Energy Nuclear Connecticut, Inc.

Millstone Power Station Unit 3

Serial No.21-005 Docket No. 50-423 Attachment 3, Page 1 of 2

!FOR INFORMATION ONLY!

Mil) 29, 2915 INDEX PAGES ARE LICENSEE CONTROLLED LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE DELETED ... ................. .... ... .. ............ .. .................... ..... ... ..... ........ .... ........ ........ ......3/4 1-24

~ Titne .... ............... ... .~ ************* *** *** *** ******** *** *******..... .................. ..... 3/41-25

~ n H:ee-Insertion Limit .. ................ .... .... ....... ......... ...... ....... ........................ 3/4 1-26 Control R:eti Insertion Limits ...................... ........ ...................................... ... .. .... .. ...3/4 1-27 3/4.2 POWER DISTRlBUTION LIMITS 3/4.2. l AXIAL FLUX DIFFERENCE..... ..... .. ............................................. .. ... ...... ...3/4 2-1 3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR - FQ(Z) ........................ ...... ......... 3/4 2-5 3/4.2.3 RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR .................................... .......... ................ ........ .............. 3/4 2-19 3/4.2.4 QUADRANT POWER TILT RATIO ..... ................. ........... ..... .. ... ................. 3/4 2-24 3/4.2.5 DNB PARAMETERS .................................. .................. ............... ................. 3/4 2-27 TABLE 3.2-1 DELETED ...... ...... ...... .. ........ ... ........ .... .................. ......... ... ... ............ ... ........... 3/4 2-28 3/4.3 INSTRUMENTATION 3/4.3. l REACTOR TRIP SYSTEM INSTRUMENTATION .......... ...... ................. ... 3/4 3-1 TABLE 3.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION ...................... .. ............ 3/4 3-2 TABLE 3.3-2 DELETED TABLE 4.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS .... ......... ........ ... ....... .............. .... .. ......... ........ .... ... ........ .... ..3/4 3-10 3/4.3.2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION .............. ............ ........ ................ ............ .......... .. ........ 3/4 3-15 TABLE 3.3-3 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION ........... ........ ....... ... ............. ...................................... ..3/4 3-17 TABLE 3.3-4 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS ....... .. .... ..... .............. .. ................. 3/4 3-26 MILLSTONE - UNIT 3 V Amendment No. -W, 6G, 89, 9+, W1,

~.2-l-8-, ~

Serial No.21-005 Docket No. 50-423 Attachment 3, Page 2 of 2

' FOR INFORMATION ONLy I LBDGR 15 MP3 005

-1/4t'il 21 , 2015 POWER DISTRIBUTION LIMITS BASES 3/4.2.2 AND 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR AND RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR TI1e limits on heat flux hot channel factor, RCS flow rate, and nuclear enthalpy rise hot channel factor ensure that: ( 1) the design limits on peak local power density and minimum DNBR are not exceeded and (2) in the event of a LOCA the peak fuel clad temperature will not exceed the 2200°F ECCS acceptance criteria limit.

Each of these is measurable but will nonually only be detennined periodically as specified in Specifications 4.2.2 and 4.2.3. This periodic surveillance is sufficient to ensure that the limits are maintained provided:

a. Control rods in a single group move together with no individual rod insertion differing by more than +/-12 steps, indicated, from the group demand position; b.

~ ~

Control ted grettf)S are sequenced with overlapping "gt'6tlf)9" as described in

..:§r>eeifieatiea 3.1.3 .6;

,-(F_SA_R_S_e_

cli_on_4-.3-.2-.2-.6--,f - / ' *- - *.- ~

c. TI1e control ~ sertion limits of Specifications 3.1.3.5 and 3.1.3.6 are maintained; and
d. TI1e axial power distribution, expressed in tenns of A,"{IAL FLUX DIFFERENCE, is maintained within the limits.

FNMI will be maintained within its lintlts provided Conditions a. through d. above are maintained. TI1e relaxation ofFNMI as a function of THERMAL POWER allows changes in the radial power shape for all permissible rod insertion limits.

TI1e FNMI as calculated in Specification 3.2.3.1 is used in the various accident analyses where FNMI influences parameters other than DNBR, e.g., peak clad temperature, and thus is the maximum "as measured" value allowed.

TI1e RCS total flow rate and FNMI are specified in the CORE OPERATING LIMITS REPORT (COLR) to provide operating and analysis flexibility from cycle to cycle. However, the minimum RCS flow rate, that is based on 10% steam generator tube plugging, is retained in the Technical Specifications.

MILLSTONE - UNIT 3 B-314 2-3 Amendment No. W, 69, i-1-7,

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