Text

License Amendment Request 01-10 Relocation of Cycle-Specific Parameters to the Core Operating Limits Report
	ML021140320
Person / Time
Site:	Seabrook
Issue date:	04/15/2002
From:	Feigenbaum T; North Atlantic Energy Service Corp
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	NYN-02022
	Download: ML021140320 (72)
	v • d • e

vW, North Atlantic Energy Service Corporation P.O. Box 300 North Seabrook, NH 03874 Atlantic (603) 474-9521 The Northeast Utilities System April 15, 2002 Docket No. 50-443 NYN-02022 United States Nuclear Regulatory Commission Attention: Document Control Desk Washington, D.C. 20555-0001 Seabrook Station License Amendment Request 0 1-10 "Relocation of Cycle-Specific Parameters to the Core Operating Limits Report" North Atlantic Energy Service Corporation (North Atlantic) has enclosed herein License Amendment Request (LAR) 01-10. License Amendment Request 01-10 is submitted pursuant to the requirements of 10 CFR 50.90 and 10 CFR 50.4.

LAR 01-10 proposes changes to the Seabrook Station Technical Specifications (TS) to relocate boron concentration limits contained in certain TSs to the Core Operating Limits Report (COLR).

Boron concentration qualifies as a cycle-specific parameter limit that is cycle dependent.

The affected TS are: 3/4.1.1.1, "Boration Control Shutdown Margin - Tavg Greater Than or Equal To 200'F;" 3/4.1.1.2, "Boration Control Shutdown Margin -

Tavg Less Than or Equal To 200'F;"

3/4.1.2.5, "Borated Water Sources - Shutdown;" 3/4.1.2.6, "Borated Water Sources - Operating;"

3/4.1.2.7, "Isolation of Unborated Water Sources;" 3/4.5.1.1, "Accumulators;" 3/4.5.4, "Refueling Water Storage Tank;" and 3/4.9.1, "Boron Concentration."

In addition, LAR 01-10 proposes changes to TS 2.1, "Safety Limits," to relocate Figure 2.1-1, "Reactor Core Safety Limits-Four Loops in Operation," to the COLR; and revise TSs 2.1.1 and 2.1.2 limiting conditions and actions so as to be consistent with the improved Standard Technical Specifications (ITS) - Westinghouse Plants, NUREG-1431, Revision 2.

Also proposed is the relocation of DNB-related parameters, specified in TS 3/4.2.5, to the COLR.

As a result of the above changes, TS 6.8.1.6, "Core Operating Limits Report," and associated TS Bases must be revised to be reflective of the above changes.

Furthermore, editorial and administrative changes to TS 6.8.1.6, consistent with ITS, are proposed as well.

The proposed changes are either based on (1) NRC Generic Letter 88-16, "Removal of Cycle Specific Parameter Limits from Technical Specifications", dated October 3, 1988; (2) the NRC staffs acceptance of WCAP-14483-A, "Generic Methodology for Expanded Core Operating Limits Report;" and/or (3) the improved Standard Technical Specifications, NUREG-1431, Revision 2.

1ý_ 0 C:

U. S. Nuclear Regulatory Commission NYN-02022/Page 2 Relocation of cycle-specific parameters from TS to the COLR (a licensee-controlled document subject to the requirements of TS 6.8.1.6 and the provisions of 10 CFR 50.59) would afford North Atlantic flexibility to revise cycle-specific parameters, in accordance with NRC-approved methodologies, without the need for a license amendment. Specifically, TS 6.8.1.6c requires copies of the COLR to be submitted to the NRC for each reload cycle, including any mid-cycle revisions or supplements thereto. Thus resources, both North Atlantic and NRC, would be saved by minimizing and/or eliminating repetitive LAR submittals associated with revising cycle-specific parameters.

The Station Operation Review Committee and the Nuclear Safety Audit Review Committee have reviewed LAR 01-10.

As discussed in the enclosed LAR Section IV, 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). North Atlantic requests NRC Staff review of LAR 01-10, and issuance of a license amendment by April 15, 2003 (see Section V enclosed).

North Atlantic has determined that LAR 01-10 meets the criterion of 10 CFR 51.22(c)(9) for a categorical exclusion from the requirements for an Environmental Impact Statement (see Section VI enclosed).

Should you have any questions regarding this letter, please contact Mr. James M. Peschel, Manager - Regulatory Programs, at (603) 773-7194.

Very truly yours, NORTH ATLANTIC ENERGY SERVICE CORP.

Ted C. Feigenbayf"*

Executive Vice President and Chief Nuclear Officer cc:

H. J. Miller, NRC Regional Administrator R.D. Starkey, NRC Project Manager, Project Directorate 1-2 G. Dentel, NRC Senior Resident Inspector Mr. Donald Bliss, Director New Hampshire Office of Emergency Management State Office Park South 107 Pleasant Street Concord, NH 03301

North Atlantic SEABROOK STATION UNIT 1 North Atlantic Energy Amendment Request.

Amendment Request:

Section I Section II Section III Section IV Section V Section VI Service Corporation pursuant to 10 CFR 50.90 submits this License The following information is enclosed in support of this License Introduction and Safety Assessment for Proposed Change Markup of Proposed Change Retype of Proposed Change Determination of Significant Hazards for Proposed Change Proposed Schedule for License Amendment Issuance And Effectiveness Environmental Impact Assessment I, Ted C. Feigenbaum, Executive Vice President and Chief Nuclear Officer of North Atlantic Energy Service Corporation 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.

Ted C. Feigen um Sworn and Subscribed Executive Vice President before me this and Chief Nuclear Officer 15th day of April, 2002 Notary Public 2

Section I Introduction and Safety Assessment for the Proposed Change

I.

INTRODUCTION AND SAFETY ASSESSMENT OF THE PROPOSED CHANGE A.

Introduction License Amendment Request (LAR) 01-10 proposes changes to the Seabrook Station Technical Specifications (TS) to relocate boron concentration limits contained in certain TSs to the Core Operating Limits Report (COLR). Boron concentration qualifies as a cycle-specific parameter limit that is cycle dependent. The affected TS are: 3/4.1.1.1, "Boration Control Shutdown Margin - Tavg Greater Than or Equal To 200'F;" 3/4.1.1.2, "Boration Control Shutdown Margin - Tavg Less Than or Equal To 200'F;" 3/4.1.2.5, "Borated Water Sources - Shutdown;" 3/4.1.2.6, "Borated Water Sources - Operating;" 3/4.1.2.7, "Isolation of Unborated Water Sources;" 3/4.5.1.1, "Accumulators;"

3/4.5.4, "Refueling Water Storage Tank;" and 3/4.9.1, "Boron Concentration."

In addition, LAR 01-10 proposes changes to TS 2.1, "Safety Limits," to relocate Figure 2.1-1, "Reactor Core Safety Limits-Four Loops in Operation," to the COLR; and revise TSs 2.1.1 and 2.1.2 limiting conditions and actions so as to be consistent with the improved Standard Technical Specifications (ITS) -

Westinghouse Plants, NUREG-1431, Revision 2.

Also proposed is the relocation of DNB-related parameters, specified in TS 3/4.2.5, to the COLR.

As a result of the above changes, TS 6.8.1.6, Core Operating Limits Report, and associated TS Bases must be revised to be reflective of the above changes. Furthermore, editorial and administrative changes to TS 6.8.1.6, consistent with ITS, are proposed as well.

The proposed changes are either based on (1) NRC Generic Letter 88-16, "Removal of Cycle Specific Parameter Limits from Technical Specifications", dated October 3, 1988; (2) the NRC staff s acceptance of WCAP-14483-A, "Generic Methodology for Expanded Core Operating Limits Report;" and/or (3) the improved Standard Technical Specifications, NUREG-1431, Revision 2.

Relocation of cycle-specific parameters from TS to the COLR (a licensee-controlled document subject to the requirements of TS 6.8.1.6 and the provisions of 10 CFR 50.59) would afford North Atlantic flexibility to revise cycle-specific parameters, in accordance with NRC-approved methodologies, without the need for license amendment submittals.

Specifically, TS 6.8.1.6c requires copies of the COLR to be submitted to the NRC for each reload cycle, including any mid cycle revisions or supplements thereto. Thus resources, both North Atlantic and NRC, would be saved by minimizing and/or eliminating repetitive LAR submittals associated with revising cycle specific parameters.

B.

Safety Assessment NRC Generic Letter 88-16 provided guidance for the preparation of a license amendment request to provide an alternative to identifying cycle-specific parameter limits within Technical Specifications.

This alternative included three separate actions to modify the plant's Technical Specifications: (1) the addition of a definition of a named formal report that includes the values of cycle-specific parameter limits that have been established using an NRC-approved methodology that are consistent with all applicable limits of the safety analysis; (2) the addition of an administrative reporting requirement to submit the formal report on cycle-specific parameter limits to the Commission for information; and (3) the modification of individual Technical Specifications to note that the cycle specific parameters shall be maintained within the limits provided in the defined report.

Amendment 9 to the Seabrook Station Operating License, dated February 18, 1992, authorized North Atlantic to implement the guidance of Generic Letter 88-16.

Subsequently, other license amendments (e.g., Amendment 76, dated October 6, 2000) authorized relocation of certain additional cycle-specific parameters to the COLR. Relocation of boron concentration limits to the COLR conforms to the NRC methodology for expanding the COLR. Cycle-specific parameters, including borated water concentrations, are generated using NRC-approved methodologies.

These methodologies are listed in Seabrook Station Technical Specification 6.8.1.6. Through issuance of Generic Letter 88-16, the NRC has determined that such cycle-specific variables may be removed from Technical Specifications and placed in a licensee-controlled Core Operating Limits Report; thus obviating prior NRC review and approval to facilitate changes. The COLR ensures that changes to the boron concentration limits will continue to be performed in accordance with NRC-approved methodologies used to derive the parameters. Thus, it is proposed that the COLR will contain all the TS-relocated boron concentration values, i.e., those values currently contained in the Limiting Condition for Operation (LCO), Action, and Surveillance Requirements1. Relocation of the values will afford North Atlantic operational flexibility to make changes to these values in accordance with the NRC-approved methodologies, as controlled by TS 6.8.1.6, without requiring a license amendment every time boron concentration values are changed.

In addition to the above proposed changes, it is proposed that limiting condition a. of TS 3/4.9.1 be deleted because it is redundant to condition b. insofar as the boron concentration value derived for condition b. is based on achieving a keff of 0.95 or less. This is currently stated in the associated Bases. Therefore, North Atlantic believes the Bases is the appropriate place for stating the basis for the boron concentration value while in Mode 6. Deletion of the kIff value from the LCO would make this LCO consistent with ITS 3.9.1 which only specifies boron concentration.

The NRC Safety Evaluation to Topical Report WCAP-14483-A found relocation of DNB-related parameters (for TS 2.1.1, TS 3/4.2.5, and TS Figure 2.1-1) to the COLR, including the replacement of more specific requirements regarding the Safety Limits in TS 2.1.1 (i.e., the fuel DNB design basis and fuel centerline melt design basis), to be acceptable. Limitations on the combination of Thermal Power, pressurizer pressure and RCS temperature will be controlled by the COLR. The more specific Safety Limit requirements of fuel DNB design basis and fuel centerline melt design basis, as well as the minimum limit for Reactor Coolant System (RCS) total flow (based on that used in the safety analysis), will be controlled by the Technical Specifications.

With the incorporation of the more specific Safety Limits requirements in TS 2.1.1 North Atlantic is proposing to fashion TS 2.1.1 and 2.1.2 to be consistent with ITS 2.1 and 2.2, (NUREG-1431, Revision 2), including the associated Bases. In addition, actions associated with violation of Safety Limits will be consistent with ITS and will be numbered as new TS 2.1.3. Fashioning these TSs to be consistent with ITS would upgrade these TSs to current regulatory standards.

The action requirements remain essentially the same with exception of referral to TS 6.6 for reporting/action requirements which ITS no longer requires. TS Task Force Traveler TSTF 5, Revision 1, justified the elimination of reporting/action requirements from ITS based on redundancy with 10 CFR 50.36, 2

which specifies the reporting/action requirements for violation of Safety Limits.

As a result of relocating the cycle-specific parameters for the affected LCOs, Action, and Surveillance Requirements, minor rearrangement of wording has been done for operator clarity.

2 TS 6.6, "Safety Limit Violation," will be revised in a future administrative LAR submittal.

Surveillance Requirements for the aforementioned TSs will continue to ensure the verification of boron concentration and DNB-related parameters to be within the specified COLR limits.

The proposed changes to relocate cycle-specific Technical Specification parameter limits to the COLR will maintain adequate controls upon these parameters during normal plant operations and anticipated operational occurrences. The subject parameter limits will be administratively controlled in accordance with Technical Specification 6.8.1.6. Specifically, TS 6.8.1.6.c requires the COLR to be submitted to the NRC each reload cycle, including any mid-cycle revisions or supplements.

Proposed is an additional administrative change to TS 6.8.1.6a to delete the specific requirement for the COLR to be maintained available in the Control Room. North Atlantic regards this requirement as an unnecessary regulatory burden which is not consistent with ITS and does not meet the requirements of 10 CFR 50.36(c)(5) pertaining to administrative controls. As with other important 3

documents for control room personnel use, North Atlantic will ensure the proper location of the COLR.

In conclusion, the cycle-specific parameter limits controlled by the subject Specifications do not need to be included within the scope of the Technical Specifications.

The subject limits are adequately controlled by the COLR.

Relocation of such cycle-specific limits from Technical Specifications to the COLR are consistent with the Commission's position established by Generic Letter 88-16, the Safety Evaluation of WCAP-14483-A, and the improved TS.

Therefore, the proposed changes to the subject Technical Specifications, Index and Bases sections do not pose a significant hazard to the public health and safety.

3 A specific technical specification does not exist requiring the TS manual (Appendix A to the Operating License) itself to be located in the Control Room.

Section II Markup of Proposed Change The Attached markup reflects the currently issued revision of the Technical Specifications and Bases listed below.

Pending Technical Specifications or Technical Specification changes issued subsequent to this submittal are not reflected in the enclosed markup.

The following Technical Specifications and Bases are included in the attached markup:

Technical Specification Index Section 2.1 Specification 2.1 Figure 2.1-1 Specification 3/4.1.1.1 Specification 3/4.1.1.2 Specification 3/4.1.2.5 Specification 3/4.1.2.6 Specification 3/4.1.2.7 Specification 3/4.2.5 Specification 3/4.5.1.1 Specification 3/4.5.4 Specification 3/4.9.1 Specification 6.8.1.6.a & b Bases Specification 2.1.1 Bases Specification 2.1.2 Bases Specification 2.2.1 Bases Specification 3/4.1.1 Bases Specification 3/4.1.2 Bases Specification 3/4.2.5 Bases Specification 3/4.9.1 Title Safety Limits Safety Limits - Reactor Core and Reactor Coolant System Pressure Reactor Core Safety Limits Four Loops in Operation Boration Control Shutdown Margin Tavg Greater Than 200'F Boration Control Shutdown Margin Tavg Less Than 200'F Borated Water Sources - Shutdown Borated Water Sources - Operating Isolation of Unborated Water Sources DNB Parameters Accumulators Refueling Water Storage Tank Boron Concentration Core Operating Limits Report Reactor Core Reactor Coolant System Pressure Limiting safety System Settings (OPAT)

Boration Systems Boration Systems DNB Parameters Boron Concentration Page(s) ii 2-1 2-2 3/4 1-1 3/4 1-3 3/4 1-11 3/4 1-12 3/4 1-14 3/4 2-10 3/4 5-1 3/45-11 3/4.9-1 6-18, 6-18A, B & E B2-1 B 2-2 B 2-5 B 3/4 1-1 B 3/4 1-3 B 3/4 2-4 B 3/4 9-1 & 9-2

SECTION PAGE 2.1 SAFETY LIMITS(S L............................

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2-1 2.1.1 REACTOR CORE.4 $............................................

2-1 2.1.2 REACTOR COOLANT SYSTEM PRESSURESL.........................

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2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS...............

2-3 TABLE 2.2-1 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS....

2-4 2.0 BASES 2.1 SAFETY LIMITS LSLS) 2.1.1 REACTOR CORE*..A-S B 2-1 2.1.2 REACTOR COOLANT SYSTEM PRESSURES.L.........................

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2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS...............

B 2-3 3.0/4.0 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.0 APPLICABILITY...............................................

3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL Shutdown Margin - Tavg Greater Than 200OF................

Shutdown Margin - Tavg Less Than or Equal to 200OF.......

Moderator Temperature Coefficient........................

Minimum Temperature for Criticality......................

3/4 0-1 3/4 1-1 3/4 1-3 3/4 1-4.

3/4 1-6 SEABROOK - UNIT 1 ii c29*

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2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS (SLs) 2.1.1 REACTOR CORE SLs In MODES 1 and 2, the combination of THERMAL POWER, Reactor Coolant System (RCS) highest loop average temperature, and pressurizer pressure shall not exceed the limits specified in the COLR; and the following SLs shall not be exceeded:

2.1.1.1 The departure from nucleate boiling ratio (DNBR) shall be maintained greater than or equal to 1.17 for the WRB-IIWRB-2/WRB-2M DNB correlations.

2.1.1.2 The peak fuel centerline temperature shall be maintained less than 5080'F, decreasing by 58°F per 10,000 MWD/MTU of burnup.

2.1.2 REACTOR COOLANT SYSTEM PRESSURE SL In MODES 1, 2, 3, 4, and 5, the RCS pressure shall be maintained less than or equal to 2735 psig.

2.1.3 SAFETY LIMIT VIOLATIONS 2.1.3.1 If SL 2.1.1 is violated, restore compliance and be in MODE 3 within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

2.1.3.2 If SL 2.1.2 is violated:

a.

In MODE 1 or 2, restore compliance and be in MODE 3 within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

"b. In MODE 3, 4, or 5, restore compliance within 5 minutes.

SEABROOK - UNIT 1 2-1 Amendment No. "

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Amendment No. 33,4&-

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SEABROOK - UNIT 1 2-2

I 2.1 SAFETY LIMITS XEATh (7 BASES SEABROOK - UNIT 1 B 2-1

~~The restrictions o/~is Safety Limit prev..et overheating of tlfuel and possible\\

cladding perforation *ht would result in the lease of fission prpucts to the reactor coolant. Overhea V g of the fuel claddingi prevented by resring fuel operation to within the nuce e boiling regime whe the heat transfer efficient is large and the cladding surf temperature is siy above the coola saturation temperature.

0 eration above the up r boundary of the ucleate boiling regim could result in exces ye cladding tempera res because of th nset of departure nucleate boiling (D) ) and the resultant s rp, reduction in he transfer coefficient.*

NB is not a directly asurable parameter uring operation an, therefore, THERM POWER and reactor coolant temperature nd pressure have en related to DNB.

his relation has been developed to pre t

and e location of DNB r axially uniform and nonuniform he ux distributions.

e local DNB heat f ratio (DNBR) is de d as the ratio of e h t flux that would se DNB at a partic r core location to th ocal heat flux and is* dicative of the ma in to DNB.

he DNB design b is is as follows: un rtainties in the DN correlation, plant op ating parameters, clear and thermal rameters, fuel fab tion parameters, and mputer codes are nsidered statisti such that there is least a 95 percent probability with 95 rcent confidence vel that DNB will n occur on the most limitin fuel rod during ndition I and 1I :v ts. This establish a design DNBR value w must b met lant safety anal s using values of i ut parameters without uncertaintie In addition, mar ' has been maintad in the design by me ng safety analysis BR limits in pe ing safety analy

s.

he curves of Fi re 2.1-1 show the ci of points of THER L POWER, Reactor Co ant System pres re, and average mperature for which e minimum DNBR is no I s than the safe nalysis DNBR Ii value, or the aver eenthalpy at the vessel exi is equal to the e alpy of saturat iquid.

Thes curves are bas on values of the e alpy rise hot channel fac r F', at RATED T RMA POWE

, for the values spe led in the COLR. The v e of F N at reduced ower is assu d to vary accordin o the expression:

A FNH AH (RTP) [1+ 0.3

]

eFH (RTP is the value at RATED HERMAL POWER, and P is t fraction of RATED T RMAL POWER.

This expressi conservatively boun the cycle specific limits specified in chnical Specifi ion 3/4.2.3 and the C R. The Safety Limits in ure 2.1-1 are also based on a ref nce cosine axial pow shape with a peak of 1.

Amendment No. 33,;';00

SAFETY LIMITS BASES 2.1.1 ACTOR Cor-fContinued)

The resu ing heat flux onditions ar more* limitinrg than ose calculated f

the ran of all contro rods fully w.hdrawn-to the max rum allowable ontrol r insertion, as ming the axi power imbalance i within the limit of the AI) and A (A functions of e Overtemperatur and Overpower AT trips.

hen th axia power imbala e is not within e tolerance, the ial pow imbalance ef t on the Ove emperature AT a Overpower AT tri will re ce the setpoi s to provide rotection consi ent with core saf y limits r cycle speci ic power dlstr butlon.

"2 REA -RCOOLANT SY PRESUR" T e restriction this Safety mit protects th tntegrity of he Reactor Coola System (RCS) rom overpres ization and the y prevents e release of 0ra nuclides cont ned in the r ctor coolant fr reaching th containment osphere.

The-re tor vessel, essuri*zer,, and e RCS pipit, valves, and f tings are design to Section I of the ASNE de for Nucl r Power Plants which.

permits maximum tr 'a -ent pressure o 110% (2135 slg) of design essure.

-The Sa ty Limit of 635 psig is, th efore., con stent with the esign criteria and sociated Co requirements.

Theen '.e RCS is hydro sted at 125*(3110 psig) o design pressure o

demonstrat integrity prio a initial ration.

Amendment No..33

SEABROOK UNIT I B 2-2

2.1 SAFETY LIMITS (SLs)

BASES 2.1.1 Reactor Core SLs BACKGROUND GDC 10 (Ref. 1) requires that specified acceptable fuel during steady state operation, normal operational transients, ar occurrences (AOOs). This is accomplished by having a departL design basis, which corresponds to a 95% probability at a 95%

criterion) that DNB will not occur and by requiring that fuel cent melting temperature.

The restrictions of this SL prevent overheating of the fuE cladding perforation, that would result in the release of fission p Overheating of the fuel is prevented by maintaining the steady below the level at which fuel centerline melting occurs. Overhe*

prevented by restricting fuel operation to within the nucleate bo coefficient is large and the cladding surface temperature is sligi temperature.

Fuel centerline melting occurs when the local LHR, or pi is high enough to cause the fuel centerline temperature to reac Expansion of the pellet upon centerline melting may cause the point of failure, allowing an uncontrolled release of activity to th Operation above the boundary of the nucleate boiling re cladding temperature because of the onset of DNB and the res transfer coefficient. Inside the steam film, high cladding tempe water (zirconium water) reaction may take place. This chemicE in oxidation of the fuel cladding to a structurally weaker form. 1 integrity, resulting in an uncontrolled release of activity to the rE The proper functioning of the Reactor Protection Systen valves prevents violation of the reactor core SLs.

APPLICABLE SAFETY ANALYSES The fuel cladding must not sustain damage as a result c reactor core SLs are established to preclude violation of the fol

a. There must be at least 95% probability at a 95% confidencE the hot fuel rod in the core does not experience DNB and

b. The hot fuel pellet in the core must not experience centerlin The Reactor Trip System setpoints (Ref. 2), in combination prevent any anticipated combination of transient conditions for temperature, pressure, RCS Flow, Al, and THERMAL POWER departure from nucleate boiling ratio (DNBR) of less than the E existence of flow instabilities.

Automatic enforcement of these reactor core SLs is pro of the RPS and the steam generator safety valves.

design limits are not exceeded id anticipated operational ire from nucleate boiling (DNB) confidence level (the 95/95 DNB erline temperature stays below the I1 and cladding, as well as possible products to the reactor coolant.

state peak linear heat rate (LHR) ating of the fuel cladding is iling regime, where the heat transfer htly above the coolant saturation ower peaking, in a region of the fuel h the melting point of the fuel.

pellet to stress the cladding to the e reactor coolant.

gime could result in excessive ultant sharp reduction in heat ratures are reached, and a cladding Il reaction results

-his weaker form may lose its

actor coolant.

r (RPS) and steam generator safety

)f normal operation and AQOs. The lowing fuel design criteria:

e level (the 95/95 DNB criterion) that ie fuel melting.

with all the LCOs, are designed to Reactor Coolant System (RCS) level that would result in a

)NBR limit and preclude the ided by the appropriate operation Amendment No. 33,76, f7

. I SEABROOK UNIT 1 B 2-1

2.1 SAFETY LIMITS (SLs)

BASES 2.1.1 Reactor Core SLs (continued) 1 r_,_

u.

The SLs represent a design requirement for establishing the RPS trip setpoints identified previously. Specification 3/4.2.5, "DNB Parameters," or the assumed initial conditions of the safety analyses (as indicated in the UFSAR, Ref. 2) provide more restrictive limits to ensure that the SLs are not exceeded.

SAFETY LIMITS The figure provided in the COLR shows the loci of points of THERMAL POWER, RCS pressure, and average temperature for which the minimum DNBR is not less than the safety analyses limit, that fuel centerline temperature remains below melting, that the average enthalpy in the hot leg is less than or equal to the enthalpy of saturated liquid, or that the exit quality is within the limits defined by the DNBR correlation.

The reactor core SLs are established to preclude violation of the following fuel design criteria:

a. There must be at least a 95% probability at a 95% confidence level (the 95/95 DNB criterion) that the hot fuel rod in the core does not experience DNB and

b. There must be at least a 95% probability at a 95% confidence level that the hot fuel pellet in the core does not experience ceerline fuel melting.

The reactor core SLs are used to define th various RPS functions such that the above criteria are satisfied during steady state operationtnormal operational transients, and anticipated operational occurrences (AOOs). To ensure thatioe RPS precludes tranien an the above criteria, additional criteria are applied to the Overtemperature and Overpower AT reactor trip functions. That is, it must be demonstrated that the average enthalpy in the hot leg is less than or equal to the saturation enthalpy and the core exit quality is within the limits defined by the DNBR correlation. Appropriate functioning of the RPS ensures that for variations in the THERMAL POWER, RCS Pressure, RCS average temperature, RCS flow rate, and Al that the reactor core SLs will be satisfied during steady state operation, normal operational transients, and AOOs.

APPLICABILITY SL 2.1.1 only applies in MODES 1 and 2 because these are the only MODES in which the reactor is critical. Automatic protection functions are required to be OPERABLE during MODES 1 and 2 to ensure operation within the reactor core SLs. The steam generator safety valves or automatic protection actions serve to prevent RCS heatup to the reactor core SL conditions or to initiate a reactor trip function, which forces the unit into MODE 3. Setpoints for the reactor trip functions are specified in Specification 3/4.3.1, "Reactor Trip System (RTS) Instrumentation." In MODES 3, 4, 5, and 6, Applicability is not required since the reactor is not generating significant THERMAL POWER.

SEABROOK UNIT 1 B 2-2 Amendment No. 33-,

2.1 SAFETY LIMITS (SLs BASES 2.1.2 Reactor Coolant System (RCS) Pressure SL BACKGROUND The SL on RCS pressure protects the integrity of the RCS against overpressurization. In the event of fuel cladding failure, fission products are released into the reactor coolant. The RCS then serves as the primary barrier in preventing the release of fission products into the atmosphere. By establishing an upper limit on RCS pressure, the continued integrity of the RCS is ensured.

According to 10 CFR 50, Appendix A, GDC 14, "Reactor Coolant Pressure Boundary," and GDC 15, "Reactor Coolant System Design" (Ref. 3), the reactor pressure coolant boundary (RCPB)

"design conditions are not to be exceeded during normal operation and anticipated operational occurrences (AOOs). Also, in accordance with GDC 28, "Reactivity Limits" (Ref. 3), reactivity accidents, including rod ejection, do not result in damage to the RCPB greater than limited local yielding.

The design pressure of the RCS is 2500 psia. During normal operation and AOOs, RCS pressure is limited from exceeding the design pressure by more than 10%, in accordance with Section III of the ASME Code (Ref. 4). To ensure system integrity, all RCS components are hydrostatically tested at 125% of design pressure, according to the ASME Code requirements prior to initial operation when there is no fuel in the core. Following inception of unit operation, RCS components shall be pressure tested, in accordance with the requirements of ASME Code, Section Xl (Ref. 5).

Overpressurization of the RCS could result in a breach of the RCPB. If such a breach occurs in conjunction with a fuel cladding failure, fission products could enter the containment atmosphere, raising concerns relative to limits on radioactive releases specified in 10 CFR 100, "Reactor Site Criteria" (Ref. 6).

APPLICABLE SAFETY ANALYSES The RCS pressurizer safety valves, the main steam safety valves (MSSVs), and the reactor high pressure trip have settings established to ensure that the RCS pressure SL will not be exceeded.

The RCS pressurizer safety valves are sized to prevent system pressure from exceeding the design pressure by more than 10%, as specified in Section III of the ASME Code for Nuclear Power Plant Components (Ref. 3). The transient that establishes the required relief capacity, and hence valve size requirements and lift settings, is a complete loss of external load without a direct reactor trip. During the transient, no control actions are assumed, except that the safety valves on the secondary plant are assumed to open when the steam pressure reaches the secondary plant safety valve settings, and nominal feedwater supply is maintained.

The Reactor Trip System setpoints (Ref. 2), together with the settings of the MSSVs, provide pressure protection for normal operation and AQOs. The reactor high pressure trip setpoint is specifically set to provide protection against overpressurization (Ref. 2). The safety analyses for both the high pressure trip and the RCS pressurizer safety valves are performed using conservative assumptions relative to pressure control devices.

SEABROOK UNIT 1 B 2-2a Amendment No.

BASES

a*

2.1.2 Reactor Coolant System (RCS) Pressure SL (continued)

More specifically, no credit is taken for operation of any of the following:

a.

Pressurizer power operated relief valves (PORVs),

b.

Steam line relief valve,

c.

Steam Dump System,

d.

Reactor Control System,

e.

Pressurizer Level Control System, or

f.

Pressurizer spray valve.

SAFETY LIMITS The maximum transient pressure allowed in the RCS pressure vessel under the ASME Code,Section III, is 110% of design pressure. The maximum transient pressure allowed in the RCS piping, valves, and fittings under USAS, Section B31.1 (Ref. 7) is 120% of design pressure.

The most limiting of these two allowances is the 110% of design pressure; therefore, the SL on maximum allowable RCS pressure is 2735 psig.

APPLICABILITY SL 2.1.2 applies in MODES 1, 2, 3, 4, and 5 because this SL could be approached or exceeded in these MODES due to overpressurization events. The SL is not applicable in MODE 6 because the reactor vessel head closure bolts are not fully tightened, making it unlikely that the RCS can be pressurized.

2.1.3 SAFETY LIMIT VIOLATIONS The following SL violation responses are applicable:

If the reactor core SL 2.1.1 is violated, the requirement to go to MODE 3 places the unit in a MODE in which this SL is not applicable. The Allowed Outage Time (Completion Time) of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months recognizes the importance of bringing the unit to a MODE of operation where this SL is not applicable, and reduces the probability of fuel damage.

If the RCS pressure SL is violated when the reactor is in MODE 1 or 2, the requirement is to restore compliance and be in MODE 3 within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . Exceeding the RCS pressure SL may cause immediate RCS failure and create a potential for radioactive releases in excess of 10 CFR 100, "Reactor Site Criteria," limits (Ref. 6). The Allowed Outage Time (Completion Time) of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months recognizes the importance of reducing power level to a MODE of operation where the potential for challenges to safety systems is minimized.

SEABROOK UNIT 1 B 2-2b Amendment No.

2.1 SAFETY LIMITS (SLs)

BASES AFETY LIMIT VIOLATIONS (continued)~ 3i If the RCS pressure SL is exceeded in MODE 3, 4, or 5, RCS pressure must be restored to within the SL value within 5 minutes. Exceeding the RCS pressure SL in MODE 3, 4, or 5 is more severe than exceeding this SL in MODE 1 or 2, since the reactor vessel temperature may be lower and the vessel material, consequently, less ductile. As such, pressure must be reduced to less than the SL within 5 minutes. The action does not require reducing MODES, since this would require reducing temperature, which would compound the problem by adding thermal gradient stresses to the existing pressure stress.

REFERENCES

1. 10 CFR 50, Appendix A, GDC 10.

2. UFSAR, Chapte 15.

3. 10 CFR 50, Appendix A, GDC 14, GDC 15, and GDC 28.

4. ASME, Boiler and Pressure Vessel Code,Section III, Article NB-7000.

5. ASME, Boiler and Pressure Vessel Code,Section XI, Article IWX-5000.

6. 10CFR100.

7. USBS B31.1, Standard Code for Pressure Piping, American Society of Mechanical Engineers, 1967.

SEABROOK UNIT 1 B 2-2c Amendment No.

LIMITING SAFETY SYSTEM SETTINGS BASES 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS (Continued)

Intermediate and Source Range. Neutron Flux The Intermediate and.Source Range, Neutron*Flux trips provide core protection during reactor startup to mitigate the consequences of an uncon trolled rod cluster control assembly bank withdrawal-from a subcritical condition.

These trips provide redundant protection to the Low Setpoint trip of the Power Range, Neutron F Iux.channels.

The Source Range channels will initiate a Reactor trip at about 10 counts per second unless manually blocked when P-6 becomes active.

The Intermediate Range channels will initiate a Reactor trip at a current level equivalent to approximately 25% of RATED THERMAL POWER unless manually blocked when P-1O becomes active.

OvertemDerature AT The Overtemperature AT trip provides-core protection to prevent DNB for all combinations of pressure, power,.coolant temperature, and axial-power distribution, provided that. the transient is slow with respect to piping transit delays from the core to the temperatufe detectors (about 4 seconds), pressure is within the range between the Pressurizer High and Low Pressure trips and power is less than the Overpower AT-trip setpoint.

The Setpoint is automatically varied with: (1) coolant temperature.to correct for temperature-induced changes in density and heat capacity of water and includes dynamic compensation for piping delays from the core to the loop temperature detectors, (2) pressurizer pressure, and (3) axial power distribution.

With normal axial power distribution, this Reactor trip limit is always below the core Safety.Limit as shown in 2--

If axial peaks are greater than design, as indicated by the difference between top.and bottom power range nuclear detectors, the Reactor trip is automatically reduced according to the notations in Table 2.2-1.

Overpower AT Tf(' c The Overpower AT.trip provides assurance of fuel integrity (e.g., no fuel pellet melting and less than 1% cladding strain) under all possible overpower conditions, limits the required range for Overtemperature AT trip, and provides a backup to the High Neutron Flux trip.

The Setpoint is automatically varied with: (1) coolant temperature to correct for temperature induced changes in density and heat capacity of water, (2) rate of change of temperature for dynamic compensation for piping delays from the core to the loop temperature detectors, and (3) axial power distribution to ensure that the allowable heat generation rate (Kw/ft) is not exceeded.

The Overpower AT trip provides protection to mitigate the consequences of various size steam breaks as reported in WCAP-9226, "Reactor Core Response to Excessive Secondary Steam Releases."

SEABROOK - UNIT 1 B 2-5 Amendment No..*J/

3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL SHUTDOWN MARGIN - T avI GREATER THAN 200°F LIMITING CONDITION FOR OPERATION 3.1.1.1 The SHUTDOWN MARGIN for four-loop operation shall be greater than or equal to the limit specified in the CORE OPERATING LIMITS REPORT (COLR).

APPLICABILITY:

MODES 1, 2*, 3, and 4.

ACTION:

1E Po With the SHUTDOWN MARGIN less than the limiting value immediately initiate and continue boration en P eiaYto 30 gpmZ1 greater than or equalthe requi-i SHUTDOWN MARGIN is restored.

0AC 4

SURVEILLANCE REQUIREMENTS 4.1.1.1.1 The SHUTDOWN MARGIN shall be determined to be greater than or equal to the limiting value:

a.

Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after detection of an inoperable control rod(s) and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter while the rod(s) is inoperable.

If the inoperable control rod is immovable or untrippable, the above required SHUTDOWN MARGIN shall be verified acceptable with an increased allowance for the withdrawn worth of the immovable or untrippable control rod(s);

b.

When in MODE 1 or MODE 2 with keff greater than or equal to 1 at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by verifying that control bank withdrawal is within the limits of Specification 3.1.3.6;

c.

When in MODE 2 with keff less than 1, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months prior to achieving reactor criticality by verifying that the predicted critical control rod position is within the limits of Specification 3.1.3.6;

d.

Prior to initial operation above 5% RATED THERMAL POWER after each fuel loading, by consideration of the factors of Specifica tion 4.1.1.1.le. below, with the control banks at the maximum inser tion limit of Specification 3.1.3.6; and

See Special Test Exceptions Specification 3.10.1.

SEABROOK - UNIT 1 Amendment No. :/

3/41-:1.

REACTIVITY CONTROL SYSTEMS BORATION CONTROL SHUTDOWN MARGIN - Tavg LESS THAN OR EQUAL TO 200*F LIMITING CONDITION FOR OPERATION 3.1.1.2 The SHUTDOWN MARGIN shall be greater than or equal to the limit spe cified in the CORE OPERATING LIMITS REPORT (COLR).

Additionally, the Reactor Cgolant System boron concentration shall be greater than or equal Vt when the reactor coolant loops are in a drained condition.

APPLICABILITY:

MODE 5.

C OL M ACTION:

With the SHUTDOWN MARGIN less than the limit specified in the COLR or the Reac tor Coolant System boron concentration less than n immediately initiate and continue borationt 0 Aa-e cat&

reater than or equal to p"

or,-elb until required SHUTDOWN MARGIN and boron concentrati re restored.

SURVEILLANCE REqUIREMENTS o

t, 4.1.1.2 The SHUTDOWN MARGIN shall be determined to be greater than or equal to the limit specified in the COLR and the Reactor Coolant System boron con-centration shall be determined to be greater than or equal t,.

when the reactor coolant loops are in a drained condition:

L LVO 1-IE COL-R.

a.

Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after detection of an inoperable control rod(s) and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter while the rod(s) is inoperable.

If the inoperable control rod is immovable or untrippable, the SHUTDOWN MARGIN shall be verified acceptable with an increased allowance for the withdrawn worth of the immovable or untrippable control rod(s); and

b.

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by consideration of the following factors:

1)

Reactor Coolant System boron concentration,

2)

Control rod position,

3)

Reactor Coolant System average temperature,

4)

Fuel burnup based on gross thermal energy generation,

5)

Xenon concentration, and

6)

Samarium concentration.

SEABROOK - UNIT 1 Amendment No.,9/

3/4 1-3

REACTIVITY CONTROL SYSTEMS BORATION SYSTEMS BORATED WATER SOURCES -

SHUTDOWN LIMITING CONDITION FOR OPERATION 3.1.2.5 As a minimum, one of the following borated water sources shall be OPERABLE:

a.

A Boric Acid Storage System with:

Oft-

1)

A minimum contained borated water volume of 6,500 gallons,

2)

A minimum boron concentration and I

3)

A minimum solution temperature of 650F.

b.

The refueling water storage tank (RWST).with:

1)

A minimum contained borated water volume of 24,500 gallons,

2)

A minimum boron concentration and

3)

A. minimumsolution temperature of 50"F.

APPLICABILITY:

MODES 5 and 6.

ACTION:

With no borated water source OPERABLE, suspend all operations involving*CORE ALTERATIONS or positive reactivity changes.

SURVEILLANCE REQUIREMENTS 4.1.2.5 The above required borated water source shall be demonstrated OPERABLE:

"a. At least once per 7 days by:

1)

.. Verifying the boron concentration of the water,

2)

Verifying the contained borated water volume, and

3)

Verifying the boric acid storage tank solution temperature when it is the source of borated water.

b.

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying the RWST temperature.

FARBROOK -

UNIT 1 3/4 1-11 Amendment No.*

REACTIVITY CONTROL SYSTEMS BORATION SYSTEMS BORATED WATER SOURCES - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.6 As a minimum, the following borated water sources shall be OPERABLE as required by Specification 3.1.2.2:

a.

A Boric Acid Storage System with:

APPLICABIL 1),

A minimum contained borated water volume of 22,000 gallons, A minimum boron concentration -!ý---

and

3)

A minimum solution temperature of 65°F.

The refueling water storage tank'(RWST) with:

I)

A minimum contained borated water volume of 477,000 gallons,

2)

A boron concentration between.99 n-

3)

A minimum solution temperature of 500F, and

4)

A maximum solution temperature of 98°F.

ITY:

MODES 1, 2, 3 and 4.

ACTION:

a.

With the Boric Acid Storage System inoperable and being used as one of the above required borated water sources, restore the system to OPERABLE.status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and borated to-a SHUTDOWN MARGIN equivalent to at least the limit specified in the CORE OPERATING LIMITS REPORT (COLR) for the above MODES at 200"F; restore the Boric Acid Storage System to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b..

With the RWST inoperable, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at and in COLD SHUTDOWN withi restore the tank to OPERABLE least HOT STANDBY within the n the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

status next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months SEABROOK - UNIT 1 Amendment No. X 3/4 1-12

REACTIVITY CONTROL SYSTEMS BORON SYSTEMS ISOLATION OF UNBORATED WATER SOURCES -

SHUTDOWN LIMITING CONDITION FOR OPERATION 3.1.2.7 Provisions to isolate the Reactor Coolant System from unborated water sources shall be OPERABLE with:

a.

The Boron Thermal Regeneration System (BTRS) isolated from the Reactor Coolant System, and

b.

The Reactor Makeup Systems inoperable except for the capability of delivering up to the capacity of one Reactor Makeup Water pump to the Reactor Coolant System.

APPLICABILITY:

MODES 4, 5, and 6 ACTION:

0L With the requirements of the abov specification not satisfied immediately suspend all operations involving ORE ALTERATIONS or positive reactivity changes and, if within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months th required SHUTDOWN MAR IN is not verified, initiate and continue boration

--r to 30 gpm a-ngreater than or equal tou the required SHUTDOWN MARGIN is res ored and the isolation provisions are restored to OPERABLE.

-r-LL-L99 PEtcFt 1,%) TYE' Co L A, ?-i-F SURVEILLANCE REQUIREMENTS 4.1.2.7 The provisions to isolate the Reactor Coolant System from unborated water. sources shall be determined to be OPERABLE at least once per 31 days by:

a.

Verifying that at least the BTRS outlet valve, CS-V-302, or the BTRS moderating heat exchanger outlet valve, CS-V-305, is closed and locked closed, and

b.

Verifying that power is removed from at least one of the Reactor Makeup Water pumps,-RMW-P-16A or RMW-P-16B.

SEABROOK-UNIT 1 3/4 1-14

POWER DISTRIBUTION LIMITS 3/4.2.5 DNB PARAMETERS LIMITING CONDITION FOR OPERATION 3.2.5 The following DNB-related parameters sh I be maintained within the following limits:

(

-eS tHANý on-EFQL T-Tt*

.*.~L,.ImAr SPECItFtC-1-ý,03TH

a.

Reactor Coolant System T...<

F P,

b.

Pressurizer Pressu re>

C.

Reactor Coolant System Flow shall be:

a-op F U' To

1.

_> 382,800 gpm*; and, Co re A0-s

2.

> 392,800 gpm***

APPLICABILITY: MODE 1.

ACTION:

With any of the above parameters exceeding its limit, restore the parameter to within its limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE REQUIREMENTS 4.2.5.1 Each of the parameters shown above shall be verified to be within its limits at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

4.2.5.2 The RCS flow rate indicators shall be subjected to CHANNEL CALIBRATION at least once per 18 months.

4.2.5.3 The RCS total flow rate shall be determined by an approved method to be within its limit prior to operation above 95% of RATED THERMAL POWER after each fuel loading. The provisions of Specification 4.0.4 are not applicable for entry into MODE 1.

Limit not applicable during either a THERMAL POWER ramp in excess of 5% of RATED THERMAL POWER per minute or a7THERMAL POWER step in excess of 10%

of RATED THERMAL POWER.

- Thermal Design Flow. An allowance for measurement uncertainty shall be made when comparing measured flow to Thermal Design Flow.

- - Minimum measured flow used in the Revised Thermal Design Procedure.

SEABROOK - UNIT 1 3/42-10 Amendment No. 12, 33,./

3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS HOT' STANDBY.

STARTUP.

AND POWER OPERATION LIMITING CONDITION FOR OPERATION 3.5.1.1 Each Reactor Coolant System (RCS) accumulator shall be OPERABLE with:

a

a.

The isolation valve opefi and power removed, bb. A contained borated water volume of between 6121 and 6596 gallons,

c. A boron concentration of between2and

d.

A nitrogen cover-pressure of between 585 and 664 psig.

APPLICABILITY:

NODES 1,'2, and 3*.

ACTION:

a.

With one accumulator inoperable, except as a result of a closed isolation valve, restore the inoperable accumulator to OPERABLE status within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or be in at least HOT STANDBY within the next

.6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce pressurizer pressure to less than 1000 psig within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b.

With one accumulatorinoperable due to the lsolation valve being closed, either-immediately open the isolation valve or be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce pressurizer pressure to less than 1000 psig Within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

c.

With one pressure or water level channel inoperable per accumulator, return the inoperable channel to OPERABLE status within 30 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months '.'

d.

With two pressure channels or two water level channels inoperable per accumulator, immediately declare the affected accumulator(s) inoperable.

SURVEILLANCE REOUIREMENTS 4.5.1.1 Each accumulator shall be demonstrated OPERABLE:

a.

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by:

1)

Verifying the contained borated water volume and nitrogen

.cover-pressure in the tanks, and

Pressurizer pressure above 1000 psig.

SEABROOK - UNIT 1 3/4 5-1 Amendment No.

BORON INJECTION SYSTEM 3/4.5.4 REFUELING WATER STORAGE TANK LIMITING CONDITION FOR OPERATION

.3.5.4 The refueling water storage tank (RWST) shall be OPERABLE with:

a.

A minimum contained borated water volume of 477,000 gillons,

b.

A boron concentration between 2 Q--affd 29*

.m o

c.

A minimum solution temperature of '50"F, and

d.

A maximum solution temperature of 98°F.

H LI tTS SPEc4F APPLICABILITY:

MODES 1, 2, 3, and-4.

ACTION:

With the RWST inoperable, restore the tank to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.5.4 The RWST shall be demonstrated OPERABLE:

a.

At least once per 7 days by:

1)

Verifying the contained borated water volume in the tank, and

2)

Verifying the boron concentration of the water.

b.

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying the RWST temperature.

CADonnv_

"NITT 1 3/4 5-11 Amendment No.A vI v

3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION LIMITING CONDITION FOR OPERATION 3.9.1 The boron concentration of all filled portions of the Reactor Coolant System and the refueling canal shall be maintained uniform and sufficient to a

e er*

ri ye ter rV bIi of eter than oM DuE to" APPLICABILITY:

MODE 6.*

ACTION:

With the requirements of the above specification n(

suspend all operations involving CORE ALTERATIONS (

changes and initiate and continue boration(:a 0o-greater than or equal to 9 U fita n t il Mik~a concentration is restored SURVEILLANCE REQUIREMENTS to greater than or equal to L a-r i E eiD

- -*'*}--'

L$-- r c

c T'-

1~

b.

Withdrawal of any full-length control rod in exces.s of 3 feet from its fully inserted position within the reactor vessel.

4.9.1.2 The boron concentration of the Reactor Coolant System and the refueling canal shall be determined by chemical analysis at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

The reactor shall be maintained in MODE 6 whenever fuel is in the reactor vessel with the vessel head closure bolts less than fully tensioned or with the head removed.

SEABROOK - UNIT 1 3/4 9-1 1103

ADMINISTRATIVE CONTROLS MONTHLY OPERATING REPORTS 6.8.1.5 Routine reports of operating statistics and shutdown experience shall be submitted on monthly basis to the U.S. Nuclear Regulatory Commission, Washington, D.C. 20555, Attn: Document Control Desk, with a copy to the NRC Regional Administrator, no later than the 15th of each month following the calendar month covered by the report.

CORE OPERATING LIMITS REPORT 6.8.1.6.a Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT prior to each reload cycle, or prior to any remaining portion of a reload cycle, for the following:

1.

Cycle dependent Overpower AT and Overtemperature AT trip setpoint parameters and function modifiers for operation with skewed axial power profiles for Table 2.2-1 of Specification 2.2.1, a-A0 OA1e0+)

,2.3 SHUTDOWN MARGlN iiN'mitfor MODES 1, 2, 3, and 4 for Speccication 3.1.1 SHUTDOWN MARGINI imijfor MODE 5 for Specification 3.1.1.2, Moderator Temperature Coefficient BOL and EOL limits, and 300 ppm surveillance limit for Specification 3.1.1.3, OF -VVErLiWAL AfUc/

'rktke t4 r-or foCtAp-r 0~ PS 1.1, 7-3.. 2w

-- (-.,

j c*A-,'

a 0 A) Cff P, 3A,2A(ck-0 t)46 F6 P- \\

Amendment No. P, 22-, 34, W6-I SEABROOK - UNIT 1 6-18

ADMINISTRATIVE CONTROLS 6.8.1.6.a. (Continued) qj 5.

Shutdown Rod Insertion limit for Specification 3.1.3.5, R od Bank Insertion lim its for S pecification 3.1.3.6, R"({ AXIAL FLUX DIFFERENCE limits for Specification 3.2.1, (2.,f Heat Flux Hot Channel Factor, Fl' and K(Z) for Specification 3.2.2, Nuclear Enthalpy Rise Hot Channel Factor, and FRTH for Specification 3.2.3.

6.8.1.6.b The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC in:

1.

WCAP-1 0266-P-A, Rev. 2 with Addenda (Proprietary) and WCAP-1 1524 A, Rev. 2 with Addenda (Nonproprietary), 'The 1981 Version of the Westinghouse ECCS Evaluation Model Using the BASH Code", March, 1987.

Methodology for Specification:

3.2.2 Heat Flux Hot Channel Factor

2.

WCAP-1 0079-P-A, (Proprietary) and WCAP-1 0080-A (Nonproprietary),

"NOTRUMP: A Nodal Transient Small Break and General Network Code", August, 1985.

Methodology for Specification:

3.2.2 Heat Flux Hot Channel Factor

3.

YAEC-1 363-A, "CASMO-3G Validation," April, 1988.

YAEC-1659-A, "SIMU LATE-3 Validation and Verification,"

September, 1988.

WCAP-11596-P-A, (Proprietary), "Qualification of the PHOENIX-P/ANC Nuclear Design System for Pressurized Water Reactor Cores",

June, 1988.

WCAP-1 0965-P-A, (Proprietary), "ANC: A Westinghouse Advanced Nodal Computer Code", September, 1986.

SEABROOK - UNIT 1 6-18A Amendment No.

Q4-'76 Hue"-

C3.4.ATr4T30.J tj' R

M4SF 1, 2 1,

JbYR 1F c-dr-Ar

ADMINISTRATIVE CONTROLS 6.8.1.6.b. (Continued)

Methodology for Specifications:

3.1.1.1 SHUTDOWN MARGIN for MODES 1, 2, 3, and 4 3.1.1.2 SHUTDOWN MARGIN for MODE 5 3.1.1.3 Moderator Temperature Coefficient 3.1.3.5 Shutdown Rod Insertion Limit 3.1.3.6 Control Rod Insertion Limits 3.2.1 AXIAL FLUX DIFFERENCE 3.2.2 Heat Flux Hot Channel Factor 3.2.3 Nuclear Enthalpy Rise Hot Channel Factor

4.

Seabrook Station Updated Final Safety Analysis Report, Section 15.4.6, "Chemical and Volume Control System Malfunction That Results in a Decrease in the Boron Concentration in the Reactor Coolant System".

Methodology for Specifications:

3.1.1.1 SHUTDOWN MARGIN for MODES 1, 2, 3, and 4 3.1.1.2 SHUTDOWN MARGIN for MODE 5

5.

YAEC-1241, '"hermal-Hydraulic Analysis of PWR Fuel Elements Using the CHIC-KIN Code", R. E. Helfrich, March, 1981.

WCAP-14565-P, (Proprietary), '"VIPRE-01 Modeling and Qualification for Pressurized Water Reactor Non-LOCA Thermal-Hydraulic Safety Analysis",

April, 1997.

Letter from T. H. Essig (NRC) to H. Sepp (Westinghouse), "Acceptance for Referencing of Licensing Topical Report WCAP-14565-P, (Proprietary), 'VIPRE-01 Modeling and Qualification for Pressurized ter Reactor on-LOCA Thermal-Hydraulic Safety Analysis", January, 1999.

2.1 FE L i 'si 0

_Methodology for Specification:

3.2.1 AXIAL FLUX DIFFERENCE 3.2.2 Heat Flux Hot Channel Factor 3.2.3 Nuclear Enthalpy Rise Hot Channel Factor 3

2

-S'-

'?ARl ETCZ9

6.

YAEC-1 849P, "Thermal-Hydraulic Analysis Methodology Using VIPRE-01 For PWR Applications," October, 1992.

WCAP-1 1397-P-A, (Proprietary), "Revised Thermal Design Procedure", April, 1989.

ON

vfr Ro&.

~~C.rA~L 6464Tr FL(Xj i.)

1jp KCV7u wZJzJ~E~LJrý f"LOA(Pl~b, L~rý IW V-L--Jj VAOR G

LI_,ý.

z-&t4v SEABROOK - UNIT 1 6-18B Amendment No. ;,

ADMINISTRATIVE CONTROLS 6.8.1.6.b.

(Continued)

15.

WCAP-9272-P-A, (Proprietary), 'Westinghouse Reload Safety Evaluation Methodology", July, 1985.

e o ohLod01ogy or Specifications:

3.1.1.1 SHUTDOWN MARGIN for MODES 1,2,3, and 4 3.1.1.2 SHUTDOWN MARGIN for MODE 5 3.1.1.3 Moderator Temperature Coefficient 3.1.3.5 Shutdown Rod Insertion Limit 3.1.3.6 Control Rod Insertion Limits 3.2.1 AXIAL FLUX DIFFERENCE 3.2.2 Heat Flux Hot Channel Factor 3.2.3 Nuclear Enthalpy Rise Hot Channel Factor 6.8.1.6.c.

The core operating limits shall be determined so that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as SHUTDOWN MARGIN, and transient and accident analysis limits) of the safety analysis are met. The CORE OPERATING LIMITS REPORT for each reload cycle, including any mid cycle revisions or supplements thereto, shall be provided upon issuance, to the NRC Document Control Desk with copies to the Regional Administrator and the Resident Inspector.

2.,. r 3.gpv

/4v F-1 2'3s~b

2.

L Ro,4m

~

rFh foancaI ofTZ6.

Amendment No. 7"-

SEABROOK - UNIT I 6-18E

3/4.1 REACTIVITY CONTROL SYSTEMS BASES 3/4.1.1 BORATION CONTROL 3/4.1.1.1 and 3/4.1.1.2 SHUTDOWN MARGIN A sufficient SHUTDOWN MARGIN ensures that:

(1) the reactor can be made subcritical from all operating conditions, (2) the reactivity transients asso ciated with postulated accident conditions are controllable within acceptable limits, and (3) the reactor will be maintained sufficiently subcritical to preclude inadvertent criticality in the shutdown condition.

SHUTDOWN MARGIN requirements vary throughout core life as a function of fuel depletion, RCS boron concentration, and RCS Tavg.

The most restrictive condition occurs at EOL, with Tavg at no-load operating temperature, and is associated with a postulated steam line break accident and resulting uncon trolled RCS cooldown.

In the analysis of this accident, a minimum SHUTDOWN MARGIN as specified in the CORE OPERATING LIMITS REPORT (COLR) is required to control the reactivity transient.

Accordingly, the SHUTDOWN MARGIN requirement is based upon this limiting condition and is consistent with FSAR safety analysis assumptions.

With Tavg less than 2000 F, the reactivity transients resulting from a postulated steam line break cooldown are minimal.

A SHUTDOWN

-- H\\MARGIN as specified in the COLR and a boron concentration of greater than Sare required to permit sufficient time for the operator to terminate a inadvertent boron dilution event with T less than 2000 F.

avg rtFz 3/4.1.1.3 MODERATOR TEMPERATURE COEFFICIENT Vke O

The limitations on moderator temperature coefficient (MTC) are provided to ensure that the value of this coefficient remains within the limiting condition assumed in the FSAR accident and transient analyses.

The MTC values of this specification are applicable to a specific set of plant conditions; accordingly, verification of MTC values at conditions other than those explicitly stated will require extrapolation to those conditions in order to permit an accurate comparison.

The most negative MTC, value equivalent to the most positive moderator density coefficient (MDC),

was obtained by incrementally correcting the MDC used in the FSAR analyses to nominal operating conditions.

These corrections

-involved subtracting the incremental change in the MDC associated with a core condition of all rods inserted (most positive MDC) to an all rods withdrawn condition and, a conversion for the rate of change of moderator density with temperature at RATED THERMAL POWER conditions.

This value of the MDC was then transformed into the limiting end of cycle life (EOL)

MTC value as specified in the COLR.

The 300 ppm surveillance limit MTC value as specified in the COLR represents a conservative value (with corrections for burnup and soluble boron) at a core condition of 300 ppm equilibrium boron concentration and is obtained by making these corrections to the limiting MTC value as specified in the COLR.

SEABROOK - UNIT 1 B 3/4 1-1 Amendment No.

REACTIVITY CONTROL SYSTEMS

/>or

o'*Ar*

BASES q__.**,*'/.

3/4.1.2 BORATION SYSTEMS (Continued) boron canabilitv reauirement occurs at EOL fi

)rium xenon conc The limitation for a maximum of one centrifugal charging pump to be OPERABLE and the Surveillance Requirement to verify all charging pumps except the required OPERABLE pump to be inoperable in MODES 4, 5, and 6 except when the reactor vessel head closure bolts are fully detensioned or the vessel head is removed, provides assurance that a mass addition pressure transient can be relieved by operation of a single PORV or an RHR suction relief valve.

As a result of this, only one boron injection system is available. This is acceptable on the basis of the stable reactivity condition of the reactor, the emergency power supply requirement for the OPERABLE charging pump and the additional restrictions prohibiting CORE ALTERATIONS and positive reactivity changes in the event the single injection system becomes inoperable.

The boron capability required below 200°F is sufficient to provide a SHUTDOWN MARGIN as specified in the CORE OPERATING LIMITS REPORT after xenon decay and cooidown from'200° F to 1400°requires n

_req cootaied ume ga onspm p

bo water from oric acid age tan r a minim cntaine ume of 24 gallons of 2700 fo The contained water volume limits include allowance for water not available because of discharge line location and other physical characteristics.

The limits on contained water volume and boron concentration of the RWST also ensure a pH value of between 8.5 and 11.0 for the solution recirculated within containment after a LOCA. This pH band minimizes the evolution of iodine and minimizes the effect of chloride and caustic stress corrosion on mechanical systems and components.

The OPERABILITY of one Boron Injection System during REFUELING ensures that this system is available for reactivity control while in MODE 6.

The limitations on OPERABILITY of isolation provisions for the Boron Thermal Regeneration System and the Reactor Water Makeup System in Modes 4, 5, and 6 ensure that the boron dilution flow rates cannot exceed the value assumed in the transient analysis.

SEABROOK - UNIT 1 B 3/4 1-3 Amendment No. g,34, 42, 7/d L

I

INSERT

© The "equivalent to" statement in the Action is a provision providing an alternate method of emergency boration via the RWST at an increased flow rate to account for the lower boron concentration within the RWST.

INSERT The "equivalent to" statement in the Action for LCO 3.1.2.7 is a provision providing an alternate method of emergency boration via the RWST at an increased flow rate to account for the lower boron concentration within the RWST.

POWER DISTRIBUTION LIMITS BASES 3/4.2.5 DNB PARAMETERS The limits on the DNB-related parameters assure that each of the parameters is maintained within the normal steady-state envelope of operation assumed in the transient and accident analyses. The limits are consistent with the updated FSAR assumptions and have been analytically demonstrated adequate to assure compliance with acceptance criteria for each analyzed transient. Operating proceduresinclude aiowances for measurement and indication uncertainty so that the limits 4ý

°for Tavg andqp for pressurizer pressure are not exceeded.

I-i'We Cox,,

The 12-hour periodic surveillance of these parameters through instrument readout is sufficient to ensure that the parameters are restored within their limits following load changes and other expected transient operation.

The periodic surveillance of indicated RCS flow is sufficient to detect only flow degradation which could lead to operation outside the specified limit.

RCS flow must be greater than or equal to, 1) the Thermal Design Flow (TDF) with an allowance for measurement uncertainty and, 2) the minimum measured flow used in place of the TDF in the analysis of the DNB related events when the Revised Thermal Design Procedure (RTDP) methodology is utilized. Measurement of RCS total flow rate is performed by performance of either a precision calorimetric heat balance or normalized cold leg elbow tap AP measurements. RCS flow measurements using either the precision heat balance or the elbow tap AP measurement methods are to be performed at steady state conditions prior to operation above 95% rated thermal power (RTP) at the beginning of a new fuel cycle. The elbow tap RCS flow measurement methodology is described in WCAP-1 5404, "Justification of Elbow Taps for RCS Flow Verification at Seabrook Station", dated April 2000.

SEABROOK - UNIT 1 B 3/4 2-4 Amendment No. 9,12, 33, 3, 7,

3/4.9 REFUELING OPERATIONS BASES 3/4.9.1 BORON CONCENTRATION The limit on the boron concentrations of the Reactor Coolant System (RCS) and the refueling canal/cavity during refueling ensures that the reactor remains subcritical during MODE 6. During refueling, the spent fuel pool water volumes and the reactor cavity water volumes will be connected when the fuel transfer gate valve is open. This configuration allows the bodies of water to be physically capable of being in contact, however, no effective mixing of the volumes occurs due to the constriction of the fuel transfer tube. The soluble boron concentration in each of these volumes is maintained greater than or equal to or equivalent to a Ke less than or equal to 0.95 when the fuel transfer gate is ope.

However, the spent fuel pool water boron O

concentration is under administrativ econtrols and not a technical specification. They are independently maintained at the ppropriate boron concentration even though no

,v intermixing of significance exists. The mixing caused by the RHR pumps (reactor cavity) or the SFP system pumps assures niformity of boron in the separate volumes.

The soluble boron concentration oets e

e core reactivity and is measured by chemical analysis of a representative sample of the coolant in each of the volumes. Plant procedures ensure the specified boron concentration in order to maintain an overall core reactivity of kef- *0.95 during fuel handling, with control rods and fuel assemblies assumed to be in the most adverse configuration (least negative reactivity) allowed by plant procedures.

GDC 26 of 10 CFR 50, Appendix A, requires that two independent reactivity control systems of different design principles be provided. One of these systems must be capable of holding the reactor core subcritical under cold conditions. The Chemical and Volume Control System (CVCS) is the system capable of maintaining the reactor subcritical in cold conditions by maintaining the boron concentration.

The reactor is brought to shutdown conditions before beginning operations to open the reactor vessel for refueling. After the RCS is cooled and depressurized and the vessel head is unbolted, the head is slowly removed to form the refueling cavity. The refueling canal and the refueling cavity are then flooded with borated water from the refueling water storage tank through the open reactor vessel by gravity feeding or by the use of the Residual Heat Removal (RHR) System pumps.

The pumping action of the RHR System in the RCS and the natural circulation due to thermal driving heads in the reactor vessel and refueling cavity mix the added solution of boric acid with the water in the refueling canal. The RHR System is in operation during refueling (see LCO 3.9.8.1, "Residual Heat Removal (RHR) and Coolant Circulation High Water Level," and LCO 3.9.8.2, "Residual Heat Removal (RHR) and Coolant Circulation -

Low Water Level") to provide forced circulation in the RCS and assist in maintaining the boron concentrations in the RCS and the refueling canal/cavity at or above the limit specified in LCO 3.9.1.

SEABROOK - UNIT 1 B 3/4 9-1

/

?1:Z1-o. 00-03/U4 CRevised by NRC letter dat

. 3/4.9 REFUELING OPERATIONS BASES During refueling operations, the reactivity condition of the core is consistent with the initial conditions assumed for the boron dilution accident in the accident analysis and is conservative for MODE 6. The boron concentration limit specified is based on the core reactivity at the beginning of each fuel cycle (the end of refueling) and includes an uncertainty allowance.

The required boron concentration and the plant refueling procedures that verify the correct fuel loading plan (including full core mapping) ensure that the Iy of the core will remain <0.95 during the refueling operation. Hence, at least 5% Ak/k margin of safety is established during refueling.

Continuation of CORE ALTERATIONS or positive reactivity additions (including actions to reduce boron concentration) is contingent upon maintaining the unit in compliance with the LCO.

During refueling operations water may be transferred to the refueling canal/cavity or the RCS from different sources. Transfers or additions of water whose boron concentration exceeds the required refueling boron concentration are acceptable.

Transfers or additions of water where the boron concentration is less than the required refueling boron concentration may be made, provided that these additions are administratively controlled to ensure that the refueling boron concentration requirements continue to be met. That is, the final concentration of boron in the total volume, after the addition of water less than the required refueling boron concentration, exceeds the required refueling boron concentration, or k*,< 0.95. Also, these administrative controls ensure such transfers or additions of water will not substantially reduce the uniformity of boron concentration in the RCS or refueling canal.

Likewise, transferring water to the RCS or the refueling canal/cavity that is lower in temperature (down to the operability requirements of the RWST in MODE 6; 50 DEG F) than the water contained in those volumes is also acceptable. These minimum requirements for boron concentration and water temperature are also applicable to other MODE 6 Technical Specification ACTIONS that limit operations involving positive reactivity additions to ensure that the reactor remains subcritical and an adequate shutdown margin is maintained.

Suspension of CORE ALTERATIONS and positive reactivity additions shall not preclude moving a component to a safe position. In addition to immediately suspending CORE ALTERATIONS or positive reactivity additions, boration to restore the concentration must be initiated immediately.

In determining the required combination of boration flow rate and concentration, no unique Design Basis Event must be satisfied. The only requirement is to restore the boron concentration to its required value as soon as possible. In order to raise the boron concentration as soon as possible, the operator should begin boration with the best source available for unit conditions.

Once actions have been initiated, they must be continued until the boron concentration is restored. The restoration time depends on the amount of boron that must be injected to reach the required concentration.

SEABROOK - UNIT 1 B 3/4 9-2 C

00-03 Zevi RC te ted 6/8/01

Section III Retype of Proposed Change The attached retype reflects the currently issued version of the Technical Specifications. Pending Technical Specification changes or Technical Specification changes issued subsequent to this submittal are not reflected in the enclosed retype.

The enclosed retype should be checked for continuity with Technical Specifications prior to issuance.

INDEX 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS SECTION PAGE 2.1 S A FETY LIM IT S (S L.......................................................................................

2.1.1 R EA CTO R C O R E S Ls.....................................................................................

2.1.2 REACTOR COOLANT SYSTEM PRESSURE SL............................................

2.1.3 SAFETY LIM IT V IO LATIO NS...........................................................................

FIGURE 2.1-1 (THIS FIGURE IS NOT USED)..................................................................

2.2 LIMITING SAFETY SYSTEM SETTINGS 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS......................

TABLE 2.2-1 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS..

BASES 2-1 2-1 2-1 2-1 2-2 2-3 2-4 2.1 SAFETY LIMITS (SL) 2.1.1 R EA CTO R C O R E S Ls.....................................................................................

B 2-1 2.1.2 REACTOR COOLANT SYSTEM PRESSURE SL............................................

B 2-2a 2.1.3 SAFETY LIM IT VIO LATIONS...........................................................................

B 2-2b 2.2 LIMITING SAFETY SYSTEM SETTINGS 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS......................

B 2-3 3.0/4.0 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.0 A P P LIC A B ILIT Y..............................................................................................

3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL Shutdown Margin - Ta,,g Greater Than 200OF...................................................

Shutdown Margin - Tavg Less Than or Equal to 200OF......................................

Moderator Temperature Coefficient............................................................

Minimum Temperature for Criticality................................................................

3/4 0-1 3/4 1-1 3/4 1-3 3/4 1-4 3/4 1-6 SEABROOK - UNIT I ii Amendment No.

2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS (SLs) 2.1.1 REACTOR CORE SLs In MODES 1 and 2, the combination of THERMAL POWER, Reactor Coolant System (RCS) highest loop average temperature, and pressurizer pressure shall not exceed the limits specified in the COLR; and the following SLs shall not be exceeded:

2.1.1.1 The departure from nucleate boiling ratio (DNBR) shall be maintained greater than or equal to 1.17 for the WRB-1/WRB-2/WRB-2M DNB correlations.

2.1.1.2 The peak fuel centerline temperature shall be maintained less than 5080'F, decreasing by 58 0F per 10,000 MWD/MTU of burnup.

2.1.2 REACTOR COOLANT SYSTEM PRESSURE SL In MODES 1, 2, 3, 4, and 5, the RCS pressure shall be maintained less than or equal to 2735 psig.

2.1.3 SAFETY LIMIT VIOLATIONS 2.1.3.1 If SL 2.1.1 is violated, restore compliance and be in MODE 3 within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

2.1.3.2 If SL 2.1.2 is violated:

a.

In MODE 1 or 2, restore compliance and be in MODE 3 within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

b.

In MODE 3, 4, or 5, restore compliance within 5 minutes.

SEABROOK - UNIT 1 Amendment No.

2-1

Figure 2.1-1 (THIS FIGURE IS NOT USED)

Amendment No. 33,-76, SEABROOK - UNIT 1 2-2

2.1 SAFETY LIMITS (SLs)

BASES 2.1.1 Reactor Core SLs BACKGROUND GDC 10 (Ref. 1) requires that specified acceptable fuel design limits are not exceeded during steady state operation, normal operational transients, and anticipated operational occurrences (AOOs). This is accomplished by having a departure from nucleate boiling (DNB) design basis, which corresponds to a 95% probability at a 95% confidence level (the 95/95 DNB criterion) that DNB will not occur and by requiring that fuel centerline temperature stays below the melting temperature.

The restrictions of this SL prevent overheating of the fuel and cladding, as well as possible cladding perforation, that would result in the release of fission products to the reactor coolant.

Overheating of the fuel is prevented by maintaining the steady state peak linear heat rate (LHR) below the level at which fuel centerline melting occurs. Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime, where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature.

Fuel centerline melting occurs when the local LHR, or power peaking, in a region of the fuel is high enough to cause the fuel centerline temperature to reach the melting point of the fuel.

Expansion of the pellet upon centerline melting may cause the pellet to stress the cladding to the point of failure, allowing an uncontrolled release of activity to the reactor coolant.

Operation above the boundary of the nucleate boiling regime could result in excessive cladding temperature because of the onset of DNB and the resultant sharp reduction in heat transfer coefficient. Inside the steam film, high cladding temperatures are reached, and a cladding water (zirconium water) reaction may take place. This chemical reaction results in oxidation of the fuel cladding to a structurally weaker form. This weaker form may lose its integrity, resulting in an uncontrolled release of activity to the reactor coolant.

The proper functioning of the Reactor Protection System (RPS) and steam generator safety valves prevents violation of the reactor core SLs.

APPLICABLE SAFETY ANALYSES The fuel cladding must not sustain damage as a result of normal operation and AOOs. The reactor core SLs are established to preclude violation of the following fuel design criteria:

a. There must be at least 95% probability at a 95% confidence level (the 95/95 DNB criterion) that the hot fuel rod in the core does not experience DNB and

b. The hot fuel pellet in the core must not experience centerline fuel melting.

The Reactor Trip System setpoints (Ref. 2), in combination with all the LCOs, are designed to prevent any anticipated combination of transient conditions for Reactor Coolant System (RCS) temperature, pressure, RCS Flow, Al, and THERMAL POWER level that would result in a departure from nucleate boiling ratio (DNBR) of less than the DNBR limit and preclude the existence of flow instabilities.

Automatic enforcement of these reactor core SLs is provided by the appropriate operation of the RPS and the steam generator safety valves.

Amendment No. 33-,7-6, SEABROOK UNIT 1 B 2-1

2.1 SAFETY LIMITS (SLs)

BASES 2.1.1 Reactor Core SLs (continued)

The SLs represent a design requirement for establishing the RPS trip setpoints identified previously. Specification 3/4.2.5, "DNB Parameters," or the assumed initial conditions of the safety analyses (as indicated in the UFSAR, Ref. 2) provide more restrictive limits to ensure that the SLs are not exceeded.

SAFETY LIMITS The figure provided in the COLR shows the loci of points of THERMAL POWER, RCS pressure, and average temperature for which the minimum DNBR is not less than the safety analyses limit, that fuel centerline temperature remains below melting, that the average enthalpy in the hot leg is less than or equal to the enthalpy of saturated liquid, or that the exit quality is within the limits defined by the DNBR correlation.

The reactor core SLs are established to preclude violation of the following fuel design criteria:

a. There must be at least a 95% probability at a 95% confidence level (the 95/95 DNB criterion) that the hot fuel rod in the core does not experience DNB and

b. There must be at least a 95% probability at a 95% confidence level that the hot fuel pellet in the core does not experience centerline fuel melting.

The reactor core SLs are used to define the various RPS functions such that the above criteria are satisfied during steady state operation, normal operational transients, and anticipated operational occurrences (AOOs). To ensure that the RPS precludes the violation of the above criteria, additional criteria are applied to the Overtemperature and Overpower AT reactor trip functions. That is, it must be demonstrated that the average enthalpy in the hot leg is less than or equal to the saturation enthalpy and the core exit quality is within the limits defined by the DNBR correlation. Appropriate functioning of the RPS ensures that for variations in the THERMAL POWER, RCS Pressure, RCS average temperature, RCS flow rate, and Al that the reactor core SLs will be satisfied during steady state operation, normal operational transients, and AOOs.

APPLICABILITY SL 2.1.1 only applies in MODES 1 and 2 because these are the only MODES in which the reactor is critical. Automatic protection functions are required to be OPERABLE during MODES 1 and 2 to ensure operation within the reactor core SLs. The steam generator safety valves or automatic protection actions serve to prevent RCS heatup to the reactor core SL conditions or to initiate a reactor trip function, which forces the unit into MODE 3. Setpoints for the reactor trip functions are specified in Specification 3/4.3.1, "Reactor Trip System (RTS) Instrumentation." In MODES 3, 4, 5, and 6, Applicability is not required since the reactor is not generating significant THERMAL POWER.

Amendment No. 33, SEABROOK UNIT 1 B 2-2

2.1 SAFETY LIMITS (SLs)

BASES 2.1.2 Reactor Coolant System (RCS) Pressure SL BACKGROUND The SL on RCS pressure protects the integrity of the RCS against overpressurization. In the event of fuel cladding failure, fission products are released into the reactor coolant. The RCS then serves as the primary barrier in preventing the release of fission products into the atmosphere. By establishing an upper limit on RCS pressure, the continued integrity of the RCS is ensured.

According to 10 CFR 50, Appendix A, GDC 14, "Reactor Coolant Pressure Boundary," and GDC 15, "Reactor Coolant System Design" (Ref. 3), the reactor pressure coolant boundary (RCPB) design conditions are not to be exceeded during normal operation and anticipated operational occurrences (AOOs). Also, in accordance with GDC 28, "Reactivity Limits" (Ref. 3), reactivity accidents, including rod ejection, do not result in damage to the RCPB greater than limited local yielding.

The design pressure of the RCS is 2500 psia. During normal operation and AQOs, RCS pressure is limited from exceeding the design pressure by more than 10%, in accordance with Section III of the ASME Code (Ref. 4). To ensure system integrity, all RCS components are hydrostatically tested at 125% of design pressure, according to the ASME Code requirements prior to initial operation when there is no fuel in the core. Following inception of unit operation, RCS components shall be pressure tested, in accordance with the requirements of ASME Code,Section XI (Ref. 5).

Overpressurization of the RCS could result in a breach of the RCPB. If such a breach occurs in conjunction with a fuel cladding failure, fission products could enter the containment atmosphere, raising concerns relative to limits on radioactive releases specified in 10 CFR 100, "Reactor Site Criteria" (Ref. 6).

APPLICABLE SAFETY ANALYSES The RCS pressurizer safety valves, the main steam safety valves (MSSVs), and the reactor high pressure trip have settings established to ensure that the RCS pressure SL will not be exceeded.

The RCS pressurizer safety valves are sized to prevent system pressure from exceeding the design pressure by more than 10%, as specified in Section III of the ASME Code for Nuclear Power Plant Components (Ref. 3). The transient that establishes the required relief capacity, and hence valve size requirements and lift settings, is a complete loss of external load without a direct reactor trip. During the transient, no control actions are assumed, except that the safety valves on the secondary plant are assumed to open when the steam pressure reaches the secondary plant safety valve settings, and nominal feedwater supply is maintained.

The Reactor Trip System setpoints (Ref. 2), together with the settings of the MSSVs, provide pressure protection for normal operation and AQOs. The reactor high pressure trip setpoint is specifically set to provide protection against overpressurization (Ref. 2). The safety analyses for both the high pressure trip and the RCS pressurizer safety valves are performed using conservative assumptions relative to pressure control devices.

SEABROOK UNIT 1 Amendment No.

B 2-2a

2.1 SAFETY LIMITS (SLs)

BASES 2.1.2 Reactor Coolant System (RCS) Pressure SL (continued)

More specifically, no credit is taken for operation of any of the following:

a.

Pressurizer power operated relief valves (PORVs),

b.

Steam line relief valve,

c.

Steam Dump System,

d.

Reactor Control System,

e.

Pressurizer Level Control System, or

f.

Pressurizer spray valve.

SAFETY LIMITS The maximum transient pressure allowed in the RCS pressure vessel under the ASME Code,Section III, is 110% of design pressure. The maximum transient pressure allowed in the RCS piping, valves, and fittings under USAS, Section B31.1 (Ref. 7) is 120% of design pressure.

The most limiting of these two allowances is the 110% of design pressure; therefore, the SL on maximum allowable RCS pressure is 2735 psig.

APPLICABILITY SL 2.1.2 applies in MODES 1, 2, 3, 4, and 5 because this SL could be approached or exceeded in these MODES due to overpressurization events. The SL is not applicable in MODE 6 because the reactor vessel head closure bolts are not fully tightened, making it unlikely that the RCS can be pressurized.

2.1.3 SAFETY LIMIT VIOLATIONS The following SL violation responses are applicable:

If the reactor core SL 2.1.1 is violated, the requirement to go to MODE 3 places the unit in a MODE in which this SL is not applicable. The Allowed Outage Time (Completion Time) of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months recognizes the importance of bringing the unit to a MODE of operation where this SL is not applicable, and reduces the probability of fuel damage.

If the RCS pressure SL is violated when the reactor is in MODE 1 or 2, the requirement is to restore compliance and be in MODE 3 within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . Exceeding the RCS pressure SL may cause immediate RCS failure and create a potential for radioactive releases in excess of 10 CFR 100, "Reactor Site Criteria," limits (Ref. 6). The Allowed Outage Time (Completion Time) of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months recognizes the importance of reducing power level to a MODE of operation where the potential for challenges to safety systems is minimized.

SEABROOK UNIT I B 2-2b Amendment No.

2.1 SAFETY LIMITS (SLs)

BASES SAFETY LIMIT VIOLATIONS (continued)

If the RCS pressure SL is exceeded in MODE 3, 4, or 5, RCS pressure must be restored to within the SL value within 5 minutes. Exceeding the RCS pressure SL in MODE 3, 4, or 5 is more severe than exceeding this SL in MODE 1 or 2, since the reactor vessel temperature may be lower and the vessel material, consequently, less ductile. As such, pressure must be reduced to less than the SL within 5 minutes. The action does not require reducing MODES, since this would require reducing temperature, which would compound the problem by adding thermal gradient stresses to the existing pressure stress.

REFERENCES

1. 10 CFR 50, Appendix A, GDC 10.

2. UFSAR, Chapters 7 and 15.

3. 10 CFR 50, Appendix A, GDC 14, GDC 15, and GDC 28.

4. ASME, Boiler and Pressure Vessel Code,Section III, Article NB-7000.

5. ASME, Boiler and Pressure Vessel Code,Section XI, Article IWX-5000.

6. 10 CFR 100.

7. USBS B31.1, Standard Code for Pressure Piping, American Society of Mechanical Engineers, 1967.

SEABROOK UNIT 1 Amendment No.

B 2-2c

LIMITING SAFETY SYSTEM SETTINGS BASES 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS (Continued)

Intermediate and Source Ranqje, Neutron Flux The Intermediate and Source Range, Neutron Flux trips provide core protection during reactor startup to mitigate the consequences of an uncontrolled rod cluster control assembly bank withdrawal from a subcritical condition. These trips provide redundant protection to the Low Setpoint trip of the Power Range, Neutron Flux channels. The Source Range channels will initiate a Reactor trip at about 105 counts per second unless manually blocked when P-6 becomes active. The Intermediate Range channels will initiate a Reactor trip at a current level equivalent to approximately 25% of RATED THERMAL POWER unless manually blocked when P-10 becomes active.

Overtemperature AT The Overtemperature AT trip provides core protection to prevent DNB for all combinations of pressure, power, coolant temperature, and axial power distribution, provided that the transient is slow with respect to piping transit delays from the core to the temperature detectors (about 4 seconds), pressure is within the range between the Pressurizer High and Low Pressure trips and power is less than the Overpower AT trip setpoint. The Setpoint is automatically varied with: (1) coolant temperature to correct for temperature induced changes in density and heat capacity of water and includes dynamic compensation for piping delays from the core to the loop temperature detectors, (2) pressurizer pressure, and (3) axial power distribution. With normal axial power distribution, this Reactor trip limit is always below the core Safety Limit as shown in the COLR. If axial peaks are greater than design, as indicated by the difference between top and bottom power range nuclear detectors, the Reactor trip is automatically reduced according to the notations in Table 2.2-1.

Overpower AT The Overpower AT trip provides assurance of fuel integrity (e.g., no fuel pellet melting and less than 1 % cladding strain) under all possible overpower conditions, limits the required range for Overtemperature AT trip, and provides a backup to the High Neutron Flux trip. The Setpoint is automatically varied with: (1) coolant temperature to correct for temperature induced changes in density and heat capacity of water, (2) rate of change of temperature for dynamic compensation for piping delays from the core to the loop temperature detectors, and (3) axial power distribution to ensure that the allowable heat generation rate (Kw/ft) is not exceeded. The Overpower AT trip provides protection to mitigate the consequences of various size steam breaks as reported in WCAP-9226, "Reactor Core Response to Excessive Secondary Steam Releases."

SEABROOK - UNIT 1 B 2-5 Amendment No. -33

3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL SHUTDOWN MARGIN -TAVG GREATER THAN 200°F LIMITING CONDITION FOR OPERATION 3.1.1.1 The SHUTDOWN MARGIN for four-loop operation shall be greater than or equal to the limit specified in the CORE OPERATING LIMITS REPORT (COLR).

APPLICABILITY:

MODES 1, 2*, 3, and 4.

ACTION:

With the SHUTDOWN MARGIN less than the limiting value, immediately initiate and continue boration equivalent to 30 gpm at a boron concentration greater than or equal to the limit specified in the COLR for the Boric Acid Storage System until the required SHUTDOWN MARGIN is restored.

SURVEILLANCE REQUIREMENTS 4.1.1.1.1 The SHUTDOWN MARGIN shall be determined to be greater than or equal to the limiting value:

a.

Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after detection of an inoperable control rod(s) and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter while the rod(s) is inoperable. If the inoperable control rod is immovable or untrippable, the above required SHUTDOWN MARGIN shall be verified acceptable with an increased allowance for the withdrawn worth of the immovable or untrippable control rod(s);

b.

When in MODE 1 or MODE 2 with keff greater than or equal to 1 at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by verifying that control bank withdrawal is within the limits of Specification 3.1.3.6;

c.

When in MODE 2 with keff less than 1, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months prior to achieving reactor criticality by verifying that the predicted critical control rod position is within the limits of Specification 3.1.3.6;

d.

Prior to initial operation above 5% RATED THERMAL POWER after each fuel loading, by consideration of the factors of Specification 4.1.1.1.1e below, with the control banks at the maximum insertion limit of Specification 3.1.3.6; and

See Special Test Exceptions Specification 3.10.1.

SEABROOK - UNIT 1 Amendment No. 9, 3/4 1-1

REACTIVITY CONTROL SYSTEMS BORATION CONTROL SHUTDOWN MARGIN -TAVG LESS THAN OR EQUAL TO 200°F LIMITING CONDITION FOR OPERATION 3.1.1.2 The SHUTDOWN MARGIN shall be greater than or equal to the limit specified in the CORE OPERATING LIMITS REPORT (COLR). Additionally, the Reactor Coolant System boron concentration shall be greater than or equal to the limit specified in the COLR when the reactor coolant loops are in a drained condition.

APPLICABILITY: MODE 5.

ACTION:

With the SHUTDOWN MARGIN less than the limit specified in the COLR or the Reactor Coolant System boron concentration less than the limit specified in the COLR, immediately initiate and continue boration equivalent to 30 gpm at a boron concentration greater than or equal to the limit specified in the COLR for the Boric Acid Storage System until the required SHUTDOWN MARGIN and boron concentration are restored.

SURVEILLANCE REQUIREMENTS 4.1.1.2 The SHUTDOWN MARGIN shall be determined to be greater than or equal to the limit specified in the COLR and the Reactor Coolant System boron concentration shall be determined to be greater than or equal to the limit specified in the COLR when the reactor coolant loops are in a drained condition:

a.

Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after detection of an inoperable control rod(s) and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter while the rod(s) is inoperable. If the inoperable control rod is immovable or untrippable, the SHUTDOWN MARGIN shall be verified acceptable with an increased allowance for the withdrawn worth of the immovable or untrippable control rod(s); and

b.

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by consideration of the following factors:

1)

Reactor Coolant System boron concentration,

2)

Control rod position,

3)

Reactor Coolant System average temperature,

4)

Fuel burnup based on gross thermal energy generation,

5)

Xenon concentration, and

6)

Samarium concentration.

SEABROOK-UNIT 1 3/4 1-3 Amendment No. -9,

REACTIVITY CONTROL SYSTEMS BORATION SYSTEMS BORATED WATER SOURCES - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.1.2.5 As a minimum, one of the following borated water sources shall be OPERABLE:

a. A Boric Acid Storage System with:

1)

A minimum contained borated water volume of 6,500 gallons,

2)

A minimum boron concentration as specified in the COLR, and

3)

A minimum solution temperature of 650F.

b.

The refueling water storage tank (RWST) with:

1)

A minimum contained borated water volume of 24,500 gallons,

2)

A minimum boron concentration as specified in the COLR, and

3)

A minimum solution temperature of 500F.

APPLICABILITY: MODES 5 and 6.

ACTION:

With no borated water source OPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes.

SURVEILLANCE REQUIREMENTS 4.1.2.5 The above required borated water source shall be demonstrated OPERABLE:

a.

At least once per 7 days by:

1)

Verifying the boron concentration of the water,

2)

Verifying the contained borated water volume, and

3)

Verifying the boric acid storage tank solution temperature when it is the source of borated water.

b. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying the RWST temperature.

SEABROOK - UNIT 1 3/4 1-11 Amendment No. 42-,

REACTIVITY CONTROL SYSTEMS BORATION SYSTEMS BORATED WATER SOURCES - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.6 As a minimum, the following borated water sources shall be OPERABLE as required by Specification 3.1.2.2:

a.

A Boric Acid Storage System with:

1)

A minimum contained borated water volume of 22,000 gallons,

2)

A minimum boron concentration as specified in the COLR, and

3)

A minimum solution temperature of 650F.

b.

The refueling water storage tank (RWST) with:

1)

A minimum contained borated water volume of 477,000 gallons,

2)

A boron concentration between the specified limits in the COLR,

3)

A minimum solution temperature of 50°F, and

4)

A maximum solution temperature of 980F.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

a.

With the Boric Acid Storage System inoperable and being used as one of the above required borated water sources, restore the system to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and borated to a SHUTDOWN MARGIN equivalent to at least the limit specified in the CORE OPERATING LIMITS REPORT (COLR) for the above MODES at 200'F; restore the Boric Acid Storage System to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b.

With the RWST inoperable, restore the tank to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SEABROOK - UNIT 1 3/4 1-12 Amendment No. 42-,

REACTIVITY CONTROL SYSTEMS BORON SYSTEMS ISOLATION OF UNBORATED WATER SOURCES - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.1.2.7 Provisions to isolate the Reactor Coolant System from unborated water sources shall be OPERABLE with:

a.

The Boron Thermal Regeneration System (BTRS) isolated from the Reactor Coolant System, and

b.

The Reactor Makeup Systems inoperable except for the capability of delivering up to the capacity of one Reactor Makeup Water pump to the Reactor Coolant System.

APPLICABILITY: MODES 4, 5, and 6 ACTION:

With the requirements of the above specification not satisfied immediately suspend all operations involving CORE ALTERATIONS or positive reactivity changes and, if within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months the required SHUTDOWN MARGIN is not verified, initiate and continue boration equivalent to 30 gpm at a boron concentration greater than or equal to the limit specified in the COLR for the Boric Acid Storage System until the required SHUTDOWN MARGIN is restored and the isolation provisions are restored to OPERABLE.

SURVEILLANCE REQUIREMENTS 4.1.2.7 The provisions to isolate the Reactor Coolant System from unborated water sources shall be determined to be OPERABLE at least once per 31 days by:

a.

Verifying that at least the BTRS outlet valve, CS-V-302, or the BTRS moderating heat exchanger outlet valve, CS-V-305, is closed and locked closed, and

b.

Verifying that power is removed from at least one of the Reactor Makeup Water pumps, RMW-P-1 6A or RMW-P-1 6B.

SEABROOK - UNIT 1 3/4 1-14 Amendment No.

POWER DISTRIBUTION LIMITS 3/4.2.5 DNB PARAMETERS LIMITING CONDITION FOR OPERATION 3.2.5 The following DNB-related parameters shall be maintained within the following limits:

a.

Reactor Coolant System Tavg is less than or equal to the limit specified in the COLR,

b.

Pressurizer Pressure is greater than or equal to the limit specified in the COLR*, and

c.

Reactor Coolant System Flow shall be:

1.

> 382,800 gpm**; and,

2.

> 392,800 gpm***

APPLICABILITY: MODE 1.

ACTION:

With any of the above parameters exceeding its limit, restore the parameter to within its limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE REQUIREMENTS 4.2.5.1 Each of the parameters shown above shall be verified to be within its limits at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

4.2.5.2 The RCS flow rate indicators shall be subjected to CHANNEL CALIBRATION at least once per 18 months.

4.2.5.3 The RCS total flow rate shall be determined by an approved method to be within its limit prior to operation above 95% of RATED THERMAL POWER after each fuel loading. The provisions of Specification 4.0.4 are not applicable for entry into MODE 1.

Limit not applicable during either a THERMAL POWER ramp in excess of 5% of RATED THERMAL POWER per minute or a THERMAL POWER step in excess of 10%

of RATED THERMAL POWER.

- Thermal Design Flow. An allowance for measurement uncertainty shall be made when comparing measured flow to Thermal Design Flow.

Minimum measured flow used in the Revised Thermal Design Procedure.

Amendment No. 2,_,

83, 77, SEABROOK - UNIT 1 3/4 2-10

3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS HOT STANDBY, STARTUP, AND POWER OPERATION LIMITING CONDITION FOR OPERATION 3.5.1.1 Each Reactor Coolant System (RCS) accumulator shall be OPERABLE with:

a.

The isolation valve open and power removed,

b.

A contained borated water volume of between 6121 and 6596 gallons,

c.

A boron concentration of between the limits specified in the COLR, and

d.

A nitrogen cover-pressure of between 585 and 664 psig.

APPLICABILITY: MODES 1,2, and 3*.

ACTION:

a.

With one accumulator inoperable, except as a result of a closed isolation valve, restore the inoperable accumulator to OPERABLE status within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce pressurizer pressure to less than 1000 psig within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b.

With one accumulator inoperable due to the isolation valve being closed, either immediately open the isolation valve or be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce pressurizer pressure to less than 1000 psig within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

c.

With one pressure or water level channel inoperable per accumulator, return the inoperable channel to OPERABLE status within 30 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

d.

With two pressure channels or two water level channels inoperable per accumulator, immediately declare the affected accumulator(s) inoperable.

SURVEILLANCE REQUIREMENTS 4.5.1.1 Each accumulator shall be demonstrated OPERABLE:

a.

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by:

1)

Verifying the contained borated water volume and nitrogen cover-pressure in the tanks, and

Pressurizer pressure above 1000 psig.

SEABROOK - UNIT 1 3/4 5-1 Amendment No. 30-,42, I

BORON INJECTION SYSTEM 3/4.5.4 REFUELING WATER STORAGE TANK LIMITING CONDITION FOR OPERATION 3.5.4 The refueling water storage tank (RWST) shall be OPERABLE with:

a.

A minimum contained borated water volume of 477,000 gallons,

b.

A boron concentration between the limits specified in the COLR,

c.

A minimum solution temperature of 500F, and

d.

A maximum solution temperature of 98 0F.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTION:

With the RWST inoperable, restore the tank to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.5.4 The RWST shall be demonstrated OPERABLE:

a.

At least once per 7 days by:

1)

Verifying the contained borated water volume in the tank, and

2)

Verifying the boron concentration of the water.

b.

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying the RWST temperature.

SEABROOK-UNIT 4

Amendment No. 42, 3/4 5-11

3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION LIMITING CONDITION FOR OPERATION 3.9.1 The boron concentration of all filled portions of the Reactor Coolant System and the refueling canal shall be maintained uniform and sufficient to ensure a boron concentration of greater than or equal to the limit specified in the COLR.

APPLICABILITY: MODE 6.*

ACTION:

With the requirements of the above specification not satisfied, immediately suspend all operations involving CORE ALTERATIONS or positive reactivity changes and initiate and continue boration equivalent to 30 gpm at a boron concentration greater than or equal to the limit specified in the COLR for the Boric Acid Storage System until the boron concentration is restored to greater than or equal to the limit specified in the COLR.

SURVEILLANCE REQUIREMENTS 4.9.1.1 Verify boron concentration is within the limits specified in the COLR prior to:

a.

Removing or unbolting the reactor vessel head, and

b.

Withdrawal of any full-length control rod in excess of 3 feet from its fully inserted position within the reactor vessel.

4.9.1.2 The boron concentration of the Reactor Coolant System and the refueling canal shall be determined by chemical analysis at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

The reactor shall be maintained in MODE 6 whenever fuel is in the reactor vessel with the vessel head closure bolts less than fully tensioned or with the head removed.

SEABROOK - UNIT 1 Amendment No.

3/4 9-1

ADMINISTRATIVE CONTROLS MONTHLY OPERATING REPORTS 6.8.1.5 Routine reports of operating statistics and shutdown experience shall be submitted on monthly basis to the U.S. Nuclear Regulatory Commission, Washington, D.C. 20555, Attn: Document Control Desk, with a copy to the NRC Regional Administrator, no later than the 15th of each month following the calendar month covered by the report.

CORE OPERATING LIMITS REPORT 6.8.1.6.a Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT prior to each reload cycle, or prior to any remaining portion of a reload cycle, for the following:

1.

Cycle dependent Overpower AT and Overtemperature AT trip setpoint parameters and function modifiers for operation with skewed axial power profiles for Table 2.2-1 of Specification 2.2.1,

2.

Cycle dependent maximum allowable combination of thermal power, pressurizer pressure and the highest operating loop average temperature (Tavg) for Specifications 2.1.1 and 2.1.2,

3.

SHUTDOWN MARGIN and minimum boron concentration limits for MODES 1, 2, 3, and 4 for Specification 3.1.1.1,

4.

SHUTDOWN MARGIN and minimum boron concentration limits for MODE 5 for Specification 3.1.1.2,

5.

Moderator Temperature Coefficient BOL and EOL limits, and 300 ppm surveillance limit for Specification 3.1.1.3,

6.

The minimum boron concentration for MODES 5 and 6 for Specification 3.1.2.5,

7.

The boron concentration limits for Modes 1, 2, 3, and 4 for Specification 3.1.2.6,

8.

The minimum boron concentration for Modes 4, 5, and 6 for Specification 3.1.2.7,

9.

Shutdown Rod Insertion limit for Specification 3.1.3.5,

10.

Control Rod Bank Insertion limits for Specification 3.1.3.6,

11.

AXIAL FLUX DIFFERENCE limits for Specification 3.2.1, Amendment No.

, -22, 34,66, SEABROOK - UNIT 1 6-18

ADMINISTRATIVE CONTROLS 6.8.1.6.a. (Continued)

12.

Heat Flux Hot Channel Factor, FRQP and K(Z) for Specification 3.2.2, RTP

13.

Nuclear Enthalpy Rise Hot Channel Factor, and F AH for Specification 3.2.3

14.

Cycle dependent DNB-related parameters for reactor coolant system average temperature (Tavg), and pressurizer pressure for Specification 3.2.5,

15.

The boron concentration limits for MODES 1, 2 and 3 for Specification 3.5.1.1,

16.

The boron concentration limits for MODES 1, 2, 3 and 4 for Specification 3.5.4,

17.

The boron concentration limits for MODE 6 for Specification 3.9.1.

6.8.1.6.b The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC in:

1.

WCAP-1 0266-P-A, Rev. 2 with Addenda (Proprietary) and WCAP-1 1524 A, Rev. 2 with Addenda (Nonproprietary), "The 1981 Version of the Westinghouse ECCS Evaluation Model Using the BASH Code", March, 1987.

Methodology for Specification:

3.2.2 Heat Flux Hot Channel Factor

2.

WCAP-1 0079-P-A, (Proprietary) and WCAP-10080-A (Nonproprietary),

"NOTRUMP: A Nodal Transient Small Break and General Network Code", August, 1985.

Methodology for Specification:

3.2.2 Heat Flux Hot Channel Factor

3.

YAEC-1 363-A, "CASMO-3G Validation," April, 1988.

YAEC-1659-A, "SIMULATE-3 Validation and Verification,"

September, 1988.

WCAP-1 1596-P-A, (Proprietary), "Qualification of the PHOENIX-P/ANC Nuclear Design System for Pressurized Water Reactor Cores",

June, 1988.

WCAP-1 0965-P-A, (Proprietary), "ANC: A Westinghouse Advanced Nodal Computer Code", September, 1986.

Amendment No. 9,- -3 70, 76, SEABROOK - UNIT 1 6-18A

ADMINISTRATIVE CONTROLS 6.8.1.6.b. (Continued)

Methodology for Specifications:

3.1.1.1 SHUTDOWN MARGIN for MODES 1, 2, 3, and 4 3.1.1.2 SHUTDOWN MARGIN for MODE 5 3.1.1.3 Moderator Temperature Coefficient 3.1.3.5 Shutdown Rod Insertion Limit 3.1.3.6 Control Rod Insertion Limits 3.2.1 AXIAL FLUX DIFFERENCE 3.2.2 Heat Flux Hot Channel Factor 3.2.3 Nuclear Enthalpy Rise Hot Channel Factor

4.

Seabrook Station Updated Final Safety Analysis Report, Section 15.4.6, "Chemical and Volume Control System Malfunction That Results in a Decrease in the Boron Concentration in the Reactor Coolant System".

Methodology for Specifications:

3.1.1.1 SHUTDOWN MARGIN for MODES 1,2,3, and 4 3.1.1.2 SHUTDOWN MARGIN for MODE 5

5.

YAEC-1 241, "Thermal-Hydraulic Analysis of PWR Fuel Elements Using the CHIC-KIN Code", R. E. Helfrich, March, 1981.

WCAP-14565-P, (Proprietary), "VIPRE-01 Modeling and Qualification for Pressurized Water Reactor Non-LOCA Thermal-Hydraulic Safety Analysis", April, 1997.

Letter from T. H. Essig (NRC) to H. Sepp (Westinghouse), "Acceptance for Referencing of Licensing Topical Report WCAP-14565-P, (Proprietary), "VIPRE-01 Modeling and Qualification for Pressurized Water Reactor Non-LOCA Thermal-Hydraulic Safety Analysis", January, 1999.

WCAP-1 5025-P, "Modified WRB-2 Correlation, WRB-2M, for Predicting Critical Heat Flux in 17X17 Rod Bundles with Modified LPD Mixing Vane Grids," February 1998.

Methodology for Specification:

2.1 Safety Limits 3.2.1 AXIAL FLUX DIFFERENCE 3.2.2 Heat Flux Hot Channel Factor 3.2.3 Nuclear Enthalpy Rise Hot Channel Factor 3.2.5 DNB Parameters

6.

YAEC-1849P, "Thermal-Hydraulic Analysis Methodology Using VIPRE-01 For PWR Applications," October, 1992.

WCAP-1 1397-P-A, (Proprietary), "Revised Thermal Design Procedure",

April, 1989.

Amendment No. 33,76, SEABROOK - UNIT 1 6-18B

ADMINISTRATIVE CONTROLS 6.8.1.6.b.

(Continued)

15.

WCAP-9272-P-A, (Proprietary), "Westinghouse Reload Safety Evaluation Methodology", July, 1985.

Methodology 2.1 3.1.1.1 3.1.1.2 3.1.1.3 3.1.2.5 3.1.2.6 3.1.2.7 3.1.3.5 3.1.3.6 3.2.1 3.2.2 3.2.3 3.2.5 3.5.1.1 3.5.4 3.9.1 for Specifications:

Safety Limits SHUTDOWN MARGIN for MODES 1, 2, 3, and 4 SHUTDOWN MARGIN for MODE 5 Moderator Temperature Coefficient Borated Water Sources - Shutdown Borated Water Sources - Operating Isolation of Unborated Water Sources - Shutdown Shutdown Rod Insertion Limit Control Rod Insertion Limits AXIAL FLUX DIFFERENCE Heat Flux Hot Channel Factor Nuclear Enthalpy Rise Hot Channel Factor DNB Parameters Accumulators for MODES 1, 2 and 3 Refueling Water Storage Tank for MODES 1, 2, 3, and 4 Boron Concentration 6.8.1.6.c.

The core operating limits shall be determined so that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as SHUTDOWN MARGIN, and transient and accident analysis limits) of the safety analysis are met. The CORE OPERATING LIMITS REPORT for each reload cycle, including any mid cycle revisions or supplements thereto, shall be provided upon issuance, to the NRC Document Control Desk with copies to the Regional Administrator and the Resident Inspector.

SEABROOK - UNIT 1 6-18E Amendment No. 7-6,

3/4.1 REACTIVITY CONTROL SYSTEMS BASES 3/4.1.1 BORATION CONTROL 3/4.1.1.1 and 3/4.1.1.2 SHUTDOWN MARGIN A sufficient SHUTDOWN MARGIN ensures that: (1) the reactor can be made subcritical from all operating conditions, (2) the reactivity transients associated with postulated accident conditions are controllable within acceptable limits, and (3) the reactor will be maintained sufficiently subcritical to preclude inadvertent criticality in the shutdown condition.

SHUTDOWN MARGIN requirements vary throughout core life as a function of fuel depletion, RCS boron concentration, and RCS Tavg. The most restrictive condition occurs at EOL, with Tavg at no-load operating temperature, and is associated with a postulated steam line break accident and resulting uncontrolled RCS cooldown. In the analysis of this accident, a minimum SHUTDOWN MARGIN as specified in the CORE OPERATING LIMITS REPORT (COLR) is required to control the reactivity transient. Accordingly, the SHUTDOWN MARGIN requirement is based upon this limiting condition and is consistent with FSAR safety analysis assumptions. With Tavg less than 2000F, the reactivity transients resulting from a postulated steam line break cooldown are minimal. A SHUTDOWN MARGIN as specified in the COLR and a boron concentration of greater than the limit specified in the COLR are required to permit sufficient time for the operator to terminate an inadvertent boron dilution event with Tavg less than 200°F.

The "equivalent to" statement in the Action is a provision providing an alternate method of emergency boration via the RWST at an increased flow rate to account for the lower boron concentration within the RWST.

3/4.1.1.3 MODERATOR TEMPERATURE COEFFICIENT The limitations on moderator temperature coefficient (MTC) are provided to ensure that the value of this coefficient remains within the limiting condition assumed in the FSAR accident and transient analyses.

The MTC values of this specification are applicable to a specific set of plant conditions; accordingly, verification of MTC values at conditions other than those explicitly stated will require extrapolation to those conditions in order to permit an accurate comparison.

The most negative MTC, value equivalent to the most positive moderator density coefficient (MDC), was obtained by incrementally correcting the MDC used in the FSAR analyses to nominal operating conditions. These corrections involved subtracting the incremental change in the MDC associated with a core condition of all rods inserted (most positive MDC) to an all rods withdrawn condition and, a conversion for the rate of change of moderator density with temperature at RATED THERMAL POWER conditions. This value of the MDC was then transformed into the limiting end of cycle life (EOL) MTC value as specified in the COLR. The 300 ppm surveillance limit MTC value as specified in the COLR represents a conservative value (with corrections for burnup and soluble boron) at a core condition of 300 ppm equilibrium boron concentration and is obtained by making these corrections to the limiting MTC value as specified in the COLR.

SEABROOK - UNIT 1 B 3/4 1-1 Amendment No. 9,

REACTIVITY CONTROL SYSTEMS BASES 3/4.1.2 BORATION SYSTEMS (Continued) boron capability requirement occurs at EOL from full power equilibrium xenon conditions.

LCO 3.1.2.6 specifies the minimum requirements for borated water sources availability during MODES 1, 2, 3 and 4. The "equivalent to" statement in the Action for LCO 3.1.2.7 is a provision providing an alternate method of emergency boration via the RWST at an increased flow rate to account for the lower boron concentration within the RWST.

The limitation for a maximum of one centrifugal charging pump to be OPERABLE and the Surveillance Requirement to verify all charging pumps except the required OPERABLE pump to be inoperable in MODES 4, 5, and 6 except when the reactor vessel head closure bolts are fully detensioned or the vessel head is removed, provides assurance that a mass addition pressure transient can be relieved by operation of a single PORV or an RHR suction relief valve.

As a result of this, only one boron injection system is available. This is acceptable on the basis of the stable reactivity condition of the reactor, the emergency power supply requirement for the OPERABLE charging pump and the additional restrictions prohibiting CORE ALTERATIONS and positive reactivity changes in the event the single injection system becomes inoperable.

The boron capability required below 200°F is sufficient to provide a SHUTDOWN MARGIN as specified in the CORE OPERATING LIMITS REPORT after xenon decay and cooldown from 200°F to 1400F. LCO 3.1.2.5 specifies the minimum requirements for borated water sources availability during MODES 5 and 6.

The contained water volume limits include allowance for water not available because of discharge line location and other physical characteristics.

The limits on contained water volume and boron concentration of the RWST also ensure a pH value of between 8.5 and 11.0 for the solution recirculated within containment after a LOCA. This pH band minimizes the evolution of iodine and minimizes the effect of chloride and caustic stress corrosion on mechanical systems and components.

The OPERABILITY of one Boron Injection System during REFUELING ensures that this system is available for reactivity control while in MODE 6.

The limitations on OPERABILITY of isolation provisions for the Boron Thermal Regeneration System and the Reactor Water Makeup System in Modes 4, 5, and 6 ensure that the boron dilution flow rates cannot exceed the value assumed in the transient analysis.

Amendment No. 9, 34, 42, 74, SEABROOK - UNIT 1 B 3/4 1-3

POWER DISTRIBUTION LIMITS BASES 3/4.2.5 DNB PARAMETERS The limits on the DNB-related parameters assure that each of the parameters is maintained within the normal steady-state envelope of operation assumed in the transient and accident analyses. The limits are consistent with the updated FSAR assumptions and have been analytically demonstrated adequate to assure compliance with acceptance criteria for each analyzed transient. Operating procedures include allowances for measurement and indication uncertainty so that the limits specified in the COLR for Tavg and for pressurizer pressure are not exceeded.

The 12-hour periodic surveillance of these parameters through instrument readout is sufficient to ensure that the parameters are restored within their limits following load changes and other expected transient operation.

The periodic surveillance of indicated RCS flow is sufficient to detect only flow degradation which could lead to operation outside the specified limit.

RCS flow must be greater than or equal to, 1) the Thermal Design Flow (TDF) with an allowance for measurement uncertainty and, 2) the minimum measured flow used in place of the TDF in the analysis of the DNB related events when the Revised Thermal Design Procedure (RTDP) methodology is utilized. Measurement of RCS total flow rate is performed by performance of either a precision calorimetric heat balance or normalized cold leg elbow tap AP measurements. RCS flow measurements using either the precision heat balance or the elbow tap AP measurement methods are to be performed at steady state conditions prior to operation above 95% rated thermal power (RTP) at the beginning of a new fuel cycle.

The elbow tap RCS flow measurement methodology is described in WCAP-1 5404, "Justification of Elbow Taps for RCS Flow Verification at Seabrook Station", dated April 2000.

Amendment No. 9,12,33,34,70,77, SEABROOK - UNIT 1 B 3/4 2-4

3/4.9 REFUELING OPERATIONS BASES 3/4.9.1 BORON CONCENTRATION The limit on the boron concentrations of the Reactor Coolant System (RCS) and the refueling canal/cavity during refueling ensures that the reactor remains subcritical during MODE 6. During refueling, the spent fuel pool water volumes and the reactor cavity water volumes will be connected when the fuel transfer gate valve is open. This configuration allows the bodies of water to be physically capable of being in contact, however, no effective mixing of the volumes occurs due to the constriction of the fuel transfer tube. The soluble boron concentration in each of these volumes is maintained greater than or equal to the limit specified in the COLR, or equivalent to a Keff less than or equal to 0.95 when the fuel transfer gate is open. However, the spent fuel pool water boron concentration is under administrative controls and not a technical specification.

They are independently maintained at the appropriate boron concentration even though no intermixing of significance exists. The mixing caused by the RHR pumps (reactor cavity) or the SFP system pumps assures uniformity of boron in the separate volumes.

The soluble boron concentration offsets the core reactivity and is measured by chemical analysis of a representative sample of the coolant in each of the volumes. Plant procedures ensure the specified boron concentration in order to maintain an overall core reactivity of keff < 0.95 during fuel handling, with control rods and fuel assemblies assumed to be in the most adverse configuration (least negative reactivity) allowed by plant procedures.

GDC 26 of 10 CFR 50, Appendix A, requires that two independent reactivity control systems of different design principles be provided. One of these systems must be capable of holding the reactor core subcritical under cold conditions. The Chemical and Volume Control System (CVCS) is the system capable of maintaining the reactor subcritical in cold conditions by maintaining the boron concentration.

The reactor is brought to shutdown conditions before beginning operations to open the reactor vessel for refueling. After the RCS is cooled and depressurized and the vessel head is unbolted, the head is slowly removed to form the refueling cavity. The refueling canal and the refueling cavity are then flooded with borated water from the refueling water storage tank through the open reactor vessel by gravity feeding or by the use of the Residual Heat Removal (RHR) System pumps.

The pumping action of the RHR System in the RCS and the natural circulation due to thermal driving heads in the reactor vessel and refueling cavity mix the added solution of boric acid with the water in the refueling canal. The RHR System is in operation during refueling (see LCO 3.9.8.1, "Residual Heat Removal (RHR) and Coolant Circulation High Water Level," and LCO 3.9.8.2, "Residual Heat Removal (RHR) and Coolant Circulation -Low Water Level") to provide forced circulation in the RCS and assist in maintaining the boron concentrations in the RCS and the refueling canal/cavity at or above the limit specified in LCO 3.9.1.

SEABROOK - UNIT 1 B 3/4 9-1 Amendment No.

3/4.9 REFUELING OPERATIONS BASES During refueling operations, the reactivity condition of the core is consistent with the initial conditions assumed for the boron dilution accident in the accident analysis and is conservative for MODE 6. The boron concentration limit specified is based on the core reactivity at the beginning of each fuel cycle (the end of refueling) and includes an uncertainty allowance.

The required boron concentration and the plant refueling procedures that verify the correct fuel loading plan (including full core mapping) ensure that the keff of the core will remain < 0.95 during the refueling operation. Hence, at least 5% Ak/k margin of safety is established during refueling.

Continuation of CORE ALTERATIONS or positive reactivity additions (including actions to reduce boron concentration) is contingent upon maintaining the unit in compliance with the LCO.

During refueling operations water may be transferred to the refueling canal/cavity or the RCS from different sources. Transfers or additions of water whose boron concentration exceeds the required refueling boron concentration are acceptable. Transfers or additions of water where the boron concentration is less than the required refueling boron concentration may be made, provided that these additions are administratively controlled to ensure that the refueling boron concentration requirements continue to be met. That is, the final concentration of boron in the total volume, after the addition of water less than the required refueling boron concentration, exceeds the required refueling boron concentration, or keff< 0.95. Also, these administrative controls ensure such transfers or additions of water will not substantially reduce the uniformity of boron concentration in the RCS or refueling canal.

Likewise, transferring water to the RCS or the refueling canal/cavity that is lower in temperature (down to the operability requirements of the RWST in MODE 6; 50 DEG F) than the water contained in those volumes is also acceptable. These minimum requirements for boron concentration and water temperature are also applicable to other MODE 6 Technical Specification ACTIONS that limit operations involving positive reactivity additions to ensure that the reactor remains subcritical and an adequate shutdown margin is maintained.

Suspension of CORE ALTERATIONS and positive reactivity additions shall not preclude moving a component to a safe position. In addition to immediately suspending CORE ALTERATIONS or positive reactivity additions, boration to restore the concentration must be initiated immediately.

In determining the required combination of boration flow rate and concentration, no unique Design Basis Event must be satisfied. The only requirement is to restore the boron concentration to its required value as soon as possible. In order to raise the boron concentration as soon as possible, the operator should begin boration with the best source available for unit conditions. The "equivalent to" statement in the Action is a provision providing an alternate method of emergency boration via the RWST at an increased flow rate to account for the lower boron concentration within the RWST.

Once actions have been initiated, they must be continued until the boron concentration is restored. The restoration time depends on the amount of boron that must be injected to reach the required concentration.

SEABROOK - UNIT 1 B 3/4 9-2 Amendment No.

Section IV Determination of Significant Hazards for the Proposed Change

IV DETERMINATION OF SIGNIFICANT HAZARDS FOR THE PROPOSED CHANGE License Amendment Request (LAR) 01-10 proposes changes to the Seabrook Station Technical Specifications (TS) to relocate boron concentration limits contained in certain TSs to the Core Operating Limits Report (COLR). Boron concentration qualifies as a cycle-specific parameter limit that is cycle dependent. The affected TS are: 3/4.1.1.1, "Boration Control Shutdown Margin - Tavg Greater Than or Equal To 200'F;" 3/4.1.1.2, "Boration Control Shutdown Margin - Tavg Less Than or Equal To 200'F;" 3/4.1.2.5, "Borated Water Sources - Shutdown;" 3/4.1.2.6, "Borated Water Sources - Operating;" 3/4.1.2.7, "Isolation of Unborated Water Sources;" 3/4.5.1.1, "Accumulators;"

3/4.5.4, "Refueling Water Storage Tank;" and 3/4.9.1, "Boron Concentration."

In addition, LAR 01-10 proposes changes to TS 2.1, "Safety Limits," to relocate Figure 2.1-1, "Reactor Core Safety Limits-Four Loops in Operation," to the COLR; and revise TSs 2.1.1 and 2.1.2 limiting conditions and actions so as to be consistent with the improved Standard Technical Specifications (ITS) - Westinghouse Plants, NUREG-1431, Revision 2.

Also proposed is the relocation of DNB-related parameters, specified in TS 3/4.2.5, to the COLR.

As a result of the above changes, TS 6.8.1.6, Core Operating Limits Report, and associated TS Bases must be revised to be reflective of the above changes. Furthermore, editorial and administrative changes to TS 6.8.1.6, consistent with ITS, are proposed as well.

The proposed changes are either based on (1) NRC Generic Letter 88-16, "Removal of Cycle Specific Parameter Limits from Technical Specifications", dated October 3, 1988; (2) the NRC staff's acceptance of WCAP-14483-A, "Generic Methodology for Expanded Core Operating Limits Report;" and/or (3) the improved Standard Technical Specifications, NUREG-1431, Revision 2.

Relocation of cycle-specific parameters from TS to the COLR (a licensee-controlled document subject to the requirements of TS 6.8.1.6 and the provisions of 10 CFR 50.59) would afford North Atlantic flexibility to revise cycle-specific parameters, in accordance with NRC-approved methodologies, without the need for license amendment submittals.

Specifically, TS 6.8.1.6c requires copies of the COLR to be submitted to the NRC for each reload cycle, including any mid cycle revisions or supplements thereto. Thus resources, both North Atlantic and NRC, would be saved by minimizing and/or eliminating repetitive LAR submittals associated with revising cycle specific parameters.

In accordance with 10 CFR 50.92, North Atlantic has concluded that the proposed change does not involve a significant hazards consideration (SHC). The basis for the conclusion that the proposed changes do not involve a SHC is as follows:

1.

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

The proposed changes to relocate cycle-specific parameters from TS to the COLR are administrative in nature and do not adversely affect accident initiators or precursors nor alter the design assumptions, conditions, and configuration of the facility or the manner in which it is operated. The proposed changes do not alter or prevent the ability of structures, systems

or components to perform their intended function to mitigate the consequences of an initiating event within the acceptance limits assumed in the Updated Final Safety Report.

The subject parameter limits will continue to be administratively controlled in accordance with Technical Specification 6.8.1.6. Specifically, TS 6.8.1.6.c requires the COLR to be submitted to the NRC each reload cycle, including any mid-cycle revisions or supplements.

Therefore, the proposed changes do not involve a significant increase in 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 accident from any previously evaluated.

The proposed changes do not alter the design assumptions, conditions, or configuration of the facility or the manner in which it is operated. There are no changes to the source term or radiological release assumptions used in evaluating the radiological consequences in the Seabrook Station Updated Final Safety Report (UFSAR). The proposed changes have no adverse impact on component or system interactions. The proposed changes will not degrade the ability of systems, structures or components important to safety to perform their safety function nor change the response of any system, structure or component important to safety as described in the Seabrook Station UFSAR. The proposed changes are administrative in nature and do not change the level of programmatic and procedural details that assure the operation of the facility in a safe manner.

Since there are no changes to the design assumptions, parameters, conditions and configuration of the facility, or the manner in which the plant is operated and surveilled, the proposed changes do not create the possibility of a new or different accident from any previously analyzed.

3.

The proposed changes do not involve a significant reduction in the margin of safety There is no adverse impact on equipment design or operation and there are no changes being made to Technical Specification cycle-specific parameter limits themselves that would adversely affect plant safety. The proposed changes are administrative in nature and imposes alternative procedural and programmatic controls on these parameter limits in accordance with the Commission's position established by Generic Letter 88-16 and the Safety Evaluation of WCAP-14483-A. Future changes to these limits will be submitted to the NRC in accordance with Technical Specification 6.8.1.6.

Therefore, relocation of the subject cycle-specific parameter limits and other proposed editorial changes, to be reflective of the relocated parameters, do not involve a significant reduction in the margin or safety provided in the existing specifications.

Based on the above evaluation, North Atlantic concludes that the proposed changes do not constitute a significant hazard.

Sections V & VI Proposed Schedule for License Amendment Issuance and Effectiveness and Environmental Impact Assessment

V PROPOSED SCHEDULE FOR LICENSE AMENDMENT ISSUANCE AND EFFECTIVENESS North Atlantic requests NRC Staff review of License Amendment Request 01-10 and issuance of a license amendment by April 15, 2003, becoming effective immediately and implemented within 90 days thereafter. The requested issuance date is based on NRC average turnaround time for non-outage related LARs.

VI ENVIRONMENTAL IMPACT ASSESSMENT North Atlantic has reviewed the proposed license amendment against the criteria of 1 OCFR51.22 for environmental considerations.

The proposed changes do not involve a significant hazards consideration, nor increase the types and amounts of effluent that may be released off-site, nor significantly increase individual or cumulative occupational radiation exposures.

Based on the foregoing, North Atlantic concludes that the proposed changes meets the criteria delineated in 10CFR51.22(c)(9) for a categorical exclusion from the requirements for an Environmental Impact Statement.

ML021140320

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