ML20055H852

Forwards Draft for New STS Section 5.0, Administrative Controls. Draft Incorporates Results of 900523-24 Meeting W/Owners Groups.Highlighted Portions Indicate Changes/ Additions

ML20055H852
Person / Time
Issue date:	07/20/1990
From:	Calvo J Office of Nuclear Reactor Regulation
To:	Hall W NUCLEAR ENERGY INSTITUTE (FORMERLY NUCLEAR MGMT &
References
NUDOCS 9007310036
Download: ML20055H852 (45)

Text

. _ _ _ - - - - _ _ _

s

?

July 20, '990 Mr. Warren J. Hall, Manager Operations, lisnegenent end 5:Jpport Se7 vices Division Nuclotr l'antgcr,.ent end Resources Cout cil 1770 Eye Street, Suite 300 Washington, DC 20006-2496

Dear Mr. Hall:

Enclosed is the first draft for the new STS section S.O Administrative Centrols. This draft incorporates the results of our meeting with NUMARC and the Owners Croups on May 23-24, 1990.

Sore additions and changes have been c4de to the section since this meeting. These portions are highlighted in the first draft for the Owr.crs Groups' review.

Sincercly, ORIGINAL. SIGNED BY JOSE A. CALVO Jose A. Calvo, Chief Technical Specifications Branch Division of Operational Events Assessment NRR

Enclosure:

As stated oc:

C. DeDeaux, CE0G Chairman DISTRIEUTION: w/o Enclosure J. Ilinds, WOG Chairwoman TEIETeT CHBerlinger P. North, B&WOG Chairman FJMiraglia FWEaranowsly J. Robertson, Acting BWROG Chairman WTRussell FGillespie E. Lczito, VEPC0 CERossi JLieberman E. Woods, SCE JGPartlow ELJordan K. Wilson, FPC DMCrutchfield WCKennedy J. Fowler, SERI SAVarga MGMalsch H. Bryan, TVA GMHolahan OTSB Members R. Ba ker, GPC ACThadani OTSB R/F BKGrimes 00EA R/F JWRoe FJCongel JHConran PDR w/ Enclosure

< Centra 14 Files?w/ Enclosure,s I

I 9007310036 900720 hill REVGPERGNgligC l

PDR 1

pum l

COCUMENT NAME:

LTR TO HALL 5.0 CONTROLS NAN i

WVM 4

l 01SB:00EA:hRR TSB:00EA:NRR OTSB:00EA:NRR C:0TSB:00EA-NYG111es

)

FMReinhart RMLobel JACelvo

'?

i

)

T/)7 /90 y

7 /1b/90

-) /t0/90 7 /2 0/90 003 1;

k 4

I g

4 t

SECTION 5.0 ADMINISTRATIVE CONTROLS t

1 l

l

Administrative Controls 5.0 5.1 RESPONSIBILITY 5.1.1 The (Plant Superintendent) shall be responsible for overall unit operation and shall delegate in writing the succession to this responsibility during his absence.

5.1.2 The (Shift Supervisor) (or during his absence from the control room, a designated individual (see Table 5.2-1)) shall be responsible for the control room command function. A management directive to this effect, signed by the [ highest level of corporate management) shall be reissued to all station personnel on an annual basis.

5.2 ORGANIZATION 5.2.1 Onsite and Offsite Oraanizations 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 inter-mediate levels to and including all operating (,rganization positions.

These relationships shall be documented and updated, as appropriate, in the form of organization charts, functional descriptions of departmental responsibilities and relationships, and job descriptions for key person-nel positions, or in equivalent forms of documentation.

These require-ments shall be documented in the FSAR.

b.

The (Plant Superintendent) shall be resaonsible for overall plant safe operation and shall have control over tiose onsite activities necessary for safe operation and maintenance of the plant.

c.

The (a specified corporate executive position) shall have corpo.' ate responsibility for overall plant nuclear safety and shall takr any mea-sures needed to ensure acceptable performance of the staff ir operating, maintaining, and providing technical support to the plant te ensure nuclear safety.

d.

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

(continued)

UNIT NAME 5-1 07/17/90

)

o Ad2inistrative Controls 5.0 5.2 ORGANIZATION (continued) 5.2.2 Unit Staff The unit staff organization shall be as follows:

a.

Each on-duty shift shall be composed of at least the minimum shift crew composition shown in Table 5.2.2-1; b.

At least one licensed Reactor Operator shall be in the control room when fuel is in the reactor.

In addition, while the unit is in [ MODE 1, 2.

3 - BWRs) (Mode 1, 2, 3, or 4 - PWRs), at least one licensed Senior Reactcr Operator shall be in the control room; J

c.

A [ Health Physics Technician) shall be on site when fuel is in the 1

reactor.

The position may be vacant for a period of time not to exceed 1

two hours in order to provide for unexpected absence provided immediate action is taken to fill the required position; d.

Either a licensed Senior Reactor Operator or licensed Reactor Operator limited to fuel handling who has no other concurrent responsibilities 1

during this operation shall be present and directly supervise all CORE ALTERATIONS; e.

Administrative procedures shall be developeo and implemented to limit the working hours of unit staff who perform safety-related functions (e.g.,

licensed Senior Reactor Operators, licensed Reactor Operators, health physicists, auxiliary operators, and key maintenance personnel).

Adequate shift coverage shall be maintained without routine heavy use of overtime.

The objective shall be to have operating personnel work a normal 8-hour day, 40-hour week while the unit is operating. However, in the event that unforeseen problems require substantial amounts of overtime to be used, or during extended periods of shutdown for refuel-ing, major maintenance, or major plant modification, on a temporary basis the following guidelines shall be followed:

1)

An individual should not be permitted to work more than 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> straight, excluding shift turnover time.

2)

An individual should not be permitted to work more than 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> in any 24-hour period, nor more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in any 48-hour period, nor more than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in any 7-day period, all excluding shift turnover time.

3)

A break of at least 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> should be allowed between work periods, including shift turnover time.

(continued)

UNIT NAME 5-2 07/17/90

Administrative Controls 5.0 5.2 ORGANIZATION (continued) 5.2.2 Unit Staff (continued) 4)

Except during extended shutdown periods, the use of ovtrtime should be considered on an individual basis and not for the entire staff on a shift.

Any deviation from the above guidelines shall be authorized in advance by the [ Plant Superintendent) or his deputy, or higher levels of management, in accordance with established procedures and with documentation of the basis for granting the deviation.

Controls shall be included in the procedures such tnt individual overtime shall be reviewed monthly by the [ Plant Superintendent) or his designee to assure that excessive hours have not been assigned.

Routine deviation from the above guidelines is not authorized.

f.

The [off shift position below) shall hold a Senior Reactor Operator license.

Operations Manager Assistant Operations Manager 9

The Shift Technical Advisor shall provide advisory technical support to the Shift Supervisor in the areas of thermal hydraulics, reactor engineering, and plant analysis with regard to the safe operation of the unit.

(continued) f l

l l

i VNIT NAME 5-3 07/17/90 3

g

Administrative Controls 5.0 TABLE 5.2.2-1 (Page 1 of 2)

MINIMUM SHIFT CREW COMPOSITION (SINGLE UNIT FACILITY)

POSITION UNIT IN MODE

1, 2, or 3 4 or 5 - BWRs]

I:1, 2, 3 or 4 5 or 6 - PWRs)

SS 1

1 SRO 1

None l

R0 2

1 A0 2

1 STA 1

None

]

l NOTES a.

Table Notation SS - Shif t Su?ervisor with a Senior Reactor Operator license SRO - IndividLal with a Senior Reactor Operator license R0 - Individua with an Reactor Operator license A0 - Auxiliary Operator STA - Shift Technical Advisor b.

The Shift Technical Advisor (STA) position may be filled by an on-shift Shift Supervisor (SS) or Senior Reactor Operator (SRO) provided the i

individual meets the Commission Policy Statement on Engineering Expertise on Shift.

c.

The shift crew compo:P. ton may be one less than the minimum require-ments of Table 5.2.2-1 for t. period of time not to exceed.two hours in order to accommodate unexpected absences of on-duty shift crew members provided immediate action is taken to restore the shift crew composi-tion to within the minimum roquirements of Table 5.2.2-1.

This provision does l

not permit any shift crew position to be unmanned upon shift change due to an oncoming shift crewman being late or absent.

(continued) i T

s UNIT NAME 5-4 07/17/90

e 8 0 Administrative Controls 5.0 TABLE 5.2.2-1 (Page 2 of 2)

MINIMUM SHIFT CREW COMPOSITION

[ SINGLE UNIT FACILITY)

NOTES (continued) d.

During any absence of the Shift Supervisor from the control room while the unit is in [ Mode 1, 2 or 3 - BWRs) [ Mode 1, 2, 3 or 4 - PWRs), an individual with a valid Senior Reactor Operator license shall be designated to assume the control room command function.

During any absence of the Shift Supervisor from the control room while the unit is in [ Mode 4 or 5 - BWRs] [ Mode 5 or 6 - PWRs), an indiv'!.'al with a valid Senior Reactor Operator license or Reactor Operator license shall be designated to assume the control room command function.

i i

l UNIT NAME 5-5 07/17/90 f

e

--e

Administrative Controls 5.0 TABLE 5.2.2-1 (Page 1 of 2)

MINIMUM SHIFT CREW COMPOSITION

[TWO UNITS WITH A COMMON CONTROL ROOM)

(TOTALS FOR BOTH UNITS) lEACH UNIT IN MODE 1, 2 OR 3 BWRs]

FOSITION lEACH UNIT IN MODE 1, 2, 3 OR 4 PWRs]

SS 1

SRO 1

R0 3

A0 3

STA 1

(ONE UNIT IN MODE 1, 2, 3 AND f

ONE UNIT IN MODE 4, MODE 5 OR DEFUELED - BWRs)

[0NE UNIT IN MODE 1, 2, 3, OR 4 AND POST. TION ONE UNIT IN MODE 5, MODE 6 OR DEFUELED - PWRs]

SS 1

SR0 1

R0 3

A0 3

STA 1

EACH UNIT IN MODE 4, MODE 5 OR DEFUELED - BWRs)

POSITION lEACH UNIT IN MODE 5, MODE 6 OR CEFUELED - PWRs)

SS 1

SRO None R0 2

A0 3

STA None I

UNIT NAM.!

56 07/17/90

e Oo Administrative Controls 5.0 TABLE 5.2.2-1 ( Page 2 of 3)

MINIMUM SHIFT CREW COMPOSITION (TWO UNITS WITH COMMON CONTROL R00M]

(TOTALS FOR BOTH VNITS)

NOTES a.

Table Notation SS -

Shift Supervisor with a Senior Reactor Operator license for each unit whose reactor contains fuel.

SRO -

Individual with a Senior Reactor Operator license for each unit whose reactor contains fuel.

Otherwise, provide an individual for each unit who holds a Senior Reactor Operator license for

)

the unit assigned. During CORE ALTERATIONS on either unit at least one licensed SRO or licensed SRO limited to fuel handling, who has no other concurrent responsibilities, must be present.

R0 -

Individual with a Reactor Operator license or a Senior Reactor f

Operator license for unit assigned.

At least one R0 shall be assigned to each unit whose reactor contains fuel and one R0

' hall be assigned as relief operator for unit (s) in (Mode 1, 2, 3 - BWRs] [ Mode 1, 2, 3 or 4 PWRs).

Individuals acting as relief operators thall hold a license for both units, f

Otherwise, provida a relief operator for each unit who holds a license for the utit assigned.

A0 -

At least one auxiliary operator shall be assigned to each unit whose reactor contains fuel.

STA -

Shift Technical Acvisor b.

The Shift Technical Advisor :STA) position may be filled by an on shift Shift Supervisor (SS) or Senior Reactor Operator (SRO) provided the individual meets the Commission Policy Statement on Engineering Expe tise I

on Shift.

l c.

The shift crew composition may be one less than the minimum rer,uirements of Table 5.2.2-1 for a period.of time not to exceed two hours in order to accommodate unexpected absence of on-duty shif t crew members provided immediate action is taken to restore the shift crew composition to within t

the minimum requirements of Table 5.2.2-1.

This provision does not permit any shift crew position to be unmanned upon shift change due to an oncoming shift crewman being late or absent.

(continued) l l

UNIT NAME 5-7 07/17/90

~

s Adainistrative Controls 5.0 TABLE 5.2.2 1 (Page 3 of 3)

MINIMUM SHIFT CREW COMPOSITION

[TWOUNITSWITHCOMMONCONTROLROOM)

NOTES (continued) d.

During any absence of the Shift Supervisor from the control room while the unit is in [ Mode 1, 2 or 3 - BWRs) [ Mode 1, 2, 3 or 4 -PWRs), an individual with a valid Senior Reactor Operator license shall be designated to assume the control room command function.

During any absence of the Shift Supervisor from the control room while the unit is in

[ Mode 4 or 5 BWRs] [ Mode 5 or 6 - PWRs), an individual with a valid senior Reactor Operator license or Reactor Operator license shall be designated to assume the control room command function.

?

UNIT NAME 5-8 07/17/90

Administrative Controls 5.0 TABLE 5.2.2 1 (page 1 of 2)

MINIMUM SHIFT CREW COMPOSITION

[TWOUNITSWITHTWOCONTROLROOMS)

+

'WITH THE OTHER UNIT IN MODE 1, 2 OR 3 BWRs)

WITH THE OTHER UNIT IN MODE 1. 2, 3. OR 4 PWRs]

POSITION UNIT IN MODE

1, 2, or 3 4 OD. 5 BWR,Q

1, 2, 3 or 4 5 or 6 PWRt i

i SS 3(a) 3(e)

SRO 1

NONE R0 2

1 A0 2

1 STA 1(*)

NONE e

'WITH THE OTHER UNIT IN MODE 4 OR 5 DEFUELED BWRs1 WITH THE OTHER UNIT IN MODE 5 OR 6 DEFUELED PWRs' POSITION UNIT IN MODE l1,2,OR3 4 OR 5 BWRs)

1, 2, 3, OR 4 5 OR 6 PWRs]

$3 t3e) 3(a)

SRO 1

NONE R0 2

1 A0 2

2*

STA 1

NONE t

UNIT NAME 59 07/17/90

Administrative Controls 5.0 TABLE 5.2.2 1 (Page 2 of 2)

MINIMUM SHIFT CREW COMPOSITION (TWO UNITS WITH TWO CONTROL ROOMS)

NOTES a.

Individual may fill the same position on the other unit if licensed for

both, b.

One of the two required individuals may fill t.'.e same position on the other unit.

c.

Table Notation SS - Shift Supervisor with a Senior Reactor Operator license.

SR0 - Individual with a Senior Reactor Operator license.

R0 - Individual with an Reactor Operator license.

E A0 Auxiliary Operator STA - Shift Technical Advisor d.

The Shift Technical Advisor (STA) position may be filled by an on-shift Shift Supervisor (SS) or Senior Reactor Operator (SRO) provided the individual meets the Commission Policy Statement on Engineering Expertise on Shift.

e.

The shift crew composition may be one less than the minimum requirements of Table 5.2.2-1 for a period of time not to exceed two hours in order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore the shift crew composition to within the minimum requirements of Table 5.2.2-1.

This provision does not permit any shift crew position to be unmanned upon shift change due to an oncoming shift crewman being late or absent, f.

During any absence of the Shift Supervisor from the control room while the unit is in (Mode 1, 2, or 3 - BWRs] (Mode 1, 2, 3 or 4 -PWRs) an individual with a valid Senior Reactor Operator license shall be designated to assu'.ie the control room command function.

During any absence of the shift supervisor from the control room while the unit is in (MODE 4 or 5 BWRs) (MODE 5 or 6 - PWRs), an individual with a valid Senior Reactor Operator license or Reactor Operator license shall be designated to assume the control room command function.

UNIT NAME 5-10 07/17/90

)

Administrative Controls 5.0 5.3 UNIT STAFF QUALIFICATIONS (Minimum qualifications for members of the unit staff shall be specified by use of an overall qualification statement referencing an ANSI Standard acceptable to the NRC staff or, alternately, by specifying individual position qualifications.

Generally, the first method is preferable; however, the second method is adaptable to those unit staffs requiring special qualification statements because of a unique organizational structure.)

Each member of the unit staff shall meet or exceed the minimum qualifications of Regulatory Guide 1.8, Revision 2,1987, (or more recent revision or ANSI Standard acceptable to the NRC staff]. Those staff not covered by this Regu-latory Guide shall meet or exceed the minimum qualifications of [ Regulations, Regulatory Guides, or ANSI Standards :.cceptabic to the NRC staff].

In addition, the Shift Technical Advisor shall meet the qualifications specified by the Commission Policy Statement on Engineering Expertise on Shift.

5.4 TRAINING A retraining and replacement training program for the unit staff shall be maintained under the direction of the [ position title) and shall meet or exceed the requirements and recommendations of Section [

] of(anANSI Standard acceptable to the NRC staff) and 10 CFR 55, and, for appropriate designated positions, shall include familiarization with relevant industry operational experience.

5.5' REVIEWS AND AUDITS

[The lidensee shall ' describe the method (s) established t'o cbnduct ihdependent

~

T reviews.. The methods may take a range of forms acceptable to; the NRC.' -These may include creating =an organizational unit', ~ a standing or ad; hoc-committee or assigning individuals capable' of conducting,these reviews >and-audits.

When an individual-performs a review function,' a cross-disciplinary review determination is'necessary.

If deemed necessary, such review shall"be performed by the review personnel of the appropriateidiscipline.; sindividifal l

reviewers shallLnot review their own work' or' work for.which:theyTnave direct _

responsibility.

Regardless of the method used, the license ~e shallcspecify the function's,; organizational arrangement, responsibilities,0 appropriate ANSI 3.1-1981,-qualifications, 'and reporting requirements of'each functional "/ANS element. or unit that contributes toltheselprocessesc '~ ' ~ ~ "

P4 views ~and sudits ofTactiVities"affseting'ipiantisafety haWtW6Tdistihet elwnts.,iThe first ofc these is the review performed by plant 1 staff pers6hM1 to assare that day-to-day activities areiconducted in a safe manner.E These; ^

are' described.in Section 5.5.1. :Theisecond of these'(described)in'Section 5.5.2) is;the -(offsite] reviewiand audit ofLfacility[activitiesfind:progr'ains affecting : nuclear safety:that are performed >independentLof'thelplent1 staff.3 The:[offsite]) review and audit'should provide _ for theninteg' ration ofjthej

~

j

'1(con.tinued)

I VNIT NAME 5 11 07/17/90

b Administrative Controls 5.0 5.5 REVIEWSANDl AUDITS.(continued) reviews and audits into.a cohesive program'to provide' senior level utility

~

management with an assessment of facility operation and recommend actions to

. improve _ nuclear safety and plant reliability.

It should include an. assessment of the. effectiveness of reviews conducted according;to_ Section 5 5.1.)

l 5.5.1 ~ plant Reviews 1

(The licensee shall describe here"the' provisions fo' ~ plant reviews r

(organization, reporting, records) and appropriate ANSI /ANS standard _f6r personnel qual _i;ication.)

a.

"Fdnetions The' (plant review method specified in;5.5.1) shall, as a' minimum, incorporate the following functions:

~~

'~

1)

Advise the (Plant Superintendent) on all matters rel' tedit6 a

nuclear safety; 2)'

Recommend to the (Plant' Superintendent) approval 'or~disipproval of items considered under Specification 5.5.1.b.1 through "

5.5.1.b.6. prior to their implementation, except as provided in Speci.fication 5.7.3; 3)

Obtain approval from the (Plant Superintendent) of ~each proposed test or experiment' and proposed changes and modifications to. unit systems or equipment that affect nuclear' safety; prior to implementation.

4)

Determine whether each item considered under'Specififations 5.5.1.b.1 through 5.5.1.b.6 constitu.tes,an unreviewedisafety question; and 5)

Notify the (Vice President Nuclear Ope' rations) of' any ' safety significantdisagreementbetweenthe-(revieworganizationor individual specified in 5.5.1) and-the (Plant Superintendent) within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

However, the (Plant Superintendent) ~sha,11 have responsibility for resolution of such' disagreements pursuant _to Specification 5.1.1.

b..

Resoonsitiilit ies The (plant review method specified ini5.5.1) shall be ~usedito1 conducQas

~

a minimum, the following performance-based plantjreviews:

1)_

_ Review of"all p' rop'osed procedures requiFed bi SpecifiEht' ion 5.7.1.and changes thereto;

'(continued)

UNIT NAME 5-12 07/17/90

Administrative Controls 5.0 I

5:51 REVIEWS.AND;fAUDITS(contjnued) 5;5.1);

P1 ant' Reviews (continOsd)

~

b..

Responsibilitiksf(continued) 2)T 'Revisw 0f 7sllTproposedipfogramsjeqUirsd';bilSpecifilationl5~.7!4 and changes,the.reto; 3)f, Review of'all" proposed changes ~and modifications"toinnit systems or equipment-that affectinuclear safety; i

4)

Reyiew ofithe(Fire';PE0tectiin Program;ian[chtsges[thefet'oj 5);

.Rsview off allLproposedLt'sstsTand[experimentsith;at(affest[nucleM safety; 6)~ LReview^of.all' proposed chan'gesltoLthese?Tsthnical._...

Specifications, their. Bases an.d:the_0perating License; c

5. 5. 2"

'rof fsitel" Review ~an'd f Audit (Ths licensee shall'descr'ibelh~eFs/the1provisibns for reviews?and a0diti i

i independent of. the' plants staff (organization, reporting, records) land appropriate ' ANSI /ANS standards' fort personnel; qualifications. :These

~

individuals may be < located onsite or offsite providedLorganizational

. independence from plant' staff = is maintained. ':The' technical review '

responsibility, 5.5.2.d, shalliinclude severaliindividuals L1ocatedionsi.te.]

I

'a.

Functions Ths (offsite' review'snd addit: pro' ision"sisp'ecifiedfini5.5;2]?.shal.l?iij v

mi.nimum incorporateithe following functi.ons:

1)' " Advise ths (Vics -Presidsnt' Nuclear 70perati6ns]Ponlill? matters related to nuclear safety and make' recommendation'stfor~' ~'~~

improving. nuclear; safety and plant l

rel,iability;

"~'

2)

Advise the m'anigcWent; 'of the" audited!6rganizationMand}ths

'[Vice President-- Nuclear Operations]u f;the audit gesults ai o

T they relate;to nuclear; safety; 3)'~ l Recommend?tojthe inanagsin6nt'of theTaUditedioigahizitioiiTihd its management,iany corrective. action toijmprove; nuclear (safety and planti. operation;

[4)" f" Notify tble '[ViceL Piesident% Nucle'arlOri'rit'16ns]"offsnj7sifeti

~

s significantdisagreement'betweenLthe'[revieworgan_izationt.or

individual specified in 5.5.2):andnthel[ organization orifuncti6n being; reviewed).within;:24 hours.

^~

(continued) i UNIT NAME 5-13 07/17/90 1

Administrative Controls 5.0 5l.5. REVI.EWS AND AUDITS 5.5;2'

' Offsite Review "and AudK(sontinUed) b; "f0ffsitel Review Resbohsibilities The' { rsview msthod's~p'e'ci fiedlin l5) $'. 2)lstia11l: tieir:esponsipKfpr3hs revi_ew of:

1); 'The7s'afetyi eva10ationi for'1)"chahies ti pro'esdUresfe4UipinintT"6V systems-and'2); tests.or experiments completed under the provision of 10 CFR*50.59,- tol verify that such actions did not constitute an" ~

unreviewed safety questi.on;

~ ' ~ ~ ' ' ' ~ ' '

~~

2)

. Proposed changesTtoTprocedUresJequipment7orl'systeins?whichlinvb1fe an'unreviewed safety question as;definediin 10 CFR.50.59; i

c Proposed' t'ests or'expbriments whi6h invo1Ve an' unr'esiewedisafety 3)

~

~

~

question as,de.fi.ned. in:10 CFR 50. 59; 4)1 License; Proposed changes to TechnicaliSpecificationsfarid th'ef0perating 5 )"

Violations'oflc6 des,1regulat' ions,"ordersflisense reqdfrements, j

and of internal l: procedures or instructions,having nuclear safety s_ignificance; 6) l All Licens.ee' EventlRepotts' requiredf by 101CFR]50(73;;;

7) l Plant ' staff perfo'rman.ce.;

l 8); ! Indications of unanticipated ~ defic'ientissin~anylaspectT f o

design orl operation of structures, systems l, ~or. components.th_at could; affect; nuclear safety;

~' ~ ~ ~

^~'

9)"

Significant" accidental, unplanned;;^or u'nconWo116dTrsdi6 active releases'_includ.ing corrective action to preve_nt;r.ecurrence.;.

10)' Significant operating abnormalities or" deviations" from normal and expected performance of equipment thatiaffectinuclear 1

safety;;

ll)lThe perform'ance of Lthe corrective _ a6_ tion " system.

~

s m_

44 Y

v 3

p

+

v dva a.vh UNIT NAME 5-14 0//17/90

Administrative Controls 5.0 l

S'.51 rey!EWSAND. AUDITS _(contin'ued) 5;5.2; f f0ffsitel Reviei and AuditT(contih0ed) c.

Aiidit Resoonsibilities Jheladditirespons1bilitiesshallfeneomplisif

~

1 )"" The conformance bf" unit ~op'er'atioh~t6 proVisi'ons~ contained

~ ithin the Technical Spectf. cations _and-.applicablej icens6 w

i conditions; i

2); iTh'eltraining' and jualificati6nsloff the.Lilnitistiff;;

3); lThslimplementatfon,of a111pF.ogramifre;qufred. bi;Specifil cat;ijon 5.7.4; 4)L TActions takenito' correct deficiencies oscarFing?in;ehUipinent,

~

structures,.. systems, components,;or me.thod.of operation that affect nuclear. safety;-and j

5) lThe 'performarice"of'~ activities requiredj tolmeetithe~requiFement( of Appendix B to 10 CFR,50, 6)1 Other acitVities'a'nd 'd6 clime 6t's asfreguestsdibfiths((Vibe President-NuclearOperat, ions);

d.-

'Teehnical~ kevies Responsibilit ies

.The ^ t echni cal.. reVi eF rssponsi bil i t ie s l s h'a11?encompa s s i 1)'

Plant operating ~ characteristics, NRCTissuances,'industFy1adVis6 ries; Licensee Event Reports and;other; sources which may;indicatelareas' '

for improying plant; safety; n

2)'

Plant operations, modificat' ions, maintenance'andLsiirveillsnde'siito

. independently verify that these activities,are: performed safely.and l

c.orrectly and that human; errors are. reduced as;much as practical; 3);. Internal'and externa 17operit'ional? experience' info.rmatio[tha_t 'may indicate 1 areas for i_mproving'plantjsafety; i

4)?.' Making deta'iled recomm'endati6nsithFough:ithe'[Vice: President?

~

" Nuclear.0perations) for revising procedures,cequipment:modificafions or other means-of improving M iear; safety 1andplant(reliabilityf

.,x s

~, -, _ _.

v. 4 1 4 e.+

d _ %.,

y 9 1

UNIT NAME 5 15 07/17/90 f

Administrative Controls 5.0 5.57REVIEWSAND' AUDITS 1(contir.ded)

$15;2f ' f 0ff 6tel' Revibw 'and' Audit 1(E6htMue'd)

$;5.3]fRecoidt Wriitin Yecords of reviews 1 and ladiitFsh'allitss""sintainid. ' Rep'oi tiToE Feid'Edi of activities shall: be forwarded to' the [Vice1 President - Nuclear Operations)'

within130l days;following compl.eti.on of;.the;reMew;.or auditc : Aslaj inimum ~

theselecords sh_all;jnclude:.

-([

Rs ful t F6 f "t hs~it tFi,t i e siEb nd u ctid [6 hd eKt EeMov i.ii ohilf SPecifichtion 5.5 bl(Re4oiimenditibnsitolthelminagimint3[thi[otgsniistib[bsihi)Jdditidi

' C 'AiGis'sessmsntT6fithelsafetylsignifib;sh6's[offthilifiK6~r;;(iildit! finding ^si c

d..

Recommesded' approvali or disipprovalfoffi(EsTebssiderid ~undei 1

~ Specification;5.5..1)b.1 through_5.5.1.b 6jland~~'~

" ' " ~ ' ' '

e.T ' DeteFminati'on of whetherTeach71teinlonsidsrsdTUndsESpe61fidtion

~ ~ ' 5. 5. ). b.. I th rough; 5. 5'.1. b. 6 con s ti t u,t e si an ; un rev.i eweda s,a fety1que stib n?

5.6 TECHNICAL SPECIFICATION BASES CONTROL Changes to the Bases of the Technical Specifications shall be made under I

appropriate administrative controls and reviewed according to Specification 5.5.1.

Licensees may make changes to Bases without prior NRC approval provided they do not involve:

1)

A change in the Technical Specifications incorporated in the License, or

.l 2)

A change to the updated FSAR that involves an unreviewed safety question as defined in 10 CFR 50.59, or 3)' " A"chinde - tolthe"wsy that" (a) 'OPERABillTYl6ri(b)Tths' Tschtiisal Specification could be met,5 applied ordinteppreted.

~ ' ~ ~ ^

Proposed changes which meet the criteria of (1) or (2) or (3) above shall be reviewed and approved by the NRC prior to implementation.

Changes to the Bases which may be implemented without prior NRC approval will be provided to the NRC et least annually.

(continued)

UNIT NAME 5 16 07/17/90

Administrative Controls 5.0 5.7 PROCEDURES, PROGRAMS AND MANUALS 5.7.1 Procedures 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 1 to NUREG-0737 as stated in Generic letter No. 82-33; c.

Security Plan implementation; d.

Emergency Plan implementation; e.

Quality Assurance for effluent and environmental monitoring; f.

Fire Protection Program implementation; g.

All programs specified in Technical Specification 5.7.4;[and]

VS CE (h.

Modification of Core Protection Calculator (CPC) Addressable Constants.

These procedures should include provisions to assure that sufficient margin is maintained in CPC Type I addressable constants to avcid excessive operator interaction with the CPCs during reactor operation.

Modifications to the CPC software (including changes of algorithms and fuel cycle specific data) shall be performed in accordance with the most recent version of "CPC Protection Algorithm Software Change Procedure,"

CEN-39(A)-P that has been detecmined to be applicable to the facility.

Additions or deletions to CPC addressable constants or changes to I

addressable constant software limit values shall-not-be implemented l

without prior NRC approval.)

1 5.7.2 Review and Aooroval Each procedure of Specification 5.7.1, and changes thereto, shall' be reviewed in acccrdance with Specifications 5.5.1, approved:by the [ Plant Superiatendent] prior to implementation and reviewed periodically as set forth in administrative procedures.

(continued)

L I

1 VNIT NAME 5 07/17/90

Administrative Controls 5.0 5.7 PROCEDURES, PROGRAMS AND MANUALS (continued) 5.7.3 Temporary Chq.ngn n

Temporary changes to procedures of Specification 5.7.1 may be made provided:

a.

The intent of the existing procedure is not altered; b.

The change is approved by two members of the plant management staff, at least one of whom holds a Senior Reactor Operator license on the unit affected; and c.

Within 14 days of implementation the change is documented and reviewed in accordance with Specifications 5.5.1 and approved by the (Plant Superintendent].

5.7.4 Proarams and Manuals The following programs shall be established, implemented, and maintained:

a.

Radiation Protection Program Procedures for personnel radiation protection shall be' prepared consistent with the requirements of 10 CFR 20 and shall be approved, maintained, and adhered to for all operations involving personnel radiation exposure.

b.

Process Control Program (PCP)

The PCP shall contain the current fora.slas, sampling, analyses, tests, and determinations to be made to ensurS that processing and packaging of solid radioactive wastes will be accomplished in such a way as to assure compliance with 10 CFR 20, 61, and 71, state regulations, burial ground requirements, and other requirements governing the disposal of solid radioactive waste.

Licensee initiated changes to the PCP:

1)

Shall be documented and records of reviews performed shall be retained. This documentation shall contain:

a)-

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

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.

(continued)

UNIT NAME 5-18 07/17/90 w

=-.

j o

Administrative Controls 5.0 5.7 PROCEDURES, PROGRAMS AND MANUALS (continued) 5.7.4 Proarams and Manuals (continued)

)

b.

Process Control Program (continued) 2)

Shall be effective after review and acceptance by the [ review method

]

specified in 5.5.1.] and the approval of the [ Plant Superintendent),

c.

Offsite Dose Calculation Manual (0DCM)

The ODCM shall contain the methodology and parameters used:

1)

In the calculation of offsite doses resulting from radioactive gaseous and liquid effluents; 2)

In the calculation of gaseous and liquid effluent monitoring alarm and trip setpoints; and 3)

In the conduct of the Environmental Radiological Monitoring Program.

The ODCM shall also contain:

1)

The Radioactive Effluent Controls and Radiological Environmental' Monitoring programs required by Technical Specification 5.7.4; and 2)

Descriptions of the information that should be included in the Annual Radiological Invirontaental Operating, and Semiannual Radioactive Effluent Release Reports required by Technical Specifications (L 9.1.3] anG { 5. 9.1.4].

Licensee-initiated changes to the LDCM:

1)

Shall be documented and records of reviews performed shall be retained. Thic decomentet)un shall contain:

(

a)

Sufficient information to support the change together with the appropriate analyse: cr evaluations justifying the change (s);

and I

b)

A determination that the change will maintain the level of l

radioactive effluent control required by 10 CFR 20.106, 40 CFR L

190, 10 CFR 53.36a, and Appendix I to 10 CFR 50 and not adversely impact the accuracy or reliability of effluent, ' dose, or setpoint calculations.

(continued) i UNIT NAME 5-19 07/17/90

o.

Administrative Controls 5.0 5.7 PROCEDURED, PROGRAMS AND MANUALS (continued) 5.7.4 Proorams and Mann]l (continued) c.

ODCM (continued) 2)

Shall become effective after review and acceptance by the (review method specified in 5.5.3) and the approval of the (Plant Superintendent]~.

3)

Shall be submitted to the Commission in the form of a complete, legible copy'of the entire ODCM as a part of or concurrent with the Semiannual Radioactive Effluent Release Report.for the period of the report in which any. change in 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 shall indicate the-date (e.g., month and year) the change was implemented.

d.

Primary Coolant Sources outside Containment This program provides controls to minimize leakage frca those portions of systems outside co'ntainment that could contain highly radioactive fluids during a serious transient or accident to levels as low as practicable.

The systems include [the recirculation spray, safety injection, chet:ical ar.d volume control, gas stripper, and hydrogen recombiners).

The program shall include the following:

1)

Preventive maintenance and periodic visual inspection requirements; and 2)

Integrated leak test requirements-for each system at refueling cycle intervals or less, e.

In-Plant Radiation Monitoring This program provides controls to ensure the capability to accurately determine the airborne iodine concentration in vital areas under accident conditions.

This program shall include the following:

1)

Training of personnel; 2)

Procedures for monitoring; and 3)

Provisions for maintenance of sampling and analysis equipment.

(continued)

J l

UNIT NAME 5 20 07/17/90

=. -

Administrative Controls 5.0 5.7 PROCEDURES, PROGRAMS AND MANUALS (continued)~

5.7 Proorams and Manuals (continued) f.

Post-Accident Sampling This program provides controls to ensure the capability to obtain and analyze reactor coolant, radioactive iodines and particulates in plant gaseous effluents, and containment atmosphere samples under accident conditions. The program shall include the following:

1)

Training of personnel; r

2)

Procedures for sampling and analysis; and 3)

Provisions for maintenance of sampling and analysis equipment, g.

Radioactive Effluent Controls Program This program is to conform 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 10 CFR 20, Appendix B, Table II, Column 2; 3)

Monitoring, sampling, and analysis of radicactive liquid and gaseous effluents in accordance with 10 CFR 20.106 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 each unit to. unrestricted areas conforming to Appendix I to 10 CFR 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; 1

(continued)

UNIT NAME 5-21 07/17/90 l

Administrative Controls 5.0 5.7 PROCEDURES, PROGRAMS AND MANUALS (continued) 5.7.4 Proarams and Manuals (continued) t g.

Radioactive Effluent Controls Program (continued) 6)

Limitations on the operability and use of the liquid and gaseous effluent treatment systems to ensure that the appropriate portionsof these systems are used to reduce releases of radioactivity when the projected doses in a 31-day period

+.'d exceed 2 percent of the guidelines for the annual dose or dos commitment conforming to Appendix I to 10 CFR 50;

'. imitations on the dose rate resulting from radioactive material

leased in gaseous effluents to areas beyond the site boundary

. forming to the dose associated with 10 CFR 20, Appendix B, Table n, Column 1; o)

Limi'

.ns on the annual and quarterly air doses resulting from noble t es released in gaseous effluents from each unit to areas beyond the site boundary conforming to Appendix 1 to 10 CFR 50; 9)

Limitations on the annual and quarterly doses to a member of the public from Iodine-131, Iodine-133, tritium, and all radionuclides in particulate form with half-liva: greater than 8 days in gaseous effluents releasec from each unit to areas beyond the site boundary conforming to Apptnuix 1 to 10 CFR 50; [and]

VS-GE

[10) Limitations on venting and purging of the Mark II containment through the StanJby Gas Treatment System to maintain releases as low as reasonably a.:hievable] [BWRs w/ Mark 11 containments]; and

11) Limitations oa the annual dose'or dose commitment to any member of the public due to releases of'racioactivity and to radiation from uranium fuel cycle sources conforming to 40 CFR 190, h.

Radiological Environmental Monitoring Program This program is for monitoring the radiation and radionuclides in the environs of the plant.

The program shall provide (1) representative measurements of radioactivity in the highest potential exposure pathways, and (2) verification of the accuracy of the effluent monitoring program and modeling of environmental exposure pathways.

The program shall (1) be contained in the ODCM, (2) conform to the guidance of Appendix I.to 10 CFR 50, and (3) include the following:

(continued) l UNIT NAME 5-22 07/17/90 1

1

Administrative Controls 5.0-l-

5,7 PROCEDURES, PROGRAMS AND MANUALS (continued) 5.7.4 Proorams and Manuals (continued) h.

Radiological Environmental Monitoring Program (continued) 1)

Monitoring, sampling, analysis, and reporting of radiation and radionuclides in the environment in accordance with the methodology and parameter. in the ODCM; 2)

A Land Use Census to ensure that changes in the use of areas at and beyond the site boundary are identified and that modifications to the monitoring program are made if required by the results of this census; and 3)

Participation in an Interlaboratory Comparison Program to ensure that independent checks on the precision and accuracy of the measurements of radioactive materials in environmental sample matrices are performed as part of the quality assurance program for environmental monitoring.

i.

Component Cyclic or Transient Limit This program provides controls to track the FSAR Section [

] cyclic and transient occurrences to ensure that components are maintained within the design limits.

j.

Containment Leak Rate Test Program This program provides controls to ensure that the containment leak rate tests are performed to ensure containment leak tightness, a requirement for OPERABILITY. The program shall include the following surveillances required by 10 CFR 50, Appendix J:

1)

Type A tests (0verall integrated containment leakage rate);

2)

Type B tests (Local penetration leak rates);

3)

Type C tests (Containment isolation valve leakage' rates);

4)

Air lock seal leakage and air lock overall leakage rates; 5)

Isolation valve and channel weld pressurization = system pressure verifications; and 6)

[

]-inch purge supply and exhaust-leakage rates.

.~.

(continued)

UNIT NAME 5-23 07/17/90

i.

Administrative Controls-5.0 l

5.7 PROCEDURES, PROGRAMS AND MANUALS (continued) 5.7.4 ProaramsandManuals(continued)

I k.

Pre stressed Concrete Containment Tendon Surveillance Program This program provides controls for monitoring any tendon degradation in l

prestressed concrete containments to ensure containment structural integrity, a requirement for OPERABILITY.

The program shall include baseline measurements prior to initial operations. The Tendon Surveillance Program shall include at least the following:

1)

Tendon lift-off to check tendon force; 2)

The number of tendons inspected for each tendon group; 3)

Tendon wire samples taken to check physical condition, tensile strength and elongation; 4)

Grease _ samples taken to check chemical. properties, physical I

appearance and presence of free water; j

5)

Measurement of grease voids; 6)

Visual inspection of end anchorage and containment 1

exterior surface for cracking and grease leakage; 7)

Procedures for establishing inspection frequencies; 8)

Acceptance criteria; 9)

The content and frequency of reporting; and

10) Remedial actions when one or more of the acceptance criteria are not met.

The Tendon Surveillance Program and all proposed changes thereto shall be l

reviewed and approved by the NRC staff prior to implementation.

1 (continued) 1 a

UNIT NAME 5-24 07/17/90

i -

Administrative Controls 5.0 5.7 PROCEDURES, PROGRAMS AND MANUALS (continued) 5.7.4 Proarams and Manuals (continued) 1.

Inservice Inspection Program This program provides controls for inservice inspection of ASME Code Class 1, 2, and 3 components. The program shall include the following:

1)

Inservice inspection of ASME Code Class 1, 2, and 3 components shall

.be performed in accordance with Section XI of the ASME Boiler and a

Pressure Vessel Code and applicable Addenda, as required by 10 CFR-50.55a(g), except where relief has been granted by the Commission pursuant to 10 CFR 50.55a(g)(6)(1) and (a)(3).

VS-CE,

[2)

Inspection of each reactor coolant pump flywheel per the W, B&W recommendations of Regulatory Position C.4 b of Regulatory Guide l.14, Revision 1, August 1975.]

3)

The provisions of SR 3.0.2 ar applicable to the above required frequencies for performing.ii..arvice inspection activities.

VS-GE

[4)

An inservice inspection program'for piping identified in NRC Generic Letter 88 01 in accordance with the NRC. staff positions on schedule, methods, personnel, and sample expansion included in this generic letter or in accordance with alternate measures approved by the NRC staff.]

5)

Nothing in the ASME Boiler and Pressure Vessel Code shall be construed to supersede the requirements of any Technical Specification.

m.

Inservice Testing Program f

This program provides controls for inservice testing of ASME Code Class -

1, 2 and 3 components. The program shall include the following:

1 1)

Inservice testing of ASME Code Class 1, 2 and 3 pumps and valves i

l shall be performed in accordance with Section XI of the' ASME Boiler i,

and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50.55a(g), except where specific written relief has been granted-by the Commission pursuant to 10 CFR.50.55a(g)(6)(1) and (a)(3).

1 (continued).

c 4

e UNIT NAME 5-25 07/17/90 t

V Administrative Controls 5.0 5.7 PROCEDURES, PROGRAMS AND MANUALS (continued) 5.7.4 Procrams and Manuals (continued) 1 m.

Inservice Testing Program (continued) 2)

Testing frequencies specified in Section XI of the ASME Boiler and Pressure' Vessel Code and applicable Addenda as follows:

ASME Boiler and Pressure Vessel Code and applicable Addenda Required frequencies for terminology for inservice performing inservice testina activities testina activities

. Weekly

-At least once per 7 days Monthly At least once per 31 days Quarterly or every 3 months At least once per 92 days Semiannually or every 6 months At least once per 184 days Every 9 months At least once ;,er 276 days:

Yearly or annually At-least once per 366 days Biennial or'every two years At least once per 731 days 3)

The provisions of. SR 3.0.2 as applicable to the above required frequencies for performing inservice testing activities.

4)

The provisions of SR 3.0.3 as applicable to inservice testing activities.

5)

Nothing in the ASME Boiler and Pressure Vessel Code shall be construed to supersede the requirements of any Technical Specification.

(n.

Steam Generator Tube Surveillance VS-B&W, This program provides controls for; monitoring steam generator tube W, CE degradation.

Each steam generator shall be demonstrated OPERABLE by meeting the requirements of Technical Specifications 5.7.4.1 and by performance of an approved augmented inservice inspection program which includes at least the following:

1)

Steam generator sample selection and inspection; 2)

Steam generator. tube sample selection and inspection; 3)

The establishment of inspection frequencies; 4)

Acceptance criteria; and 5)

The content and frequency of reports.

(continued)

UNIT NAME 5-26 07/17/90

&' i' Administrative Controls i

5.0 5.7 PROCEDURES, PROGRAMS AND MANUALS (continued) 5.7.4-Proarams and Manuals (continued) n.

Steam Generator Tube Surveillance Program (continued)-

The Steam Generator Tube Surveillance Program and all proposed changes thereto shall be reviewed and approved by the NRC Staff prior to implementation.]

[o. Secondary Water Chemistry VS-W,Ci This program provides controls for monitoring secondary water chemistry to inhibit steam generator tube degradation.

The program shall include:

1)

Identification of a sampling schedule for the critical variables and control points for these variables; 2)

Identification of the procedures used to measure the values of the critical variables; 3)

Identification of process sampling poin, which shall include monitoring the discharge of the condent b ' umps for evidence of concenser in-leakage; 4)

Procedures for the recording and-management of data; 5)

Procedures defining corrective actions for all off-control point chemistry conditions; and 6)

A procedure identifying:

(a) the authority responsible for the interpretation of the data, and (b)' the sequence and timing of administrative events required to initiate corrective action.]

& lVdntil'at;ilon]li)MTsi[finMEogfani A!pFogram :shalliblei esthblistisaltodhipliindhBWhif61)1bWincjifedul^ Fed testing!ofs fiJtersfiniaccorda' ceWit((Regb)atoMGujds015.52]jRevjijjhD n

_orMNSl;..N510-) 9.80]..

jf vin}pljiRpenitfiil1onMg/bypas(Woit}QhtMJjfR(D0f}llMiQ 2)T"Wh1FcedehettitioOEUMiMdfTcMidMisifst3iEfiM 3RJMilhyl{i6didijehitfit1;6nEtbiG6DMihirjji))isanijiliI i

sMfibviritesidWiisup?Jgssgid

~

5)D6 stir 3aehteiG (continued)

UNIT NAME 5-27 07/17/90

Administrative Controls 5.0-5.7 PROCEDURES, PROGRAMS AND MANUALS (continued) 5.7.4 Proarams and Manuals (continued) q..

Explosive Gas and Storage Tank Radioactivity Monitoring Program This program provides assurance that (1) the con;entration of potentik1ly explosive gas mixtures contained in.the (waste gat holdup system].is maintained below the' flammability limits of hydrogen and oxygen; (2) in the event of an uncontrolled ' release of gaseous waste storage tank contents, the resulting offsite radiological consequences will not exceed a small fraction of the dose reference values in 10 CFR 100; and (3) in the event of an uncontrolled release of outdoor liquid storage tank-contents, the resulting concentrations would be 'ess than the limits specified in 10 CFR 20 at the nearest potable or surface water supply in an unrestricted area.

The program shall include:

1)

The limits for the concentration of hydrogen and oxygen it. the (waste gas holdup system] and a surveillance program te ensure the limits are maintained.

Such limits shall be appropriate to the system's design criteria; i.e., whether or not the system is designed to withstand a hydrogen explosion.

2)

The limits for the quantity of radioactive gas contained in.each gas storage tank and a surveillance program to ensure the limits are maintained.

3)

The limits for the quantity of radioactive material contained in unprotected outdoor tanks and a surveillance program to ensure the limits are maintained.

The limits specified in this' program and-any proposed changes thereto shall be reviewed and approved by the NRC staff prior-to implementation.

5.8 OPERABILITY DEFINITION IMPLEMENTATION RULES Compliance with the intent of the Technical Specifications requires adherence to the rules of this section regarding the implementation of the definition of l

OPERABILITY, l

(continued)

L UNIT NAME 5-28 07/17/90

-l g

~

Administrative C(ntrols 5.0 5.8 OPERABILITY DEFINITION IMPLEMENTATION RULES (continued) 5.8.1 Inocerability Rules a.

Discussion A system, subsystem, train, component, or device (hereafter referred to as the supported system) is inoperable when it is not capable cf performing its specified function (s). The capability to perform its specified function (s) require that all its support systems have the capability to perform their related support function (s).

Support systems are (from the OPERABILITY definition) all necessary attendant instrumentation, controls, electrical power, cooling or seal water, lubrication or other auxiliary equipment that are required for the supported system to perform its specified function (s).

A support system is inoperable when it is not-capable of performing its related support function (s).

b.

Rules I)

Determining the OPERABILITY of support and supported systems is an ongoing and continuous decision making process.

For the most part this process includes following procedures governing the day-to-day operation of the facility; performance of procedures.to implement surveillance requirements, inservice testing and inspection programs, programs prescribed by 5.7.4, and maintenance; routine plant walkdowns or tours, and observations of control room indications. Many procedures contain explicit acceptance criteria for OPERABILITY.

In scme cases, maintaining a system OPERABLE-requires adherence to th9 system operating procedure.

In addition to the above proactive process for determining OPERABILITY, there is a reactive process which takes place as part of the corrective action for a nonconforming or degraded condition on a support or supported system.

Upon identification of such a condition, promptly determine the affect of the condi'. ion upon the OPERABILITY of the affected support or supported system.

2)

Upon determining that a support or supported system is inoperable immediately declare the system inoperable (whether by nilure to meet a procedural 0PERABILITY acceptance criteria, or be ause of a nonconforming condition which makes the system inoperable, or some other reason).

(continued)

UNIT NAME 5-29 07/17/90

h6 Administrative Controls 5.0 5.8 OPERABillTY DEFINITION IMPLEMENTATION RULES (continued) p; E.8.1 Inocerability Rules (continued) 4 b.

Rules (continued) 3)

Upon declaring a support system inoperable,. immediately declare its supported system (s) inoperable.

Exceptions to this rule (such as delaying the supported system (s) inoperability declaration) are provided in certain support system LCOs, and 5.8.3.2 and Table 5.8-1.

The justification for these_ exceptions, usually contained in the suppuit r supported system Bases OPERABILITY discussion, must address why the supported system remedial actions are not applicable during the. allowed delay time.

4)

Intentional delay in declaring a support or supported system inoperable, once a determination of inoperability has been made is prohibited.

5)

An LCO Completion Time interval begins upon identification that the associated LC0 Condition exists.

For a sbpport or supported system LCO, this is the time the system is declared inoperable.

6)

The Completion Time for accomplishing the actions to restore a support system to OPERABILITY must not.be greater than the most limiting completion time of all the supported systems that are made inoperable when the support system.is declared inoperable.

7)

All changes to the OPERABILITY discussions in the LCO Bases for support ar.d supported systems and Table 5.8-1 delay times require-prior NRC review and approval, but not a license amendment.

All other changes to 5.8 require a license amendment.

5.8.2 Suocorted System inoperability Rule 5.8.2.1 Upon declaring a supported system inoperable, immediately enter the LC0 for the supported system, for all LC0 Conditions that apply.

Perform each Required Action within the allowed Completion Time.

5.8.2.2 Upon failure to perform tae Required Actions to restore the supported system to OPERA!!ILITY (and any other remedial actions specified) any time befora or at the end of the allowed Completion Time, take alternate actions specified in the LC0 such as bringing the plant to a MODE where the supported system LC0 is not applicable.

(continued)

UNIT NAME 5-30 07/17/90 1

J

p Administrative Controls a

5.0-5.8 OPERABILITY DEFINITION IMPLEMENTATION RULES (continued) 5.8.3 Suonort System Inocerability Rules (Assuming no loss of functional capability) 5.8.3.1 Suocort Systems in Technical Soecifications (TS)

Upon declaring a TS support system inoperable:

a.

Imc.ediately declare its supported systems inoperable and follow 5.8.2

'anless justified differently in c below.

b.

Immediately enter the LC0 of the support system for all onditions that apply.

Perforn the Required Actions within the specified Completion Times.

if th'e support system LC0 specifically permits it, delay declaring the c,

supported system (s) inoperable for the time allowed as long as (1) and (2) below are immediately verified and continue to be met.

Otherwise, immediately deciere the supported systems inoperable and follow 5.8.2.

1)

The justification fe the delay as describeo in the support system LC0 Bases is valid for the circumstances; and 2)

Redundant and/or diverse support and supported systems are OPERABLE.

d.

Upon failure to perform the Required Action to restore OPERABILITY of the TS support system (and any other remedial actions specified) any time before or at the end of the allowed Completion Time, take the action speci fier; in-the LC0.

5.8.3.2

.csooort Systems Outside TS a.

Upon declaring a non-TS support system inoperable:

j 1)

Immediately declare its supported systems inoperable and follow l

5.8.2 (; or 1

2)

If permitted by Table 5.8-1, delay declaring the supported systems inoperable for the time allowed as long as (1) and (2) below are immediately verified and continue to be met.

Otherwise, immediately.

declare the supported systems inoperable and follow 5.8.2.-

a)

The justification for the delay as described in the supported system LC0 Bases is valid for the circumstances; and b)

Redundant and/or diverse support and supported systems are OPERABLE.

(continued)

UNIT NAME 5-31 07/17/90

o*

s.

Adminictrative Controls 5.0 5.8 OPERABILITY DEFINITION IMPLEMENTATION RULES (continued) l 5.8.3 Suonort System Inocerability Rules (continued)

]

5.8.3.2 Sucoort Systems Outside TS (continued) b.

Upon failure to perform the required action to restore OPERABILITY to the non-TS support system (and any other remedial actions specified), any time before or at the end of the allowed delay time, either:

1)

Take the plant to a MODE in which the supported system LC0 is nct I

applicable, within the shortest shutdown completion time allowed among the supported system LCOs; or l

2)

Dv.lare the supported system (s) inoperable and follow 5.8.2.]

5.8.4 Ryle) Reaardina loss of Function-a.

A loss of function exists when all redundant and identified diverse systems designed to perform that function are inoperable, b.

A supported system LCO usually includes the overall knowledge to ascertain the loss of functional capability as well as the actions to mitigate it. When a supp_ ort and its supported systems are declared inoperable at the same time, the loss of functional capability need only be addressed as directed by the supported systems' LCOs.

When a support system is declared inoperable and the supported systems c.

are not immediately declared inoperable, the actions associated with restoring the support system to OPERABLE status must include verification-of OPERABILITY of:

1)

Redundant and diverse support system (s) in the opposite train; and 2)

Redundant and diverse supported system (s) in the opposite train; and 3)

All other support systems, for which a delay is permitted before declaring their supported systems inoperable, _in the opposite train.

If any of the above support or supported systems are found inoperable, a loss of functional capability condition may exist.

d.

Upon determining that a loss of functional capability condition exists, immediately accomplish all actions that mitigate the loss of functional capability.

Usually these actions are specified in the support or supported system LCO.

If not, take the actions specified in 5.8.3.2.b.

(continued)

UNIT NAME 5-32 07/17/90

(

i.

Administrative Controls c

5.0 5.8 OPERABILITY DEFINITION IMPLEMENTATION RULES (continued) 5.8.5 SuoDort and Subcorted Systems Association The licensee shall describe here the approach it established to associate TS and non TS support systems with TS supported systems.

(continued)

UNIT NAME 5-33 07/17/90

'T

Administrative Controls 5.0 Table 5.8-1 i

SUPPORT SYSTEMS NOT ADDRESSED BY TECHNICAL SPECIFICATIONS-ALLOWED DELAY TIMES AND FOLLOW-0N ACTIONS I

SAMPLE-Required Action Inoperable Allowed Upon Failure To Support Delay.

' Restore Operability Supported-3ystems &

System Time '

by End of Delay Related LC0(s)

Time 1.

Train A

[X] hours Follow 5.8.3.2(b)(2)

-LC0 3.8.2.1 Emergency-A.C. Distribution -

Switchgear Operating-I z

Room Vent-Train A Emergency ilation 4160 Vac Switchgear System i

t 3.

Train A

[X] hours Follow 5.8.3.2(b)(1)

LC0 3.8.2.3

(

Battery Room D.C. Distribution -

Ventilation Operating 1

System Train A Battery l

-1 i

UNIT NAME 5-34 07/17/90 l

I

11 c.

Administrative Controls 4

5.0

)

5.9 REPORTING REQVIREMENTS 5.9.1 Routine Reoorts i

The following reports shall be submitted in accordance with 10 CFR 50.4.

5.9.1.1 Startuo Report A summary report of plant startup and power escalation testing shall be submitted following:

(1) receipt of an Operating License, (2) amendment to the lic ense involving a planned increase in power level, (3) installation of fuel thet has a different design or has been manufactured by a different fuel supplier, and (4) modifications that may have significantly altered the nuclear, ti.armal, or hydraulic performance of the unit.

The initial Stai+up Report shall address each of the startup tests identified in Chapter 14 of the Final Safety Analysis Report and shall include a descrip-I tion 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.

Subsequent Startup Reports shall address startup tests that are necessary to demonstrate the acceptability of changes and modifications.

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 avents have been completed.

5.9.1.2 Annual Reports

...................................-NOTE-----

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

The submittal should combine those sections that are com. ion to all units at the station.

Annual Reports covering the activities of the unit as described below for the previous calendar year shall be submitted by March 31 each year. The initial report shall be submitted by March 31 of the year following initial criticality.

(continued)

UNIT NAME 5-35 07/17/90

j Administrative Controls 5.0 5.9 REPORTING REQUIREMEN'IS (continued) 5.9.1 Routine Report 1 (continued)

- 5.9.1.2 Annual Reports (continued)

Reports required on an annual basis shall include:

a.

Occupational Radiation Exposure Report 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 e

inspection, routine maintenance, special maintenance (describe mainten-ance], waste processing, and refuel _ing).. This tabulation supple-ments the requirements of Section 20.407 of 10 CFR 20. 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.

VS W-

b. Coolant Radioactivity Report The results-of specific activity analyses in which the primary coolant exceeded the limits of (appropriate Technical ~Specifiestion Section).

The following information shall be' included:

(1) Reactor power history starting 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> 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 <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> prior.to the first sample in which the limit was exceeded; (4)' Graphic of the I-131 concentration

(#Ci/gm) and one other radioiodine nsotope concentration (#Ci/gm) as a function of time for the duration of the specific activity above the steady-state level; and (5) The -time duration when the specific activity of the primary coolant exceeded the radioiodine limit.

(c. Any other unit unique reports required 01 an annual basis].

(continued) l

' UNIT NAME 5-36 07/17/90

)

l i

i m

a

~ -

.---r

.n 9

Administrative Controls 5.0 5.9 REPORTING REQUIREMENTS (continued) 5.9.1 Routine Reports (continued) 5.9.1.3 Annual Radiolooical Environmental Operatino ReDort

....................................N0TE---------...---....---------.........

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.

The Annual Radiological Environmental Operating Report covering the operation of the unit during the previous calendar year shall be submitted by May 15 of each year. The report shall include summaries, interpretations, and analyses of trends of the results of the Radiological Environmental Monitoring Program for the reporting period.

The material provided shall be consistent with the objectives outlined in (1) the Offsite Dose Calculation Manual (0DCM) and (2)

Section IV.B.2, IV.B.3, and IV.C of Appendix.I to CTR 50.

The" Ah60il 7 Rad i bl opiial ? E n M6'nme n t'al?0iieFa t'ihg E R e'FoVti?Ihal l Si nEl ndilt hs fe'sultsjof[analysestofj allt radiologic'al environmentabsamplesiandf of;alli environmental radiation 1 measurements;takenLduring:ithejperiod!pursuantitofth_e l ocati onsl speci fi ed ki n / the ntabl e cand t figures % the10f fsi te J Do'.ie ! Cal bul ation Manual,cas well; as summarizedianditabulateddesults?offthe'setanalysessand~

ineasurementsiin the' formattof cthe: table:in)the$ Radiblog'icah Asiessment(:Bfa'#Eli Technica14 Position,dRevisioni l3Novembes-)19796illnsthecev~entithatssome" ~ '

i nd i v i du al t re's ul t s 1 a rei no t Java i l a bl eI forii ncl us i on fwi t hf t he kepo r t e thiIIeji6M

~

s h al l i be Fs ubmi t ted Tnot i ng j a nd Le~x~p_1 a i n i ngithe7 resso n s % fodthe l mi's s i ng fres01 ts

,TheLmissingidatatshal] Lbeysubmitted as esconla's!possible;jnja7supplementari 'j report';

5.9.1.4 Semiannual Radioactive Effluent Release Report

.................................... NOTE-------------------------------------

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; however, for units with separate radwaste systems, the submittal shall specify the releases of radioactive material from each unit.

The Semiannual Radioactive Effluent Release Report covering the operation of the unit during the previous 6 months of operation shall be submitted within 60 days after January 1 and July 1 of each year.

The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit.

The material provided shall be (1) consistent with the objectives outlined in the ODCM and Process Control Program (PCP); and (2) in conformance with'10 CFR 50.36a and Section IV.B.1 of Appendix 1 to 10 CFR 50.

(continued)

UNIT NAME 5-37 07/17/90

Administrative Controls 5.0 l

- 5.9 REPORTING REQUIREMENTS (continued) 5.9.1 Routine Reoorts (continued) 5.9.1.5 Monthly Ooeratina Reports i

Routine reports of operating statistics and shutdown experience [, including documentation of all challenges to the PORVs or safety valves.] shall be submitted on a monthly basis no later than the 15th of each month following the calendar month covered by the report.

5.9.1.6 CORE OPERATING LIMITS REPORT (COLR)

Core operating limits shall be. established prior to each reload _ cycle, or a.

prior to any remaining portion of a reload cycle, for the following:

1)

[The individual specifications that address core operating limits must be referenced here.]

and shall-be documented in the CORE OPERATING LIMITS REPORT (COLR).

b.

The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically. those-described in the following documents:

1)

(Identify the Topical Report (s) by number, title, date, and NRC staff approval document, or identify the staff Safety Evaluation Report for a plant-specific methodology by NRC letter and date.]

The core operating limits shall be determined such that all applicable c.

limits (e.g., fuel thermal-mechanical limits, core thermal hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, transient analysis limits, and accident analysis limits) of the safety analysis are

met, d.

The COLR, including any mid-cycle revisions or supplements shall be provided, upon issuance for each reload cycle, to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.

5.9.1.7 RC? PRESSVRE AND TEMPERATURE LIMITS REPORT Reactor Coolant System (RCS) pressure and temperature limits including heatup and cooldown rates, criticality, and hydrostatic and leak tests limits shall be established and documented in the PRESSURE AND TEMPERATURE LIMIT 3 REPORT and submitted to the NRC staff for approval prior to initiation of a new reactor vessel fluency period.

(If desired, the individual specification that-addresses the reactor vessel pressure and temperature limits and the heatup and cooldown rates may be referenced.) The analytical methods used to deter-mine the pressure and temperature limits including the heatup and cooldown (continued)

UNIT NAME 5-38 07/17/90 l

l

+

.s Administrative Controls 5.0 5.9 REPORTING REQUIREMENTS (continued) 5.9.1 Routine Reports (continued) 5.9.1.7 RCS PRESSURE AND TEMPERATURE LIMITS REPORT (continued) rates shall be those previously reviewed and approved by NRC in [ identify the Topical Report (s) by number, title, date, and NRC staff approval document, or identify the staff's safety evaluation report for a plant specific methodology y

by NRC letter and 6 te].

The reactor vessel pressure and temperature limits including heatup axJ cooldown rates shall be determined so that all applicable i

limits (e.g., heatup limits, cooldown limits, inservice leak and hydrostatic testing limits) of the analysis are met.

The PRESSURE AND TEMPERATURE LIMITS REPORT, including revisions or supplements thereto shall be reviewed and approved by the NRC staff prior to= implementation.

Copies shall be provided, for each reactor vessel fluency period, in accordance with 1. CFR 50.4.

5.9.2 Special Reports

[Special Reports may be required covering inspection, test, and maintenance activities. These special reports are determined on an individual basis for each unit and their preparation and submittal are designated in the Technical Specifications.]

Special Reports shall be submitted to the Regional Administrator of the Regional Office of the NRC within the time period specified for each report.

[The following Special Reports must be submitted:]

a.

In the event an ECCS system is actuated and injects water into the Reactor Coolant System, a Special Report shall be-prepared and submitted within 90 days describing the circumstances of the actuation and the total accumulated actuation cycles to date.

The current value of the usage factor for each affected safety injection nozzle shall be provided in this Special Mport whenever its value exceeds 0.70.

ti.1 DIfJniiliidividiialiEriiirginip!Diiisil?GsneFutVE{(EDG)liehiesisisisl[f6HETE

' ~ "mo'reWalidJfailures LinLthe1 ast 251deniands h'atDtini these1fallUreRandfanyin6H

Val id f f ai l ureif experi encediby EthatiEDGlin(th l rep 6rted!within330'daysb:ReportsfonEEDGifailuFes!shalliinblUds[thi

~

ziM.n fo rma ti on i re c omme nd e dli n; Reg ul,a to ryf o s i t i o nj c (5$fdsg u] at 9hRegision'3.;

If a preplanned alternate method of monitoring post accident parameters c.

must be initiated per Specification (3.3.3], a report for the inoperable system shall be submitted within 14 days following the event.

The report shall outline the action taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status.

)

(continued)

UNIT NAME 5-39 07/17/90

Administrative Controls 5.0 5.9 REPORTING REQUIREMENTS (continued) 5.9.2 Soecial ReDorts (continued)

'd.E (Thsl NRC?ihil l ? ie' i nformsd Ni th in F24T hiu rs "of{d iscoveryJ s f"AliesEtistilj.

anomaly 11nvolving aldisparitsof 2l1% Ak/klinicoreireactivityisinswhich

thefcause"cannotubeidetermined.L11n addition,fthaLNRCsshall(beninformed-k, ithi.n 24

hou'rs 'of 2 discovery:ofJat (Quadrant!PoWebTiltERatioz(QPTR)T~~

Tilt;(T )f210.10);;foWer;Tiltiniaxim%jjmitior1Azimuthaljfowsh 094cr?Quadra'nt q

5.10 RECORD RETENTION In addition to the applicable record retention requirements of Title 10, Code of Federal Regulations, the following records shall be retained for at least the minimum period indicated.

5.10.1 The following records shall be retained for at least 3 years:

a.

All License Event Reports required by 10 CFR 50.73; b.

Records of changes made to the procedures required by Specification 5.7.1; c.

Records of radioactive shipments; 5.10.2 The following records shall be retained for at least 5 years:

a.

Records and logs of unit operation covering time interval at each power level; b.

Records and logs of principal maintenance activities, inspections, repair, and replacement of principal items of equipment related to nuclear safety; c.

Records of surveillance activities, inspections, and calibrations required by the Technical Specifications [and the Fire Protection Program);

d.

Records of sealeo source and fission detector leak tests and results; and e.

Records of annual physical inventory of all sealed source material of record.

(continued)

UNIT NAME 5-40 07/17/90

Administrative Controls 5.0

~

5.10 RECORD RETENTION (continued) 5.10.3 The following records shall be retained for the duration of the unit Operating License:

c a.

Records and drawing changes reflecting unit design modifications made to systems and equipment described in the Final Safety Analysis Report; i

b.

Records of new and irradiated fuel inventory, fuel transfers, and assembly burnup histories; c.

Records of radiation exposure' for all individuals entering radiation control areas; d.

Records of gaseous and liquid radioactive material released to 1

the environs; i

e.

Records of transient or operational cycles for those unit components identified in [FSAR Section X];

f.

Records of reactor tests and experiments; g.

Records of training and qualification for current members of the unit staff; h.

Records of inservice inspections performed pursuant to the Technical Specifications; 1

1.

Records of quality assurance activities required by the Operational Quality Assurance Manual (not listed in Technical Specification 5.10.1 and which are classified as permanent records by applicable regulations, codes and standards);

j j.

Records of reviews performed for changes made to procedures or-equipment or reviews of tests-and experiments pursuant to 10 CFR 50.59; k.

Records of the reviews and audits specified in Technical Specifications 5.5.1 and 5.5.2; JI 1.

Records of the service lives of all hydraulic and mechanical snubbers required by (document where snubber requirements relocated to] including the date at which the service life 3

commences and associated installation and maintenance records; (m.

Records of secondary water sampling and water quality;]

(continued)

UNIT NAME 5-41 07/17/90

1 Administrative Controls i

o 5.0 5.10 RECORED RETENTION (continued) 4 5.10.3 (continued) n.

Records of analyses required by the Radiological Environmental Monitoring Program that would permit evaluation of the accuracy of the analysis at a later date. This should include procedures effective at specified times and QA records showing that these procedures were followed; o.

Records of reviews performed for changes made to the Offsite Dose Calculation Manual (ODCM) and the Process Control Program (PCP);

p.

Records of pre-stressed concrete containment tendon surveillances; and q.

Records of steam generator tube surveillances.

i i

5.11 HIGH RADIATION AREA 5.11.1 Pursuant to paragraph 20.203(c)(5) of 10 CFR 20, in lieu of the I

" control device" or " alarm signal" required by paragraph 20.203(c),

each high radiation area, as defined in 10 CFR 20, in which the-intensity of radiation is greater than 100 mrem /hr but less than 4

1000 mrem /hr at 45cm (18in.) from the radiation source or from eny surface which 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 j

-1 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 mrem /hr, provided they are otherwise following plant radiation protection procedures for entry into such high 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 radiation monitoring device which continuously indicates the a.

radiation dose rate in the area; or b.

A radiation monitoring device which continuously integrates the radiation dose rate in the area' and alarms when a preset integrated dose is received.

Entry into such areas with this i

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 (continued)

UNIT NAME 5-42 07/17/90 l

.o o,

I-Administrative Controls 5.0 5.11 HIGH RADIATION AREA (continued) 5.11.1 (continued) 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 i

i l

and shall perform periodic radiation surveillance at the l

frequency.specified by the (Radiation Protection Manager) in the RWP.

5.11.2 In addition to the' requirements of Technical Specification 5.11.1, areas accessible to personnel with radiation levels greater than 1000 mrem /hr at 45cm (181n.) from the radiation source or from any surface which 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-Foreman on duty o,i health physics supervision.

Doors shall remain locked except during periods of access by personnel under an approved RWP which 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 perforr-within the area.

For individual high radiation areas accessible to personnel with-1 radiation levels of within large areas, greater than 1000 mrem /hr that are located such as reactor primary 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.

L L

i b

UNIT NAME 5-43 07/17/90 L

l l

,-_ ~

-