ML20198E783

Proposed Tech Specs Pages to LAR 98-09,re Editorial & Administrative Changes to TS

ML20198E783
Person / Time
Site:	Seabrook
Issue date:	12/16/1998
From:	NORTH ATLANTIC ENERGY SERVICE CORP. (NAESCO)
To:
Shared Package
ML20198E775	List: ML20198E770 ML20198E778 ML20198E783
References
NUDOCS 9812240144
Download: ML20198E783 (36)

Text

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I!!D.EJ LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQU SECTION

.P. AGE FIGURE 3.4-1 DOSE EQUIVALENT I-131 REACTOR COOLANT SPECIFIC ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL POWER WITH THE REACTOR COOLANT SPECIFIC DOSE EQUIVALENT I-131............. ACTIVITY > 1pCi/ gram

....................... 3/4 4-28 TABLE 4.4-3 REACTOR COOLANT PR0 GRAM. . . . . . . . . . . . SPECIFIC ACTIVITY SAMPLE AND ANALYSIS 3/4.4.9

....................................... 3/4 4-29 PRESSURE / TEMPERATURE LIMITS General.................................................. 3/4 4-30 FIGURE 3.4-2 REACTOR COOLAN M YSTEM HEATUP L APPLICABLEUPTOgEFPY............ IMITATIONS-

..................... 3/4 4-31 1 FIGURE 3.4-3 REACTOR COOLA YSTEM C00LDOWN L APPLICABLE UP TO EFPY.............. ................... IMITATIONS - 3/4 4-32

//. l Pressurizer.

Overpressure.............................................

Protection Systems.......................... 3/4 4-33 3/4 4-34

,._ FIGURE 3.4-4 T

RCS COLD DVERPRESSURE PR0'ECTION3/4 SETPOINTS...........

4-36 3/4.4.10 STRUCTURAL INTEGRITY....................

3/4.4.11 ................ 3/4 4-37 REACTOR COOLANT SYSTEM VENTS............ ................ 3/4 4-38 3/A.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS '

Hot Standby, Startu Shutdown. . . . . . . . . . . p, and Power Operation. . . . . . . . . . . . . . . . 3/4 5-1 3/4.5.2

...................................... 3/4 5-3 ECCS SUBSYSTEMS - T 3/4.5.3 ECCS SUBSYSTEMS - TavgGREATER LESS THAN350*F,.............F.... THAN OR EQUAL TO 350* 3/4 S-4 avg ..... 3/4 5-8 ECCS Subsystems - T,yg Equal To or Less Than 200'F.......

3/4.5.4 3/4 5-10 REFUELING WATER STORAGE TANK............................. 3/4 5-11 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT Containment Integrit Containment Leakage.y.................................... 3/4 6-1

~ ..................................... 3/4 6-2 SEABROOK - UNIT 1 vi 9e12240144 9s1216 9I PDR ADOCK0500gj43 P gggg gn, '7o

la <

!e i IND11

. . . 6.0 ADMINISTRATIVE CONTROLS A,,

SICIION gig I 6.9 RECORD RETENTION . . . . . . . . . . . . . . . . . . . . . . 6-19 6.10 RADIATION PROTECTION PROGRAM ............... 6-20 6.11 HIGH RADIATION AREA . . . . . . . . . . . . . . . . . . . . 6-20 l

l 6.12 PROCESS CONTROL PROGRAM (PCP) ............... 6-21 i

( 6.13 0FFSITE DOSE CALCULATION MANUAL (nXM) .......... 6-22 6.14 MAJOR CHANGES TO LIOUID. GASEOUS. AM) SOLID l RADWASTE TREATMENT SYSTEMS ................ 6-23 L . I C C o MAlw+>t mmc Mn wn@ Pe%W . .

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er SEABROOK - UNIT I xv AmendmentNo.[

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LIMITING SAFETY SYSTEM SETTINGS fs BASES 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETp0INTS (Continued)

Undervaltaos and UWrfraarr. v - Reactor Coolant P- Busses ,

The Undervoltage and Underfrequency Reactor Coolant Pump Bus trips provide core protection against DNB as a result of complete loss of forced coolant flow. The specified Setpoints assure a Reactor trip signal is generated before the Low Flow Trip setpoint is reached. Time delays are incorporated in the Underfrequency and Undervoltage trips to prevent spurious Reactor trips from momentary electrical power transients. For undervoltage, the delay is set so that the time required for a signal to reach the Reactor trip breakers following the staultaneous trip of two or more reactor coolant pump bus circuit breakers shall not exceed 1.5 seconds. For underfrequency, l the b.re.

trip delay is set so that the time required for a signal to reach the Reactor exce s after the Underfrequency Tri Setpoint is reached shall not 0g Reactor second. On decreasing power t Undervoltage and Underfrequency .,

1 oolant Pump sus trips are automatically blocked by p-7 a power level of approximately 105 of RATED THERMAL POWER with a turbine impuls(e chamber pressure at approximately 105 of full power equivalent); and on increasing reinstated automatically by p-7. power, the Undervoltage and Underfrequency Reactor Turbine Trio A Turbine trip initiates a Reactor trip. On decreasing power, the Reactor trip from the Turbine trip is automatically blocked by P-g (a power level of approximately 205 of RATED THERMAL POWER); and on increasing power, the Reactor trip from the Turbine trip is reinstated automatically by P-9.

Safety Infection Innut from ESF System instrumentation, the ESF automatic actuation logic channels will initiate a Reactor trip upon any signal which initiates a Safety Injection.

The ESF instrumentation channels that initiate a Safety Injection signal are shown in Table 3.3-3.

Reactor Tria System Interlocke The Reacter Trip System interlocks perfore the following functions:

P-6 On incrwasin Range trip (g power, P-6 allows the manual block of the Sourcei.e., p On decreasing power, Source Range Level trips are automatically reactivated and high voltage is restored.

.s SEA 8 ROOK - UNIT I B 2-7 Amendment No.

POWER DISTRIBUTION LI'MITS

- 3/4.2.4 OUADRANT POWER TILT RATIO LIMITING CONDITION FOR OPERATION 3.2.4 The QUADRANT POWER TILT RATIO shall not exceed 1.02.

APPLICABILITY: MODE 1, above 50% of RATED THERMAL POWER *.

ACTION:

With the QUADRANT POWER TILT RATIO determined to exceed 1.02:

a.

Within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> reduce THEPr.AL POWER at least 3% from RATED THERMA POWER for each 1% of indicated QUADRANT POWER TILT RATIO in exces of I and similarly reduce the Power Range Neutron Flux-High Trip Setpoints within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

A.e

b. '

Within F

24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and every 7 days thereafter, verify that F (Z) and 4l.1.are within 2.2 and their limits by performing Surveillance Requirements 4.2.3.2. f THERMAL POWER and setpoint reductions shall /

then be in accordance with the ACTION statements of Specifications 3.2.2 and 3.2.3.

SURVEILLANCE REOUIREMENTS 4.2.4.1 ~

The QUADRANT POWER TILT RATIO sh'all be determined 'to be within the limit above 50% of RATED THERMAL POWER by:

a. Calculating the ratio at least once per 7 days when the alarm is OPERABLE, and b.

Calculating the ratio at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> during steady-state operation when the alarm is inoperable.

4.2.4.2 The QUADRANT POWER TILT RATIO shall be determined to be within the limit when above 75% of RATED THERMAL POWER with one Power Range channel inoperable by using the Incore Detector System to confirm indicated QUADRANT POWER TILT RATIO at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by either:

a. Using the four pairs of symmetric detector locations or

b. Using the Incore Detector System to monitor the QUADRANT POWER TILT RATIO subject to the requirements of @ctf4eek3.3.3.2.

l Teepiscrt. RcquetmeoT lk2.o3

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cSee Special Test Exceptions Specification 3.10.2 SEABROOK - UNIT 1 3/4 2-9 AmendmentNo.)

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m.Q. POWER DISTRIBUTION LIMITS y.. .

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BASES I

• 3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR and NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR (Continued)

F" above are, will be maintained within its limits provided Conditions a. through d.

maintained. The design limit DNBR includes margin to offset any rod bowI penalty. Margin is also maintained between the safety analysis limit DNBR and th~e design limit DNBR. This margin is available for plant design flexibility.

When an F measurement is taken, an allowance for both measurement error and manufacturing t,olerance must be made. An allowance of 5% is appropriate for a full-core map taken with the movable incore detectors, while 5.21% is. appropriate for surveillance results determined with the fixed incore detectors.. A 3%

allowance is appropriate for manufacturing tolerance.

For operation with the Fixed Incore Detector System (FIDS) Alarm OPERABLE, {

the cycle-dependent normalized axial peaking factor, K(I), specified in COLR accounts for axial power shape sensitivity in the LOCA analysis. Assurance that the Fo (Z) limit on Specification 3.2.2 is met during both normal operation and in the event of xenon redistribution following power changes is provided by the FIDS Alarm through the plant process computer. This assures that the consequences of a f.e LOCA would be within specified acceptance criteria.

For operation with the FIDS Alarm inoperable, the cycle-dependent normalized axial peaking factor, K(Z), specified in COLR accounts for possible xenon redistribution following power changes in addition to axial power Mape sensitivity in th9 LOCA analysis. This assures that the conseque..es of a LOCA i would be within specified acceptance criteria. ,

When RCS F" is measured, no additional allowances are necessary prior to comparisonwithSeestablishedlimit. A bounding measurement error of 4.13% for F", has been allowed for in determination of the design DNBR value.

3/4.2.4 OUADRANT POWER TILT RATIO g The purpose of this specification is to detect gross changes 'n core power distribution between monthly Incore Detector System surveillance During normal operation the QUADRANT POWER TILT RATIO is set equal to(Tgfdhce acceptability of1 core peaking factors has been established by review of incore surveillances. The limit of 1.02 is established as an indication that the power distribution has changed enough to warrant further investigation.

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

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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 J;he transient and accident analyses.fThe ipn"its are onsistent ith

[

, al FSAT ssumptipos anftgve been palytical)y demon ated adeq te to mainta aminiumDNBRff1.30 hroughoutAach analfzed tra ient. Op ating proc ures in ude all ances f r measure 4ent and hdicatio uncertain so tha the imits o 594.3*F or T, nd 2205 sig for essuriz pressure re not l e eeded.

k T nt err of 2.4% for RCS otal fio rate is bp ed up er-formin a preci d using e resul to normal ze the flow rate, indicator /smeasureUn heat / alanceich migh not b r

. Poterytial foul ng of th feedwatpfventuri det cted coul bias t b result rom the ecision/heat bala in a gencon-se ative m ner. refore, penalty of 0.1% f for undete ed foul fig of t f edwater fnturi applie Any fogling wh h might bi the R flow te easurem great than 0 %canb/detectedpbymonitoringand ending ario plant p ormanc paramet rs. Iffetected,/actionsha}Tlbeta n befo

{ perfo ng sub quent prlcision dat balande measurem6nts, i. ., eit r the -

I effe of the fouling f all be antified'and comp sated f in th RCS ow' rat measur ent or e ventu shall tyd cleaned eliming e the oulir3 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 pecified limit.

I gg 33 M p h b W-

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SEABROOK - UNIT 1 B 3/4 2-4 AmendmentNo.)4

,- r POWER DISTRIBUTION LIMITS BASES 3/4.2.5 DNB PARAMETERS

-The44mits-en-the-DNB-related-par-ameters-assure-that-each-ef-the par-ameters-

-i: ?.:intained-within the remal stead"-st:te envelep:-Of Operat4en-as:tred in_j;"

d transient-and_eccident enelysey [ TFe' limits are. consistent with the updated FSAR k 4YsGiiiptions and have been analytically demonstrated adequate to assure compliance with acceptance criteria for each analyzed transient. Operating procedures-include allowances for measurement and . Jication uncertainty so that the limits

) of 594.3*F for T,, and 2185 psig for pressurizer pressure are not exceeded. )

j 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 arilysis of DNB related events when the Revised (Thermal Design Procedure (RTDP) methodology is utilized.

--The-12-heur-per-iedio-survetMance-of-these-parameters-through4nstru :nt-

-r-eadou t-i s-suff4 ci ent-to-en su re-that 'the-paramet e r: :r: r::ter:d within their

-Hmits-fellowing-lead-ch:nge: :nd other exp:04ed tr:r.:icut Op: ration.

-The-per4edic :urveill:nce cf indicatM ".CS #1ce 1: zuf#istent te deteet only

-flow-degradat4en-which-could4ead-to-eperat-ten-out:1de the :p:04ficd 'imW p p 4 M M P cq L M /f + >M M e M e M cnR 33- / P od L AR. 33 -2 0-

& 0 + P * *""'

SEABROOK - UNIT 1 B 3/4 2-4 Amendment No.,33'

1 REACTOR COOLANT SYSTEM j

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BASES 3/4.4.8 SPECIFIC ACTIVITY (Continued)

The sample analysis for determining the gross specific activity and E can exclude the radioiodines because of the low reactor coolant limit of 1 microcurie / gram DOSE EQUIVALENT I-131, and because, if the limit is exceeded, the radioiodine level is to be determined every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. If the gross specific activity level and radiciodine level in the reactor coolant were at their limits, the radioiodine contribution would be approximately 1L In a release of reactor coolant with a typical mixture of radioactivity, the actual radio-iodine contribution would probably be about 20L The exclusion of radio-nuclides with half-lives less than 10 minutes from these determinations has been made for several reasons. The first consideration is the difficulty to identify short-lived radionuclides in a sample that requires a significant time to collect, transport, and analyze. The second consideration is the predictable delay time between the postulated release of radioactivity from the reactor coolant to its release to the environment and transport to the SITE BOUNDARY, which is relatable to at least 30 minutes' decay time. The choice of 10 minutes for the half-life cutoff was made because of the nuclear characteristics of the typical reactor coolant radioactivity. The radio-nuclides in the typical reactor coolant have half-lives of less than 4 minutes or half-lives of greater than 14 minutes, which allows a distinction between the radionuclides above and below a half-life of 10 minutes. For these reasons the radionuclides that are. excluded from consideration are expected to decay c

to very low levels before they could be transported from the reactor coolant to the SITE BOUNDARY under any accident condition.

q upo theaboveconsi'derations%rexclu'd'ngcertknradionclidesf m thb sample nalysia the allbwable timg of 2 ho rs between sample aking an com\p ting t e initi 1 analysi is based upon a pical tKne necess ry to pe - t form esampting,tr&qsportthesample, nd perfo <n the anglysis of bout 90 min s. Af ter 90 m'4 utes, th gross coynt shout ( be madh in a rep oducible ometry f sampl and co nter having reprod'ugible beta or gadboa self-s ielding p perties. The c nter s uld be re' set to aNeproducit le effikency ve us s t energy. It q not n essary to identify specif14 nuclidy. The diochem' cal determination e sampl within ypical bf nucl des sis huntin sho (f6dowing be based s on mult'iple sam ling counting ofsless thh 1 hou of about 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />,about1%,gabo(1 wee g andabo(1 month -

Reducing T,yg to less than 500*F prevents the release of activity should a steam generator tube rupture, since the saturation pressure of the reactor coolant is below the lift pressure of the atmospheric steam relief valves.

The Surveillance Requirements provide adequate assurance that excessive specific activity levels in the reactor coolant will be detected in sufficient time to take corrective action. A reduction in frequency of isotopic analyses following power changes may be p" issible if justified by the data obtained.

SEABROOK - UNIT 1 B 3/4 4-6 p. h g pur a. N

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3/4.5 EMERGENCY CORE COOLING SYSTEMS

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l BASES l

3/4.5.1 ACCUMULATORS l The OPERABILITY of each Reactor Coolant System (RCS) accumulator ensures that a sufficient volume of borated water will be immediately forced into the reactor core through each of the cold legs in the event the RCS pressure falls below the pressure of the accumulators. This initial surge of water into the core provides the initial cooling mechanism during large RCS pipe ruptures.

The limits on accumulator volume, boron concentration, and pressure ensure that the assumptions used for accumulator injection in the safety analysis are met.

'/he accumulator power-operated isolation valves are considered to be

" operating bypasses" in the context of IEEE Std. 279-1971, which requires that bypasses of a protective function be removed automatically whenever permissive conditions are not met.,s In addition, as these accumulator isolation valves fail to meet single-fail ure criteria, removal of power to the valves is required. Qgg l The limits for operation with an accumulator inoperable for any reason except an isolation valve closed minimizes the time exposure of the plant to a LOCA event occurring concurrent with failure of an additional accumulator which may result in unacceptable peak cladding temperatures. If a closed isolation valve cannot be immediately opened, the full capability of one accumulator is not available and prompt action is required to place the reactor in a mode where this capability is not required.

3/4.5.2 and 3/4.5.3 ECCS SUBSYSTEMS The OPERABILITY of uo independent ECCS subsystems ensures that sufficient emergency core cooling capability will be available in the event of a LOCA assuming the loss of one subsystem through any single-failure consideration.

Either subsystem operating in conjunction with the accumulators is capaele of supplying sufficient core cooling to limit the peak cladding temperatures within acceptable limits for all postulated break sizes ranging from the double ended break of tha largest RCS cold-leg pipe downward. In addition, each ECCS subsystem provides long-term core cooling capability in the recirculation mode during the accident recovery period.

With the RCS temperature below 350*F, one OPERABLE ECCS subsystem is acceptable without single failure consideration on the basis of the stable reactivity condition of the .sactor and the limited core cooling requirements.

The limitation for a maximum of one centrifugal charging pump to be OPERABLE and the Surveillance Requirement to verify all charging pum'ps and safety injection pumps except the required OPERABLE charging pump to be in-operable in MODES 4 and 5 and in MODE 6 with the reactor vessel he'ad on pro-vides assurance that a mass addition pressure transient can be reli'eved by the operation of a single PORV or RHR suction relief valve.

i SEABROOK - UNIT 1 B 3/4 5-1 kf l k m A mcur M O.

INSERT In Modes 1,2,3, and in Mode 4 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> of entry into Mode 3 from 4, the accumulator isolation valves are open with their power removed whenever pressurizer pressure is greater than 1000 psig.

ADMINISTRATIVE CONTROLS 6.2.3 INDEPENDENT TECHNICAL REVIEWS S A TechnicaltheReview encompass Program following shall Technical be established,ibilities.

Review respons implemented and maintained FUNCTION 6.2.3.1 The Technical Review Program responsibilities shall encompass:

a. NRC issuances industry advisories, Licensee Event Reports, and other sourcesthatmayIndicateareasforimprovingplantsafety- '

I

b. Internal and external operating e indicate areas for improving plant sagperience ety; information that may

c. Plant operating characteristics plant operations, hat thesemodifications, endently t maintenance and surveillance activities are performed safel to verify inde!y and that human errors are reduced as much as practical, yandand correct

d. Making detailed recommendations to the Senior Site Official for procedure revisions equipment modifications or other means of improving nuclear safety and lant reliability.

The Technical Review Program shall utilize several on-site personnel who are independent of the plant management chain to perform the reviews.

RECORDS 6.2.3.2 Written records of technical reviews shall be maintained. As a

' minimum these records shall include the results of the activities conducted, t'e n status of recommendations made pursuant to Specification 6.2.3.1 and an assessment of company operations related to the reviews performed. A copy of the monthly Technical Review Program report shall be provided to the Senior Site Official.

OVALIFICATIONS 6.2.3.3 Personnel performing reviews pursuant to Technical Specification 6.2.3.1 shall have either a bachelor's degree in engineering or related science a and at least 2 years professional level experience, at least 1 year of which shall be in the nuclear field, or equivalent education and experience as f -

defined in ANSI /ANS 3.1, 1981, Section 4.1.

6.2.4 SHIFT TECHNICAL ADVISOR 6.2.4.1 The Shift Technical Advisor shall provide advisory technical support to the Control Room Commander in the areas of thermal hydraulics, reactor engi-neering, and plant analysis with regard to the safe operation of the station.

6.3 TRAININ W@FG770N MWA G STI USEh {

6.3.1 A tra11iMg and eplac rstpf)sh 1 be intal d un r the nt liegnsed d rection trgining,Jaininp/ogram f the fiana rf6r and th stat shal()nN me6t or exceed he re irem ts and ecommen tions # Sect 5. of A I Kl8.1 971.a the s pple tal re iremen speciffed in REG-) 21, ds 11 incl de famyfiariz ion wJth relevant ind ryopefationalexpe/ience .

SEABROOK - UNIT 1 6-5 Amendment No.

ADMINISTRATIVF CONTR01 S 6.4.1.7 The SDRC shall:

a.

Recomend in writing to the Station Director approval or disapproval of items considered under Specification 6.4.1.6a. through d:  ;

4 b.

Render determinations in writing with regard to whether or not each item considered under Specification 6.4.1.6a @iF5s5fiXTconstitutes an unreviewed safety question: and 2 }

g, Q J, c.

Provide written notification within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to the Executive between the 50RC and the Station Director howe Director shall have responsibility for resolution of such disagreements pursuant to Specification 6.1.1.

RECORDS  !

6.4.1.8 minimum, responsibility document the results of all SORC activities pe provided to the. provisions of these Technical Specifications. Copies shall be c) o 6.4 2 Executive Vice President & Chief Nuclear Officer{and t STATION DUAL 1FIFD RFVIFWFR PROGRAM FUNCTION

,- . m ?.

whereby required reviews of designated procedures o required and by Sacification 6.4.1.6.a are performed by Station Qualified Reviewers revi. ews approved by the by'the 50RC. designated department heads. These reviews are in lieu of i evaluation must be reviewed by the 50RC.However, procedures which require RFSPONSIBIlITIES '

6.4.2.2 The Station Qualified Reviewer Program shall:

a.  !

Provide for the review of designated procedures, programs, and changes thereto by a Qualified Revie who prepared the procedure, program,wer(s) or change. other than the individual

b.  !'

Provide for cross-disciplinary review of procedures, programs, and changes thereto when organizations other than the pr

~ organization are affected by the procedure, program,eparing or change.

c. l Ensure cross-disciplinary reviews are performed by a Qualified by cognizant department heads as having spec to assess a particular procedure, program or change. Cross-disciplinary reviewers may function as a comittee.

~.S . .

I SEABROOK - UNIT 1

, 6-8 AmendmentNo.44.[

. ,. _ _ - - - l ADMfNTSTRATIVE CONTR01 S

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d. Provide for a screening of designated procedures programs and

-.. changes thereto to determine if an evaluation should be performed in i e, . 'mM ' accordance with the provisions of'10 CFR 50.59 to verify that an unreviewed safety question does not exist. This screening will be grueumy 1 trained and qualified in performing 10 CFR e.'

Provide for written recomendation by the Qualified Reviewer (s) to .V the responsible department head for approval or disapproval of procedures and programs considered under Specification 6.4.1.6a and

/h i V) l that the procedure or program was screened by a qualified individual and found not to require a 10 CFR 50.59 evaluation.

6.4.2.3 If the responsible departmeht head determines that a new arogram. I procedure. or change thereto requires a 10 CFR 50.59 evaluation.'tiat designated ^

department head will ensure the required evaluation is performed to determine if the new procedure, Tne new procedure, program, or change involves an unreviewed safety question, y1 program, or change will then be forwarded with the 10 CFR '

50.59 evaluation to 50RC for review. -

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6.4.2.4 Personnel recomended to be Station Qualified Reviewers shall be designated in writing by the Station Director for each procedure, program, or class of procedure or program within the scope of the Station' Oualified Reviewer h' Program.

6.4.2.5 Temporary procedure changes shall be made in accordance with Specification 6.7.3 with the exception that chang'es to procedures for which

1. 1 reviews are assigned to Qualified Reviewers will be reviewed and approved as

=< described in Specification 6.4.2.2. ,

RECORDS 6.4 2.6 The review of procedures and programs performed under the Station Qualified Reviewer Program shall be documented in accordance with administrative

• procedures.

TRAINING AND OUALIFICATION 6.4.2.7 The training and qualification requirements of personnel designated as a Qualified Reviewer in accordance with the Station Qualified Reviewer Program shall be in accordance with administrative procedures. Qualified reviewers shall have:

a. A Bachelors degree in engineering, related science or technical discipline. and two years of nuclear power plant experience:

OR

b. Six years of nuclear power plant experience:

j'f ,

c. An ecuivalent combination of education and experience as approved by the cesignated department head. v

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SEABROOK - UNIT 1 6-BA .

Amendment No. 34. ss I

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ADMINISTRATIVF CONTROLS

. .7.p.  :., RECORDS w .. .,

6.4.3.9 indicated below; Records of NSARC activities shall be prepared and distributed as a.

Minutes of each NSARC meeting shall be prepared and forwarded to days followin.g each meeting; Executive Vice President J &

b.

Reports of reviews encompassed by Specification 6.4.3.7 shall be I included in the minutes where applicable or fomarded under separate 30 working days following completion of the review: andcov c.

Audit reports encompassed by Specification 6.4. 8 shall be 1 forwarded to the Executive Vice President & Chief Nuclear Offic and to the management positions responsible for the areas audited '

within organization. 30 days after conpletion of the audit by the auditing 6.5 RFPORTABLF FVFNT ACTION The following actions shall be taken for REPORTABLE' EVENTS:

a.

The Comission shall be notified and a report submitted pursuant to

. , ):li the requirements of Section 50.73 to 10 CFR Part 50, and b.

Each REPORTABLE EVENT shall be reviewed by the SORC and the results of this review shall be submitted to the NSARC and the Executive Vice President & Chief Nuclear Officer.

S.6 SAFFTY t THTT VI0l ATION i

violated: The following actions shall be taken in the event a Safety Limit is a.

The NRC Operations Center shall be notified by telephone as soon as J

>ossible and in all cases within I hour. The Executive Vice 3

) resident within 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />s: & Chief Nuclear Officer and the NSARC shall b e notified i j V

b.

A Safety Limit Violation Re) ort shall be prepared. The report shall be reviewed by the SORC. T1is report shall describe: (1) applicable circumstances preceding the violation. (2) effects of the v1olation

- upon facility components, systems action taken to prevent recurrence,:or structures, and (3) corrective c.

The Safety Limit Violation Report shall be submitted to the Comission.

Nuclear Officer the NSARC.

within 14 daysand theviolation:

of the Executiveand Vice President & ' Chief

.f.r d.

"., 0>eration tle Commission. of the station shall not be resumed until authorized by r

SEABROOK - UNIT 1 6-11 Amendment No. 34

ADMINISTRATIVE CONTROLS 6.8.1.6.a. (Continued)

5. Shutdown Rod Insertion limit for Specification 3.1.3.5,

6. )

Control Rod Bank Insertion limits for Specification 3.1.3.6,

7. AXIAL FLUX ' DIFFERENCE limits for Specification 3.2.1,

8. Heat Flux Hot Channel Factor, F"'l and K(Z) for Specification 3.2.2,

9. Nuclear Enthalpy Rise Hot Channel Factor, and F"'[, for Specificktion 3.2.3.

[

The CORE OPERATING LIMITS REPORT shall be maintained a~vailable in the Control Room.

[

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

3 g

1. (Nonproprietary),

WCAP-10266-R-A, "The !,981 Rev. 2 with Version of theAddenda ECCS Evalua andfy (Proprietary)

Westinghouse WCA Model Using the BASH Code",Qst,1986f mucy, t989 Methodology for Specification:

3.2.2 -

Heat Flux Hot Channel Factor

2. WCAP-10079-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-1363-A, "CASMO-3G Validation," April 1988.

YAEC-1659-A, " SIMULATE-3 Validation and Verification," September 1988.

L Methodology for Specifications: '

3.1.1.1 -

SHUTDOWN MARGIN for M00ES 1, 2, 3, and 4 3.1.1.2 SHUTDOWN MARGIN for MODE 5  !

3.1.1.3 -

Moderator' Temperature Coefficient t 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 j 3.2.3 -

Nuclear Enthalpy Rise Hot Channel Factor f  ;

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 (

SEABROOK - UNIT 1 6-18A Amendment No. I

ADMINISTRATIVE CONTROLS

_, 6.8.1.6.b. (Continued)

~ .

10. YAEC-1855PA,"Seabrook Station Unit 1 Fixed Incore Detector System '

4 Analysis " October 1992 Methodology for Specification:

3.2.1 - AXIAL FLUX DIFFERENCE l 3.2.2 - Heat Flux Hot Channel Factor l l

3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor l l 11. YAEC-1624P " Maine Yankee RPS Setpoint Methodology Using Statistical Combination of Uncertainties - Volume 1 - Prevention of Fuel Centerline Melt," March 1988

, 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

12. NYN-95048. Letter from T. C. Feigenbaum (NAESCo) to NRC. " License Amendment Request 95-05: Positive Moderator Temperature Coefficient". May 30, 1995  ;

l Methodology for Specification.  ;

, 3.1.1.3- Moderator Temperature Coefficient ~

^

13. WCAP-12610-P-A " VANTAGE + Fuel Assembly Reference Core Report". /

April 1995. (Westinghouse Proprietary) [

Methodology for S>ecification:

3.2.2- Heat lux Hot Channel Factor 6.8.1.6.c. The core operating limits shall be determined so that all a)plicable limits (e.g. , fuel thermal-mechanical limits, core tiermal-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 l.IMITS 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.

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@s SEABROOK - UNIT 1 6-18C AmendmentNo.[

. ---4 - - _

I SECTION III Retype of Proposed Changes The attached retype reflects the currently issued version of the Technical Specifications. Pending i Technical Specification changes or Technical Specification changes issued subsequent to this i submittal are not reflected in the enclosed retype. The enclosed retype should be checked for i continuity with Technical Specifications prior to issuance.

1 Page 8

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION A fLAGE FIGURE 3.4-1 DOSE EQUIVALENT I-131 REACTOR COOLANT SPECIFIC ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL POWER WITH THE REACTOR COOLANT SPECIFIC ACTIVITY > 1pCi/ gram DOSE EQUIVALENT I-131 .. . .. . . 3/4 4-28 TABLE 4.4-3 REACTOR COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM . . . ... ... . . . 3/4 4-29 3/4.4.9 PRESSURE / TEMPERATURE LIMITS General . . . ... . .. .. . 3/4 4-30 FIGURE 3.4-2 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS -

APPLICABLE UP TO 11.1 EFPY . .. 3/4 4-31 l FIGURE 3.4-3 REACTOR COOLANT SYSTEM C00LDOWN LIMITATIONS -

APPLICABLE UP TO 11.1 EFPY . .... .. 3/4 4-32 l Pressurizer . . . . . . . 3/4 4-33 Overpressure Protection Systems .. 3/4 4-34 FIGURE 3.4-4 RCS COLD OVERPRESSURE PROTECTION SETPOINTS .. . 3/4 4-36 3/4.4.10 STRUCTURAL INTEGRITY ...... . 3/4 4-37 3/4.4.11 REACTOR COOLANT SYSTEM VENTS . ... ... . 3/4 4-38 3/4.5 EMERGENCY CORE COOLING SYSTEMS l 3/4.5.1 ACCUMULATORS Hot Standby, Startup, and Power Operation .. . 3/4 5-1 Shutdown .. .. ..... . . . 3/4 5-3 3/4.5.2 ECCS SUBSYSTEMS - T ,, GREATER THAN OR EQUAL TO 350 F . 3/4 5-4 3/4.5.3 ECCS SUBSYSTEMS - T,,, LESS THAN 350 F .. ... . 3/4 5-8 ECCS Subsystems - T Equal To or Less Than 200 F 3/4 5-10 3/4.5.4 REFUELINGWATERSTOMGETANK .. .... .. 3/4 5-11 I

3/4.6 CONTAINMENT SYSTEMS 1

3/4.6.1 PRIMARY CONTAINMENT l

Containment Integrity ....... .. 3/4 6-1 l Containment Leakage . . . ..... . .. 3/4 6-2 i

SEABROOK - UNIT 1 vi Amendment No.

4 ,

INDEX 6.0 ADMINISTRATIVE CONTROLS SECTION PAGE 6.9 PECORD RETENTION . . . . . . . . .. . .. 6-19

- 6.10 RADIATION PROTECTION PROGRAM . . ..... . 6-20 6.11 HIGH RADIATION AREA . . . . . ... .... 6-20 6.12 PROCESS CONTROL PROGRAM (PCP) ... . . 6-21 6.13 0FFSITE DOSE CALCULATION MANUAL (0DCM) . . . 6-22 6.14 MAJOR CHANGES TO LIQUID. GASEOUS. AND SOLID RADWASTE TREATMENT SYSTEMS . ... . . .. 6-23 6.15 CONTAINMENT LEAKAGL RATE TESTING PROGRAM ,.. . 6-24 I

SEABROOK - UNIT 1 xv Amendment No. 64,

=

t.

LIMITING SAFETY SYSTEM SETTINGS l

BASES 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS (Continued)

Undervoltage and Underfreauency - Reactor Coolant Pumo Busses The Undervoltage and Underfrequency Reactor Coolant Pump Bus trips provide core protection against DNB as a result of complete loss of forced coolant flow.

The specified Setpoints assure a Reactor trip signal is generated before the Low ,

l Flow Trip Setpoint is reached. Time delays are incorporated in the Underfrequency l and Undervoltage trips to prevent spurious Reactor trips from momentary electrical '

power transients. For undervoltage. the delay is set so that the time required for a signal to reach the Reactor trip breakers following the simultaneous trip i

, of two or more reactor coolant pum) bus circuit breakers shall not exceed 1.5 ,

t seconds. For underfrequency, the celay is set so that the time required for a l l signal to reach the Reactor trip breakers after the Underfrequency Trip Setpoint is reached shall not exceed 0.6 second. On decreasing power the Undervoltage and l l Underfrequency Reactor Coolant Pump Bus trips are automatically blocked by P-7 (a power level of approximately 10% of RATED THERMAL POWER with a turbine impulse chamber pressure at approximately 10% of full power equivalent): and on increasing power, the Undervoltage and Underfrequency Reactor Coolant Pump Bus trips are l reinstated automatically by P-7. '

i Turbine Trio A Turbine trip initiates a Reactor trip. On decreasing power, the Reactor trip from the Turbine trip is automatically blocked by P-9 (a power level of approximately 20% of RATED THERMAL POWER): and on increasing power, the Reactor trip from the Turbine trip is reinstated automatically by P-9.

1 Safety Iniection Inout from ESF '

If a Reactor trip has not already been generated by the Reactor Trip System l instrumentation the ESF automatic actuation logic channels will initiate a Reactor trip upon any signal which initiates a Safety Injection. The ESF instrumentation channels that initiate a Safety Injection signal are shown in l

Table 3.3-3.

Reactor Trio System Interlocks The Reactor Trip System interlocks perform the following functions:

P-6 On increasing power. P-6 allows the manual block of the Source Range trip (i.e. prevents premature block of Source Range trip). On trips are automatically decreasing power. Source Range Level reactivated and high voltage is restored.

i l

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SEABROOK - UNIT 1 B 2-7 Amendment No. 34

POWER DISTRIBUTION LIMITS l l 3/4.2.4 OUADRANT POWER TILT RATIO l

l l LIMITING CONDITION FOR OPERATION I l

l 3.2.4 The OUADRANT POWER TILT RATIO shall not exceed 1.02.

l APPLICABILITY: MODE 1. above 50% of RATED THERMAL POWER *. l l

ACTION:

With the OUADRANT POWER TILT RATIO determined to exceed 1.02:

a. Within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> reduce THERMAL POWER at least 3% from RATED THERMAL POWER for each 1% of indicated QUADRANT POWER TILT RATIO in excess of 1 and similarly reduce the Power Range Neutron Flux-High Trip Setpoints within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

b. Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and every 7 days thereafter, verify that F (Z) a and FL l are within their limits by performing Surveillance Requirements 4.2.2.2 and 4.2.3.2. THERMAL POWER and setpoint reductions shall then be in accordance with the ACTION statements of Specifications 1

3.2.2 and 3.2.3.

SURVEILLANCE REQUIREMENTS l-1 4.2.4.1 The QUADRANT POWER TILT RATIO shall be determined to be within the limit above 50% of RATED THERMAL POWER by:

a. Calculating the ratio at least once per 7 days when the alarm is OPERABLE. and

b. Calculating the ratio at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> during steady-state operation when the alarm is inoperable.

4.2.4.2 The QUADRANT POWER TILT RATIO shall be determined to be within the limit when above 75% of RATED THERMAL POWER with one Power Range channel inoperable by using the Incore Detector System to confirm indicated QUADRANT POWER TILT RATIO at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by either:

l a. Using the four pairs of symmetric detector locations or l b. Using the Incore Detector System to monitor the OUADRANT POWER TILT ,

RATIO subject to the requirements of Technical Requirement TR20-3.3.3.2.

l

• See Special Test Exceptions Specification 3.10.2 SEABROOK - UNIT 1 3/4 2-9 Amendment No. N. 33.

-l l

!' POWER DISTRIBUTION LIMITS I BASES 3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR and NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR (Continued) l FL will be maintained within its limits provided Conditions a. through

d. above are maintained. The design limit DNBR includes margin to offset any rod bow penalty. Margin is also maintained between the safety analysis limit l DNBR and the design limit DNBR. This margin is available for plant design flexibility.

When an F a measurement is taken, an allowance for both measurement error and manufacturing tolerance must be made. An allowance of 5% is appropriate l for a full-core map taken with the movable incore detectors, while 5.21% is .

appropriate for surveillance results determined with the fixed incore detectors. A 3% allowance is appropriate for manufacturing tolerance.

For o)eration with the Fixed Incore Detector System (FIDS) Alarm OPERABLE. t1e cycle-dependent normalized axial peaking factor KCZ). specified in COLR accounts for axial power shape sensitivity in the LOCA analysis.

Assurance that the Fa (Z) limit on Specification 3.2.2 is met during normal operation and in the event of xenon redistribution following power changes is provided by the FIDS Alarm through the plant process computer. This assures that the consequences of a LOCA would be within specified acceptance criteria.

! For operation with the FIDS Alarm inoperable. the cycle-dependent normalized axial peaking factor. K(Z), specified in COLR accounts for possible xenon redistribution following power changes in addition to axial power shape l sensitivity in the LOCA analysis. This assures that the consequences of a LOCA would be within specified acceptance criteria.

When RCS F" comparison with die is measured, establishedno additional limit. allowances A bounding are necessary measurement frior to error o 4.13%

for FL has been allowed for in determination of the design DNBR value.

3/4.2.4 OUADRANT POWER TILT RATIO i

The purpose of this specification is to detect gross changes in core power distribution between monthly Incore Detector System surveillances.

During normal operation the QUADRANT POWER TILT RATIO is set equal to 1.0 once acceptability of core peaking factors has been established by review of incore surveillances. The limit of 1.02 is established as an indication that the power distribution has changed enough to warrant further investigation. I l

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SEABROOK - UNIT 1 B 3/4 2-3 Amendment No. 9. 12, 27. 33.

I.

• 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 of 594.3*F for Tm and 2185 psig for pressurizer pressure are not exceeded.

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 i 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 l utilized.

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.

I l

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i SEABROOK - UNIT 1 B 3/4 2-4 Amendment No. 9. 12. 33,31,

i O n

REACTOR COOLANT SYSTEM i

l BASES 3/4.4.8 SPECIFIC ACTIVITY (Continued) l The sample analysis for determining the gross specific activity and E can i exclude the radiciodines because of the low reactor coolant limit of 1 microcurie / gram DOSE EQUIVALENT I-131 and because. if the limit is exceeded, the radiciodine level is to be determined every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. If the gross specific activity level and radiciodine level in the reactor coolant were at their limits, the radiciodine contribution would be approximately 1%. In a release of reactor coolant with a typical mixture of radioactivity, the actual radio-iodine contribution would probably be about 20%. The exclusion of radio-l nuclides with half-liver less than 10 minutes from these determinations has '

l been made for several reasons. The first consideration is the difficulty to l identify short-lived radionuclides in a sample that requires a significant time to collect, transport, and analyze. The second consideration is the ,

predictable delay time between the )ostulated release of radioactivity from the  !

l reactor coolant to its release to t1e environment and transport to the SITE BOUNDARY. which is relatable to at least 30 minutes' decay time. The choice of 10 minutes for the half-life cutoff was made because of the nuclear

characteristics of the typical reactor coolant radioactivity. The radio- l l nuclides in the typical reactor coolant have half-lives of less than 4 minutes l or half-lives of greater than 14 minutes, which allows a distinction between i the radionuclides above and below a half-life of 10 minutes. For these reasons l the radionuclides that are excluded from consideration are expected to decay to l l very low levels before they could be transported from the reactor coolant to '

l the SITE BOUNDARY under any accident condition.

l .

Reducing T y to less than 500 F prevents the release of activity should a steam generator tube rupture, since the saturation pressure of the reactor coolant is below the lift pressure of the atmospheric steam relief valves. The l Surveillance Requirements provide adequate assurance that excessive specific activity levels in the reactor coolant will be detected in sufficient time to

, take corrective action. A reduction in frequency of isotopic analyses l following power changes may be permissible if justified by the data obtained.

i l

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t SEABROOK - UNIT 1 B 3/4 4-6 Amendment No.

l

3/4.5 EMERGENCY CORE COOLING SYSTEMS BASES 3/4.5.1 ACCUMULATORS The OPERABILITY of each Reactor Coolant System (RCS) accumulator ensures that a sufficient volume of borated water will be immediately forced into the reactor core through each of the cold legs in the event the RCS pressure falls below the pressure of the accumulators. This initial surge of water into the core provides the initial cooling mechanism during large RCS pipe ruptures.

The limits on accumulator volume, boron concentration, and pressure ensure that the assumption: used for accumulator injection in the safety analysis are met.

In MODES 1 and 2. the accumulator power-operated isolation valves are l considered to be " operating bypasses" in the context of IEEE Std. 279-1971, which requires that bypasses of a protective function be removed automatically whenever permissive conditions are not met. In MODES 1. 2. 3. and in MODE 4 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> of entry into MODE 3 from 4. the accumulator isolation valves are open with their power removed whenever pressurizer pressure is greater than 1000 psig. In addition, as these accumulator isolation valves fail to meet single-failure criteria. removal of power to the valves is required. l The limits for operation with an accumulator inoperable for any reason except an isolation valve closed minimizes the time exposure of the plant to a LOCA event occurring concurrent with failure of an additional accumulator which may result in unacceptable peak cladding temperatures. If a closed isolation valve cannot be immediately opened. the full capability of one accumulator is not available and prompt action is required to place the reactor in a mode where this capability is not required.

3/4.5.2 and 3/4.5.3 ECCS SUBSYSTEMS The OPERABILITY of two independent ECCS subsystems ensures that sufficient emergency core cooling capability will be available in the event of a LOCA assuming the loss of one subsystem through any single-failure consideration. Either subsystem operating in conjunction with the accumulators ,

is capable of su) plying sufficient core cooling to limit the 3eak cladding I' temperatures witlin acceptable limits for all postulated breac sizes ranging from the double-ended break of the largest RCS cold-leg pipe downward. In addition, each ECCS subsystem provides long-term core cooling capability in the recirculation mode during the accident recovery period.

With the RCS temperature below 350 F. one OPERABLE ECCS subsystem is  !

acceptable without single failure consideration on the basis of the stable reactivity condition of the reactor and the limited core cooling requirements. ,

I The limitation for a maximum of one centrifugal charging pump to be j OPERABLE and the Surveillance Requirement to verify all charging pumps and j safety injection pumps except the required OPERABLE charging pum) to be in-operable in MODES 4 and 5 and in MODE 6 with the reactor vessel lead on aro-vides assurance that a mass addition pressure transient can be relieved )y the j operation of a single PORV or RHR suction relief valve. j l;

SEABROOK - UNIT 1 B 3/4 5-1 Amendment No. j i

r ,

ADMINISTRATIVE CONTROLS 6.2.3 INDEPENDENT TECHNICAL REVIEWS A Technical Review Program shall be established, implemented and maintained to encompass the fM lowing Technical Review responsibilities.

FUNCTION 6.2.3.1 The Technical Review Program responsibilities shall encompass:

a. NRC issuances, industry advisories, Licensee Event Reports, and other sources that may indicate areas for improving plant safety:

b. Internal and external operating experience information that may  ;

indicate areas for improving plant . safety: '

c. Plant operating characteristics, plant operations, modifications, maintenance and surveillance to verify independently that these activities are performed safel much as practical. and y and correctly and that human errors are reduced as

d. Making detailed recommendations to the Senior Site Official for procedure revisions, equipment modifications or other means of improving nuclear safety and plant reliability.

The Technical Review Program shall utilize several on-site personnel who are independent of the plant management chain to perform the reviews.

RECORDS 6.2.3.2 Written records of technical reviews shall be maintained. As a minimum these records shall include the results of the activities conducted, the status of recommendations made pursuant to Specification 6.2.3.1 and an assessment of com)any operations related to the reviews performed. A copy of the monthly Tec1nical Review Program report shall be provided to the Senior Site Official.

QUALIFICATIONS 6.2.3.3 Personnel performing reviews pursuant to Technical Specification 6.2.3.1 shall have either a bachelor's degree in engineering or related science and at least 2 years professional level experience, at least 1 year of which shall be in the nuclear field, or equivalent education and experience as defined in ANSI /ANS 3.1. 1981. Section 4.1 6.2.4 SHIFT TECHNICAL ADVISOR 6.2.4.1 The Shift Technical Advisor shall provide advisory technical support to the Control Room Commander in the areas of thermal hydraulics reactor engi-neering, and plant analysis with regard to the safe operation of the station.

6.3 TRAINING 6.3.1 (THIS SPECIFICATION NUMBER IS NOT USED)

SEABROOK - UNIT 1 6-5 Amendment No. 34.--36.

ADMINISTRATIVE CONTROLS 6.4.1.7 The SORC shall:

a. Recommend in writing to the Station Director approval or disapproval of items considered under Specification 6.4.1.6a. through d:

b. Render determinations in writing with regard to whether or not each item considered under Specification 6.4.1.6a. , b. and d. constitutes an unreviewed safety question; and

c. Provide written notification within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to the Executive Vice President & Chief Nuclear Officer and the NSARC of disagreement between the SORC and the Station Director however, the Station Director shall have responsibility for resolution of such i disagreements pursuant to Specification 6.1.1.

RECORDS l

1 6.4.1.8 The SORC shall maintain written minutes of each SORC meeting that, at a  !

minimum, document the results of all SORC activities performed under the l responsibility provisions of these Technical Specifications. Copies shall be l provided to the Executive Vice President & Chief Nuclear Officer and the NSARC. 1 6 4.2 STATION OUALIFIED REVIEWER PROGRAM FUNCTION l 6.4.2.1 The Station Director may establish a Station Qualified Reviewer Program whereby required reviews of designated procedures or classes of procedures ,

recuired by S)ecification 6.4.1.6.a are performed by Station Qualified Reviewers '

anc approved )y the designated department heads. These reviews are in lieu of reviews by the SORC However, procedures which require a 10 CFR 50.59 evaluation must be reviewed by the SORC.

RESPONSIBILITIES 6.4.2.2 The Station Qualified Reviewer Program shall:

a. Provide for the review of designated 3rocedures, programs, and changes thereto by a Qualified Reviewer (s) otler than the individual who prepared the procedure, program, or change,

b. Provide for cross-disciplinary review of procedures, programs, and changes thereto when organizations other than the preparing ,

organization are affected by the procedure, program, or change.  !

c. Ensure cross-disciplinary reviews are performed by a Qualified Reviewer (s) in affected disciplines, or by other persons designated by cognizant department heads as having specific expertise required to  ;

assess a particular procedure, program or change. Cross-disciplinary 1 reviewers may function as a committee.

SEABROOK - UNIT 1 6-8 Amendment No. 34,-45,

ADMINISTRATIVE CONTROLS

d. Provide for a screening of designated procedures, programs and changes thereto to determine if an evaluation should be performed in accordance with the provisions of 10 CFR 50.59 to verify that an unreviewed safety question does not exist. This screening will be performed by personnel trained and qualified in performing 10 CFR 50.59 screenings.

l

e. Provide for written recommendation by the Qualified Reviewer (s) to the responsible department head for approval or disapproval of procedures and programs considered under Specification 6.4.1.6a and that the procedure or program was screened by a qualified individual and found not to require a 10 CFR 50.59 evaluation.

6.4.2.3 If the responsible department head determines that a new arogram.

procedure, or change thereto requires a 10 CFR 50.59 evaluation, tlat designated department head will ensure the required evaluation is performed to determine if l the new procedure, l The new procedure, program program or change or change involves will then an unreviewed be forwarded with the safety 10 CFRquestion.

50.59 evaluation to SORC for review.  ;

l 6.4.2.4 Personnel recommended to be Station Qualified Reviewers shall be designated in writing by the Station Director for each procedure, program or class of procedure or program within the scope of the Station Qualified Reviewer Program.

6.4.2.5 Temporary procedure changes shall be made in accordance with Specification 6.7.3 with the exception that changes to procedures for which reviews are assigned to Qualified Reviewers will be reviewed and approved as described in Speci fication 6.4.2.2.

RECORDS 6.4.2.6 The review of procedures and programs performed under the Station Qualified Reviewer Program shall be documented in accordance with administrative procedures.

TRAINING AND OUALIFICATION 6.4.2.7 The training and qualification requirements of personnel designated as a Qualified Reviewer in accordance with the Station Qualified Reviewer Program shall be in accordance with administrative procedures. Qualified reviewers shall have:

a. A Bachelors degree in engineering. related science, or technical discipline, and two years of nuclear power plant experience; OR

b. Six years of nuclear power plant experience:

OR

c. An ecuivalent combination of education and experience as approved by the cesignated department head.

SEABROOK - UNIT 1 6-8A Amendment No. 34.55.

ADMINISTRATIVE CONTROLS RECORDS 6.4.3.9 Records of NSARC activities shall be prepared and distributed as indicated below:

l

a. Minutes of each NSARC meeting shall be prepared and forwarded to Executive Vice President & Chief Nuclear Officer within 30 working days following each meeting:

b. Reports of reviews encompassed by Specification 6.4.3.7 shall be

included in the minutes where applicable or forwarded under separate l cover to the Executive Vice President & Chief Nuclear Officer within 30 working days following completion of the review; and

c. Audit reports encompassed by Specification 6.4.3.8 shall be forwarded l to the Executive Vice President & Chief Nuclear Officer and to the management positions responsible for the areas audited within 30 days after completion of the audit by the auditing organization.

l 6.5 REPORTABLE EVENT ACTION The following actions shall be taken for REPORTABLE EVENTS:

l a. The Commission shall be notified and a report submitted pursuant to the requirements of Section 50.73 to 10 CFR Part 50, and

, b. Each REPORTABLE EVENT shall be reviewed by the SORC and the results of this review shall be submitted to the NSARC and the Executive Vice President & Chief Nuclear Officer.

6.6 SAFETY LIMIT VIOLATION The following actions shall be taken in the event a Safety Limit is violated:

a. The NRC Operations Center shall be notified by telephone as soon as possible and in all cases within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. The Executive Vice President

& Chief Nuclear Officer and the NSARC shall be notified within 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />s:

b. A Safety Limit Violation Re) ort shall be prepared. The report shall

, be reviewed by the SORC. T1is report shall describe: (1) applicable circumstances preceding the violation, (2) effects of the violation upon facility components, systems, or structures, and (3) corrective action taken to prevent recurrence:

c. The Safety Limit Violation Report shall be submitted to the Commission, the NSARC, and the Executive Vice President & Chief Nuclear Officer within 14 days of the violation: and

d. Operation of the station shall not be resumed until authorized by the Commission. j SEABROOK - UNIT 1 6-11 Amendment No. 34,65.

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ADMINISTRATIVE CONTROLS l 6.8.1.6.a. (Continued)

5. Shutdown Rod Insertion limit for Specification 3.1.3.5.

6. Control Rod Bank Insertion limits for Specification 3.1.3.6,

7. AAfAL FLUX DIFFERENCE limits for Specification 3.2.1.

8. Heat Flux Hot Channel Factor. F7 and K(Z) for Specification 3.2.2,

9. Nuclear Enthalpy Rise Hot Channel Factor, and F*n for Specification l 3.2.3.

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The CORE OPERATING LIMITS REPORT shall be maintairied available in the Control Room.

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-10266-P-A, Rev. 2 with Addenda (Pro)rietary) and WCAP-11524-A.

Rev. 2 with Addenda (Nonproprietary). "Tle 1981 Version of the i Westinghouse ECCS Evaluation Model Using the BASH Code". March, 1987.

Methodology for Specification:

j 3.2.2 -

Heat Flux Hot Channel Factor

, 2. WCAP-10079-P-A (Proprietary) and WCAP-10080-A (Nonpro)rietary),

l "NOTRUMP: A Nodal Transient Small Break and General ietwork Code".

l August, 1985 Methodology for Specification:

3.2.2 -

Heat Flux Hot Channel Factor

3. YAEC-1363A -

"CASM0-3G Validation," April 1988. j YAEC-1659-A, " SIMULATE-3 Validation and Verification," September l l 1988.

Methodology for Specifications: 1 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 l

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 l 4. Seabrook Station Updated Final Safety Analysis Report, Section i 15.4.6, " Chemical and Volume Control System Malfunction That Results i 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 i 3.1.1.2 -

SHUTDOWN MARGIN for MODE 5 SEABROOK - UNIT 1 6-18A Amendment No. 9, 33, l

ADMINISTRATIVE CONTROLS 6.8.1.6.b. (Continued) ,

10. YAEC-1855PA. "Seabrook Station Unit 1 Fixed Incore Detector System Analysis." October 1992 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

11. YAEC-1624P. " Maine Yankee RPS Setpoint Methodology U. sing Statistical Combination of Uncertainties - Volume 1 - Prevention of Fuel Centerline Melt." March 1988 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 i 12. NYN-95048. Letter from T. C. Feigenbaum (NAESCo) to NRC. " License Amendment Request 95-05: Positive Moderator Temperature Coefficient".

May 30. 1995 Methodology for Specification:

3.1.1.3- Moderator Temperature Coefficient

13. WCAP-12610-P-A. " VANTAGE + Fuel Assembly Reference Core Report April 1995. (Westinghouse Proprietary)

Methodology for Specification:

3.2.2 -

Heat Flux Hot Channel Factor 6.8.1.6.c. The core o>erating limits shall be determined so that all applicable limits (e.g. , fuel tiermal-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 I thereto, shall be provided upon issuance. to the NRC Document Control Desk with l copies to the Regional Administrator and the Resident Inspector.  ;

i SEABROOK - UNIT 1 6-18C Amendment No. 9. 33. A1.52.

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i Section IV Determination of Significant IIazards for Proposed Changes 1

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• s IV. DETERMINATION OF SIGNIFICANT HAZARDS FOR PROPOSED CHANGES License Amendment Request (LAR) 98-09 proposes several editorial and administrative changes to the Seabrook Station Technical Specifications (TS). The proposed changes are either to revise references and statements that are inaccurate or provide relief from administrative controls which provide insignificant safety benefit.

Index Page vi Figures 3.4-2 and 3.4-3 Index Page xv 6.0 Administrative Controls Bases 2.2.1 Reactor Trip System Instrumentation Setpoints 4.2.4.2b Determination of Quadrant Power Tilt Ratio Bases 3/4.2.4 Quadrant Power Tilt Ratio Bases 3/4.2.5 DNB Parameters Bases 3/4.4.8 Specific Activity Bases 3/4.5.1 Accumulators 6.4.1.7b. SORC Responsibilities 6.4.2.2d. Station Qualified Reviewer Program 6.3.1 Training 6.4.3.9c. Records of NSARC 6.8.1.6.b.1 Core Operating Limits Report )

6.8.1.6.b.10 Core Operating Limits Repon l In accordance with 10CFR50.92, North Atlantic has reviewed the proposed changes and has concluded that they do 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 design basis accidents are not affected by the proposed editorial and administrative changes.

The proposed changes do not change the level of programmatic controls or the procedural details currently in place. The proposed changes do not revise the station design, the response of the station to transients nor the manner in which the station is operated, therefore, these changes have no adverse affect to the safe operation of the station. The proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

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2. The proposed changes do not create the possibility of a new or different kind of accident from any previously analyzed.

l The proposed changes do not alter the design assumptions, conditions, configuration of the facility or the manner in which the plant is operated. There are no changes to the source term, containment isolation or radiological release assumptions used in evaluating the radiological j consequences in the Seabrook Station UFSAR. Existing system and component redundancy is not being changed by the proposed changes. The proposed changes have no adverse affect on component or system interactions. The proposed changes are editorial and administrative in nature and do not change the level of programmatic controls and procedural details associated with the aforementioned technical specifications. Therefore, since there are no changes to the design assumptions, conditions, 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 kind of accident from any previously analyzed.

3. The proposed changes do not involve a significant reduction in a margin of safety.

l There are no changes being made to the Technical Specification safety limits or safety system j settings that would adversely affect plant safety. The changes do not affect the operation of structures, systems or components nor do they introduce administrative changes to plant procedures that could affect operator response during normal, abnormal or emergency situations.

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

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

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i Proposed Schedule for License Amendment Issuance and Effectiveness {

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EnvironmentalImpact Assessment j Page 12

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V. PROPOSED SCIIEDULE FOR LICENSE AMENDMENTISSUANCE AND l EFFECTIVENESS I i

North Atlantic requests NRC Staff review of License Amendment Request 98-09 and issuance of a license amendment by June 15, 1999, having immediate effectiveness and implementation required )

within 90 days.

VI. ENVIRONMENTAL IMPACT ASSESSMENT North Atlantic has reviewed the proposed license amendment against the criteria of 10CFR51.22 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 offsite, 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) and 10CFR51.22(c)(10) for a categorical exclusion from the requirements for an Environmental impact Statement.

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