SBK-L-07042, License Amendment Request 07-01 Setpoint Change for Reactor Trip System Interlock P-9

License Amendment Request 07-01 Setpoint Change for Reactor Trip System Interlock P-9

ML070920139
Person / Time
Site:	Seabrook
Issue date:	03/29/2007
From:	St.Pierre G Florida Power & Light Energy Seabrook
To:	Document Control Desk, NRC/NRR/ADRO
References
SBK-L-07042
Download: ML070920139 (16)

Text

FPL Energy Seabrook Station FPlL EnergR PO. Box 300 Seabrook, NH 03874

.Seabrook Station (603) 773-7000 March 29, 2007 SBK-L-07042 Docket No. 50-443 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001 Seabrook Station License Amendment Request 07-01 "Setpoint Change for Reactor Trip System Interlock P-9" Pursuant to 10 CFR 50.90, FPL Energy Seabrook, LLC (FPL Energy Seabrook) has enclosed License Amendment Request (LAR) 07-01. This LAR proposes a change to the Seabrook Station Technical Specifications (TS) to increase the power level required for a reactor trip following a turbine trip (P-9 setpoint). The current TS requires a reactor trip when a turbine trip signal occurs and reactor power is greater than 20% rated thermal power (RTP). FPL Energy Seabrook proposes to change the P-9 setpoint to 45% RTP. The proposed change will decrease unnecessary challenges to the reactor protection system and increase plant availability when the cause of a turbine trip from low power can be promptly corrected.

As discussed in the enclosed LAR, the proposed change does not involve a significant hazard consideration pursuant to 10 CFR 50.92. A copy of this letter and the enclosed LAR has been forwarded to the New Hampshire State Liaison Officer pursuant to 10 CFR 50.91(b). FPL Energy Seabrook has determined that LAR 07-01 meets the criteria of 10 CFR 51.22(c)(9) for a categorical exclusion from the requirements for an Environmental Impact Statement. The Station Operation Review Committee and the Company Nuclear Review Board have reviewed this LAR.

Attachment 1 to this letter contains a notarized affidavit for this submittal, and attachment 2 provides FPL Energy Seabrook's evaluation of the proposed change. Attachment 3 provides a markup of the technical specifications showing the proposed change. Enclosed in attachments 4 and 5 are copies of LTR-SCS-07-14-P Attachment, January 2007, "Engineering Report Seabrook Nuclear Power Station Turbine Trip without Reactor Trip Transient from the P-9 Setpoint Analysis"(Proprietary) and LTR-SCS-07-14-NP Attachment, January 2007, "Engineering Report Seabrook Nuclear Power Station Turbine Trip without Reactor Trip Transient from the P-9 Setpoint Analysis" (Non-Proprietary), respectively. Also enclosed in attachment 6 is Westinghouse authorization letter CAW-07-2232 with accompanying affidavit, Proprietary Information Notice, and Copyright Notice.

Acc) an FPL Group company contains information proprietary to Westinghouse Electric Company LLC, and it is supported by an affidavit in attachment 6 signed by Westinghouse, the owner of the information. The affidavit sets forth the basis on which the information may be withheld from public disclosure by the Commission and addresses with specificity the considerations listed in paragraph (b) (4) of Section 2.390 of the Commission's regulations. Accordingly, it is respectfully requested that the information that is proprietary to Westinghouse be withheld from public disclosure in accordance with 10 CFR Section 2.390 of the Commission's regulations. Correspondence with respect to the copyright or proprietary aspects of the items listed above or the supporting Westinghouse affidavit should reference CAW-07-2232 and should be addressed to J. A. Gresham, Manager, Regulatory Compliance and Plant Licensing, Westinghouse Electric Company LLC, P.O. Box 355, Pittsburgh, Pennsylvania 15230-0355.

FPL Energy Seabrook requests NRC Staff review and approval of LAR 07-01 with issuance of a license amendment by March 31, 2008 and implementation of the amendment within 90 days. The requested approval date supports implementation of the amendment following startup from refueling outage 12, which is scheduled for the spring of 2008.

No commitments are made in this submittal. Should you have any questions regarding this letter, please contact Mr. James M. Peschel, Regulatory Programs Manager, at (603) 773-7194.

Very truly yours, FPL Energy Seabrook, LLC.

Gene St. Pierre Site Vice President Attachments:

1. Notarized Affidavit

2. Evaluation of the Proposed Change

3. Proposed Technical Specification Markup

4. Engineering Report Seabrook Nuclear Power Station Turbine Trip without Reactor Trip Transient from the P-9 Setpoint Analysis"(Proprietary)

5. Engineering Report Seabrook Nuclear Power Station Turbine Trip without Reactor Trip Transient from the P-9 Setpoint Analysis" (Non-Proprietary)

6. Application for Withholding Proprietary Information from Public Disclosure

cc: S. J. Collins, NRC Region I Administrator G. E. Miller, NRC Project Manager, Project Directorate 1-2 G.T. Dentel, NRC Senior Resident Inspector Mr. Christopher M. Pope, Director Homeland Security and Emergency Management New Hampshire Department of Safety Division of Homeland Security and Emergency Management Bureau of Emergency Management 33 Hazen Drive Concord, NH 03305

F ,PL ,Enrgy Seabrook Station AFFIDAVIT The following information is enclosed in support of this License Amendment Request:

• Attachment 2 Evaluation of the Proposed Change

• Attachment 3 Proposed Technical Specification Change Mark-up

• Attachment 4 Engineering Report Seabrook Nuclear Power Station Turbine Trip without Reactor Trip Transient from the P-9 Setpoint Analysis (Proprietary)

• Attachment 5 Engineering Report Seabrook Nuclear Power Station Turbine Trip without Reactor Trip Transient from the P-9 Setpoint Analysis (Non-Proprietary)

• Attachment 6 Application for Withholding Proprietary Information from Public Disclosure I, Gene F. St. Pierre, Site Vice President of FPL Energy Seabrook, LLC hereby affirm that the information and statements contained within this License Amendment Request are based on facts and circumstances which are true and accurate to the best of my knowledge and belief.

Sworn and Subscribed before me this

) day of*

'0', A /

ie St. Pierre Vice President

Attachment 2 Evaluation of the Proposed Change

Subject:

License Amendment Request 07-01, Setpoint Change for Reactor Trip System Interlock P-9

1. DESCRIPTION

2. PROPOSED CHANGE

3. BACKGROUND

4. TECHNICAL ANALYSIS

5. REGULATORY ANALYSIS 5.1 No Significant Hazards Consideration 5.2 Applicable Regulatory Requirements / Criteria

6. ENVIRONMENTAL CONSIDERATION

7. REFERENCES 1

1.0 DESCRIPTION

In LAR 07-01, FPL Energy Seabrook, LLC (FPL Energy Seabrook) requests an amendment to Facility Operating License NPF-86 for Seabrook Station. The proposed change modifies the Technical Specifications (TS) by increasing the setpoint for the reactor trip system interlock, P-9, from 20% rated thermal power (RTP) to 45% RTP. The proposed change will decrease the likelihood of unnecessary challenges to the reactor trip system and consequently, increase plant availability.

2.0 PROPOSED CHANGE

This change revises TS Table 2.2-1, Reactor Trip System Instrumentation Trip Setpoints; Functional Unit 18d, Power Range Neutron Flux, P-9. The current Trip Setpoint of

<20% of RTP is changed to <45% of RTP, and the Allowable Value is changed from

<22.1% of RTP to <45.3% of RTP. As a result, the anticipatory reactor trip on turbine trip will be blocked following turbine trips that occur at power levels less than or equal to 45% of RTP.

A change to the Bases for TS 2.2.1, Reactor Trip System Instrumentation Setpoints, to incorporate the revised P-9 setpoint will be required for this amendment. The change will be made in accordance with the Seabrook Station Technical Specification Bases Control Program.

3.0 BACKGROUND

3.1 System Description Reactor Trip System Interlock P-9 The P-9 interlock blocks a reactor trip following a turbine trip when the plant is below approximately 20 percent of full power. The turbine trip is sensed by either: (1) all turbine main stop valves closed or (2) low hydraulic fluid pressure in the turbine emergency trip system in two out of three pressure sensors. Blocking of the reactor trip following a turbine trip occurs when three out of four neutron flux power range signals are below the setpoint. Thus, below the P-9 setpoint, the reactor will be allowed to operate with the turbine tripped. When power increases to the P-9 setpoint, as sensed by two out of four of the neutron flux power range signals, the interlock enables the reactor trip following turbine trip.

Seabrook Station is designed with 50% load rejection capability. The three systems that act to mitigate this transient are the pressunizer pressure control system, rod control system, and steam dump control system, which are described in the Seabrook Station Updated Final Safety Analysis Report (UFSAR) chapter 7 (Reference 1).

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Pressurizer Pressure Control System The pressurizer pressure control system controls reactor coolant system pressure by using either the heaters (in the water region) or the spray (in the steam region) of the pressurizer plus steam relief for large transients. The electrical immersion heaters are located near the bottom of the pressurizer. The spray nozzles are located on the top of the pressurizer.

A portion of the heater group is proportionally controlled to correct small pressure variations. These variations are caused by heat losses, including heat losses due to a small continuous spray. The remaining (backup) heaters are turned on when the pressurizer pressure controller signal demands approximately 100 percent proportional heater power. Spray is initiated when the pressure controller spray demand signal is above a given setpoint. The spray rate increases proportionally with increasing spray demand signal until it reaches a maximum value. Steam condensed by the spray reduces the pressurizer pressure. In addition, the power-operated relief valves may prevent unnecessary challenges to the pressurizer safety valves during some positive pressure transients.

Rod Control System The rod control system enables the plant to follow load changes automatically, including the ability to accept a step load increase or decrease of 10 percent and ramp increases or decreases of 5 percent per minute within the load range of 15 percent to 100 percent power, without reactor trip, steam dump, or pressure relief (subject to possible xenon limitations). The system is also capable of restoring reactor coolant average temperature to within the programmed temperature dead band following a change in load.

The rod control system controls the reactor coolant average temperature by regulation of control rod bank position. An additional control input signal is derived from the reactor power versus turbine load mismatch signal. This additional control input signal improves system performance by enhancing response and reducing transient peaks.

Steam Dump Control System The steam dump system (SDS), in conjunction with the rod control system, enables the plant to accept a 50 percent loss of net load without tripping the reactor. The automatic SDS is able to accommodate this abnormal load rejection and to reduce the effects of the transient imposed upon the reactor coolant system. Bypassing main steam directly to the condenser maintains a steam load on the primary system. The rod control system, with the ability to accept a 10 percent step load decrease, can then reduce the reactor temperature to a new equilibrium value without causing overtemperature and/or overpressure conditions. The nominal steam dump design steam flow capacity is 40 percent of full load steam flow at full load steam pressure. If the difference between the reference Tavg based on turbine impulse chamber pressure and the lead/lag compensated average Tavg exceeds a preset value corresponding to a 10 percent step load decrease or a sustained ramp load decrease of 5 percent/minute, a demand signal will actuate the steam dump to maintain the reactor coolant system temperature within a control range until a new equilibrium condition is reached.

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3.2 Condition the ProposedChange is Intended to Resolve The proposed change raises the threshold power level above which the reactor will automatically trip in the event of a turbine trip from 20% RTP to 45% RTP. FPL Energy Seabrook expects that this change will result in a reduction in the likelihood of unnecessary reactor trips when operating at power levels below 45% RTP. Further, this change may increase plant availability when the cause of a turbine trip can be promptly corrected.

4.0 TECHNICAL ANALYSIS

4.1 Introduction Seabrook Station is a Westinghouse, four loop pressurized water reactor designed with 50% load rejection capability. The Westinghouse design criterion is that the design basis turbine load rejection should not actuate a reactor trip if all control systems function properly. A plant-specific analysis performed by Westinghouse Electric Company, LLC evaluated the impact of increasing the P-9 setpoint from 20% RTP to 45% RTP. This best-estimate analysis evaluated the impact of a turbine trip without reactor trip to assess the potential actuation of the pressurizer power operated relief valves (PORV). The analysis also included sensitivity studies to assess the affects of degraded control systems on the best estimate analysis results, an assessment of impact on the safety analyses for inadvertent operation of emergency core cooling system during power operation, and an evaluation of the non-LOCA design basis accident analyses. The report of the analysis is included in attachment 4, "Engineering Report Seabrook Nuclear Power Station Turbine Trip without Reactor Trip Transient from the P-9 Setpoint Analysis"(Proprietary).

4.2 Evaluation of P-9 Setpoint Increase The plant-specific analysis applied the initial conditions and assumptions in the current licensing basis for Seabrook. The methods and computer codes are the same as those used for Seabrook's stretch power uprate (Reference 2) and the P-9 analyses for other Westinghouse plants. These methods have been reviewed and approved by the NRC as part of the licensing submittal for those plants.

Analysis Inputs and Assumptions To assess the potential actuation of the pressurizer PORVs, simulated cases and conditions were selected in order to determine maximum pressurizer pressure after initiation of the transient. In support of this objective, the following key inputs and assumptions were used to ensure a conservatively high prediction of pressurizer pressure.

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" Turbine trip from initial primary and secondary side conditions of 45% of full power of 3678 MWt.

" Full power vessel average temperature, full power feedwater temperature, nominal steam pressure and 0% steam generator tube plugging.

" Steam dump capacity input based on the nominal steam pressure and 10% SG tube plugging and low Tavg.

• Best estimate beginning-of-life reactivity parameters (for conservative overall fuel to coolant heat transfer coefficients).

" Automatic rod control, pressurizer pressure control, steam dump control (loss of load controller) and steam generator level control (not explicitly modeled) functioning per design; (sensitivity studies assumed various degrees of degraded system operation).

" All pressurizer heaters functioning at total capacity and pressurizer level program varying linearly as a function of Tavg.

Analysis Approach, Methodology, Computer Codes and Results The methodology used to assess the Seabrook P-9 setpoint change is consistent with the established Westinghouse methodology that has been used successfully on other similar projects. The analyses and evaluations were performed in conformance with applicable Westinghouse and industry codes, standards, and regulatory requirements. Conservative inputs, assumptions and other analytical conditions were applied to assess the impact of the P-9 setpoint change on the operation of the systems and the pressurizer PORVs. In summary, the analysis was performed using:

" NRC-approved techniques; same as those used for the current Seabrook analyses as described in the Seabrook UFSAR.

" the same Westinghouse methodology and associated computer codes in licensing recent similar P-9 setpoint changes for other plants.

• currently approved techniques to demonstrate compliance with the licensing criteria and standards (no new methods have been introduced into the analysis and initial conditions and assumptions from the current Seabrook licensing basis are unchanged).

• the lumped loop version of the LOFTRAN computer code, which simulates the overall thermal-hydraulic and nuclear response of the nuclear steam supply system as well as the various control and protection systems.

" The same LOFTRAN model as developed for the Stretch Power Uprate program was used in this analysis. The steam dump valve model was refined to model all four banks explicitly and the initial conditions were revised to reflect the 45% power level.

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Following the Three Mile Island accident, the NRC expressed a concern regarding implementation of plant modifications that could increase the probability of a stuck-open PORV. NUREG-0737, Clarification of TMI Action Plan Requirements, (Reference 3),

Item II.K.3.10 establishes the NRC's position:

Position The anticipatorytrip modification proposed by some licensees to confine the range of use to high-power levels should not be made until it has been shown on a plant-by-plant basis that the probabilityof a small-break loss-of-coolant accident (LOCA) resultingfrom a stuck-open power-operatedreliefvalve (PORV) is substantiallyunaffected by the modification.

To address the NRC position, the plant-specific analysis demonstrates that the PORVs will not be challenged for a best estimate simulation (all control systems performing as designed) of a turbine trip without reactor trip with a P-9 setpoint of 45% RTP. The pressurizer PORV setpoint is 2400 psia, and the best estimate analysis resulted in a maximum pressurizer pressure of 2304 psia.

Additionally, a sensitivity study considered the effects of degraded control systems, although an analysis of degraded control systems is not required to satisfy the NRC position above. Consistent with the results of similar analyses for other four-loop Westinghouse plants, the Seabrook analysis of degraded control systems showed acceptable results (the PORVs were not challenged) for all scenarios evaluated except for the case involving a complete failure of all condenser steam dump valves. Failing all condenser steam dump valves results in challenging the PORVs and the steam generator (SG) safety valves. The 10% capacity of the atmospheric steam dump valves is insufficient to prevent the SG safety valves from opening, and steam will be released from the pressurizer PORV. However, overfilling of the pressurizer will not occur and this Condition 2 event will not initiate a Condition 3 event. An analysis of the case with all steam dumps failed determined the maximum power level at which the pressurizer PORVs and SG safety valves are not challenged. At 30% power, the pressurizer PORVs will not open but the SG safety valves will open. At 20% power, neither the PORVs nor the SG safety valves will open.

4.3 Summary of Technical Evaluation A turbine trip causes a reactor trip during operation above the P-9 setpoint. This reactor trip on turbine trip provides additional protection and conservatism beyond that required for protection of public health and safety and is included as good engineering practice and prudent design. However, the safety analyses in chapter 15 of the UFSAR do not take credit for this reactor trip. Increasing the P-9 setpoint will not challenge the pressurizer PORVs or SG safety valves for the best estimate scenario. This change has no impact on the safety analyses for inadvertent operation of the emergency core cooling system during power operation. Increasing the P-9 setpoint is acceptable with regard to the non-LOCA 6

design basis accidents. The consequences of loss of load / turbine trip events below 50%

power without the anticipatory reactor trip are bounded by the full power analyses presented in the Seabrook Station UFSAR, sections 15.2.3 and 15.3.2.

4.4 Precedent The NRC has previously issued amendments for similar submittals that requested an increase in the power level required for a reactor trip resulting from a turbine trip:

Donald C. Cook Units 1 and 2 Amendments 297 and 278 (Reference 4)

Byron / Braidwood Units 1 and 2 Amendments 13 and 3 (Reference 5)

Indian Point Unit 3 Amendment 192 (Reference 6)

North Anna Units 1 and 2 Amendments 119 and 103 (Reference 7)

Salem Units 1 and 2 Amendments 85 and 58 (Reference 8) 5.0 REGULATORY SAFETY ANALYSIS 5.1 No Significant Hazards Consideration

1. The proposed changes do not involve a significant increase in the probability or consequences of an accidentpreviously evaluated.

The analysis of the proposed change included an evaluation of loss of load / turbine trip transient. With systems functioning as designed, the proposed change to the P-9 setpoint does not impact accident analyses previously evaluated in the Updated Final Safety Analysis Report (UFSAR). In the best estimate case (normal plant conditions; all control systems functioning per design), the pressurizer power operated relief valves (PORV) and the steam generator safety valves are not challenged following the turbine trip without reactor trip. Consequently, the proposed change does not adversely affect the probability of a small break loss of coolant accident due to a stuck-open PORV. The sensitivity study that assessed the affects of degraded control systems found that a failure of all condenser steam dump valves resulted in challenging the PORVs and the steam generator (SG) safety valves. However, overfilling of the pressurizer will not occur and this Condition 2 event will not initiate a Condition 3 event. The challenge to the PORVs with all steam dump banks failed does not violate design or licensing criteria. Therefore, the proposed setpoint change does not significantly increase the probability or consequences of an accident previously evaluated.

2. The proposed changes do not create the possibility of a new or different kind of accidentfrom anypreviously evaluated.

The proposed setpoint change does not create the possibility of a new or different kind of accident than any accident previously evaluated in the FSAR. No 7

new accident scenarios, failure mechanisms or limiting single failures are introduced as a result of the proposed change. The proposed Technical Specification changes have no adverse effects on any safety-related system and do not challenge the performance or integrity of any safety-related system. The revised setpoint for the P-9 function ensures that accident/transient analyses acceptance criteria continue to be met. This change makes no modifications to the plant that would introduce new accident causal mechanisms and has no affect on how the trip functions operate upon actuation. Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any previously evaluated.

3. The proposedchanges do not involve a significant reduction in the margin of safety.

The proposed Technical Specification changes do not involve a significant reduction in a margin of safety. The analyses supporting the proposed change to the P-9 setpoint demonstrate that margin exists between the setpoint and the corresponding safety analysis limits. The calculations are based on plant instrumentation and calibration/functional test methods and include allowances associated with the setpoint change. The results of analyses and evaluations supporting the proposed change demonstrate acceptance criteria continue to be met.

The reactor trip on turbine trip provides additional protection and conservatism beyond that required for protection of public health and safety; the safety analyses in chapter 15 of the UFSAR do not take credit for this reactor trip. Therefore, the proposed changes do not involve a significant reduction in the margin of safety Based on the above, FPL Energy Seabrook concludes that the proposed amendment presents no 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.

5.2 Applicable Regulatory Requirements / Criteria 10 CFR 50.36(c)(2)(ii) (Reference 9) stipulates that a technical specification limiting condition for operation of a nuclear reactor must be established for each item meeting one or more of the following criteria:

(A) Criterion 1. Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

(B) Criterion2. A process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

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(C) Criterion 3. A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

(D) Criterion 4. A structure, system, or component which operating experience or probabilistic risk assessment has shown to be significant to public health and safety.

The requirement for a reactor trip on turbine trip at power levels greater than the P-9 setpoint is controlled by Technical Specification (TS) 3/4.3.1, Reactor Trip System Instrumentation, and TS 2.2.1, Reactor Trip System Instrumentation Setpoints. The reactor trip system instrumentation satisfies criterion 3 above. The reactor trip initiated by a turbine trip anticipates the loss of secondary heat removal resulting from a turbine trip. Tripping the reactor minimizes the pressure / temperature transient following a turbine trip when operating above the P-9 setpoint. The proposed change retains the P-9 function in the TS; therefore, the proposed change satisfies 10 CFR 50.36.

In conclusion, based on the considerations discussed previously, (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.

6.0 ENVIRONMENTAL CONSIDERATION

The proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent 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 of environmental assessment need be prepared in connection with the proposed amendment.

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7.0 REFERENCES

1. Seabrook Station Updated Final Safety Analysis Report, Chapter 7, section 7.7.

2. Seabrook Station Unit 1 - Issuance of Amendment RE: 5.2% Power Uprate (TAC NO. MC 23464), February 28, 2005.

3. NUREG 0737, Clarification of TMI Action Plan Requirements, November 1980.

4. Donald C. Cook Nuclear Plant, Units 1 and 2 (DCCNP-1 and DCCNP-2) - Issuance of Amendments Regarding Reactor Trip System Instrumentation (TAC NO.

MD0496 and MD 0497), October 30, 2006.

5. Byron Station Units 1 and 2, Amendment 13; Braidwood Station Unit 1, Amendment 3, December 8, 1987.

6. Indian Point Nuclear Generating Unit No. 3 - Issuance of Amendment RE: Reactor Trip of Turbine Trip (TAC NO. MA4696), September 8, 1999.

7. North Anna Units 1 and 2 - Issuance of Amendments RE: Direct Reactor Trip of Turbine Trip Blocked Below 30% of Rated Thermal Power (TAC Nos. 69800 and 69801), July 18, 1989.

8. Salem Nuclear Generating Station, Unit Nos. 1 an 2 - Technical Specification Changes - Trip Reduction P-7 Permissive to P-9 Permissive (TAC NO. 66039 and 66040), June 27, 1988.

9. 10 CFR 50.36, Technical Specifications.

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Attachment 3 Proposed Technical Specification Change (mark-up)

Refer to the attached markup of the proposed change to the Technical Specifications. The attached markup reflects the currently issued revision of the Technical Specifications. At the time of submittal, the Technical Specifications were revised through Amendment 113. Pending Technical Specifications or Technical Specification changes issued subsequent to this submittal are not reflected in the enclosed markup.

Listed below are the license amendment requests that are awaiting NRC approval and may impact the currently issued version of the Technical Specifications.

LAR Title FPL Energy Seabrook Date of SBK Letter No. Submittal NONE The following Technical Specifications are included in the attached markup:

Technical Specification Title Page Table 2.2-1 Reactor Trip System Instrumentation 2-6 Trip Setpoints

TABLE 2.2-1 (continued)

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS SENSOR TOTAL ERROR FUNCTIONAL UNIT ALLOWANCE (TA) Z S TRIP SETPOINT ALLOWABLE VALUE

18. Reactor Trip System Interlocks

a. Intermediate Range N.A N.A N.A >1 x 10-10 amp 6 x 10-11 amp Neutron Flux, P-6

b. Low Power Reactor Trips Block, P-7

1) P-10input N.A. N.A. N.A _<10% of RTP* <12.1% of RTP*

2) P-13 input N.A N.A. N. A <1 0% RTP* Turbine _<12.3% of RTP* Turbine Impulse Pressure Impulse Pressure Equivalent Equivalent

c. Power Range Neutron N.A N.A N.A <50% of RTP* <52.1 % of RJI.

Flux, P-8

d. Power Range Neutron N.A. N.A N.A 0/o of RTP* o of RTP*

Flux, P-9

e. Power Range Neutron N.A. N.A. >10% of RTP* >7.9% of RTP*

Flux, P-10

f. Turbine Impulse Chamber N.A N.A. N.A <1 0% RTP* Turbine <1 2.3% RTP* Turbine Pressure, P-13 Impulse Pressure Impulse Pressure Equivalent Equivalent

19. Reactor Trip Breakers N.A N.A N.A N.A N.A

20. Automatic Trip and Interlock N.A N.A. N.A N.A N.A Logic

• RTP = RATED THERMAL POWER SEABROOK - UNIT 1 2-6