ML17263A790

Proposed TS Section 6.0, Administrative Controls

ML17263A790
Person / Time
Site:	Ginna
Issue date:	09/27/1994
From:	ROCHESTER GAS & ELECTRIC CORP.
To:
Shared Package
ML17263A791	List: ML17263A790 ML17309A555
References
NUDOCS 9409290184
Download: ML17263A790 (101)

Text

ATTACHMfNT B Marked Up Copy of R.f. Ginna Nuclear Power Plant Technical Specification 6.0 Included pages:

ll 3.1-21 3.5-2a 3.6-3 3.16-2 6.2-1 6.2-2 6.2-3 6.2-4 6.4-1 6.5-1 6.5-2 6.5-3 6.5-4 6.5-4a 6.5-5 6.5-6 6.5-7 6.5-8 6.5-8a 6.5-9 6.5-10 6.5-11 6.5-12 6.6-1 6.7-1 6.8-1 6.8-2 6.9-4 6.9-6 6.9-7 6.10-1 6.10-2 6.10-3 6.11-1 6.13-1 6.13-2 6 '5-1 6.16-1 6.17-1 9409290l84 PDR ADOCK

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4.8 4.9 4.10 4.11 4.12

4. 13

4. 14

4. 15 4.16 TABLE OF CONTENTS (cont'd)

Auxiliary Feedwater Systems Reactivity Anomalies Environmental Radiation Survey Refueling Effluent Surveillance Radioactive Material Source Leakage Test Shock Suppressors (Snubbers)

Deleted Overpressure Protection System

~acae 4.8-1 4.9-1 4.10-1 4 '1-1 4 '2-1 4'3-1 4 '4-1 4 '5-1 4.16-1 5.0 DESIGN FEATURES 5.1 5.2 5.3 5.4 5.5 Site Containment Design Features Reactor Design Features Fuel Storage Waste Treatment Systems 5.1-1 5.2-1 5.3-1 5.4-1 5.5-1 6.0 ADMINISTRATIVE CONTROLS

( bc.he.4h) 6.1 6.2 6.3 6.4 6.5 6.6 6.7 6.8 6.9 6.10

6. 12

6. 13

6. 14

6. 15 6.16 6.17 Responsibility Organization 6.2.1 Onsite and Offsite Organization 6.2.2 Facility Staff Station Staff Qualification Trainin ew and Audit 6.5.1 t Operation Review Commi (PORC) 6.5.2 Nuclear S

udit and Review NSARB)

Quality Assurance Group Safety Limit Violation Procedures Reporting Requirements 6.9.1 Routine Report:s 6.9.2 Unique Reporting Requirements (Deleted)

High Radiation Area (Delet:ed)

Offsite Dose Calculation Manual Process Control Program Major Changes t.o Radioactive Waste Treatment Systems 6.1-1 6.2-1 6.2-1 6.2-2

6. 3-3.

6. 4-1

6. 5-1 6.5-1 6 5-5 6.5-11 6'-1 6.7-1 6.8-1 6.9-1 6.9-1 6.9-3 6.10-1 6 11-1 C.iX-.l 6 ~ 13-1 G. lI-1

6. 15-1 6.16-1

6. 17-1 Amendment No. gg, 49 ll

Maximum Coolant Activity Specifications Whenever the reactor is critical or the reactor coolant average temperature is greater than 500 F:

a.

The total specific activity of the reactor coolant shall not exceed 84/E pCi/gm, where E is the average beta and gamma energies per disintegration in Mev.

3.1.4.2 3.1.4.3 b.

The I-131 equivalent of the iodine activity in the reactor coolant shall not exceed 0.2 pCi/gm.

c.

The I-131 equivalent of the iodine activity on the secondary side of a steam generator shall not exceed 0.1 pCi/gm.

If the limit of 3.1.4.1.a is exceeded, then be subcritical with reactor coo'ant average temperature less than 500 F

within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

a.

If the I-131 equivalent activity in the reactor coolant exceeds the limit of 3.1.4.1.b but is less than the allowable limit shown on Figure 3.1.4-1, operation may continue for up to 168 hours0.00194 days <br />0.0467 hours <br />2.777778e-4 weeks <br />6.3924e-5 months <br />.

e I-131 equivalent activity in the reactor coolant exc the limit of

.4.1.b for more than 500 hours0.00579 days <br />0.139 hours <br />8.267196e-4 weeks <br />1.9025e-4 months <br /> in nsecutive 6-month

period, then pre and submit a report the Commis s pursuant to Specification 6.9.2.

Amendment No. 27 F 1-21

3.5.5.2 If the setpoint for a radioactive effluent monitor alarm and/or trip is found to be higher than required.

one of the following three measures shall be taken immediately:

(i) the setpoint shall be immediately corrected without declaring the channels inoperable; or immediately suspend the release of effluents monitored by the effected channel; or.

(iii) declare the channel inoperable.

3.5.5.3 If the number of channels which are operable is found to be less than required, take the action shown in Table 3.5-5.

Exert best efforts to return the instruments to OPERABLZ status within 31 days and, if

. 2..V>i unsuccessful, explain in the next Radio-3.5.6 3.5.6.1 active Effluent Release Report why the inoperability was not corrected in a timely manner.

Control Room HVAC Detection Systems During all modes of plant operation, detection systems for chlorine gas, ammonia gas and radioactivity in the control room HVAC intake shall be operable with setpoints to isolate air intake adjusted as follows:

3.5-2a Amendment No. 29

Basis:

The reactor coolant system conditions of cold shutdown assure that no steam will be formed and hence there would be no pressure buildup in. the containment if the reactor coolant system ruptures.

The shutdown margins are selected based on the type of activities that are being carried out.

The (2000 ppm) boron concentration provides shutdown margin which precludes criticality under any circumstances.

When the reactor head is not. to be removed, a cold shutdown margin of 14k/k precludes criticality in any occurrence.

Regarding internal pressure limitations, the containment design pressure of 60 psig would not be exceeded if the internal pressure before a major steam break accident were as much as 1 psig.'"

The containment is designed to withstand an internal vacuum of 2.5 psig.<'>

The 2.0 psig vacuum is specified as an operating limit to avoid any difficulties with motor cooling.

In order to minimize containment leakage during a design basis accident involving a

significant fission product

release, penetrations not required for accident mitigation are provided with isolation boundaries.

These isolation boundaries consist of either passive devices or active automatic valves and are listed in a procedure un er t e control o Closed manual valves, deactivated automatic va ves secured xn their closed position (including check valves with flow through the valve secured),

blind flanges and closed systems are considered passive devices.

Automatic isolation valves designed to close following an accident without operator action, are considered active devices.

Two isolation devices are provided for each mechanical penetration, such that no single credible failure or malfunction of an active component can cause a loss of isolation, or result in a leakage rate that exceeds limits assumed in the safety analyses<'>.

In the event that one isolation boundary is inoperable, the affected penetration must be isolated with at least one boundary that is not affected by a single active failure.

Isolation boundaries that meet this criterion are a closed and deactivated automatic containment isolation valve, a closed manual valve, or a blind flange.

The opening of closed containment isolation valves on an intermittent basis under administrative control includes the following considerations:

(1) stationing an individual qualified in accordance with station procedures, who is in constant communication with the control room, at the valve controls, (2) instructing this individual to close these valves in an accident situation, and (3) assuring that environmental conditions willnot preclude access to isolate the boundary and that this action will prevent the release of radioactivity outside the containment.

Amendment No.

4S 54

3. 6-3

6.9-2 when averaged over any calendar quarter, a

Special Report shall be submitted to the Commission within thirty days which includes an evaluation of any release conditions, environmental factors or other aspects which caused the reporting levels of Table 6.9-2 to be exceeded.

When mo. e than one of the radionuclides in Table 6.9-2 are detect d in the sampling medium, this eport shal'e submitted if:

concentration 1

+

concentration (2

+

....> 1.0 lima.t level (1) limit level (2)

When radionuclides othe than those in Table 6.9-2 are deflected and are the result of plant effluents, this report shall be submitted if the potential annual dose to an individual is greater than the calendar year limit of Specifications 3.9.1.2.a or 3.9.2.2.b.. This report is not required if the measured level of radioactivity was not the result of plant effluents; however, in such an event, the condition shall be reported and d scribed in the Annual Radiological Environmental Operating Report.

3.16.1.4 If milk or fresh leafy vegetable samples are unavailable for more than one sample period from one or more of the sampling locations indicated by the

ODCM, a dis-C. P.Vi i cussion shall be included in the Radioactive Re.4e.o.~

Effluent Report which ident'=ies th cause of the unavailab'lity of sa...~les and ident,"'fies locations for 3-16-2

6.2 ORGANIZP TION 6.2.1

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Onsite and Offsite Or anization An onsite and an offsite.organization shall be established t

for unit operation and corporate management.

The onsite and offsite organization 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 from the highest management levels through intermediate levels to and including all Plant management positions.

Those relationships shall be documented and

updated, as appropriate, in the form of organization charts.

These organization charts will be documented in the UFSAR and updated in accordance with~sW~ +~Ma~

b.

The Senior Vice President, shall have corporate responsibility for overall Plant nuclear

safety, and shall take any measures needed to, assure acceptable performance of the staff in operating, maintaining, and providing technical support in the Plant so that continued nuclear safety is assured.

c.

The Plant

Manager, Ginna Station shall have f

responsibility for overall unit operation and shall have control over those resources necessary for safe operation and maintenance of the Plant.

Amendment No. P, )f, 38 6.2-1

Insert 1

An alternate title may be designated for this position in accordance with 10 CFR 50.54(a)(3).

All requirements of these Technical Specifications apply to the position with the alternate title as apply with the specified title.

Alternate titles shall be specified in the Updated final Safety Analysis Report.

d.

The persons responsible for tne training, health physics and quality assurance funct'ns may repor" to an appropriate manager onsite, bu" shall have direct access to responsible corporate management at a level where action appropriate to the mitigation of training, health physics and quality assurance concerns can be accomplished.

6.2.2 Pacilit Staff in'.u4 The Facility organization shall the following:

2.

Each on duty shift shall be composed of at least th l

'nimum shift crew composition shown in Table 6 -l.

C.'Z. i'i b.

At le t one licensed Operator shall be in he control r when fuel is in the reac r.

c.

At least two 1'nsed Operators s ll be present in the control room d ing reac r start-up, scheduled reactor shutdown and du recover from reactor trips caused by tra ients emergencies.

d.

All core alter mons shall be dz ctly supervised by either a 1'nsed Senior Reactor Ope tor or Senior React Operator Limited to Fuel Handlin ho has no o

er concurrent responsibilities during this operation.

An individual qualified in radiation protection pro-cedures shall be on site when fuel is in the reactor.

6'-2 Amendment,'(p.

28

jI

Insert 2

a.

An auxiliary operator shall be assigned to the shift crew with fuel in the reactor.

An additional auxiliary operator shall be assigned to the shift crew above Cold Shutdown.

b.

At least one licensed operator shall be present in the control room when fuel is in the reactor.

In addition, above Cold Shutdown, at least one licensed Senior Reactor Operator (SRO) shall be present in the control room.

C.

Shift crew composition may be less than the minimum requirements of 10 CFR 50.54(m)(2)(i) and Specifications 6.2.2.a and 6.2.2.f for a period of time not to exceed 2

hours in order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to 'estore shift crew composition to within the minimum requirement.

Adequate shift coverage shall be maintained without routine heavy use of overtime. Administrative procedures shall be developed and implemented to limit the working hours of unit staff who perform safety-related functions l

including senior reactor operators, reactor operators,

+

health physicists, auxiliary operators, and key maintenance personnel.

Changes to the guidelines for the administrative procedures shall be submitted to the NRC for review.

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Mih Amendment No.

49 6.2-3

Insert 3

An alternate title may be designated for this position.

All requirements of these Technical Specifications apply to the position with the alternate title as apply with the specified title.

Alternate titles shall be specified in the Updated Final Safety Analysis Report.

Table 6.2-1 MINIMUMSHIFT CREN COMPOSITION IT ON NUMBER OF INDIVIDUALS REQUIRED TO FILL OSITION RCS Above Cold Shutdown Cold Shu own

& Refueling SS SRC RO AO STA 1

None 1

1 None SS - Shift Supervi r with a Senior Reac r Operators License SRO - Individual with Senior Reactor 9 erators License RO - Individual with a eactor Opera mrs License AO Auxiliary Operator STA - Sh'ft Technical Advis Except for the Shift Supervi or, e Shift Crew Composition may be one less than the minimum requ'ments f Table 6.2-1 for a period of time not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in der to ac mmodate unexpected absence of on duty shift crew members ovided imme 'ate action is taken to restore

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the Snift Crew Composit'on to within th minimum requirements of Table 6.2-1..his provisio does not permit a

shift crew position to be unmanned upon shift ange due to an onco

'ng shift crewman being late or absent.

During any absen of the Shift Supervisor fr the Control Room while the unit is abo Cold Shutdown, an individual other than the Shift Technical Advi r) with' valid SRO license shal be designated to assume the Co trol Room command function.

During ny absence of the Shi t Superv sor from the Control Room while the un't is in Cold Shu"down o

Refueling, an individual with a valid SR or RO license shall be esignated to assume the Control Room comman function.

6.2-4 Amendment t(o. gg, gg, 28

TRAINING A

ret raining and replacement

.training program for the facility staff shall be maintained under the direction of the Division Training Manager and shall meet or exceed the req irements and recommendations of Section

5. 5 of ANSI Y'.1-1971 and Appendix A of 10 CFR Part 55.

The training program shall meet or exceed NFPA No. 27, 1975 Se "ion 40, except that {1) training for salvage operations neec not be provided and (2) the Fire Brigade training ses

'ons shall be held at least quarterly.

Drills are considered to be training sessions.

6.4-1

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.5 REVIEW AND AUDIT 6.5.1 6.5.1.1 fety-related matters.

One of these will be n on-site ope tions review

group, the Plant Oper ions Review Committ (PORC).

A second is the Qualit Assurance (Q.A.)

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group, re onsible for the audi of safety-related activities as ciated with plant operations.

A third is the independent a it and rev ew group, the Nuclear Sa ety Audit and Review Boa (N

).

This group is responsible for the periodic rev e

of the activities of the Plant Operations Review Committe for directing audits and evaluating the 'esults, and f the management evaluation t

adequacy of the Quality Assurance of the s

tus and progr P

OPERATIONS REVIEW COMMITTEE ( PORC CTICH he Plant Operations Review Committee shal function to advise the Plant

Manager, Ginna Station on a

matters related to nuclear safety and for referral of appr riate matters to the Nuclear Safety Audit and Review Board.

Three separate organizational units shall be establis d

for the purpose of review and audit of plant operati s and 5

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Corrected Amendment Ho. 15,+

6. 5-1

COMPOSITION

5. 1.

The PORC shall be composed of the; Chairman:

Superintendent, Ginna Production Vice Chairman/Member:

Superintendent, Ginna Su port Services Vice Chairman/Member:

Operations Manager M

er: Maintenance Manager Me r: Maintenance Planning/Schedulin Manager Membe

. Instrument

& Control Superv'r Member:

echnical Manager Member:

Re ults and Test Superv'sor Member:

Reac r Engineer Member: Health hysics and hemistry Manager Member:

Nuclear A suranc Manager Member: Quality Con Engineer ALTERNATES 6.5.1.3 MEETING Alternate memb s shall b

designated by name, in writing, by the Chair n.

UENCY 6.5.1.4 The PORC shall meet at least once per calendar month and as conve d by the PORC Chairman.

QUORUM 6.5.1.5 A quorum of the PORC shall consist of t e Chairman or Vice Chairman and four members including alte nates.

No more-than two shall be alternates.

Amendment No. X5, 32 6.5-2

SPONSIBILITIES 6.

.1.6 The PORC shall be responsible for:

a.

Review of

1) all procedures required by Specif'tion 6.8 and changes thereto,

2) any other proposed p ocedures or changes thereto as determined by the Pl t Manager, Ginna Station to affect nuclear safety.

b.

eview of all proposed tests and experi nts that affect nu ear safety.

c.

Revie of all proposed change to the Technical Specifi tions.

d.

Review of ll proposed changes or modifications to plant systems or e ipment that fect nuclear saf ety.

e.

Investigation of all iolations of the Technical Specifications a

sh 1 prepare and forward a report covering evaluatio and recommendations to prevent recurrence to the Sen r Vice President, Production and Engineering, a

to the Chairman of the Nuclear Safety Audit and Re ew Board.

f.

Review of acility operations to detect potential safety r

hazards.

g.

Perfo mance of special reviews a d investigations and re rts thereon as re~ested hy t e Chairman of the clear Safety Audit and Review Board.

Amendment No. gg, 3S 6.5"3

h.

Review of the Plant Security Plan and shall subm'ecommended changes to the Chairman of.the Nuc ar Safety Audit and Review Board.

Review of the Radiation Emergency Plan an shall submit recommended changes to the Chairman of e Nuclear fety Audit and Review Board.

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Rev'ew of implementing procedures or the Plant Secure y Plan and the Radiation ergency Pian and proposed hanges thereto.

k.

Review of a l Reportable Ev ts.

Review of the ire Prote ion Program and Implementing Procedures and s

mitt of recommended Program changes to the Chairman of e Nuclear Safety Audit and Review Board (NSARB).

C. 3.iv AUTHORITY 6.5.1.7 The PORC shall:

b.

Recomme in writing to the Pla

Nanager, Ginna Stati approval or disapproval of 'tems considered und r 6.5.1.6(a) through (d) and (l) bove.

nder determinations in writing with r ard to whether or not each item considered under 6.5.1.6(

through (d) and (1) above constitutes an unreviewed s fety question as defined in 10 CFR Section 50.59.

Amendment No.

49 6.5-4

Provide immediate written notification to the Senior Vic President, Production and Engineering, and the Nucl ar Safety Audit and Review Board of disagreement b tween the PORC and the Plant Manager, Ginna Station;

owever, t e Plant

Manager, Ginna Station s all have re ponsibility for resolution of such isagreements pur uant to 6.1.1 above.

RECORDS 6.5.1.8 The PO C shall maintain written inutes of each meeting and co 'es shall be prov ed to the Senior Vice Presiden

, Production and ngineering, the Chairman of the Nucle r Safety Aud t and Review

Board, and such others as t e Chairma may designate.

Amendment No.

32, 38 6.5-4a

..C... "!L:CLFAP. res F:Tq AU",rrT AND }r.. VrLAr EG;:;i'rgVEAEE}

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6.5..1 The YSARB sha.'1 'unction to provide independent revic v and audit o dcsignat'.d activities in the. areas oi:

a

~

uclear pov:er plant opera:ions b.

nu lear en Tineering c.

chemi.~ and rad ochemistry d.

meta llur.

e.

instrumenta ion and control radicle":ca 1

.- fety g.

mechanica! and e ectrical engin ering h.

quality a s s..rance

p. "ctices CC'.ii!'i "-I'rIQ.J 6.5.2.?

The compositio of -he NSARB sha be established as folio'vs.

a.

Chairman an Vice Chairman appoint d by name by thc Chairman oi the l3o d and Chief E::.ecutive C'ffice of the Corpo.ation or officer f his desirnation.

At 1

st four tech"..'cally'qualified persons;s'r'ot member" c

the plant stafi to provide e.:pertisc in the fun tio'i:ilarea."

described in 6.5.2.1.

6. 5-5

C MPOSITION (Continued

)

c.

At least one quali ied non-company affiliated technical consultant and others as required.

Duly J

appointed consultant members shall have e

al vote with company affiliated members of the oard.

d.

Three members from the staff of the

. E. Ginna uclear Po~er Plant.

C.. 3.iV e.

Me ers in (b) and (d) above to e designated by the airman of the Board and hief Executive Office ALTERNATES 6.5.2.3 Alternate members hall e appointed in writing by the NSARB Chairman to serve on a te rary basis;

however, no more than two alternates shall part'ci ate in NSARB activities at any c ne time.

QUALIFICATIONS 6.5.2.4 The minim cualifications o

t e Nuclear Safety Audit and Review Board w'th regard to the individu members shall be maintained at a evel ~qual to or higher than e following:

a.

Reactor Eng 'eering Engineer

.".g g "aduate or equivalent w th over eight years ex.'er 'nce in the nuclear power eld and over four: e mrs respons ible engineering nagement.

6.5-6 Amendment No.

ALIFICATIONS (Continued (b)

UtilityOperations gineering graduate or equivalent with over eight years exper ence in ilityoperations and with over four years responsible gineering manage ent.

(c)

Reactor sics Physics gra te or equivalent with over five years xperience in reactor physic work.

(d)

Heat and Fluid Fl Engineering or Physi graduate or equivalen with four years experience in heat.

and fluid flow alysis.

C. 3'"

(e)

Environmental Analysis Engineering graduate or equi lent wi h over five years experience in environmental hazard analysis.

(f)

Reactor Control and Instrumenta o

Engineering graduate or equiv ent wi nuclear engineering.

over five years experience in (g)

Power Plant Operations Engineering graduate equivalent with over ve years experience in power plant operati ns.

(h)

Safety Analysis Engineering gr duate or equivalent with over five ye experience in nuclear eng eering.

(i)

Chemistry and Radiochemistry Engine ing graduate or equivalent with over five years experi ce in nuc ar engineering diological Safety Engineer graduate or equivalent with over five years experience in heal physics and/or radiological safety.

6. S-7

ETING FRE UENCY 6.

2.5 At least semi-annually and as required on call of the I

Chairman.

QUORUM 6.5. 2.6 A quorum shall consist of a majority of t 1

a will include the Chairman or Vice C

principals irman. At least one f the quorum shall be a non-comp ny affiliated techni 1 consultant and no more t n a minority of the quorum s ll be members of the P ant staff.

REVIEW 6.5.2.7 The NSARB sha review:

a.

The safety e aluation for 1) changes to procedures, I

equipment or s tern as described in the safety analysis report d 2) tests or experiments completed

'nder the prov'sion f 10CFR Section 50.59 to I

verify that uch aetio s did not constitute an unrevi ewe sa fe ty ques ti n.

b.

Propose changes to proced es, equipment or syst is which have been dete ined by the PORC to in lve an unreviewed safety qu stion as defined

'n 10CFR Section 50.59.

c Proposed tests or experiments which ave been determined by the PORC to involve an u eviewed safety cuestion as defined in 10CFR Secti n 50.59.

d.

Proposed changes in Technical Specification or licenses.

6.5-8 Amendment No.

2

e.

Violations of appl icable statutes,

codes, regulati l

orders, Technical Speci fications, 1 icense requi

ments, I

or of internal procedures or instructions hav'ng n clear sa fe ty s ignificance.

Sig ificant operating abnormalities or viati'ons from ormal and expected performance f plant equipme t that affect nuclear safe g.

All Repor ble Events.

6.5-8a Amendment No.

21

Continue h.

Any indication of an unanticipated deficiency in aspect of deslgTL or operation of safety r

structurem,

systems, or componanti.

eport3 and meeting minutes of the Plant Operations Re iew Committee.

AUDITS 6,5.2.8 The NS shall direct the establi nt of an audit program d evaluate audits perf ed to ensure oafI facility op ration.

Aud1ta sha encompass:

a.

The con8orman of facility ration to a11 provisions contained with n the T

hnical Speci ficati.oaa an4 applicable license coa iona at least once per Year.

b.

The perfoxmanca, ning and quality icat fons of the operating and tec c

Staff at least once a year.

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d.

The resultl of a aetio taken to correct deficieaciaa occurriag in f ci1Lty ~i at, structuroa, 3yitema or method of o ration that aff t nuclear safety at leal once per ix maths The pe oraumce of aL1 activities ired hy the Qualitjr A,ss ance Program for R. S. Qiana Nu car &ever Plant t It thta criteria of Appoadhc 8, 10 CFR 0, ac toast one per 24 monthl.

The Radiation Emergency Plan and haplementL proceduroa at least at the frequency repaired by 10 CFR 0.54(t) ~

The Station Security Elan and imylementiee gcoc uxia at least at the frepuaacy required hy 10 CFP.

73.(0t 6.5 9

Amendment No.

46

(S

UDITS (Continued) g.

The Facility Fire Protection Program and implement' procedures at least once per two years.

h.

An independent fire protection and loss p

vention program inspection and audit performed at le t once per 12 months utilizing either qualified of ite licensee ersonnel or an outside fire protectio firm.

An inspection and audit of the fire rotection and loss prev ntion program performed by on-licensee personnel t

at lea t once per 36 months The personnel may be I

represent tives of ANI, an 'urance brokerage firm, or C. 3.'iV other quali ied individu s.

j.

The radiologi al envi onmental monitoring program and the results the of at least once per 12 months.

k.

The Offsite Dos Calculation Manual and implementing procedures at east nce per 24 months.

1.

The Process ontrol Pr ram and implementing procedures at least nce per 24 mon s.

m.

Any her area of fac ity operation considered app opriate by the NSARB or e Senior Vice President, oduction and Engineering.

Amendment No. jig, 38 6.5-10

AUTHORITY 6.5.2.9 a.

The Chairman of the Nuclear Safety Audit and Review B ard is responsible to the President on all activit' for which the Review Board is responsible.

RECORDS b.

The NSARB shall rePort to and advise the enior Vice President, Production and Engineering, on ose areas of responsibility specified in Sections 6.5 2.7 and 6.5.2.8.

6 '.2.10 distribu ed as indicated below:

a.

Minute shall be recorded of 1 meetings of this Board..

Copies o

the minutes shal be forwarded within 14 days following e ch meeting t the Corporate Chairman of the Boarc, Senior Vice Pre ident, Production and Engineering and such othe the Chairman of the NSARB may designate.

b.

Reports of rev'ews compassed by Section 6.5.2.7 e,f,g and h above shall be prepared, approved and forwarded to the Se >or Vice Presi ent, Production and Engineering within 4 days following mpletion of the review.

c.

Audi reports encompassed by ection 6.5.2.8 above, shall b

forwarded to the Senior Vice resident, Production and ngineering and to the managemen positions responsible for the areas audited within 30 da after completion of the audit.

endment No./P,

)$,38 6.5-11

PR EDURES 6.5.

.11 written administrative procedures for committee oper ion',

shall be prepared and maintained describing the me od of submission and the content of presentations t the committee, provisions for use of subcommittee, review l4 nd approval by members of written committe evaluations an recommendations, distribution of minu es, and such I

oth matters as may be appropriate.

6.5.3 6.5.3.1 UALI Y ASSURANCE GROUP The org nization, qualifications, esponsibilities and training f quality assurance pe sonnel for audits of safety rel ed activities are escribed in the Quality Assurance Pr ram.

6.5<<12

6.6.1

~

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REPORTABLE EV19PZ ACTlON e following actions shall be taken for Reportable nts:

a.

The mmission shall be notified and a re t submitted tI t'ursuant to e requirements of Se on 50.73 to 10 CFR Part 50, and the results this review NSARB nd the Senior Vice ngineering.

b.

Each Reportable Eve shal shall n

reviewed by the PORC and submitted to the President, Pr ction and

4. 4

( Mhc~Q)

SkewJt', )

Amendment No.

~

6.7 SAFETY LIMIT VIOLATION The following actions shall be taken in the event a Safety

(.~.iii Limit is violated:

e')

a.

The provisions of 10 cFR ~seer~ 60.36(c) (1) (i) shall he complied with immediately.

b.

The Safety Limit viol e

eported to the MAevwvr Senior Vice President,

, to'the l

3. i )((

C., Z.iV Cgf;m t44Jt~

W~63C C ~ ~ and to the

'mmediately.

HRC A Safety Limit Violation Report sh red.

The o~6& ~ieau N~c9n.o~

report shall be reviewed by the This report shall d.

describe (1) applicable circumstances preceding the violation, (2) effects of the violation upon facility components, systems or structures, and (3) corrective act' taken to prevent recurrence.

The Safety Limit Uio hall be submitted to QP(')sCiaK ~vi~

evviva'v1ovi the

, the

, and the Senior Vice President, violation.

within two weeks of the

~ gdrth~ D ++CIAO~

M~~ Qi Amendment No-ZZ, 38 6.7-1

I

6.8 PROCEDURES 6.8.

C. l. iV Written procedures shall be established,"

implemented, and No~i maintained covering the actxvities

• h 1'

l d

d

~ Reguletozy Guide 1. 33,

~we on Z,, g pp~g~X 1

%derv~ i~18.

C.2..iV

c. z..v 6.

2 6.8e3 Fire Protection Program implementation.

The radiological environmental monitoring program.

Offsite Dose Calculation Manual implementation.

Process Control Program implementation.

Each procedure and administrative policy of 6.8. 1 ve, the changes

thereto, shall be

. reviewed by th ORC anc appr ed by the Plant

Manager, Ginna S

ion prior tc

'mplemen".a 'n and reviewed periodic y as set forth in the applicable p

cedures.

Temporary changes to ce res of 6.8.1 above may be made provided:

a.

The intent be The c

ge the original pro dures is not altered.

is approved by two me rs of the plant" agement

staff, at least one of whom the Shift Supervisor who holds a

Senior Reactor Ope tor'icense.

Amendment tio. ~

~

~

6.8-1

c ~

T e documented, reviewed by th ',

and N'c-'I:"'pproved by the Plant g

tion;within 10 p ementation.

and directions from the reactor, and the results of the participation in an interlaboratory comparison program.

ti Radioactive Effluent Release Reoort Routine radioactive effluent release reports covering the operation of the unit during the previous~

months of operation shall be submitted A

of each year.

This report shall include' summa~,

on a quarterly basis, of the quantities of radioactive liquid and gaseous effluents and solid waste released as outlined in Regulatory Guide 1.21, Revision 1.

The

'oactive effluent release re ays of 1 shall include an assessment.

of radiation doses from the radioactive liquid and gaseous effluents released from the unit during each of the previous four calendar quarters as outlined in Regulatory Guide 1.21, Revision 1.

In addition, the site boundary maximum noble gas gamma ai" and beta air doses shall be evaluated.

The assessment of radiation doses shall be performed in accordance with the ODCM.

This same report shall include an annual summary of hourly meteorological data collect d over the previous calendar year.

Alternatively, the licensee has the option of retaining this summary on site in a file that shall be provided to the NRC upon request.

h y

ll '-*l.

location(s) identified by the land use census which 6.8-4

6.9.2 6.9.2.1 6.9.2.2 Uni e Reoortin Re uirements 1

Annually:

Results of required leak test performed on sources if the tests reveal the presence of 0.005 microcurie or more of removable contamination.

Annually:

A tabulation on an annual basis of the I

number of station, utility and other personnel (including contractors) receiving exposures greater than 100 6.9.2.3 mrem/yr and their associated man-rem exposure according to work and job functions, e.g.,

reactor operations and surveillance, in-service inspection, routine maintenance, special maintenance (describe maintenance),

waste processing, and refueling.

The dose assignment to various duty functions may be estimates based on pocket 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 shall be assigned to specific major work functions.

(NOTE:

This tabulation supplements the requirements of Section 20.407 of 10CFR Part 20.)

Annually:

The results of specific activity ana s in I'hich the primary coolant exceeded the 1's of Specifx

'on 3.1.4.1.a and b.

following information C.2.

shall be included:

ctor 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 e first s e in which the limit was excee; (2) Results of the last x ic analysis fo adioiodine performed prior to exceeding the 6.9-6 Amendnent Ho.

27

re ed to e date and re its of analyses while the limit was exceeded and result lysis after the radioiodine activity was less than the imit.

Each result should includ time of sampling d the radioiodine centrations; (3) 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 limi as exceeded; (4) Graph of the I

I-131 concentration one ot radioiodine isotope Clean-up system flow hi ry start'oncentration as function of time for e duration of the specific ac vity above the steady-state leve and (5)

The time ration when the specific activity of the rimary oolant exceeded the radioiodine limit.

6.9.2.4 Reactor Overpressure Protection System Operation The report shall describe the circumstances initiating the transient, the effect of the PORVs or vent.(s) on the transient In the event either the PORVs or the RCS vent(s) are used to mitigate a

RCS pressure transient, a Special Report shall be prepared and submitted to the Commission within thirty days.

and any other corrective action necessary to

prevent, 6.9.2.

recurrence.

0t l7 CLCCo ('CLRHC.C, ecial reports shall be submitted upi.

OFF.'0. 0 C'. 2.. ic with' time period specified for each report.

Amendment No.

27 F 9-7 C

I

c 6 10 6.10.1 RECORD RETENTION In accordance with Rochester Gas a

pow level.

c nd Electric Corpoi'~a son

'.". '.h,Q;>~c'gC"p policy, operating charts for the first year '

. oper ron

~ *>> c will be permanently stored.

The following records shall be retained for

-leas ve years:

h ~

't'. -g~ -,

4 a.

Records and logs of facility opera on, including p wer levels and periods of ope tion at each

+h ~hVii b.

Recor and logs of princi al maintenance activities, 6.1

.2 c.

Reportable Eve Reports.

Specif'cations.

e.

Rec ds of reactor test or experiments.

f.

ecords of changes made i the Operating Procedures.

results. "'"

chh all sealed the, duration Records of sealed source le k tests and Records of annual physical in entory of h.

source material of record.

The following records shall be retaine for of the Facility Operating License:

d drawing changes reflecting facility ifications made to systems and equipment in the Final Safety Analysis Rep I

h Amendment'o.4 a.

Records an design mod described 6 10-1 includin inspection, r air, substitution or 1'

~

replacemen of princ'l items of equipment pertaining to nuclear s fety vc'PcI d.

Records of rve lance activities, inspections, and calib ations r ired by these Technical hh'

)

'$ (hfc ii'Cl'NC h'.

s-built

/new and into the a

cords of irradiated fuel invent y~

fue14'ra sfers<

and assembly burnup historic c.

Reco s of plant radiation and contam nation survey d.

Records f off-site environmental monitoring e.

Records of r diation exposure of all plant

'ersonnel>

inc uding all c tractors and visitors to the plant who enter r iation control areas.

-p '1 f.

Records of radioa ivi in liquid and gaseous material released t he environmental and e.

b surveys.

c.G.vis 4'adioactive waste i

ents.

4~ 4>>>>4~>>>>>>1 ~Cai~14aaaaXar~~~>>

g~

~ 414~g~~

"L114,g gy1~14 "4 P ~

- ~~4.

11~~i. <g~~g~1. y 4 I'~4'4-"~9~4,@~i

g P

t changes shall also be periodically incorpora44te g.

Records of trans'ent or operational cycles for those facility components designed for limited number of tr nsients or cyc es.

h.

Records of training and qual'cation for current I4 4 P

4,4 j.

R cords of reviews performed f station echnical and opera tio sta ff members.

i.

Recor of in-service inspection peformed pursuant to ese Technical Specifications.

14 411 or cha ges made to 1"

g '4=

iews of ests and

'ection 5 -59 C and the NS RB I

1 ctivities'as r quired 6.10 1 Amen'dment No.

pp I

Records of the service lives of all hydraul'i'echanical snubbers listed in the Inservice ace n pectxon A'4 Program including the date at which the se ice life ommences and associated installation and ma nte'nance cords'e Q,y))

Am dment f(o. 37

~A"

'i>) ~,

ll

.6.10"3 k

3 RADIATION PROTECTION PROGRAM 3

control procedures shall be prepaxed aad made ava le to adiation exposure show perm ble radiation exposuii, of 10 CPR Part. 20.

The radiatioa ad maintaiae meet the requirements ceptioas set forth in Section 6.13 hese The'rogram shall be adhered to for all oper ons Radiac at the station.

These proceduxes and shall be consistent with the requir protection program shall be or xed of 10 CPR Part 20, wit Technical Spe cacions.

all station personnel or her persons who may be subject 3t tai

"'mro g personnel radiation exposuxe.

3 3>>l l>>

3>>

," 31 3>>

~/

"3

~Ca Bhaa.~u.)

a>>

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3 3>>3 l 6'11-1 3

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't -': "

i:. a f-,-<~""

3 3

3,

>>a AN/

)a a,

.~,'

3 3'I"

6 13 KEGH MDZATZOH AREA.

6.13.1 In lieu of the "contxol device" or "alamo signal" 20.203(c) (2) of 10 CFR Part 20:

quiring issuance of a individual or group of individuals permitted to

~8 1

required hy paragraph a.

Each High Radiation Area in which the intensity of radiation is 1000 mrem/hr or less shall be barri'-..:,"-":,<~'~-;".'aded and conspicuously posted as a high radiation.":;,"..

area and entrance thereto shall be controlled by; re-R~m~o m gg ~p3 Work Permi.*

Any enter such areas shall be provided with one or more of the following:

A radiation monitoring device which con-tinuously indicates the radiation dose rate in' I",M ~

integrates the radiation dose rate in the area and alarms when a preset integrated dose is received.

Entry into such areas with this may be made after the dose rate monitoring device levels in the area h

nel have been made have been established and person knowledgeable of them.

(3)

A qualified health physicist~(i.e., qualified the area.

(2)

A radiation monitoring device which continuously in radiation pxotection procedures) with requirement protection protection 1 be exempt from rmance of their a

they are followi ry into high rad personnel shal during the perfo duties, providing procedures for ent C.m the issuance ssigned radiation ng plant radiation iation areas-.

6 '3-1 I4

Q within locked doors shall be maintained under the administrative control of the Shift Supervisor on duty.

W"Wi"Ciaa4'k M~akl ~hQ~ SI~lCCkl55k<~~

a-radiation dose rate monitoring'evice who is

'I responsible for providing positive.control over the activities the area and.'ho will

.f

'erformperiodic radiation surveillance at "*,the frequency specifi.ed in the HPWP.

The survei.llance frequency will be established by a=- plant chealth physicist.

Each High Radiation Area in which the intensity of radiation is greater than 1000 mrem/hr shall.he sub)ect

='o the provisions of 6.13.1 a.

above, and in addition locked doors shall be provided to prevent unauthorized entry into these areas and the keys to unlock these 5

Amendment No; 32 6.13-2

6.15 II L!t-tt I

~tlt~atlIIIIISIttBIItttlttt tt.

I I

I p

'I

~

'ffsite Dose Calculation Manual (ODCM)

ODCM shall be made by the following I

changes shall be submitted to the Radioactive Eff3;uent the period in which the change(s) to support" the 6.15.1 Any changes to I

the method:

Eicense e initiated 6.15.1. a Commission with the

.Yi 4 for all Release Report.

was made and sh contain:

fficient (i) su detailed information ly or setpoint determinations; and rationale for the change.

(ii) a determination that the change will not reduce the accuracy or reliability of dose calculations C. 3. iV

~

~

t

6. 15.1.b (iii). documentation of the fact that the change has been reviewed and found acceptable by the(gQ.

licensee initiated changes shall.become effective after C.. Z. iv review and acceptanc'e by the by the licensee.

on a date specified

+nCi"d4 ~Vw~ RncQon Z -t I

I

'6.15-1.

0 6.16 Process Control Program (PCP)

/

Any changes to the PCP shalL be MA' the following mitted to.'the Effluent'~"..

change(s)'o support the 6.16.1 made b I~

method:

6.16.1.a Licensee initiated changes shall be sub mmission with the ease Report for the adioactive which the R

in Co Rel C.~.Via period.

was made and shall contain:

sufficiently detailed information rationale for the change; (ii) a determ'nation that the change will not reduce the overall conformance of the solidified waste product to existing criteria for solid wastes; and

'.'E.H 6.16.1.b Q,3.aV

~~g>~ ~<~AA&bA.

(iii) documentation of the fact that the change has been rev'ewed and found acceptable by the SI F

Licensee initiated changes shall become effective after t

t reviekc and acceptance by the($ GQ n a date specified t

by the licensee.

k

~

~

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C

, tki I I I

"6.16-1 E

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0 6.17 Ma'or Chances to Radioact've Waste Treatment S stems FUNCTION (Liquid, Gas ous and Solid) 6.17.1

. 6.17.2 The radioact'v waste treatment systems (liquid, gaseous and solid) are Mose systems defined in Technical Specificat'on 5.5.

Major changes to the radioactive w'aste systems (licpud and gaseous) shall be reported by the following method.

For the pu=ose o= this specification, "major changes" is def'" " '." S"ec'=ication 6.17.3 be ow.

6.17.2.1

~ V'LL The Commission sh ll be informed of all major changes by the '".con of a suitable discussion or by reference to a suitable discussion of each change in the Radioact.'ve =f luent Release Report for the period in which the c.".rages were made.

The discussion of each change shall contain-.

a) a sum~a=~

o" Ae evaluation that led to the determination that "".

change could be made (in accordance with 10 C:-R 50.54).

b) suf"icient detailed information to support the reason cr Me change; c) a deta 1 d d scription of the equipment, components and proc sses involved and the 'nterfaces with othe plan" systems; 6.17-1 a

j

Proposed ATTACHMENT C Revised R.E.

Ginna Nuclear Power Plant Technical Specification 6.0 Revise the pages as follows:

Remove ll 3.1-21 3.5-2a 3.6-3 3.16-2 6.2-1 6.2-2 6.2-3 6.2-4 6.4-1 6.5-1 6.5-2 6.5-3 6.5-4 6.5-4a 6.5-5 6.5-6 6.5-7 6.5-8 6.5-8a 6.5-9 6.5-10 6.5-11 6.5-12 6.6-1 6.7-1 6.8-1 6.8-2 6.9-4 6.9-6 6.9-7 6,10-1 6.10-2 6.10-3 6.11-1 6.13-1 6.13-2 6.15-1 6.16-1 6.17-1 Inser t ll 3,1-21 3,5-2a 3.6-3 3.16-2 6.2-1 6.2-2 6.2-3 6,4-1 6.5-1 6.6-1 6.7-1 6.8-1 6.9-4 6.9-6 6.9-7 6.10-1 6.11-1 6.13-1 6.13-2 6,15-1 6.16-1 6.17-1

r 1

4.8 4.9 4.10 4.11 4.12 4.13 4.14 4.15 4.16 TABLE OF CONTENTS (cont'd)

Auxiliary Feedwater Systems Reactivity Anomalies Environmental Radiation Survey Refueling Effluent Surveillance Radioactive Material Source Leakage Test Shock Suppressors (Snubbers)

Deleted Overpressure Protection System Pacae 4.8-1 4.9-1 4.10-1 4.11-1 4.12-1 4.13-1 4.14-1 4.15-1 4.16-1 5.0 DESIGN FEATURES 5.1 5.2 5.3 5.4 5.5 Site Containment Design Features Reactor Design Features Fuel Storage Waste Treatment Systems

5. 1-1 5.2-1 5.3-1 5.4-1 5.5-1 6.0 ADMINISTRATIVECONTROLS 6.1 6.2 6.3 6.4 6.5 6.6 6.7 6.8 6.9

6. 10

6. 11

6. 12

6. 13

6. 14 6.15 6.16

6. 17 Responsibility Organization 6.2.1 Onsite and Offsite Organization 6.2.2 Facility Staff Station Staff Qualifications Training (Deleted)

(Deleted)

Safety Limit Violation Procedures Reporting Requirements 6.9.1 Routine Reports 6.9.2 Unique Reporting Requirements (Deleted)

(Deleted)

(Deleted)

High Radiation Area (Deleted)

Offsite Dose Calculation Manual Process Control Program Major Changes to Radioactive Waste Treatment Systems

6. 1-1 6.2-1 6.2-1 6.2-2 6.3-1 6.4-1 6.5-1 6.6-1 6.7-1 6.8-1 6.9-1 6.9-1 6.9-3 6.10-1 6.11-1 6.12-1 6.13-1 6.14-1 6.15-1 6.16-1 6.17-1 Amendment No.

gp, 3.i Proposed

3.1.4

~

~

Maximum Coolant Activit S ecifications

3. 1.4. 1 a 4 C ~

3. 1.4. 2 Whenever the reactor is critical or the reactor coolant average temperature is greater than 500 F:

The total specific activity of the reactor coolant shall not exceed 84/E pCi/gm, where E is the average beta and gamma energies per disintegration in Mev.

The I-131 equivalent of the iodine activity in the reactor coolant shall not exceed 0.2 pCi/gm.

The I-131 equivalent of the iodine activity on the secondary side of a steam generator shall not exceed 0.1 pCi/gm.

If the limit of 3.1.4.1.a is

exceeded, then be subcritical with reactor coolant average temperature less than 500'F within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

3.1.4.3 a

~

If the I-131 equivalent activity in the reactor coolant exceeds the limit of 3.1.4.1.b but is less than the allowable limit shown on Figure 3.1.4-1, operation may continue for up to 168 hours0.00194 days <br />0.0467 hours <br />2.777778e-4 weeks <br />6.3924e-5 months <br />.

F 1 21 Proposed

I 4 L'E

3.5.5.2 If the setpoint for a radioactive effluent monitor alarm

~

~

~

and/or trip is found to be higher than required, one of the following three measures shall be taken immediately:

(i) the setpoint shall be immediately corrected without declaring the channels inoperable; or immediately suspend the release of effluents 3.5.5.3 3.5.6 3.5.6.1 monitored by the effected channel; or (iii) declare the channel inoperable.

If the number of channels which are operable is found to be less than required, take the action shown in Table 3.5-5.

Exert best efforts to return the instruments to OPERABLE status within 31 days and, if unsuccessful, explain in the next Radioactive Effluent Release Report why the inoperability was not corrected in a timely manner.

Control Room HVAC Detection Systems During all modes of plant operation, detection systems for chlorine gas, ammonia gas and radioactivity in the control room HVAC intake shall be operable with setpoints to isolate air intake adjusted as follows:

Amendment No.

gp 3.5-2a Proposed

4 r

~ 4

(

~ I'

Basis:

The reactor coolant system conditions of cold shutdown assure that no steam will be formed and hence there would be no pressure buildup in the containment if the reactor coolant system ruptures.

The shutdown margins are selected based on the type of activities that-are being carried out.

The (2000 ppm) boron concentration provides shutdown margin which precludes criticality under any circumstances.

When the reactor head is not to be removed, a cold shutdown margin of ll~k/k precludes criticality in any occurrence.

Regarding internal pressure limitations, the containment design pressure of 60 psig would not be exceeded if the internal pressure before a major steam break accident were as much as 1 psig.<"

The containment is designed to withstand an internal vacuum of 2.5 psig.">

The 2.0 psig vacuum is specified as an operating limit to avoid any difficulties with motor cooling.

In order to minimize containment leakage during a

design basis accident involving a

significant fission product

release, penetrations not required for accident mitigation are provided with isolation boundaries.

These isolation boundaries consist of either passive devices or active automatic valves and are listed in a

procedure under the control of the Quality Assurance Program.

Closed manual valves, deactivated automatic valves secured in their closed position (including check valves with flow through the valve

,secured),

blind flanges and closed systems are considered passive devices.

Automatic isolation valves designed to close following an accident without operator action, are considered active devices.

Two isolation devices are provided for each mechanical penetration, such that no single credible failure or malfunction of an active component can cause a loss of isolation, or result in a

leakage rate that exceeds limits assumed in the safety analyses@.

In the event that one isolation boundary is inoperable, the affected penetration must be isolated with at least one boundary that is not affected by a single active failure.

Isolation boundaries that meet this criterion are a closed and deactivated automatic containment isolation valve, a closed manual valve, or a blind flange.

The opening of closed containment isolation valves on an intermittent basis under administrative control includes the following considerations:

(1) stationing an individual qualified in accordance with station procedures, who is in constant communication with the control room, at the valve controls, (2) instructing this individual to close these valves in an accident situation, and (3) assuring that environmental conditions will not preclude access to isolate the boundary and that this action will prevent the release of radioactivity outside the containment.

Amendment No.

3.6-3 Proposed

A k

6.9-2 when averaged over any calendar quarter, a Special Report shall be submitted to the Commission within thirty days which includes an evaluation of any release conditions, environmental factors or other aspects which caused the reporting levels of Table 6.9-2 to be exceeded.

When more than one of the radionuclides in Table 6.9-2 are detected in the sampling medium, this report shall be submitted if:

concentration 1

+

concentration 2

+

....> 1.0 limit level (1) limit level (2)

When radionuclides other than those in Table 6.9-2 are detected and are the result of plant effluents, this report shall be submitted if the potential annual dose to an individual is greater than the calendar year limit of Specifications 3.9.1.2.a or 3.9.2.2.b. This report is not required if the measured level of radioactivity was not the result of plant effluents; however, in such an event, the condition shall be reported and described in the Annual Radiological Environmental Operating Report.

3.16.1.4 If milk or fresh leafy vegetable samples are unavailable for more than one sample period from one or more of the sampling locations indicated by the

ODCM, a discussion shall be included in the Radioactive Effluent Release Report which identifies the cause of the unavailability of samples and identifies locations, for 3.16-2 Proposed

Ar C

g

6.2 6.2.1 ORGANIZATION Onsite and Offsite Or anization An onsite and an offsite organization shall be established for unit operation and corporate management.

The onsite and offsite organization shall include the positions for activities affecting the safety of the nuclear power plant.

a ~

b.

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

Those relationships shall be documented and

updated, as appropriate, in the form

.,of organization charts.

These organization charts will be documented in the UFSAR and updated in accordance with 10 CFR 50.71.

The Senior Vice President, Customer Operations, shall have corporate responsibility for overall Plant nuclear safety, and shall take any measures needed to assure acceptable performance of the staff in operating, maintaining, and providing technical support in the Plant so that continued nuclear safety is assured.

An alternate title may be designated for this position in accordance with 10 CFR 50.54(a)(3).

All requirements of these Technical Specifications apply to the position with the alternate title as apply with the specified title.

Alternate titles shall be specified in the Updated Final Safety Analysis Report.

Amendment No. gp,gg,pp 6.2-1 Proposed

c ~

d.

The Plant

Manager, Ginna Station shall have responsibility for overall unit operation and shall have control over those resources necessary for safe operation and maintenance of the Plant.

The persons responsible for the training, health physics and quality assurance functions may report to an appropriate manager

onsite, but shall have direct access to responsible corporate management at a

level where, action appropriate to the mitigation of training, health physics and quality assurance concerns can be accomplished.

6.2.2 Facilit Staff The Facility organization shall include the following:

a

~

b.

C.

An auxiliary operator shall be assigned to the shift crew with fuel in the reactor.

An additional auxiliary operator shall be assigned to the shift.

crew above Cold Shutdown.

At least one licensed operator shall be present in the control room when fuel is in the reactor.

In

addition, above Cold

Shutdown, at least one licensed Senior Reactor Operator (SRO) shall be present in the control room.

Shift crew composition may be less than the minimum requirements of 10 CFR 50.54(m)(2)(i) and Specifications 6.2.2.a and 6.2.2.f for a period of time not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore shift crew composition to within the minimum requirement.

Amendment No.

gg

6. 2-2 Proposed

l'

d.

e.

An individual qualified in radiation protection procedures shall be on site when fuel is in the reactor.

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

Administrative procedures shall be developed and implemented to limit the working hours of unit staff who perform safety-related functions including senior reactor operators, reactor operators, health physicists, auxiliary operators, and key maintenance personnel.

Changes to the guidelines for the administrative procedures shall be submitted to the NRC for review.

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

The STA shall be assigned to the shift crew above Cold Shutdown.

An alternate title may be designated for this position.

All requirements of these Technical Specifications apply to the position with the alternate title as apply with the specified title.

Alternate titles shall be specified in the Updated Final Safety Analysis Report.

Amendment No.

6.2-3 Proposed

TRAINING 6.4.2 A retraining and replacement training program for the facilitystaff shall be maintained under the direction of the Division Training Manager and shall meet or exceed the requirements and recommendations of Section 5.5 of ANSI N18.1-1971 and Appendix A of 10 CFR Part 55.

The training program shall meet or exceed NFPA No.

27, 1975 Section 40, except that (1) training for salvage operations need not be provided and (2) the Fire Brigade training sessions shall be held at least quarterly.

Drills are considered to be training sessions.

An alternate title may be designated for this position.

All requirements of these Technical Specifications apply to the position with the alternate title as apply with the specified title.

Alternate titles shall be specified in the Updated Final Safety Analysis Report.

Amendment No.

6.4-1 Proposed

6.5 (Deleted)

(Intentionally Left Blank)

Amendment No. g,gP,Pg,Pg, PPIPPIPP Proposed

6.6 (Deleted)

(Intentionally Left Blank)

Amendment No. P,gg,gg Proposed

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Tl II

SAFETY LIMIT VIOLATION The following actions shall be taken in the event a

Safety Limit is violated:

a.

The provisions of 10 CFR 50.36(c)(1)(i)(A) shall be complied with immediately.

b.

The Safety Limit violation shall be reported to the Senior Vice President, Customer Operations, to the offsite review

function, and to the NRC immediately.

c.

A Safety Limit Violation Report shall be prepared.

d.

The report shall be reviewed by the onsite review function. 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.

The Safety Limit Violation Report shall be submitted to the NRC, the offsite review function, and the Senior Vice President, Customer Operations within two weeks of the violation.

An alternate title may be designated for this position in accordance with 10 CFR 50.54(a)(3).

All requirements of these Technical Specifications apply to the position with the alternate title as apply with the specified title.

Alternate titles shall be specified in the Updated Final Safety Analysis Report.

Amendment No. gg,gP 6.7-1 Proposed

PROCEDURES Written procedures shall be established, implemented, and maintained covering the following activities:

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b.

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d.

e.

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

Fire Protection Program implementation.

The radiological environmental monitoring program.

Offsite Dose Calculation Manual implementation.

Process Control Program implementation.

Amendment No.

6.8-1 Proposed

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and directions from the reactor, and the results of the participation in an interlaboratory comparison program.

Radioactive Effluent Release Re ort Routine radioactive effluent release reports covering the operation of the unit during the previous twelve months of operation shall be submitted by May 1 of each year.

This report shall include a

summary, on a quarterly

basis, of the quantities of radioactive liquid and gaseous effluents and solid waste released as outlined in Regulatory Guide 1.21, Revision 1.

This report shall include an assessment, of radiation doses from the radioactive liquid and gaseous effluents released from the unit during each of the previous four calendar quarters as outlined in Regulatory Guide 1.21, Revision 1.

In addition, the site boundary maximum noble gas gamma air and beta air doses shall be evaluated.

The assessment of radiation doses shall be performed in accordance with the ODCM.

This same report shall include an annual summary of hourly meteorological data collected over the previous calendar year.

Alternatively, the licensee has the option of retaining this summary on site in a file that shall be provided to the NRC upon request.

Also, the report shall include any nearby location(s) identified by the land use census which 6.9-4 Proposed

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6.9.2 Uni ue Re ortin Re uirements

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6.9.2.1 Annually: Results of required leak test performed on sources if the tests reveal the presence of 0.005 microcurie or more of removable contamination.

6.9.2.2 Annually: 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, in-service inspection, routine maintenance, special maintenance (describe maintenance),

waste processing, and refueling.

The dose assignment to various duty functions may be estimates based on pocket dosimeter,

TLD, or film badge measurements.

Small exposures totalling less than 204 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 shall be assigned to specific major work functions.

(NOTE:

This tabulation supplements the requirements of Section 20.407 of 10CFR Part 20) 6.9.2.3 (Deleted)

Amendment No.

$ 7 6.9-6 Proposed

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6.9.2.4 Reactor Overpressure Protection System Operation

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In the event either the PORVs or the RCS vent(s) are used to mitigate a

RCS pressure transient, a Special Report shall be prepared and submitted to the Commission within thirty days.

The report shall describe the circumstances initiating the transient, the effect of the PORVs or vent(s) on the transient and any other corrective action necessary to prevent recurrence.

Amendment No.

$ 7 6.9-7 Proposed

bC

(Deleted)

(Intentionally Left Blank)

Amendment No. ),PP,)7 Proposed

(Deleted)

(Intentionally Left Blank) 6.11-1

4'

6.13 HIGH RADIATION ARE 6.13.1 In lieu of the "control device" or "alarm signal" required by paragraph 20.203(c)(2) of 10 CFR Part 20:

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Each High Radiation Area in which the intensity of radiation is 1000 mrem/hr or less shall be barri-caded and conspicuously posted as a high radiation area and entrance thereto shall be controlled by requiring issuance of a

Radiation Work Permit (RWP).

Any individual or group of individuals permitted to enter such areas shall be provided with one or more of the following:

(1)

A radiation monitoring device which con-tinuously indicates the radiation dose rate in the area.

(2)

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 monitoring device may be made after the dose rate levels in the area have been established and personnel have been made knowledgeable of them.

Radiation Protection personnel shall be exempt from the RWP issuance requirement during the performance of their assigned radiation protection

duties, providing they are following plant radiation protection procedures for entry into high radiation areas.

An alternate title may be designated for this position.

All requirements of these Technical Specifications apply to the position with the alternate title as apply with the specified title.

Alternate titles shall be specified in the Updated Final Safety Analysis Report.

6.13-1 Proposed

b.

(3)

A Qualified health physicist (i.e., 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 who will perform periodic radiation surveillance at the frequency specified in the HPWP.

The surveillance frequency will be established by

~ a plant ealth ysicist Each High Radiation Area in which the intensity of radiation is greater than 1000 mrem/hr shall be subject to the provisions of 6.13.1 a.

above, and in addition locked doors shall be provided to prevent unauthorized entry into these areas and the keys to unlock these locked doors shall be maintained under the administrative control of the Shift Supervisor on duty.

An alternate title may be designated for this position.

All requirements of these Technical Specifications apply to the position with the alternate title as apply with the specified title.

Alternate titles shall be specified in the Updated Final Safety Analysis Report.

6.13-2 Proposed

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Offsite Dose Calculation Manual (ODCM)

Any changes to the ODCM shall be made by the following method:

6.15.1.a Licensee initiated changes shall be submitted to the Commission with the Radioactive Effluent Release Report for the period in which the change(s) was made and shall contain:

(i) sufficiently detailed information to support the rationale for the change.

(ii) a determination that the change will not reduce the accuracy or reliability of dose calculations or setpoint determinations; and (iii) documentation of the fact that the change has been reviewed and found acceptable by the onsite review function.

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6.15.1.b Licensee initiated changes shall become effective after review and acceptance by the onsite review function on a date specified by the licensee.

Proposed

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Process Control Pro ram (PCP)

Any changes to the PCP shall be made by the following method:

6.16.1.a Licensee initiated changes shall be submitted to the Commission with the Radioactive Effluent Release Report for the period in which the change(s) was made and shall contain:

(i) sufficiently detailed information to support the rationale for the change; (ii) a determination that the change will not reduce the overall conformance of the solidified waste product to existing criteria for solid wastes; and (iii) documentation of the fact that the change has been reviewed and found acceptable by the onsite review function.

6.16.1.b Licensee initiated changes shall become effective after review and acceptance by the onsite review function on a date specified by the licensee.

6.16-1 Proposed

6.17 Ma'or Chan es to Radioactive Waste Treatment S stems

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(Liquid, Gaseous and Solid)

FUNCTION

6. 17. 1

6. 17.2 The radioactive waste treatment systems (liquid, gaseous and solid) are those systems defined in Technical Specification 5.5.

Major changes to the radioactive waste systems (liquid and gaseous) shall be reported by the following method.

For the purpose of this specification, "major changes" is defined in Specification 6.17.3 below.

6.17.2.1 The Commission shall be informed of all major changes by the inclusion of a suitable discussion or by reference to a suitable discussion of each change in the Radioactive Effluent Release Report for the period in which the changes were made.

The discussion of each change shall contain:

a) a summary of the evaluation that led to the determination that the change could be made (in accordance with 10 CFR 50.59);

b) sufficient detailed information to support the reason for the change; c) a detailed description of the equipment, components and processes involved and the interfaces with other plant systems;

6. 17-1 Proposed