Text

License Amendment Request 02-10, Administrative Changes to Technical Specification Section 6
	ML030030550
Person / Time
Site:	Seabrook
Issue date:	12/23/2002
From:	Warner M; North Atlantic Energy Service Corp
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	NUREG-1431, Rev 2, NYN-02126
	Download: ML030030550 (53)
	v • d • e

FPL Energy FPL Energy Seabrook Station Seabrook Station P.O. Box 300 Seabrook, NH 03874 (603) 773-7000 December 23, 2002 Docket No. 50-443 NYN-02126 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C. 20555-0001 Seabrook Station License Amendment Request 02-10 "Administrative Changes To Technical Specification Section 6" FPL Energy Seabrook, LLC (FPLE Seabrook) has enclosed herein License Amendment Request (LAR) 02-10.

LAR 02-10 is submitted pursuant to the requirements of 10 CFR 50.90 and 10 CFR 50.4.

LAR 02-10 proposes changes to the Seabrook Station Technical Specifications (TS) Index and TS 6.0, Administrative Controls.

The purpose of LAR 02-10 is to update the Technical Specifications to adopt portions of NUREG-1431, Revision 2 ("Standard Technical Specifications, Westinghouse Plants"). In addition, changes are also proposed in accordance with the guidance in Administrative Letter 95-06 ("Relocation Of Technical Specification Administrative Controls Related To Quality Assurance") and the requirements of 10 CFR 50.36(c)(2)(ii) and 10 CFR 20.

Section I of the LAR provides details of these changes.

The Station Operation Review Committee and the Nuclear Safety Audit Review Committee have reviewed LAR 02-10.

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). FPLE Seabrook requests NRC Staff review of LAR 02-10, and issuance of a license amendment by April 30, 2003 (see Section V enclosed). FPLE Seabrook requests these changes in less than the one year normally afforded for NRC review because the changes are administrative in nature and will afford increased organizational flexibility and efficiency at an earlier date.

U. S. Nuclear Regulatory Commission NYN-02126/Page 2 FPLE Seabrook has determined that LAR 02-10 meets the criterion of 10 CFR 51.22(c)(10) 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, Regulatory Programs Manager, at (603) 773-7194.

Very truly yours, FPL Energy Seabrook, LLC.

Mark E. Warner Site Vice President Seabrook Station cc:

H. J. Miller, NRC Regional Administrator R. D. Starkey, NRC Project Manager, Project Directorate 1-2 G. T. 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

FPL Energy Seabrook Station SEABROOK STATION UNIT 1 FPL Energy Seabrook, LLC submits this License Amendment Request pursuant to 10CFR50.90. The following information is enclosed in support of this License Amendment Request:

S 0

0 0

0 Section I Section H Section mI Section IV Section V 0

Section VI Introduction and Safety Assessment for Proposed Changes Markup of Proposed Changes Retype of Proposed Changes Determination of Significant Hazards for Proposed Changes Proposed Schedule for License Amendment Issuance And Effectiveness Environmental Impact Assessment I, Mark E. Warner, Site Vice President of FPL Energy Seabrook, LLC hereby affirm that the information and statements contained within this License Amendment Request are based on facts and circumstances which are true and accurate to the best of my knowledge and belief.

Sworn and Subscribed before me this

_____ day of December, 2002 4

No&

Pulic Mark E. Warner Site Vice President

p SECTION I INTRODUCTION AND SAFETY ASSESSMENT FOR PROPOSED CHANGES Page I of 7

I.

INTRODUCTION AND SAFETY ASSESSMENT OF PROPOSED CHANGES A.

Introduction License Amendment Request (LAR) 02-10 proposes changes to the Seabrook Station Technical Specifications (TS) Index and TS 6.0, Administrative Controls.

The purpose of LAR 02-10 is to revise the Technical Specifications Section 6 to: (1) relocate administrative requirements discussed in Administrative Letter 95-06 "Relocation Of Technical Specification Administrative Controls Related To Quality Assurance" to a licensee controlled document, (2) change the title of the senior onsite official and (3) reflect changes in 10 CFR 20.

Utilizing the guidance in Administrative Letter 95-06, this LAR discusses the transfer of requirements from the Technical Specifications to the Operational Quality Assurance Program.

The requirements being transferred are: Independent Technical Reviews, Review and Audit, specifics related to the review of procedures and programs, and Records Retention.

Changes in the title of the senior onsite official from "Senior Vice President and Chief Nuclear Officer" to "Site Vice President" does not affect the onsite reporting responsibility or chain of command. The responsibility of this individual remains unchanged.

Seabrook Station has been complying with the requirements in the revised 10 CFR 20, the references in the TS had not been updated. This change brings the TS into consistency with 10 CFR 20.

Table I provides a tabulation of and justification for the proposed changes.

B.

Safety Assessment for Proposed Changes Utilizing the guidance in Administrative Letter 95-06, this LAR discusses the transfer of requirements from the Technical Specifications to the Operational Quality Assurance Program.

The OQAP is incorporated into the Updated Final Safety Analysis Report (UFSAR) Chapter 17.

Changes to the UFSAR are controlled in accordance with the requirements of 10 CFR 50.59 and 10 CFR 50.71(e). The OQAP change control process is contained in 10 CFR 50.54(a). The requirements are to be transferred intact simultaneously with implementation of the proposed changes to TS Section 6.0.

The relocated requirements are not required to be in TSs. 10 CFR 50.36c(2)(ii) contains the requirements for items that must be in TSs. This regulation provides criteria that can be used to determine the requirements that must be included in the TSs. Items not meeting the criteria can be relocated from TSs to a Licensee controlled document. The Licensee can then change the relocated requirements, if necessary, in accordance with 10 CFR 50.59.

This will result in significant reductions in time and expense to modify requirements that have been relocated while not adversely affecting plant safety.

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

Page 2 of 7

This criterion addresses instrumentation installed to detect excessive RCS leakage.

TS 6.2.3, "Independent Technical Reviews", TS 6.4, "Review and Audit", TS 6.7.2 through 6.7.5 (specific descriptions of the procedure review and approval process), and TS 6.9, "Records Retention", do not cover installed instrumentation that is used to detect, and indicate in the control room, a significant degradation of the reactor coolant pressure boundary. The listed TSs do not satisfy Criterion 1.

Criterion 2 A process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

The purpose of this criterion is to capture those process variables that have initial values assumed in the design basis accident and transient analyses, and which are monitored and controlled during power operation. This criterion also includes active design features (e.g., high pressure/low pressure system valves and interlocks) and operating restrictions (pressure/temperature limits) needed to preclude unanalyzed accidents and transients.

TS 6.2.3, "Independent Technical Reviews", TS 6.4, "Review and Audit", TS 6.7.2 through 6.7.5 (specific descriptions of the procedure review and approval process), and TS 6.9, "Records Retention", are not concerned with a plant system.

They are administrative programs. Therefore, the TSs being relocated are not a process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

TS 6.2.3, "Independent Technical Reviews", TS 6.4, "Review and Audit", TS 6.7.2 through 6.7.5 (specific descriptions of the procedure review and approval process), and TS 6.9, "Records Retention" do not satisfy Criterion 2.

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

The purpose of this criterion is to capture only those structures, systems, and components that are part of the primary success path of the safety analysis (an examination of the actions required to mitigate the consequences of the design basis accidents and transients). The primary success path of a safety analysis consists of the combinations and sequences of equipment needed to operate, so that the plant response to the design basis accidents and transients limits the consequences of these events to within the appropriate acceptance criteria. Also captured by this criterion are those support and actuation systems that are necessary for items in the primary success path to successfully function.

TS 6.2.3, "Independent Technical Reviews", TS 6.4, "Review and Audit", TS 6.7.2 through 6.7.5 (specific descriptions of the procedure review and approval process), and TS 6.9, "Records Retention" are administrative programs. As a Page 3 of 7 A

result, they are not a structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

TS 6.2.3, "Independent Technical Reviews", TS 6.4, "Review and Audit", TS 6.7.2 through 6.7.5 (specific descriptions of the procedure review and approval process), and TS 6.9, "Records Retention" do not satisfy Criterion 3.

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

The purpose of this criterion is to capture only those structures, systems, and components that operating experience or probabilistic risk assessment has shown to be significant to public health and safety.

Requirements proposed for relocation do not contain constraints of prime importance in limiting the likelihood or severity of the accident sequences that are commonly found to dominate risk.

TS 6.2.3, "Independent Technical Reviews", TS 6.4, "Review and Audit", TS 6.7.2 through 6.7.5 (specific descriptions of the procedure review and approval process), and TS 6.9, "Records Retention" are not a structure, system, or component which operating experience or probabilistic safety assessment has shown to be significant to the public health and safety. They have also not been modeled in the current Seabrook Station Probabilistic Safety Study (SSPSS). A review of industry operating experience did not produce any examples where the administrative programs in the commercial nuclear power industry has had a significant adverse affect on public health and safety.

TS 6.2.3, "Independent Technical Reviews", TS 6.4, "Review and Audit", TS 6.7.2 through 6.7.5 (specific descriptions of the procedure review and approval process), and TS 6.9, "Records Retention" do not meet Criterion 4.

The requirements contained in TS 6.2.3, "Independent Technical Reviews", TS 6.4, "Review and Audit", TS 6.7.2 through 6.7.5 (specific descriptions of the procedure review and approval process), and TS 6.9, "Records Retention", do not meet the 10 CFR 50.36c(2)(ii) criteria for items that must be in TSs. Therefore, relocating these requirements from the Seabrook Station Technical Specifications to a licensee controlled document will not adversely affect public health and safety. The relocation of this information maintains the consistency with NUREG-1431.

Any change to these requirements is made in accordance with 10 CFR 50.59 and 10 CFR 50.54(a).

Changes in the title of the senior onsite official from "Senior Vice President and Chief Nuclear Officer" to "Site Vice President" does not affect the onsite reporting responsibility or chain of command. The responsibility of this individual remains unchanged.

Seabrook Station has been complying with the requirements in the revised 10 CFR 20, the references in the TS had not been updated. This change brings the TS into consistency with 10 CFR 20.

Page 4 of 7

The proposed changes discussed in this LAR are editorial and administrative in nature and reflect the current configuration of the plant. The proposed changes do not affect nor modify the physical configuration, the operation, maintenance and management of the facility nor the manner in which it responds to normal, transient or accident conditions. Thus, the changes are an enhancement and do not affect plant safety.

FPLE Seabrook concludes that based upon the justifications presented in Table 1 as well as the Determination of No Significant Hazards for Proposed Changes, presented in Section IV, that the proposed changes do not adversely affect or endanger the health or safety of the general public or involve a significant safety hazard.

Page 5 of 7

It Table 1 Tabulation of Proposed Changes Item Marked-Up Description Justification Page #

I xiii, xiv, xv TS Index modified to reflect the proposed changes Editorial.

herein or other minor corrections.

2 6-1, Changed "Executive Vice President & Chief Nuclear Reflects change is position title. Wording is 6-11 Officer" to "Site Vice President" consistent with ANS Standard for use of generic titles I

in TS Section 6.

3 6-5, Relocated the requirements in TS 6.2.3, Independent Follows the guidance in AL 95-06 to relocate 6-6, Technical Reviews and TS 6.4, "Review and Audit" administrative requirements to a licensee-controlled 6-7, to a licensee-controlled document (Operational document. The requirements will be relocated intact 6-8 Quality Assurance Program (OQAP)). TS 6.4 to the OQAP. Changes to the OQAP are documented and includes TS 6.4.1, "Station Operation Review and controlled in accordance with 10CFR50.59 and 6-8A Committee (SORC)," and TS 6.4.2, "Station 10CFR50.54(a). Items being relocated are not Qualified Reviewer Program."

required to be in TS based upon the criteria contained in 10 CFR 50.36.

4 6-12 Relocated the requirements in TS 6.7.2 through TS Follows the guidance in AL 95-06 to relocate 6.7.5, which deal with specifics of the procedure administrative requirements to a licensee-controlled review and approval process, to a licensee-controlled document. The requirements will be relocated intact document (Operational Quality Assurance Program to the OQAP. Changes to the OQAP are documented (OQAP)).

and controlled in accordance with IOCFR50.59 and 10CFR50.54(a). Items being relocated are not required to be in TS based upon the criteria contained in 10 CFR 50.36.

5 6-14A and Revised TS 6.7.6g to reflect IOCFR20 references.

Seabrook Station has been complying with the 6.15 Revise footnote to reflect 10CFR20 references, requirements in the revised 10 CFR 20, the references in the TS had not been updated. This change brings the TS into consistency with 10CFR20.

Revise TS 6.7.6g.4) to reflect Seabrook Station as Editorial.

only one unit.

II Page 6 of 7

Table 1 Tabulation of Proposed Changes (continued)

Item Marked-Up Description Justification Page #

6 6-19 Relocated the requirements in TS 6.9, "Records Follows the guidance in AL 95-06 to relocate and Retention," to a licensee controlled document administrative requirements to a licensee-controlled 6-20 (Operational Quality Assurance Program (OQAP)).

document. The requirements will be relocated intact to the OQAP. Changes to the OQAP are documented and controlled in accordance with 10CFR50.59 and 10CFR50.54(a). Items being relocated are not required to be in TS based upon the criteria contained in 10 CFR 50.36.

7 6-20 TS 6.11, "High Radiation Area," changes to reflect While Seabrook Station has been complying with the and current 10 CFR 20 references.

requirements in the revised 10CFR20, the references 6-21 in the TS had not been updated. These changes bring the TS into consistency with 10CFR20.

8 6-21 Change "Shift Superintendent" to "Shift Manager" Editorial change to reflect the actual position title at Seabrook Station.

Revise 6.12, "Process Control Program (PCP)" to The requirements in TS 6.9.3 have been relocated to reference the Operational Quality Assurance Program the OQAP.

9 6-22 Revise 6.12.2 (Continued) to 6.12 (Continued)

An editorial correction that should have been done for License Amendment 66, which revised TS 6.12.

Revise 6.13a, "Offsite Dose Calculation Manual The requirements in TS 6.9.3 have been relocated to (ODCM)" to reference the Operational Quality the OQAP.

Assurance Program (OQAP)

Revised TS 6.13a.2) to reference 10 CFR 20.1302.

While Seabrook Station has been complying with the requirements in the revised 10CFR20, the references in the TS had not been updated. This change brings I the TS into consistency with 10CFR20.

Page 7 of 7

SECTION II MARKUP OF PROPOSED CHANGES Refer to the attached markup of the proposed changes to the Technical Specifications. The attached markup reflects the currently issued revision of the Technical Specifications listed below. Pending Technical Specifications or Technical Specification changes issued subsequent to this submittal are not reflected in the enclosed markup.

The following Operating License and Technical Specification changes are that are included in the attached markup are delineated in Table 1 in Section I

INDEX 5-0 DESIGN FEATURES

,SECTION PAG 5.3 REACTOR CORE 5.3.1 FUEL ASSEMBLIES 5-9 5.3.2 CONTROL ROD ASSEMBLIES......

.5-9 5.4 REACTOR COOLANT SYSTEM 5.4.1 DESIGN PRESSURE AND TEMPERATURE..............

5-9 5.4.2 VOLUME...................

5-9 5.5 (THIS SPECIFICATION NUMBER IS NOT USED)..............

5-9 5.6 FUEL STORAGE 5.6.1 CRITICALITY...

5-10 5.6.2 DRAINAGE............

5-10 5.6.3 5-10 5.7 COMPONENT CYCLIC OR TRANSIENT LIMIT..................

5-10 TABLE 5.7-1 COMPONENT CYCLIC OR TRANSIENT LIMITS............ 5-11 6.0 ADMINISTRATIVE CONTROLS 6.1 RESPONSIBILITY............

6-1 6.2 ORGANIZATION............

6-1 6.2.1 OFFSITE AND ONSITE ORGANIZATIONS.....

6-1 6.2.2 STATION STAFF-1--.

6-2 FIGURE 6.2-1 (This figure number Is not used)

..4.. *..

6"ý.

FIGURE 6.2-2 (This figure number is not used)--

6-3 TABLE 6.2-1 MINIMUM SI CREW COMPOSITION 4..

-4 6.2.3 IN ENDE J~&ETY O1fNEERING GR (ISEG)

+

f n

obun tlon....

6-5

,sponsbt ilies....

6-5 Records 6.2.4 SHIFT TECHNICAL ADVISOR.......................

6-5 6.3 6-5 Amendment No. 61-xlll SEABROOK - UNIT 1

INDEX 6.0 ADMINISTRATIVE CONTROLS PAGE

,, m-C

t 5ft-0F tM ?To H A* w, f,,s 90or * ! q)......

6-6 SECTION I

6.5 6.6 6.7 6.8 6.8.

.3 TyIS SP*,IFIC~klION N~436MB*R IS IDIOT pSEQV)

.REPORTABLE EVENT ACTION

................................ 6-11 SAFETY LIMIT VIOLATION..........................................

6-11 PROCEDURES AND PROGRAMS.......................................................... 6-12 REPORTING REQUIREMENTS 1 ROUTINE REPORTS...............................................................................

6-14D Startup Report.......................................................................................

6-14D Annual Reports...................................................................................... 6-15 Annual Radiological Environmental Operating Report....................................

6-16 Annual Radioactive Effluent Release Report................................................

6-17 Mont*ly Operating Reports.......................................................................

6-18 CORE OPERATING LIMITS REPORT........................................................

6-18 6.8.2 SPECIAL REPORTS 6-19 SEABROOK - UNIT 1 xiv Amendment No. 34, 9 I

INDEX 6.0 ADMINSTRATIVE CONTROLS SECTION PAGE 6.9*

............ ~ ~~~~..............................

9 6.________9cte~~~

kn~ slTcs 2 6-19 66.10 RADIATION PROTECTION PROGRAM

............ 6-20 6.11 HIGH RADIATION AREA....................................................................... 6-20 6.12 PROCESS CONTROL PROGRAM (PCP).................................................... 6-21 6.13 OFFSITE DOSE CALCULATION MANUAL (ODCM).......................................

6-22 6.14 MAJOR CHANGES TO LIQUID, GASEOUS, AND SOLID RADWASTE TREATMENT SYSTEMS 6-23 6.15 CONTAINMENT LEAKAGE RATE TESTING PROGRAM................................ 6-24 (1

Amendment No. 24.SEABROOK - UNIT I xv

60 ADMINTSTRATTVF CMfRTROLr 0

6 RESPDNSTBTL TrY 6.1.1. The Station Director shall be responsible for overall station operation and shall delegate In writing the succession to this responsibility during his absence.

b 6.1.2 The Shift Manager (or during his absence from the control room, a designated Individual) shall be responsible for the control roomy; a

b_

AAA ll mnt d y to this effect, signed by the shall be reissued to all station person n

an.annual,S,..._*

e,..

(j ORIANTIZA'TOf 6.2.1 OFSTTF AND 03NSTTF ORIGAITATTIAS*

Onsite and offsite organizations shall be established for unit operation and corporate management, respectively.

The onsite and offsite organizations shall Include the positions for activities affecting the safety of ihe nuclear power plant.

a.

Lines of authority, responsibility, and communication shall be established and defined for the highest management levels through intermediate levels to and including all operating organization positions.

These relationships shall be documented and updated, as appropriate, in the form of organimation charts, functional descriptions for departmental responsibillties and relationships.

and Job descriptions-for key personnel positions, or in equivalent forms of documentation.

These requirements shall be documented in the FSAR and updated in accordance with the requirements of 10 CFR 50.71.

b.

The Station Director shall be responsible for overall unit safe operation and shall have control over those onsite activities necessary nnp.at en.

maintenance of the plant.

c.

" rshall have corporate respons 1 ty or overall plant nuc ear safety and shall take any measures needed to ensure acceptable performance of the staff In operating. maintaining, and providing technical support to the plant to ensure-nuclear safety.

d.

The individuals who train the operating staff and those who carry out health physics and quality assurance functions may report to the appropriate onsite manager: however, they shall have sufficient organizational freedom to ensure their independeftce from operating pressures.

SEABROK

- UNIT 1 6-1 "Amendment No. 5

ADMINISTRATIVE CONTROLS 6.2.3 t8-E0f'E, HN 'L RE =-!_EA ('Ti sSpec; 4"-,*e-, A/u,,,Jr iS 4

st is' )

A Technical Review Program shall be established, implemented and maintained to encompass the following Technical Review responsibilities.

/

FUNCTION 6.2.3.1 Th echnical Review Progra responsibilities shall encompass:

a NRC Issuances, Industry dvisories, Licensee EvepReports, and other sources that may Indicate areas for proving plant safety;

b. Internal and exte operating experience ormation that may indicate areas for for Improving p t safety;

c. Plant ope rag characteristics, plan perations, modifications, intenance and survellaa to verify independentatat these activities are p formed safely and corre and that human errors re reduced as much as patical, and

d. M *ng detailed recommen tions to the Senior Site icial for procedure viIons, equip7ment mo ications or other means Improving nuclear safety and plant reliability.

The chnical Review Prog shall utilize several on personnel who are indepe nt of the plant management In to perform the revIe RECORDS 6.2.3.2 Written cords of technical revie s shall be maintained. As a inimum these records shall In de the results of the a

ies conducted, the status recommendations made pursuan o Specification 62.3.1 d an assessment of comp riy operations related to the riews rformed. A copy of th monthly Technical Review ram report shall be provided the Senior Site Official.

QUAL ]CATIONS 6.2.3.3 Personnel perfo ing reviews pursuant to chnical Specrfication.2.3.1 shall have either a bachelor's egree in engineering or ted science and at st 2 years professional level ex nce, at least 1 year of Ich shall be in the nu ear field, or equivalent education and experience as define in ANSIlANS 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 engineering, and plant analysis with regard to the safe operation of the station.

6.3 TRAINING 6.3.1 (THIS SPECIFICATION NUMBER IS NOT USED)

Amendment No. 34,35, S

(I SEABROOK - UNIT 1 6-5

J.

AflMTIJT.7DATw

rnrrDrI e 5ROR 1 AT1OI OPFRATTON RFVTFW COMMTTTFF ('

r6.4.1.1 The SORC S function to advise ttion Director on all matte rel~ated to nuclea safety.

r 6.4.1.2 S ORPC shall. as a mini

, be composed of Chairman and nine Indlvi als who collectively have xperlence and tise i h olwn

are :..

.I h olwn Nuclear Power P ntAdministratly otrols Mechanical a*,tenance Electrical I*ntenance Instr*e tion & Control Chemlstr Healt ics Oper Ions Te nical SupportfEn neering actor Engineerin The Stat n Director shall rve as Chairman of SORC and shall appoint the SORC m rs in wrln.

rs shall have inimum of eight year power pla eprience of hi a minimum of thr yars shall be nule, power ience.

At leastne member shall ha an SRO license for brook Stati ALIERATIS 6.4`1.3 All al rnate members sh be appointed in ing by the SORC Chairman to s e on a terpDra asis and shall hay qualifications equivalent to those of members.

6.4.e SOIRC shal mpeet at least on r calendar month and as dby th OCChairman or neof his designa alternate(s).

6.4. 1._5 The rum of the SRC ssary for the performan of the SORC responsibil yand authori isions f these Technica pecifications shall consist t

Chairman or of his designated alte e(s and fficient SORC rs including al mnates to equal -at least 5 prcent of the SORC CoMPosi ion.

SEABROO*.-

UNIT I Amendmeent No..44, AnMTIJTC'MATTVr 9-nMTD 4 It to. 41t kou" V1 a I

V L.*..=J 1 6-6

"I.

ADJMNrSTRATSTVF CONTROIS RESONTBUTTF I I4 ý ýIsC 4I i

6.4.1.6 The.SORC shall be responsible for:

"a.

Review of:

(

all proposed procedures required by Specification 6.7 and ch es thereto. (2) all roPosed programs reuired by Specfic* On 6.7 and changes t reto. and (3) any other proposed proced es or changes thereto s determined by the Station Director Toaf ect nuclear safety.

ocedures and progra required by S

fication 6.7 that ar designated for revi and approval by the Stion Qualified Revi r Program in accord ce with Specification

.4.2 do not require C review.

b Review of all pro ed tests and expe.r nts that af.fect nuclear safety:.

c.

Review of all roposed changes to pendix WA" Technica Specificati

d.

Review all proposed chang or mnodi fi cati ons t station systems or equi nt that affect n ear safety:

e. In stigatlon of all vi atlons of the Tech Ica Specifications.

e ludin the prepara on and forwarding o reports covering evalua ion and recommendat s to prevent recu ence. to the Executive Vice President & C ef Nuclear officer nd to the Nuclear Safety Audit Review Com*i tee(NSARC):

"Review of all PORTABLE EVENTS:

g.

Review of tion operations detect potential hazards o nuclear safety:

Pe.rfo nce of special ews. investigations, r nalyses and repo thereon as req d by the Station Di or or the NSARC:

1.

N.used:

J.

Not used:

Review of a nyccidental, unplanned.

r uncontrolled radioactive release inc ding the preparation

-reports covering evalt tion.

recommend ions. and disposition f the corrective acti to prevent recur and the forwardi these reports to th -ecutive Vice Presi

&nChief Nuclear Of cer and to the NSR

1.

Rev' of changes to the OCESS CONTROL P OFFSITE DOSE TION MAUAL. an he Radwaste Trea System; and Me.

eview of the Fire ection Program implementing instructions and submittal of ommended Fire Pr ection Program changes to the NSARC..

SEABROOK - UNIT I Amendment No. ;4.,9.r, 6-7

ADMINISTRATIVE CONTROLS "h

posc.,,

Io,.,Y.

6.4.1.7 The SORC shall:

a. Reco end In writing to the t on Director approval or disapproval of Items co ered under Spec fico

.4.1.6a. through d;

b.

ender determinations writing with regard whether or not each item considered under Speiication 6.4.1.6a., b.

d d. constitutes a need for a license amenden d

"c. Provide written otification within 24 h rs to the Exe bce President &

Chief Nuclea icer and the NSARC *f disagreement b een the SORC and the Station rector however, the tion Direr a

respons y for resolutio fsuch disagreements rsuant to: Specf n 6A.1.1

RECORDS, 6.4.1.8 e SORC shall maintal wdtten minutes each SORC mee that, at a mnimum ocument the results all SORC acti performed under e ressibility provisis of these Technical ciffications.

les shall be provi to the Executive Vice Ident & Chief Nuclea icer and the C.

2 STATION QUA-IF! D REVIEWER ROGRAM FUNCTION

./

6.42.1. The on Director establish a tion Qualified Re wer Program whereby requi reviews of des nated procedur or classes of p ures required by Specification

.4.1.6.a are pef-ed by Statio ualifled Reviewers nd approved by the designated epartment hea These rev

, are In lieu of iews by the SORC.

However, cedures wh require a 10 R 50.59 evaluatir st be reviewed by the 6.4.2.2 The S on Qualified vewer Program sb

a. P de for the rew of designated p cdures, programs, and anges thereto b Qualified ewer(s) other tha e Individual Who pre d the procedure,

,, gram, or ge.

Provide V cross-disipinay eviw of procedures rograms, and changes thereto *en orgazton er tan the preparin rganization areafy "the p ure, program, or nge.

C.

ure cross-disciplin revrlws are by a Qualifle r(s) In ected disciplines, r by other persons signated by Izant dpartment eads as having ecfic expertise re ireci to assess rticuar

cedure, pr~ogramor chn

. Cross-disciplina rvewers may f n as a committee.

Amendment No. 345.7-0, '+9 SEABROOK - UNIT 1 6-8

IS

ADMINISTRATIVE CONTROLS dProvide for a screening of designated procedures, programs and changes theretoto determinO an evaluation should be performed In accordance with the provisions of 10 50.59 to verify that a need for a license amendment does not exist This creening will be perfo ed by personnel trained and qualified In performing I CFR 50.59 screenings

e.

Provide or written recomme n by the Qu ed Reviewer(s) to the respo ible department head or approval or dis proval of procedures and pro ms considered under peclfication 6.4.1.

and that the procedure or P

m was screened a qualified individ and found not to require a CFR 50.59 evaluation S

6.4..3 If the responsible de rtment head dete ines that a new progra

rocedure, or ge thereto requires a CFR 50.59 evalu on, that designated dep ent head ensure the required e luation Is perform to determine If the rJw procedure, ram, or change Invol a need for a I *nse amendment. The/hew procedure. I program, or change will n be forwarded the 10 CFR 50.59 eval

on to SORG for review.

6A.2.4 Personn recommended to Station Qualified Revie rs shall be designated In writing by th

. n Director for ach procedure, program, r class of procedure or program within scope of the Stati Qualified Reviewer Prog m.

6.4.2.5 T

porary procedure anges shall be made in ccordance with Specification 6.7.3 withte exception that anges to procedures for ich reviews are assignid to Qualified evewers wil be rewed and approved as d r'bed In Specification 6.

REG DS 61/2.6 The review procedures nd program performed under the S tion Qualified Reviewer Program sh be documented In acco nce with administrative rocedures.

TRINING AND 0 LFOCATION 6.A2.7 The Ining and qualification reuirements of pte nnel designated as a Qualified Revi r in accordance with St on Qualified r Program shall be In accordance administrative procedr es. Qualified revee shall have:

a.

A Bachelors degree In e glneering, related s; Ince, or technical dis pline, and two years of nddear r plant experience; OR

b. Six years of nud r power plant exe" roe; OR V

c. An equivalent combination of education and experience as approved by the designated department head.

Amendment No. 34r -,4

, 19 -;

SEABROOK-UNIT 1 6-8A

f

(ý{ADMINISTRATIVE CONTROLS 6.4.3-TRIS SPE.tMATION NUMBER ISJT USED

,--. ~

PAGE INTENTiONALLY BLANK EUAmendment No. 34.,55, 6-8B,

$ EABROOK - UNIT 1,

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

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 E=eeT

I Vice President &-Chief Nuclea: OfF!.

" "I 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 tel one as soon as possible and In all cases within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . The e

Vice President &A C'h,,f Nu-clea Ofcrý.,and the NSARC shall be notified within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ;

b.

A Safety Umit Violation Report shall be prepared. The report shall be reviewed by the SORC. This report shall describe: (1) applicable circumstances preceding the violation, (2) effects of the violation upon facility components, systemsor 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 ice President &-Gi.'-ca--:.c-I within 14 days of the violation;a

d.

Operation of the station shall not be resumed until authorized by the Commission.

Amendment No. 340-,,*.

SEABROOK - UNIT I 6-11

ADMTNTSITRTTVF CONTROLS 6-7 PR0CMDURFS AND PR{(RAMS 6.7.1 Written-procedures shall be established. Implemented. and maintained covering the activities referenced belpw:

a.

The applicable procedures reco nded In Appendix A of Regulatory Guide 1.33. Revision 2. February 1978:

b.

The emergency operating procedures required to Implement the requirements of NUREG- 077 anVSRG-7 requ remen*.

.o~ H IRE.G-037 and Supplement 1 to NUREG-0737 as state in Generic Letter No. 82-33:

u

c.

Not used:

d.

N" t used:

e.

PROCESS CONTROL PRO3M Implementation:

f.

OFFSITE DOSE CALCULATION MANUAL Implementation:

g.

Quality Assurance Program for effluent and environmental monitoring:

h.

Fire Protection Program implementation: and I.

Technical Specification Improvement Program implementation.

6.7.2 The Station Dimctor may designate specific procedures and programs or classes of procedug and programs to be revie*ed in accordance with t~he Station Qualified Revewi Program I n lieu of review by the SORC.

The review per the Qualified Rev er Program shall in accordance with Specification 6.4.2.

6.7L3 P edures and progr isted in Specific Ion 6.7.1. and nges th6.

shall be approv the Station Dir r or by cogniza*department he*

r rdrectorswho.a esignated as theý ýproval Authority the Station or. as specifi n administrative 6dures.

The oval Authority for each procedure and ogram or class of edure and progr shall be specified in adnilnistratv rocedures.

6.7.4 Each cedure of Specif tion 6.7.1. and c n~es thereto. s 1 be revi ewed b he SORC and shal approved by t tation Director or be revied nd approved in a rdance With the S ion Qualified i ewer Program, pr~ior Implementation.

ach.-procedure of ci fi cati on 6.7 shall be re periodically set forth in ade i s rative proc es.

6.7.5 Chanes to eures of Speci tion 6.7.1 be made prior to review provded:

a.

Intent of the iginal procedur is not altered:

b The change is proved by two rs of the pl an a nagement staff.

at least o

om holds enior Operator I se: and

c.

The ch e Is documente. reviewed by the S.

and approved e

Stt n *recor. or viewed and apprv in accordance the Ion Qalifi viewer Program, i

n 14 days of

" mple ation.

Amendment No. 24. ag.t9 I J I

I SEABROOK - UNIT 1 6-12

ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS 6.7.6 (Continued)

a.

Radioactive Effluent Controls Proaram A program shall be provided conforming with 10 CFR 50.36a for the control of radioactive effluents and for maintaining the doses to MEMBERS OF THE PUBLIC from radioactive effluents as low as reasonably achievable. The program (1) shall be contained In the ODCM, (2) shall be implemented by operating procedures, and (3) shall include remedial actions to be taken whenever the program limits are exceeded. The program shall include the following elements:

1)

Umitations on the operability of radioactive liquid and gaseous monitoring instrumentation including surveillance tests and setpoint determination in accordance with the methodology in the ODCM,

2)

Umitations on th rtansou concentrations of radioactive material released in liquid effluen to UNRESTRICTED-AREASconforminp o 10 A-~()

ppendix 8, 2,b 4-unZ a 10FR P0;. loot -a

3)

Monitoring, sampling, and analysis of radioactive liquid and gaseous effluents in accordance with 10 CFR 20.

and with the methodology and parameters in the ODCM,

('i

4)

Umitations on the annual and quarterly doses or dose commitment to a MEMBER OF THE PUBLIC from radioactive materials in liquid effluents released from nit to UNRESTRICTED AREAS conforming to Appendix It 0 CFR Part 50,

5)

Determination ;cumulative and projected dose contributions from radioactive effluents for the current calendar quarter and current calendar year in accordance with the methodology and parameters in th ODCM at least every 31 days,

6)

Limitations on the operability and use of the liquid and gaseous effluent treatment systems to ensure that the appropriate portions of these systems are used to reduce releases of radioactivity when the projected doses in a 31-day period would exceed 2 percent of the guidelines for the annual dose or dose commitment conforming to Appendix I to 10 CFR Part 50, SEABROOK - UNIT I

(

r.,/-

6-14A Amendment No. 66-.

ADMINISTRATIVE CONTROLS The Startup Report shall address each of the tests identified in the Final Safety Analysis Report and shall include a description of the measured values of the operating conditions or characteristics obtained during the test program and a comparison of these values with design predictions and specifications.

Any corrective actions that were required to obtain satisfactory operation shall also be described.

Any additional specific details required in license condi tions based on other commitments shall be included in this report.

Startup Reports shall be submitted within:

(1) 90 days following completion of the Startup Test Program, (2) 90 days following resumption or commencement of commercial power operation, or (3) 9 months following initial criticality, whichever is earliest.

If the Startup Report does not cover all three events (i.e., initial criticality, completion of Startup Test Program, and resumption or commencement of commercial operation), supplementary reports shall be submitted at least every 3 months until all three events have been completed.

ANNUAL REPORTS*

6.8.1.2 Annual Reports covering the activities of the station as described below for the previous calendar year shall be submitted prior to March 1 of each year.

The initial report shall be submitted prior to March 1 of the year following initial criticality.

Reports required on an annual basis shall include:

a. A tabulation on an annual basis of the number of station, utility, and other personnel (including contractors) receiving exposures greater than 100 mrem/yr and their associated man-rem exposure according to work and job functions** (e.g., reactor operations and surveillance, inservJce inspection, routine maintenance, special maintenance [describe maintenance], waste processing, and refueling).

The dose assignments to various duty functions may be estimated based on pocket dosimeter, thermoluminescent dosimeter (TLD),

or film badge measurements.

Small exposures totalling less than 20% of the individual total dose need not be accounted for.

In the aggregate, at least 80% of the total whole-body dose received from external sources should be assigned to specific major work functions;

b.

The results of specific activity analyses in which the primary coolant exceeded the limits of Specification 3.4.8. The following information shall be included:

(1) Reactor power history starting 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months prior to the first sample in which the limit was exceeded (in graphic and tabular format); (2) Results of the last isotopic analysis for radiolodine performed prior to exceeding the limit, results of analysis while limit was exceeded and results of one analysis after the radioiodine activity was reduced to less than limit.

Each result should include date and time of sampling and the radioiodine concentrations; (3) Clean-up flow history starting 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months prior to the first sample in which the limit was exceeded; (4) Graph of the 1-131 concentration (pCi/gm) and one other radio iodine isotope concentration (pCi/gm) as a function of time for the

A single submittal may be made for a multiple unit station.

The submittal should combine those sections that are common to all unilat the statijgn.

- This tabulation supplements the requirements of

.O CFR ZCr.KUU UNII 6-15

ADMINISTRATIVE CONTROLS SPECIAL REPORTS 6.8.2 Special reports shall be submitted to the U.S. Nuclear Regulatory Commission, Washington, D.C. 20555, Attn:

Document Control Desk, with a copy to the NRC Regional Administrator within the time period specified for each report.

(-i c i

6.9 ECO 0 RET T j

6.9.

_1 n add ion to he applicab e record retention requirements of Title 10, Code f Feel'1 Regu tYons, the ollowing recor s shall be re ained for at lea the mi imum p od indicat d.

.9.2 The ollowin records s 11 be retaine for at least years:

a.

Records nd logs of station opera on covering ime interval each

/

power 1 vel; R.

Record and logs principal intenance act vities, insp tions, repai, and repl ement of pri cipal items o equipment r ated to nucl ar safety; C. All REPORTABL EVENTS;

d.

R ords of s rveillance a ivities, insp ctions, and alibrations r quired by hese Techni al Specificati ns;

e.

ecords of changes mad to the proced res required ySpecifi cation 6

.1;

f. Records of radioac ye shipments;

Recor s of sealed source and fiss on'detector eak tests and r ults; and

h.

R ords of ann al physical In ntory of al sealed source m erial f record.

6.9.3 T e following cords shall b retained fo the duration of the station Operating License:

a.

Records nd drawing ch ges reflect g station desig modificatio s

.inade to systems and e lpment desc bed in the Fina Safety Anal sis Report

b.

Recor s of new an irradiated el inventory, fu transfers and ass. bly burnup stories;

c.

Re rds of radation expos e for all Indivi als enteri radiation c trol areas

d.

ecords of aseous and iquid radioactive material r eased to t environs;

e.

Records of transient or operational cycles for those station components' identified in Table 5.7-1;

-4 SEABROOK - UNIT I 6-19

~4 A

4

tWM, a

L k

ADMINISTRATIVE CONTROLS RECORD EtETENTION 6.9.3 (Co nued) f ecords o ct tests and xperiments;

. Records of raining and qu ification for cu nt members the station sta

h. Records fInservice ins ecions. perform d pursuant to ese Technical pecifications

1. Reco of quality as rance acivte quired by Operational Q lity Assuran Man 1;

j. Re rds of re performed for anges made procedur.es or uipment or ews of tests nd experimen ursuant to 10 FR 50.59;

k.

ecords of etings of the S RC and the N C;

Records the service liv* of all hydrauli nd mechanical nubbers req red by Speclfie on 3.7.7 incldi the date at ich th a

and associ ed installation a ma/ntenan recor

m. Re s of secondlar ter samplin and-water qual*, and

n. R rds.of analyse required by e Radiological vironmental o*nltoring gram at would permit luation of e accuracy of analysis at a ter date.

is sho

dude procedu seffective a pecitled times d QA records howing tha ese procedures we followed.

o. Records of re iews perfo ed for changes ade to the 0 SITE DOS CALCULATI N MANUAL and the PROC S CONTROL PROGRAM 6.10 RADIATION PROTECTION PROGRAM 6.10.1 Prooedures for personnel radiation protection shall be prepared consistent with the requirements of 10 CFR Part 20 and shall be app r.

adhered to for all operations Involving personnel radiation exposu SHIGH RADIATIONAR a.,

o.*1 6.11.1 Pursuant to paragraph 0 CFR Part 20, in lieu of the "control device" or "alarm signar required by paragraph 29.29e,each high radiation area, as defined in 10 CFR/

IZ Part 20, In which the intensity of radiation Is equal to or less-than 1000 mR/h at 45-n o(484n.)

10 from the radiation source or from any surface that the radiation penetrates shall be barricaded an conspicuously posted as a high radiation area and entrance thereto shall be controlled by requiring Issuance of a Radiation Work Permit (RWP). Individuals qualified in radiation protection procedures (e.g., Health Physics Technician) or personnel continuously escorted bysuch individuals may be exempt from the RWP Issuance requirement during the performance of their assigned duties In high radiation areas with exposure rates equal to or less than 1000 mR/h, provided they are otherwise following plant radiation protection procedures for entry into such high SEABROOK - UNIT I 6-20 Amendment No. If")

ADMINISTRATIVE CONTROLS HIGH RADIATION AREA 6.11.1 (Continued) radiation areas. Any individual or group of Individuals permitted to enter such areas shall be provided with or accompanied by one or more of the following:

a.

A radiation monitoring device that continuously indicates the radiation dose rate in the area; or

b.

A radiation monitoring device that continuously integrates the radiation dose rate in the area and alarms when a preset integrated dose is received. Entry into such areas with this monitoring device may be made after the dose rate levels in the area have been established and personnel have been made knowledgeable of them; or

c.

An individual qualified in radiation protection procedures with a radiation dose rate monitoring device, who Is responsible for providing positive control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified in the Radiati Sao 6.112 In addition to the requirements of Spec.fic ras accessible to personnel with radiation levels greater than 1000 mR/h at 44E-emn-(48W from the radiation source or from I

any surface that the radiation pbnetrates shall be provided with locked doors to prevent unauthorized entry, and the keys shall be maintained under the administrative control of the

.Sh'i on duty and/or health physics supervision. Doors shall remain locked PIG lt during periods of.access by personnel under an approved RWP that shall specify the dose rate levels in the immediate work areas and the maximum allowable stay time for PIG norindividuals in that area. In lieu of the stay time speciffication of the RWP, direct or remote (such as dosed circuit TV cameras) continuous surveillance may be made by personnel qualified in radiation protection procedures to provide positive exposure control over the activities being performed within the area.

For individual high radiation areas accessible to personnel with radiation levels of greater than 1000 mR/h that are located within large areas, such as PWR containment, where no enclosure exists for purposes of locking, and where no enclosure can be reasonably constructed around the individual area, that individual area shall be-barricaded, conspicuously posted, and a flashing light shall be activated as a warning device.

6.12 PROCESS CONTROL PROGRAM (PCP)

Changes to the PCP:

a.

Shall be documented and records of reviews performed shall be retained as required by This documentation shall contain:

1)

Sufficient information to support the change together with the appropriate analyses or evaluatioris justifying the change(s) and SEABROOK - UNIT 1I-2 Amendment No. 22,e&

ADMINISTRATIVE CONTROLS PROCESS CONTROL PROGRAM (PCP) 6.12(Continued)

2)

A determination that the change will maintain the overall conformance of the solidified waste product to existing requirements of Federal, State, or other applicable regulations.

b.)

Shall become effective after review and acceptance by the SORC and approval of the Station Director.

6.13 OFFSITE DOSE CALCULATION MANUAL (ODCM)

Changes to the ODCM:

e-QofA*.AAL QqAtf'

a.

Shall be document nd records of reviews performed shall be retained as required by e

his documentation shall contain:

1)

Sufficient information to support the change together with the appropriate analyses or evaluations juq he change(s) and

2)

A determination that the change will in

.i the level of radioactive effluent control required by 10 CFR 2O.T, 40 CFR Part 190, 10 CFR 50.36a, and Appendix Ito 10 CFR Part 50 and not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations.

b.

Shall become effective after review and acceptance by the SORC and the approval of the Wtation Director.

c.

Shall be submitted to the Commission in the form of a complete, legible copy of the entire ODCM as part of or concurrent with the Annual Radioactive Effluent Release Report for the period of the report in which any change to the ODCM was made. Each change shall be identified by markings in the margin of the affected pages, clearly indicating the area of the page that was changed, and each affected page shall indicate the revision number the change was implementrd.

Amendment No. 2 Xr6 6 SEABROOK - UNIT 1 6-22

SECTION HI RETYPE OF PROPOSED CHANGES Refer to the attached retype of the proposed changes to the Technical Specifications.

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 5.0 DESIGN FEATURES SECTION PAGE 5.3 REACTOR CORE 5.3.1 FUEL ASSEMBLIES.............................................................................................

5-9 5.3.2 CONTROL ROD ASSEMBLIES...........................................................................

5-9 5.4 REACTOR COOLANT SYSTEM 5.4.1 DESIGN PRESSURE AND TEMPERATURE......................................................

5-9 5.4.2 VO LUM E...............................................................................................................

5-9 5.5 (THIS SPECIFICATION NUMBER IS NOT USED)....................................................

5-9 5.6 FUEL STORAGE 5.6.1 CRITICALITY........................................................................................................

5-10 5.6.2 DRAINAG E...........................................................................................................

5-10 5.6.3 CAPAC ITY............................................................................................................

5-10 5.7 COMPONENT CYCLIC OR TRANSIENT LIMIT........................................................

5-10 TABLE 5.7-1 COMPONENT CYCLIC OR TRANSIENT LIMITS........................................

5-11 6.0 ADMINISTRATIVE CONTROLS 6.1 RESPONSIBILITY.......................................................................................................

6-1 6.2 ORGANIZATION.........................................................................................................

6-1 6.2.1 OFFSITE AND ONSITE ORGANIZATIONS........................................................

6-1 6.2.2 STATION STAFF..................................................................................................

6-2 FIGURE 6.2-1 (THIS FIGURE NUMBER IS NOT USED)...................................................

6-3 FIGURE 6.2-2 (THIS FIGURE NUMBER IS NOT USED)...................................................

6-3 TABLE 6.2-1 MINIMUM SHIFT CREW COMPOSITION....................................................

6-4 6.2.3 (THIS SPECIFICATION NUMBER IS NOT USED)............................................

6-5 6.2.4 SHIFT TECHNICAL ADVISOR............................................................................

6-5 6.3 (THIS SPECIFICATION NUMBER IS NOT USED)....................................................

6-5 SEABROOK - UNIT I XIII==

Amendment No. 6R0,

INDEX 6.0 ADMINISTRATIVE CONTROLS SECTION PAGE 6.4 (THIS SPECIFICATION NUMBER IS NOT USED)....................................... 6-6 6.5 REPORTABLE EVENT ACTION...........................................................

6-11 6.6 SAFETY LIMIT VIOLATION..................................................................

6-11 6.7 PROCEDURES AND PROGRAMS............................................................ 6-12 6.8 REPORTING REQUIREMENTS 6.8.1 ROUTINE REPORTS.............................................................................. 6-14D Startup Report....................................................................................... 6-14D Annual Reports...................................................................................... 6-15 Annual Radiological Environmental Operating Report.................................... 6-16 Annual Radioactive Effluent Release Report................................................ 6-17 Monthly Operating Reports....................................................................... 6-18 CORE OPERATING LIMITS REPORT........................................................ 6-18 6.8.2 SPECIAL REPORTS.............................................................................. 6-19 SEABROOK - UNIT 1 xiv Amendment No. 34,67,73,

INDEX 6.0 ADMINSTRATIVE CONTROLS SECTION PAGE 6.9 (THIS SPECIFICATION NUMBER IS NOT USED)........................................ 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 OFFSITE DOSE CALCULATION MANUAL (ODCM)..................................... 6-22 6.14 MAJOR CHANGES TO LIQUID, GASEOUS, AND SOLID RADWASTE TREATMENT SYSTEMS....................................................... 6-23 6.15 CONTAINMENT LEAKAGE RATE TESTING PROGRAM............................... 6-24 Amendment No. 34-70, SEABROOK - UNIT I XV

6.0 ADMINISTRATIVE CONTROLS 6.1 RESPONSIBILITY 6.1.1 The Station Director shall be responsible for overall station operation and shall delegate in writing the succession to this responsibility during his absence.

6.1.2 The Shift Manager (or during his absence from the control room, a designated individual) shall be responsible for the control room command function. A management directive to this effect, signed by the Site Vice President shall be reissued to all station personnel on an annual basis.

6.2 ORGANIZATION 6.2.1 OFFSITE AND ONSITE ORGANIZATIONS Onsite and offsite organizations shall be established for unit operation and corporate management, respectively. The onsite and offsite organizations shall include the positions for activities affecting the safety of the nuclear power plant.

a.

Lines of authority, responsibility, and communication shall be established and defined for the highest management levels through intermediate levels to and including all operating organization positions. These relationships shall be documented and updated, as appropriate, in the form of organization charts, functional descriptions for departmental responsibilities and relationships, and job descriptions for key personnel positions, or in equivalent forms of documentation. These requirements shall be documented in the FSAR and updated in accordance with the requirements of 10 CFR 50.71.

b.

The Station Director shall be responsible for overall unit safe operation and shall have control over those onsite activities necessary for safe operation and maintenance of the plant.

c.

The Site Vice President shall have corporate responsibility for overall plant nuclear safety and shall take any measures needed to ensure acceptable performance of the staff in operating, maintaining, and providing technical support to the plant to ensure nuclear safety.

d.

The individuals who train the operating staff and those who carry out health physics and quality assurance functions may report to the appropriate onsite manager; however, they shall have sufficient organizational freedom to ensure their independence from operating pressures.

SEABROOK - UNIT 1 Amendment No.6,=,

6-1

ADMINISTRATIVE CONTROLS 6.2.3 (THIS SPECIFICATION NUMBER IS NOT USED) 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 engineering, and plant analysis with regard to the safe operation of the station.

6.3 TRAINING 6.3.1 (THIS SPECIFICATION NUMBER IS NOT USED)

Amendment No. 34-,3570, I

SEABROOK - UNIT I 6-5

ADMINISTRATIVE CONTROLS 6.4 (THIS'SPECIFICATION NUMBER IS NOT USED)

Amendment No. *34.,6 SEABROOK - UNIT 1 6-6

ADMINISTRATIVE CONTROLS THIS PAGE INTENTIONALLY BLANK Amendment No.,34,3,66, SEABROOK - UNIT 1 6-7

ADMINISTRATIVE CONTROLS THIS PAGE INTENTIONALLY BLANK Amendment No. 34,55,70,70, SEABROOK - UNIT 1 6-8

ADMINISTRATIVE CONTROLS THIS PAGE INTENTIONALLY BLANK Amendment No. 34,5,70,79, o

SEABROOK - UNIT I 6-8A

ADMINISTRATIVE CONTROLS THIS PAGE INTENTIONALLY BLANK Amendment No. 34,567, ADMINISTRATIVE CONTROLS o

I I

I SEABROOK - UNIT 1 6-813

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

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 Site Vice President.

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 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . The Site Vice President and the NSARC shall be notified within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ;

b.

A Safety Limit Violation Report shall be prepared. The report shall be reviewed by the SORC. This 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 Site Vice President within 14 days of the violation; and

d.

Operation of the station shall not be resumed until authorized by the Commission.

Amendment No. 34,55,67, I

I I

SEABROOK-UNIT 1 6-11

ADMINISTRATIVE CONTROLS 6.7 PROCEDURES AND PROGRAMS 6.7.1 Written procedures shall be established, implemented, and maintained covering the activities referenced below:

a.

The applicable procedures recommended in Appendix A of Regulatory Guide 1.33, Revision 2, February 1978;

b.

The emergency operating procedures required to implement the requirements of NUREG-0737 and Supplement I to NUREG-0737 as stated in Generic Letter No. 82-33;

c.

Not used;

d.

Not used;

e.

PROCESS CONTROL PROGRAM implementation;

f.

OFESITE DOSE CALCULATION MANUAL implementation;

g.

Quality Assurance Program for effluent and environmental monitoring;

h.

Fire Protection Program implementation; and

i.

Technical Specification Improvement Program implementation.

Amendment No. 34,3Q,5, SEABROOK - UNIT I 6-12

ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS 6.7.6 (Continued)

g.

Radioactive Effluent Controls Program A program shall be provided conforming with 10 CFR 50.36a for the control of radioactive effluents and for maintaining the doses to MEMBERS OF THE PUBLIC from radioactive effluents as low as reasonably achievable. The program (1) shall be contained in the ODCM, (2) shall be implemented by operating procedures, and (3) shall include remedial actions to be taken whenever the program limits are exceeded. The program shall include the following elements:

1)

Limitations on the operability of radioactive liquid and gaseous monitoring instrumentation including surveillance tests and setpoint determination in accordance with the methodology in the ODCM,

2)

Limitations on the concentrations of radioactive material released in liquid effluents to UNRESTRICTED AREAS, conforming to ten times the concentration values in Appendix B, Table 2, Column 2, to 10 CFR 20.1001-20.2402,

3)

Monitoring, sampling, and analysis of radioactive liquid and gaseous effluents in accordance with 10 CFR 20.1302 and with the methodology and parameters in the ODCM,

4)

Limitations on the annual and quarterly doses or dose commitment to a MEMBER OF THE PUBLIC from radioactive materials in liquid effluents released from the unit to UNRESTRICTED AREAS conforming to Appendix I to 10 CFR Part 50,

5)

Determination of cumulative and projected dose contributions from radioactive effluents for the current calendar quarter and current calendar year in accordance with the methodology and parameters in the ODCM at least every 31 days,

6)

Limitations on the operability and use of the liquid and gaseous effluent treatment systems to ensure that the appropriate portions of these systems are used to reduce releases of radioactivity when the projected doses in a 31-day period would exceed 2 percent of the guidelines for the annual dose or dose commitment conforming to Appendix I to 10 CFR Part 50, SEABROOK - UNIT I 6-14A Amendment No. 66,

ADMINISTRATIVE CONTROLS The Startup Report shall address each of the tests identified in the Final Safety Analysis Report and shall include a description of the measured values of the operating conditions or characteristics obtained during the test program and a comparison of these values with design predictions and specifications. Any corrective actions that were required to obtain satisfactory operation shall also be described. Any additional specific details required in license conditions based on other commitments shall be included in this report.

Startup Reports shall be submitted within: (1) 90 days following completion of the Startup Test Program, (2) 90 days following resumption or commencement of commercial power operation, or (3) 9 months following initial criticality, whichever is earliest. If the Startup Report does not cover all three events (i.e., initial criticality, completion of Startup Test Program, and resumption or commencement of commercial operation), supplementary reports shall be submitted at least every 3 months until all three events have been completed.

ANNUAL REPORTS*

6.8.1.2 Annual Reports covering the activities of the station as described below for the previous calendar year shall be submitted prior to March 1 of each year. The initial report shall be submitted prior to March 1 of the year following initial criticality.

Reports required on an annual basis shall include:

a.

A tabulation on an annual basis of the number of station, utility, and other personnel (including contractors) receiving exposures greater than 100 mrem/yr and their associated man-rem exposure according to work and job functions" (e.g., reactor operations and surveillance, inservice inspection, routine maintenance, special maintenance [describe maintenance], waste processing, and refueling). The dose assignments to various duty functions may be estimated based on pocket dosimeter, thermoluminescent dosimeter (TLD), or film badge measurements. Small exposures totaling less than 20% of the individual total dose need not be accounted for. In the aggregate, at least 80%

of the total whole-body dose received from external sources should be assigned to specific major work functions;

b.

The results of specific activity analyses in which the primary coolant exceeded the limits of Specification 3.4.8. The following information shall be included: (1)

Reactor power history starting 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months prior to the first sample in which the limit was exceeded (in graphic and tabular format); (2) Results of the last isotopic analysis for radioiodine performed prior to exceeding the limit, results of analysis while limit was exceeded and results of one analysis after the radioiodine activity was reduced to less than limit. Each result should include date and time of sampling and the radioiodine concentrations; (3) Clean-up flow history starting 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months prior to the first sample in which the limit was exceeded; (4) Graph of the 1-131 concentration (t.Ci/gm) and one other radio iodine isotope concentration (gCVgm) as a function of time for the

A single submittal may be made for a multiple unit station. The submittal should combine those sections that are common to all units at the station.

"**This tabulation supplements the requirements of 10 CFR 20.220b.

SEABROOK - UNIT I 6-15 Amendment No.

ADMINISTRATIVE CONTROLS SPECIAL REPORTS 6.8.2 Special reports shall be submitted to the U.S. Nuclear Regulatory Commission, Washington, D.C. 20555, Attn: Document Control Desk, with a copy to the NRC Regional Administrator within the time period specified for each report.

6.9 (THIS SPECIFICATION NUMBER IS NOT USED)

SEABROOK - UNIT 6

91 le 6-19 Amendment No.

ADMINISTRATIVE CONTROLS 6.10 RADIATION PROTECTION PROGRAM 6.10.1 Procedures for personnel radiation protection shall be prepared consistent with the requirements of 10 CFR Part 20 and shall be approved, maintained, and adhered to for all operations involving personnel radiation exposure.

6.11 HIGH RADIATION AREA 6.11.1 Pursuant to paragraph 20.1601(c) of 10 CFR Part 20, in lieu of the "control device" or "alarm signal" required by paragraph 20.1601(a) and (b), each high radiation area, as defined in 10 CFR Part 20, in which the intensity of radiation is equal to or less-than 1000 mR/h at 30 cm (12 in.) from the radiation source or from any surface that the radiation penetrates shall be barricaded and conspicuously posted as a high radiation area and entrance thereto shall be controlled by requiring issuance of a Radiation Work Permit (RWP). Individuals qualified in radiation protection procedures (e.g., Health Physics Technician) or personnel continuously escorted by such individuals may be exempt from the RWP issuance requirement during the performance of their assigned duties in high radiation areas with exposure rates equal to or less than 1000 mR/h, provided they are otherwise following plant radiation protection procedures for entry into such high SEABROOK - UNIT 1 I

6-20 Amendment No.6-,

4 ADMINISTRATIVE CONTROLS HIGH RADIATION AREA 6.11.1 (Continued) radiation areas. Any individual or group of individuals permitted to enter such areas shall be provided with or accompanied by one or more of the following:

a.

A radiation monitoring device that continuously indicates the radiation dose rate in the area; or

b.

A radiation monitoring device that continuously integrates the radiation dose rate in the area and alarms when a preset integrated dose is received. Entry into such areas with this monitoring device may be made after the dose rate levels in the area have been established and personnel have been made knowledgeable of them; or

c.

An individual qualified in radiation protection procedures with a radiation dose rate monitoring device, who is responsible for providing positive control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified in the Radiation Work Permit.

6.11.2 In addition to the requirements of Specification 6.11.1, areas accessible to personnel with radiation levels greater than 1000 mR/h at 30 cm (12 in.) from the radiation source or from any surface that the radiation penetrates shall be provided with locked doors to prevent unauthorized entry, and the keys shall be maintained under the administrative control of the Shift Manager on duty and/or health physics supervision. Doors shall remain locked except during periods of access by personnel under an approved RWP that shall specify the dose rate levels in the immediate work areas and the maximum allowable stay time for individuals in that area. In lieu of the stay time specification of the RWP, direct or remote (such as closed circuit TV cameras) continuous surveillance may be made by personnel qualified in radiation protection procedures to provide positive exposure control over the activities being performed within the area.

For individual high radiation areas accessible to personnel with radiation levels of greater than 1000 mR/h that are located within large areas, such as PWR containment, where no enclosure exists for purposes of locking, and where no enclosure can be reasonably constructed around the individual area, that individual area shall be barricaded, conspicuously posted, and a flashing light shall be activated as a warning device.

6.12 PROCESS CONTROL PROGRAM (PCP)

Changes to the PCP:

a.

Shall be documented and records of reviews performed shall be retained as required by the Operational Quality Assurance Program (OQAP). This documentation shall contain:

1)

Sufficient information to support the change together with the appropriate analyses or evaluations justifying the change(s) and Amendment No. 22,66, SEABROOK - UNIT 1 6-21

ADMINISTRATIVE CONTROLS PROCESS CONTROL PROGRAM (PCP) 6.12 (Continued)

2)

A determination that the change will maintain the overall conformance of the solidified waste product to existing requirements of Federal, State, or other applicable regulations.

b.)

Shall become effective after review and acceptance by the SORC and approval of the Station Director.

6.13 OFFSITE DOSE CALCULATION MANUAL (ODCM)

Changes to the ODCM:

a.

Shall be documented and records of reviews performed shall be retained as required by the Operational Quality Assurance Program (OQAP). This documentation shall contain:

1)

Sufficient information to support the change together with the appropriate analyses or evaluations justifying the change(s) and

2)

A determination that the change will maintain the level of radioactive effluent control required by 10 CFR 20.1302,40 CFR Part 190, 10 CFR 50.36a, and Appendix I to 10 CFR Part 50 and not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations.

b.

Shall become effective after review and acceptance by the SORC and the approval of the Station Director.

c.

Shall be submitted to the Commission in the form of a complete, legible copy of the entire ODCM as part of or concurrent with the Annual Radioactive Effluent Release Report for the period of the report in which any change to the ODCM was made. Each change shall be identified by markings in the margin of the affected pages, clearly indicating the area of the page that was changed, and each affected page shall indicate the revision number the change was implemented.

Amendment No. 22,66, SEABROOK - UNIT I 6-22

SECTION IV DETERMINATION OF NO SIGNIFICANT HAZARDS FOR PROPOSED CHANGES Page I of 3

IV.

DETERMINATION OF NO SIGNIFICANT HAZARDS FOR PROPOSED CHANGES License Amendment Request (LAR) 02-10 proposes changes to the Seabrook Station Technical Specifications (TS) Index and TS 6.0, Administrative Controls.

The purpose of LAR 02-10 is to revise the Technical Specifications Section 6 to: (1) relocate administrative requirements discussed in Administrative Letter 95-06 "Relocation Of Technical Specification Administrative Controls Related To Quality Assurance" to a licensee controlled document, (2) change the title of the senior onsite official and (3) reflect changes in 10 CFR 20.

Utilizing the guidance in Administrative Letter 95-06, this LAR discusses the transfer of requirements from the Technical Specifications to the Operational Quality Assurance Program.

The requirements being transferred are: Independent Technical Reviews, Review and Audit, specifics related to the review of procedures and programs and Records Retention.

Changes in the title of the senior onsite official from "Senior Vice President and Chief Nuclear Officer" to "Site Vice President" does not affect the onsite reporting responsibility or chain of command. The responsibility of this individual remains unchanged.

Seabrook Station has been complying with the requirements in the revised 10 CFR 20, the references in the TS had not been updated. This change brings the TS into consistency with 10 CFR 20.

In accordance with 10 CFR 50.92, FPLE Seabrook has concluded that the proposed changes 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 proposed changes to the Seabrook Station TS do not adversely affect accident initiators or precursors nor alter the design assumptions, conditions, and configuration of the facility or the manner in which the plant is operated and maintained. In addition, the proposed changes do not affect the manner in which the plant responds in normal operation, transient or accident conditions nor do they change any of the procedures related to operation of the plant. The proposed changes do not alter or prevent the ability of structures, systems and components (SSCs) to perform their intended function to mitigate the consequences of an initiating event within the acceptance limits assumed in the Updated Final Safety Analysis Report (UFSAR).

The proposed changes are administrative and editorial for the purpose of correcting or updating TS to reflect current NRC and industry initiatives.

The proposed changes do not affect the source term, containment isolation or radiological release assumptions used in evaluating the radiological consequences of an accident previously evaluated in the Seabrook Station UFSAR. Further, the proposed changes do not increase the types and amounts of radioactive effluent that may be released offsite, Page 2 of 3

nor significantly increase individual or cumulative occupational/public radiation exposures.

Therefore, it is concluded that these proposed revisions do not involve a significant increase in the probability or consequence 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 to the Seabrook Station TS do not change the operation or the design basis of any plant system or component during normal or accident conditions.

The proposed changes do not include any physical changes to the plant. In addition, the proposed changes do not change the function or operation of plant equipment or introduce any new failure mechanisms. The plant equipment will continue to respond per the design and analyses and there will not be a malfunction of a new or different type introduced by the proposed changes.

The proposed changes are administrative in nature and only correct, update and clarify the Seabrook Station Technical Specifications to reflect NRC guidance, i.e., AL 95-06.

The proposed changes do not modify the facility nor do they affect the plant's response to normal, transient or accident conditions. The changes do not introduce a new mode of plant operation. The changes are an enhancement and do not affect plant safety. The plant's design and design basis are not revised and the current safety analyses remains in effect.

Thus, these proposed revisions to the Seabrook Station TS do not create the possibility of a new or different kind of accident from any accident previously evaluated.

3.

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

The proposed changes are administrative changes to the Seabrook Station Technical Specifications.

The safety margins established through Limiting Conditions for Operation, Limiting Safety System Settings and Safety Limits as specified in the Technical Specifications are not revised nor is the plant design or its method of operation revised by the proposed changes.

Thus, it is concluded that these proposed revisions to the Seabrook Station TS do not involve a significant reduction in a margin of safety.

Based on the above evaluation, FPLE Seabrook concludes that the proposed changes to the Seabrook Station TS do not constitute a significant hazard.

Page 3 of 3

'Z

SECTIONS V AND VI PROPOSED SCHEDULE FOR LICENSE AMENDMENT ISSUANCE AND EFFECTIVENESS AND ENVIRONMENTAL IMPACT ASSESSMENT Page 1 of 2

V.

PROPOSED SCHEDULE FOR LICENSE AMENDMENT ISSUANCE AND EFFECTIVENESS FPLE Seabrook requests NRC review of License Amendment Request 02-10, and issuance of a license amendment by April 30, 2003, having immediate effectiveness and implementation within 60 days. FPLE Seabrook requests these changes in less than the one year normally afforded for NRC review because the changes are administrative in nature and will afford increased organizational flexibility and efficiency at an earlier date.

VI.

ENVIRONMENTAL IMPACT ASSESSMENT FPLE Seabrook has reviewed the proposed license amendment against the criteria of 10 CFR 51.22 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, FPLE Seabrook concludes that the proposed changes meet the criterion delineated in 10 CFR 51.22(c)(1 0) for a categorical exclusion from the requirements for an Environmental Impact Statement.

Page 2 of 2

ML030030550

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