SBK-L-16206, Submittal of Changes to Technical Specification Bases Dated 12/21/2016

Submittal of Changes to Technical Specification Bases Dated 12/21/2016

ML17005A277
Person / Time
Site:	Seabrook
Issue date:	12/21/2016
From:	Browne K NextEra Energy Seabrook
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
SBK-L-16206
Download: ML17005A277 (4)

Text

NEXTera ENERGY~

SEABROOK December 21, 2016 Docket No. 50-443 SBK-L-16206 U.S. Nuclear Regulatory Commission Attn.: Document Control Desk Washington, DC 20555-0001 Seabrook Station Submittal of Changes to the Seabrook Station Technical Specification Bases NextEra Energy Seabrook, LLC submits the enclosed changes to the Seabrook Station Technical Specification Bases. The changes were made in accordance with Technical Specification 6.7.6.j., "Technical Specification (TS) Bases Control Program." Please update the Technical Specification Bases as follows:

REMOVE INSERT B 3/4 1-4 B 3/4 1-4 B 3/4 9-2c B 3/4 9-2c Should you have any questions concerning this submittal, please contact me at (603) 773-7932.

Sincerely,

~brook,LLC Kenneth Browne Licensing Manager cc: D. Dorman, NRC Region I Administrator J. Poole, NRC Project Manager, Project Directorate 1-2 P. Cataldo, NRC Senior Resident Inspector f ()0 I

_Ne_x_tE_ra_E_n_er_gy_S_e_ab_r_oo_k,_L_LC~~~~~~~~~~~~~~~~~~~~~~~~~~-Jl.)f-f PO Box 300, Seabrook, NH 03874

Enclosure to SBK-L-16206

• REACTIVITY CONTROL SYSTEMS BASES 3/4.1.3 MOVABLE CONTROL ASSEMBLIES The specifications of this section ensure that: (1) acceptable power distribution limits are maintained, (2) the minimum SHUTDOWN MARGIN is maintained, and (3) the potential effects of rod misalignment on associated accident analyses are limited. OPERABILITY of the control rod position indicators is required to determine control rod positions and thereby ensure compliance with the control rod alignment and insertion limits. Verification that the Digital Rod Position Indicator agrees with the demanded position within+/- 12 steps at 24, 48, 120, and 228 steps withdrawn forthe Control Banks and 18, 210, and 228 steps withdrawn for the Shutdown Banks provides assurances that the Digital Rod Position Indicator is operating correctly over the full range of indication. Since the Digital Rod Position Indication System does not indicate the actual shutdown rod position between 18 steps and 210 steps, only points in the indicated ranges are picked for verification of agreement with demanded position.

The ACTION statements which permit limited variations from the basic requirements are accompanied by additional restrictions which ensure that the original design criteria are met. Misalignment of a rod requires measurement of peaking factors and a restriction in THERMAL POWER. These restrictions provide assurance of fuel rod integrity during continued operation. In addition, those safety analyses affected by a misaligned rod are reevaluated to confirm that the results remain valid during future operation.

The maximum rod drop time restriction is consistent with the assumed rod drop time used in the safety analyses. Measurement with rods at their individual mechanical fully withdrawn position, Tavg greater than or equal to 551°F and all reactor coolant pumps operating ensures that the measured drop times will be representative of insertion times experienced during a Reactor trip at operating conditions.

The fully withdrawn position of shutdown and control banks can be varied between 225 and the mechanical fully withdrawn position (up to 231 steps), inclusive. Westinghouse guidance allows a selected bank to be withdrawn one step beyond the mechanical full out position for an indicated 232 steps; however, to avoid misalignment of the rod control system bank overlap unit, control bank D step counters, the P/A converter, and MPCS position indication, withdrawal should be limited to 231 steps. The 225 to 231 step interval allows axial repositioning to minimize RCCA wear.

Control rod positions and OPERABILITY of the rod position indicators are required to be verified in accordance with the Surveillance Frequency Control Program with more frequent verifications required if an automatic monitoring channel is inoperable. These verification frequencies are adequate for assuring that the applicable LCOs are satisfied.

For Specification 3.1.3.1 ACTIONS b. and c., it is incumbent upon the plant to verify the trippability of the inoperable control rod(s). Trippability is defined in Attachment C to a letter dated December 21, 1984, from E. P. Rahe (Westinghouse) to C. 0. Thomas (NRC).

This may be by verification of a control system failure, usually electrical in nature, or that the failure is associated with the control rod stepping mechanism. In the event the plant is unable to verify the rod(s) trippability, it must be assumed to be untrippable and thus falls under the requirements of ACTION a. Assuming a controlled shutdown from 100% RATED THERMAL POWER, this allows approximately 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for this verification.

SEABROOK - UNIT 1 B 3/4 1-4 Amendment No. 8, 74, BC 14 05, 15-05

- 3/4.9 REFUELING OPERATIONS BASES 3/4.9.2 INSTRUMENTATION (Continued)

SURVEILLANCE SR4.9.2.a SR 4.9.2.a is the performance of a CHANNEL CHECK, which is a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that the two indication channels should be consistent with core conditions. Changes in fuel loading and core geometry can result in significant differences between source range channels, but each channel should be consistent with its local conditions. Operations procedures verify operation of the audible count rate function.

The surveillance frequency is controlled under the Surveillance Frequency Control Program.

SR4.9.2b.

SR 4.9.2.b is the performance of a CHANNEL CALIBRATION. This SR is modified by a Note stating that neutron detectors are excluded from the CHANNEL CALIBRATION.

The CHANNEL CALIBRATION for the source range neutron flux monitors consists of obtaining the detector plateau or preamp discriminator curves, evaluating those curves, and comparing the curves to the manufacturer's data. The surveillance frequency is controlled under the Surveillance Frequency Control Program.

REFERENCES

1. 10 CFR 50, Appendix A. GDC 13, GDCP 26, GDC 28, and GDC 29.

2. FSAR, Section 15.4.6 3/4.9.3 DECAY TIME The minimum requirement for reactor subcriticality prior to movement of irradiated fuel assemblies in the reactor vessel ensures that sufficient time has elapsed to allow the radioactive decay of the short-lived fission products. This decay time i~ consistent with the assumptions used in the safety analyses.

SEABROOK - UNIT 1 B 3/4 9-2c Amendment No. 93, BC 14 05, 15-06