Text

Environ Protection Plan (Nonradiological),App B to Ol.Draft Tech Specs Encl
	ML20137X316
Person / Time
Site:	Vogtle
Issue date:	02/28/1986
From:	; GEORGIA POWER CO.
To:	;
Shared Package
ML20137X311	List: ML20137X305; ML20137X316;
References
	NUDOCS 8603060059
	Download: ML20137X316 (700)
	v • d • e

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g Vogtle

Electric Generating Plant I Unit 1 and Unit 2

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,I I Environmental Protection 1g Plan I

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!g CCCXET ?'05. EO-424, 50-425

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E?T.'IPO!"E? CAL PROTECTIC!! PLAtl
(l'0!lPADICLOGICAL )
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, l V0GTLE ELECTRIC' GENERATING PLANT UNITS 1 AND 2 l
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I ENV!R0t1 MENTAL PROTECTION PLAN (N0tRADIOLOGICAL) l 1
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TABLE OF C0flTENTS Sec tion Page 1.0 Objectives of the Envi ronmental Protection Plan. . .. . . . . . . 1 -1
2. 0 Envi ro nmen tal Pro tec tion Is sues. . . . . . . . . . . . . . . . . . . . . . . . . . 2-1 2.1 AcuaticIssues........................................... 2-1 2.2 Te r r e s t r i a l I s s u e s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-1 l
2.3 llo i s e I s s u e s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-1
3. 0 Consistency Requirements................................. 3-1 3.1 Pl a n t De si gn a nd 0perati on. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-1 3.2 Pecortin3 Delated to the NPDES Permit and State Certifications...................................... 3-2
3. 3 Changes Required for Coroliance with Other Environmental Pegulations........................................ 3-3
1. 0 E nvi ronnen tal Condi ti on s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-1 4.1 unusual or Inportant Envi ronmental Events. . . . . . . . . . . . . . . 4-1 4.? En v i ro n e n t al "o n i to ri n g. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-1 5.0 itministrative Procedures............................... 5-1 5.1 Poview and Aurt1t........................................ 5-1 5.2 R ec o rd s P e te n ti o n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-1 -
5. 3 CNnges i n Envi ronnental Protec tion Pl an. . . . . . . . . . . . . . . . 5-2 5.4 3' a nt )epo rti ng Requ i reme n ts. . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-2 k.
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1.0 Objectives of the Environmental Protection Plan l l
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! The Environmental Protection Plan (EPP) is to provide for protection of j nonradiological environmental values during operating of the nuclear facility. l The principal objectives of the EPP are as folicws:
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I l (I) 'lerify that the facility is operated in an ervironmentally acceptable i
canner, as established by the Final Environmental Statenent - Operating License Stage (FES-OL) and other flRC environmental impact assessments. I l I
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) (2) Coordinate fiPC recuirements and maintain consistency with other Federal,
.g 1 State and local requirenents for environrental protection.
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l (3) Keep imC inforced of the environnental effects of facility operation and 1
! of actions taken to control those effects.
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2.0 Environmental Protection Issues In the FES-OL dated Parch,1985, the staff considered the environmental impacts associated with the operation of the two unit Vogtle Electric Generating Plant (VEGP). CL. tain environmental issues were identified which required study or license cenditions to resolve environmental concerns and to assure adequate protection of the environment.
f 2.1 Aquatic Issues Po specific aquatic issues were raised by the ?!PC staff in the FES-OL, Ccmpliance with the !!PDES permit will assure adequate protection of the aquatic envi ronmen t.
2.2 Terrestrial Issues
?'o s;ecific terrestrial issues were identified by the ilPC staff in the FES-OL. .
Issues raised during the ASLB licensing hearings relative to cooling tower emissions were fully resolved during that process and no further conditions are required.
I 2.? :'oise Issues License conditions relative to noise associated with transmission facilities ar? specified in Section 4.2.4 of this EPD.
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3.0 Consistency Requirerents 3.1 Plant Design and Operation I The licensee may make changes in plant design or operation or perforn tests or experiments affecting the environment provided such activities do not involve an unreviewed environmental cuestion and do not involve a change in the EPP*.
Changes in plant design or operation or performance of tests which do not affect the environment are not subject to the requirements of this EPP.
Activities governed by Section 3.3 are not subject to the requirerents of this Section.
A proposed change test or experirent shall be deemed to involve an unreviewed environmental question if it concerns: (1) a matter which may result in a significant increase in any adverse environnental impact previously evaluated in the FES-OL, environmental irpact appraisals, or in any decisions of the V,onic Safety and Licensing Board; or (2) a significant change in effluents or power level; or (3) a ratter, not previously reviewed and evaluate ' in the docurents specifi M in (1) of this Subsection, which may have a significant I durse envi ror.nentsl impact.
I lefore engaging in additional construction or operational activities which nay significantly affect the enviror.nent, the licensee shall prepare and record an environmental evaluation of such activity. Activities are excluded from this requirerent if all measureele nonradiological environmental effects are
'This provision does not relieve the 1fcensee of the requirements of 10 CFR IC.50, 3-1
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I I confined to the on-site areas previously disturbed during site preparation and plant construction. When the evaluation indicates that such activity involves an unreviewed environmental question, the licensee shall provide a written evaluation of such activity and obtain prior !!PC approval. When such activity involves a change in the EPP, such activity and change to the EPP may be implemented only in accordance with an appropriate license amendrent as set forth in Section 5.3 of this EPP.
I I The licensee shall maintain records of changes in plant design or operation and of tests and experiments carried out pursuant to this Subsection. These records shall include written evaluations which provide bases for the determination that the change, test, or experiment does not involve an unreviewed environmental question or constitute a decrease in the effectiveness of this EPP to meet the objectives specified in Section 1.0. The licensee shall include as part of this Annual Environmental Operating Report (per suosection 5.4.1) briaf descriptions, analyses, interpretations, and
> valuations of such changes, tests, and experiments.
I ?.? 7eporting Related to the l'PDES Permit and State Certification Changes to, or renewals of, the NPDES Permits or the State certification shall l be reported to the NRC within 30 days following the date the change or renewal is approved. If a permit or certification, in part or in its entirety, is appealed and stayed, the IPC shall be notified within 30 days following the date the stay is grantedi 3-2 I
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The licensee shall notify the NPC of changes to the effective NPDES Permit proposed by the licensee by providing NRC with a copy of the proposed change l imediately after it is submitted to the permitting agency. The licensee shall provide the NPC a copy of the application for renewal of the NPOES Permit at the sane time the application is submitted to the permitting agency.
I Changes ?ecuired for Corpliance with Other Environnental Regulations I 3. 3 "Mnges in olent design or operation and performance of tests or experirents which are recuired to achieve compliance with other Federal, State, and local environcwntal regulaticns are not subject to the requirements of Section 3.1.
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I i 4.0 Environnental Conditions 4.1 Unusual or Important Environmental Events Any occurrence of an unusual or important event that indicates or could result '
in significant environmental irpact causally related to plant operation shall be recorded and reported to the NPC within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months followed by a written report per Subsection 5.4.2. The following are examples: excessive bird impactica events, onsite plant or animal disease outbreaks, nortality or unusual occurrence of any species protected by the Endancered Species Act of 1973, fish kills, increase in nuisance organisms or conditions, and unanticipated or emergency discharge of waste water or chemical substances.
I No route monitoring programs are required to implement this condition.
I 4.2 Envi ra r.nen t 31 ."o ni to ri ng I 4.'.1 Aquatic "onitoring
'he certifications and nernits required under tre Clean IIater Act provide echanf tms for protecting wter qualf t;' and, Indirectly, aquatic biota. The
'OC 7111 rely on the decisions made by the State of Georgia, under the 3otacrity of the Clean 'Jater Act, for any requirenents for aquatic nonitoring.
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4 4.2.? Terrestrial Monitoring i
f.'o tcrrestrial monitoring is reouired.
4. 2. 3 "ainterance of Transmission Line Corridors T.e use of Serbicides .vitbin the Vcqtle Electric Generating Plant transmission 1
iine ccericoes ( H GP-Thairz.nn, VEGP Scherer, Georgia side of VEGP-South -
Carol 1M Elect"f C Jnd Cas, <1nd VEGP-Coshen) shall confom to the approved use of selected 5ereictces is registered by the Environmental Protection Agency a .1 ". 3peroved by the State of Georgia autrorities .ird applied as directed on the herbicide label, il Se:ords shall be raintainerf in accordance with EPA or Otate of Cecrgia reuf rererts ey the licensee's Transmission Ocorating and ".11ntenance
"$'ir:rere carcernf ra berbisive u:e. Such records shall be made readily 1v11 ac13 to the 'TC mc1 recua:t, There shall be no routine reporting
. ?'uff:ra'.t m :-i.trr attk, this condition. ,
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E 4.2.3.1 Ebenezer Creek I
Any routine maintenance involving triming of the trees within the National Natural Landmark area necessary to maintain conductor clearance shall be done by hind (Section 5.2.2, FES-OL).
4.?.3.2 Francis Plantation I deutine maintenance involving trirming of the trees within the National Register cf Historic Places property necessary to raintain conductor clearance shall be denu by hand (.'>enorandum of Agreenent between Advisory Council on Historic Proservation, U.S. Nuclear Pegulatory Comission, State Historic Preservathit officer for Georgia and Georgia Power Company).
I 4.2.3.3 C:lltural Properties Along Transmission Line Corridors hutine raintenaNe activities in these areas will be in accordance with the Fia11 Cultur31 Nsource Treatment Plans, d.?.? Moise Monitorinn "rp*aints received by Georgia Power Company regarding noise :.lcn3 the high voltaw tranwission lines (VEGP-Goshen, VEGP-Scherer, VEGP-Thairy., and Sececia cise of VEGP-5CEG) so4 a report of the actions taken in rewc,1se to ory e;rplaints shall be se:bn(tted to the !!PC staff in the annual recort.
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5.0 Administrative Procedures I
5.1 Review and Audit The licensee shall provide for review and audit of conpliance with the EPP.
The audits shall be conducted independently of the individual or groups responsible for performing the specific activity. A description of the organization structure utilized to achieve the independent review and audit function and results of the audit activities shall be maintained and nade available for inspection.
I 5.2 Pecords Petention Pecords and logs relative to the envircnnental aspects of station operation shall be made and retained in a nanner convenient for review and inspection.
Thesa records and logs shall be made available to ffRC on request.
I hcceds of nodifications to station structures, systens and components determined to potentially affect the continued protection of the environnent s%11 be retained for the life of the station. All other records, data and logs relating to this EPP shall be retained for five years or, where applicable, in accordance with the requiremcnts of other agencies.
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5.3 Changes in Environmental Protection Plan I
Requests for changes in the EPP shall include an assessment of the environmental inpact of the proposed change and a supporting justification.
Implementation of such changes in the EPP shall not cornence prior to NRC approval of the proposed changes in the forn of a license amendment I incorporating the appropriate revision to the EPP.
I 5.4 Plant Peporting Requirenents I 5.4.1 Poutine Peports I An Annual Environmental Operating Peport describing implementation of this EPP for the previous year shall be submitted to the NPC prior to May 1 of each year. The initial report shall be submitted prior to l'ay 1 of the year following issuance of the operating license.
I The report shall include summaries and analyses of the results of the environmental protection activities required by Subsection 4.2 of this EPP for the r? port period, including a corparison with related preoperational studies, )
operational cont >ols (as appropriate), and previous nonradiological environmental monitoring reports, and an assessment of the observed impacts of the plant operation on the environment. If harmful effects or evidence of trends toward irreversible danage to the environment are observed, the I licensee shall provide a detailed analysis of the data and a proposed course of mitigating action.
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The Annual Environmental Operating Peport shall also include:
I (1) A list of the EPP noncompliances and the corrective actions taken to remedy them.
I (2) A list of all changes in station design or operation, tests, and experiments nade in accordance with Subsection 3.1 which involved a notentially significant unreviewed environmental cuestion.
(3) A list of noncoutine reports submitted in accordance with Subsection 5.4.2.
I In the event that some results are not available by the report due date, the report shall be submitted noting and explaining the missing results. The missing results shall be submitted as soon as possible in a supplementary re p o r*t.
I 5.4.2 Penroutine Reports
written report shall be subnitted to the !!PC within 30 days of occurrence of a noncoutine event. The report shall (a) describe, analyze, and evaluate the event, includina extent and magnitude of the impact,.and plant operating characteristics, (5) describe the probable cause of the event, (c) indicate the action taken to correct the report event, (d) indicate the corrective I
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action taken to preclude repetition of the event and to prevent similar occurrences involving similar components or systens, and (e) indicate the agencies notified and their preliminary responses.
l Events reportable under this subsection which also require reports to other r ederal, State or local agencies shall be reported in accordance with those reporting recuirements in lieu of the requirements of this subsection. The f.'PC shall be provided with a copy of such report at the same time it is subnitted to the other agency.
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-r i 3/4.7 PLANT SYSTEMS
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v 3/4.7.1 TURBINE CYCLE SAFETY VALVES LIMITING CONDITION FOR OPERATION 3.7.1.1 All main steam line Code safety valves associated with each steam generator of = =i::?:t:d re::ter = Int l=p shall be OPERABLE with lift settings as specified in Table S-7-ih 3.7-l. .'
APPLICABILITY: MODES 1, 2, and 3.
ACTION:
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' eper:b!:, :p:r:tien in "_GC,ES ______1, ,_ 2, =d,3 =,y proceed provided, that
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' V /' g " T ' *, n=t S hours-and-in-COLD-SHUTDOWN-within-the-following-30 h=r:.
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eper:ti e =d with-ene- *-mece-main :t: = lin: 0-d; 2:fety c.iv;;
as+ostated-with-an-operating-loop-inoperabley-operati= ' "a"ES 1,
[ 2, =d 3-may-proceed-providedr-that-within-4 hours, either the inocerable-valve-is-restared t OPER.^.SLE :t:t;; Or th: "== "=;;
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Neutron Fhx High Trip ":tpeint-ts reduced per Table- 3.7-2; cther iee, w __m , <-- -> - - . -
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\c. The peowis4:n: ef Specift:ati n 3.0.' =; n:t :pplict!:.
SURVEILLANCE REQUIREMENTS 4.7.1.1 No additional requirements other than those required by Specification 4.0.5.
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i V0GTLE - UNIT 1 3/4 7-1 1
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Insert for page 3/4 7-1 MODE 1:
a. With 4 reactor coolant loops and associated steam generators in operation and with one or more main steam line Code safety valves inoperable, operation in MODE 1 may proceed provided that within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , either _
the inoperable valve is restored to OPERABLE status or the Power Range Neutron Flux High Trip Setpoint is reduced per Table 3.7-2; otherwise reduce thermal power to less than 5% of RATED THERMAL POWER within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .
b. The provisions of Specification 3.0.4 are not applicable.
MODES 2 and 3:
a. With no main steam line Code safety valves in each steam generator OPERABLE, either restore at least one main cteam line Code safety valve to OPERABLE status in each steam generator within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .
b. The provisions of Specification 3.0.4 are not applicable.
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l TABLE 3.7h STEAM LINE SAFETY VALVES PER LOOP VALVE NUMBER LIFTSETTING[11% ORIFICE SIZE SG-t SG-2 Scr-3 SG--I
1. nv seei ni r e,i nn-vi ou e c3 a? ?A psig is .o i.,. z-i
2. tuvsea psv se rv kn Now,est ?cce psig is. e in "
3. tsv. wet ,psv tear ow3e2lpsnes3 !2io psig '
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4.wtrecs en,.in av,cu pusep i2 zo psig is . e in '
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The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure.
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V0GTLE - UNIT 1 3/4 7-t
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.L x TABLE 3.7 ') MAXIMUM ALLOWABLE POWER RANGE NEUTRON FLUX HIGH SETPOINT WITH INOPERABLE STEAM LINE SAFETY VALVES DURING-W- LOOP OPERATION i
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MAXIMUM NUMBER OF INOPERABLE MAXIMUM ALLOWABLE POWER RANGE SAFETY VILVES ON ANY NEUTRON FLUX HIGH SETPOINT OPERATING STEAM GENERATOR (PERCENT OF RATED THERMAL POWER) 1 '[)877 2 E443GS
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3 VCGTLE - UNIT 1 3/47 d
PLANT SYSTEMS
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AUXILIARY FEEDWATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.1.2 At least three independent steam generator auxiliary feedwater pumps and associated flow paths shall be OPERABLE with:
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a. Two motor-driven auxiliary feedwater pumps, each capable of being powered from separate emergency busses, and
b. One steam turbine-driven auxiliary feedwater pump capable of being powered from an OPERABLE steam supply system.
APPLICABILITY: MODES 1, 2, and 3.
ACTION:
a. With one auxiliary feedwater pump inoperable, restore the required auxiliary feedwater pumps to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . '
i O b. With two auxiliary feedwater pumps inoperable, be in at least HOT STAND 2Y witin 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .
c. With three auxiliary feedwater pumps inoperable, immediately initiate corrective action to restore at least one auxiliary feedwater pump to OPERABLE status as soon as possible.
SURVEILLANCE REOUIREMENTS 4.7.1.2.1 Each auxiliary feedwater pump shall be demonstrated OPERABLE:
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a. At least once per 31 days on a STAGGERED TEST BASIS by:
1) Verifying that each motor-driven pump develops a discharge pressure of greater than or equal to /jLgp psig at a flow of greater than or equal to tys gpm; ( pr.g,+o, pr-3,41,1 r-tric/, j ff-iSio . PZ-fiz 1, PZ~ S/2.9)
2) Verifying that the steam turbine-driven pump d'avelops a discharge pressure of greater than or equal to /G5S psig at
, a flow of greater than or equal to /6o gpm when the secondary steam supply pressure is greater than 9 r e psig. The provisions of Specification 4.0.4 are not applicable for entry into MODE 3; O
V0GTLE - UNIT 1 3/4 7-4 -
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O PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
3) Verifying that each non-automatic valve in the flow path that is not locked, sealed, or otherwise secured in position is in its correct position; and
4) Verifying that each automatic valve in the flow path is in the fully open position whenever the Auxiliary Feedwater System is pi;;;d in cuts.T.ati; ;;ntrei er when ab;ve 10% RAT:0 Tll 2"AL PCWtR. in .rw,dly for a.uiNary feedwaJer au.ny,ajic. inifinfios e r wi, e n above so % 2MGo mfMAf44. Pcwft .
b. At least once per 18 months during shutdown by:
1) Verifying that each automatic valve in the flow path actuates to its correct position upon receipt of an Auxiliary Feedwater Actuation test signal, and
2) Verifying that each auxiliary feedwater pump starts as designed automatically upon receipt of an Auxiliary Feedwater Actuation test signal.
4.T.1.2.2* An a iliarj feed.;;ter flow p;th t: ;;;h :t;= g=:r;t;r ;h:11 b; demonstr;t;d CPEPlaLE following each COLO S:'UT00' N cf scoatei thor. 30 uog pri;r t; cat ring "00: 2 by-veci-fying-normel-fic to eech steaa. goc..cetec.
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This is e,,plicable only for plants that do not u;; auxili;ry feed ;tcr f;r ST?9 TUP /5 HUT 00',l" cp;r;ti;n;.
V0GTLE - UNIT 1 3/4 7-5
l PLANT SYSTEMS CONDENSATE STORAGE TANK LIMITING CONDITION FOR OPERATION V4 0cl
, 3.7.1.3 -The fondensate storage tank (CST); shall be OPERABLE with a contained water volume of at least3accee gallons of water.
h ( ca s. t2. %K t r-Sic /)
APPLICABILITY: MODES 1, 2, and 3.
ACTION:
y4 cot With 4ho CST, inoperable, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months either:
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a. Restore the CST,to OPERABLE status or be in at least HOT STANOBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , or mo t csr v4cc.1
b. DemonsthatetheOPERABILITYofsthe[alternatewatersource]as a backu) supply to the auxiliary feedwater pumps and restore the CST; to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .
SURVEILLANCE REQUIREMENTS V4cci 4.7.1.3.1 The CSTt shall be demonstrated OPERABLE at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by verifying the contained water volume is within its limits when the tank is the supply source for the auxiliary feedwater pumps.
c gr v4cc2 .
4.7.1.3.2 The [elternete weter seurc Q shall be demonstrated OPERABLE at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months bys[::thod depend:nt upon alternet; :;;urce] whenever //
the [elternete weter scurce] is the supply source for the auxiliary feedwater pumps. \
i veroYy ino th e cen ta.ined. wa./er* vo/ame 4 a f le.t.s t J Yo, c o o ja //e n s (6 9 12, T,) (l Z- fic +),
O V0GTLE - UNIT 1 . 3/4 7-6
. _ _ . _ _ -~ _ -
PLANT SYSTEMS SPECIFIC' ACTIVITY LIMITING CONDITION FOR OPERATION 3.7.1.4 The specific activity of the Secondary Coolant System shall be less than or equal to 0.1 microcurie / gram DOSE EQUIVALENT I-131.
APPLICABILITY: MODES 1, 2, 3, and 4.
ACTION:
With the specific activity of the Secondary Coolant System greater than 0.1 microcurie / gram DOSE EQUIVALENT I-131, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 nours.
SURVEILLANCE REQUIREMENTS 4.7.1.4 The specific activity of the Secondary Coolant System shall be determined to be within the limit by performance of the sampling and analysis program of Table 4.7-1.
s e
O V0GTLE - UNIT 1 3/4 7-7
TABLE 4.7-1 Os SECONDARY COOLANT SYSTEM SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM TYPE OF MEASUREMENT SAMPLE AND ANALYSIS AND ANALYSIS FREQUENCY
1. Gross Radioactivity At least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .
Determination"
2. Isotopic Analysis for DOSE a) .0nce per 31 days, when-EQUIVALENT I-131 Concentration ever the gross radio-activity determination indicates concentrations greater than 10% of the allowable limit for radioicdines.
b) Once per 6 months, when-ever the gross radio-activity determination indicates concentrations less than or equal to 10%
of the allowable limit O for radioiodines.
/l4'
A gross radioactivity analysis shall, consist of the quantitative measurement of the total specific activity cf the secondary coolant except for radio-nuclides with half-lives less than ti minutes. Determination of the contributors to the gross specific activity shall be based upon those energy peaks identifiable with a 95% confidence level.
O_
V0GTLE - UNIT 1 3/4 7-8
PLANT SYSTEMS O,rS sysic,,,s(cs,,sssrin so. ,na.i., svan ,
MAIN STEAM LINE ISOLATION VALVES . .
i.sols.h'an vs.lVe.(AtSt V) An d. t f.s
' assoeio.+eol bypass va.Ive.(MSie v1pe r-LIMITING CONDITION FOR OPERATION inu, ff,, ,
i Mo 3.7.1.5 -Gash main steam line isolation volv '"
!V) shall be OPERABLE.4 APPLICABILITY: MODES 1, 2, and 3.
ACTION:
me n .- 4.
rm .
uith cne MSIV ineperable but egen, POWER OPERATION :y centinue previded the ineperable valve is-rectered te OPER?.SLE :t tu: . th'-
heur:; otherai : be in HOT STANDSY within the n ;;t S heur :nd i- MCT SH"TOCh" with'- the fellet.in;; 5 h:ur:.
"00ES 2 nd 3:
With en: "SIV inoperable, :ub cquent oper:ti n in ."00E 2 er 3 ;;y pr;;;;d p = 4 dad +ka isolatic.' velve is maintained cle::d. Oth;rai :, be in ll0T STANDSY eithi- the next S heur and in HOT SH"TOOWN within the fell wing S heur:.
A Rep /a c e. wi/h inserf fa $9 e S/+ TA*'
SURVEILLANCE REQUIREMENTS
.Lnd. Aff/8 V 4.7.1.5 Each MSIV shall be demonstrated OPERABLE by verifying full closure within r seconds when tested pursuant to Specification 4.0.5. The provisions of Specification 4.0.4 are not applicable for entry into MODE 3.
+ dn C/G*gedLC inain steam /ta,e ise/afien sys/cin ca ru cois.sitt of an Ot' EAR 8LE MstV and an in opera.ble bu.f closed a.s. socia.ted MS /A V provided the inoperab/c .Msid y i,s ,na.intairrez closei.
V0GTLE - UNIT 1 3/4 7-9 nwwe-w ,e ,. v-- --a
Insert for Page 3/4 7-9
.s MODE 1:
a. With two main steam line isolation systems in any steam line inoperable; POWER OPERATION may/ continue provided each MSIV in the affected steam line is open and at least one main steam line isolation system in the affected steam line is restored to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . Otherwise, reduce power to less than or equal to 5% of RATED THERMAL POWER within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ,
b. With one main steam line isolation system inoperable, power operation may continue provided the MSIV in the affected isolation system is open and the inoperable system is restored to OPERABLE status within 7 days. Other-wise, reduce power to less than or equal to 5% of RATED THERMAL POWER within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .
! MODES 2 and 3:
1
a. With two main steam line isolation systems in any steam line inoperable, subsequent operation in MODES 2 or 3 may proceed provided at least one main steam line isolation system in the affected steam line is maintained closed.
The provisions of Specification 3.0.4 are not applicable. Otherwise, be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and H0T SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .
b. With one main steam line isolation system inoperable but closed, subsequent operation in MODES 2 or 3 may proceed provided that the isolation system in O' the affected steam line is maintained closed. The provisions of Specification 3.0.4 are not applicable.
c. With one main steam line isolation system inoperable but open, either close the OPERABLE isolation system or restore the inoperable system to OPERABLE status within 7 days. Otherwise be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .
PLANT SYSTEMS 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION LIMITING CONDITION FOR OPERATION l 3.7.2 The temperatures of both the reactor)and secondary coolants in the steam generators shall be greater than T70fF when the pressure of either coolant in the steam generator is greafer than (200fpsig. #
j /
APPLICABILITY: At all times.
j ACTION:
' With the requirements of the above specification not satisfied:
a. Reduce the steam generator pressure of the applicable side to less than or equal to (200,}'psig within 30 minutes, and
b. Perform an engineering evaluation to determine the effect of the overpressurization on the structural integrity of the
! steam generator. Determine that the steam generator remains acceptable for continued operation prior to increasing its temperatures above 200*f. '
O SURVEILLANCE REQUIREMENTS 4.7.2 The pressure in each side of the steam generator shall be determined to be less than(2003'psig at least once per hour when the temperature of either the reactor or sei:endary coolant is less than 470}'F.
O V0GTLE - UNIT 1 3/4 7-10
. _ - - - . ~ _ - - . _
PLANT SYSTEMS i O U. 3/4.7.3 COMPONENT COOLING WATER SYSTEM LIMITING CONDITION FOR OPERATION ha. ins 3.7.3 At least two independent component cooling water 4eope shall be OPERABLE /
WNh a} / eat /wo pump.s MODES 1, 2, 3, and 4.per /YAin.
. APPLICABILITY:
ACTION:
i h nin With only one component cooling water kop OPERABLE, restore at least two Nuoneeps to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .
SURVEILLANCE REOUIREMENTS
%'ns 4.7.3 At least two component cooling water leeps shall be demonstrated OPERABLE:
a. At least once per 31 days by verifying that each valve ( :ne:1, l power-eperated, er :ute stic) :ervicing :sfety- :1:ted :quipx nt that is not locked, sealed, or otherwise secured in position is in its correct position; and
b. At least once per 18 months during shutdown, by verifying tha I
i
\1) Seh autc= tic valv: :ervicing :sfety-related equip ent actuetes te it correst pe:ition er test signal, ead Gt M ach Component Cooling Water System pump starts automatically on a - test signal.
t
$4.$ Gj *ecNCn t
. O_
e V0GTLE - UNIT 1 3/4 7-11
PLANT SYSTEMS O Noc teart sa r2ner cooDMG- W4rst (NscW) sysrr&1 3/4.7.4 SERVICE WATER SYST:"
LIMITING CONDITION FOR OPERATION 3.7.4 At least two independent NscW scrn ::trainsw:ter 1::p: shallbeOPERABLE/
with af /ea.sf ha puns per ka.in.
APPLICABILITY: MODES 1, 2, I, and 4.
ACTION:
,% cd fra.in tr'a in s With only one ser" ice ester leep OPERABLE, restore at least two leeps to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .
SURVEILLAtiCE RE0VIREMENTS NscW frains 4.7.4 At least two sern.ce water 100p shall be demonstrated OPERABLE:
a. At least once per 31 days by verifying that each valve (manual, power-operated, or automatic) servicing safety-related equipment that is not locked, sealed, or otherwise secured in position is in its correct position; and
b. At least once per 18 months during shutdown, by verifying that:
1) Each automatic valve servicing safety-related equipment actuates to its correct position on a , test signal, and Safe /y Zgee:Mn
2) Each Service Water System pump starts automatically on a test signal.
Safe'y hjec-%
i V0GTLE - UNIT 1 3/4 7-12 i
PLANT SYSTEMS s
3/4.7.5 ULTIMATE HEAT SINK [0PTIOM^']
LIMITING CONDITION FOR OPERATION 3.7.5 The ultimate heat sink shall be OPERABLE with:
.<,,e r.a c e y ,. ,r* ,
. :_::= u:ter hvel at- cr ebev; chvetion " ;n R ; L;.;1,
/<w d 'e' USC': ht=, =d 31r 7-13
?n aver g: u:t;r temperatur; cf h:: th= Or :;=1 t:
F.
(b.
APPLICABILITY: MODES 1, 2, 3, and 4.
ACTION:
With the requirements of the above specification not satisfied, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .
SURVEILLANCE REQUIREMENTS 4.7.5 The ultimate heat sink shall be determined OPERABLE::t h=t =:: p:r 24-heurs by verifying-the average-weter te.,per;tura and water hvci t; b:
Mthin th& "~4'!_
i
~L ' b r
, W st. F o#
C r9 C 4 ,$d f A Y b & t./Vf ,hY ye f t' f / /pe ~ nt> hje g, gt.stg e ')j 4g i
vo fer d e < * / a. a.d wwie e fe v n.oe.r n ia re. fo b e <Jo' fir,'r>
- n e ,', li m ,' .-s .
b- l 8 s2 $ f fYY $$ V& fl' fl)d f fkk M s"r1,'pp164eyf f71% d.
SlT mf ! ^r, d r e a a ,J e .t' r,u . <,xe ei fans s w + a,,x epa ,afe ;'e ,.tr
!+' t s t /5 mi ss a.les .
J V0GTLE - UNIT 1 3/4 7-13 1
,_ _ _ . _ - _ _ . . . .._ . . _ _ _ _ . . . _ _ _ _ _ _ .._..m, _ ._. ., _ . . _ _ _ . . _ . _ . _ . . ~ . . . _ _
~
1.
(3 V Insert for Page 3/4 7-13
a. Two OPERABLE Nuclear Service Cooling Water (NSCW) tower basins each with:
1. A minimum water level (LI-1606 and LI-1607) at or above plarit elevation of 217 ' 3" (73*.' of span).
2. A maximum water temperature (TE-1642 and TE-1643) of 900F.
b. Two OPERAELE trains of NSCW tower fans and spray cells. The required number of fans and spray cells per train is a function of ambient wet bulb tem;:erature and shall be in accordance with the following:
Ambient Wt Bulb Minimum Required Minimum Required Temperature OPERABLE Fans OPERABLE Fans
>550F 4 4 Between 40U F and C00F 3 3*
<400F 0 2 2*
Between 15 F and 650F 3 4 2 4 1150F
Spray required to the cells with OPERABLE fans and spray to other cells isolated, i
l O !
I k----------
,-- \
I PLANT SYSTEMS s3/4.7.6 FLOOD PROTECTION [0PTIONAL*]
LIMITING CONDITION FOR OPERATION
'N 3.7.6 Flood protection shall be provided for all Safety-Related Systems, components, and structures when the water level of the (usua}ly the ultimate heat sink] exceeds Mean Sea Level, USGS datum, at/ .
/
APPLICABILITY: At all times.
ACTION-With the water lev at above elevation Mean Sea Level, USGS datum:
/
/
a. [Be in at lea HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months /and in at least COLD SHUTOOWN w'Kthin the following 30 hour3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months s], and
b. Initiate and comp te within ho s, the following flood protection measuresis
1. [ Plant dependent], and
,i
,- )
2. [ Plant dependent].
w .-
SURVEILLANCE REOUIREMENTS
/
4.7.6 The water level at
\
shall be determ(ned to be within the limits by:
a. Measurement at 1 ast once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Nhen the water level is below elevation Mean Sea Level, USGS dathm, and Measurement
' \
at least once per 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months when the water level is equal
b. f to or abov4 elevation Mean Sea Level, USGS datum.
/
/ 'N s
/ 'N p x
^ Inis specification not requirec if the facility design has adequate passive floc Ccontrol protection features sufficient to accommodate the Design Basis Flood identified in Reculatory Guide 1.59, Auaust 1973.
s j^'\
v / ,
/ ~
V0GTLE - UNIT 1 3/4 7-14
~
inifia)e .2 nd snainfain a ra.ftn of fhe remaintne opu$$4f Con +rel p geom EMergeney Altra.+ien Sy s+em in the Eme,yency 6 ede. C Nier /sej Q PLANT SYSTEMS
& pn.rner r?cd a
3/4.7.7 CONTROL ROOM EMERGENCY AIR CLEANUP SYSTEM LIMITING CONDITION FOR OPERATION i, fi/tra.ticn 3.7.+ Two independent Control Room Emergency Air Cleer.up Systems shall be OPERABLE.
APPLICABILITY: AH-M00E-S. wop g j,2,4 ,,,s 4. g ,o g , s a.,,x g, d u i,,y m e v es e n t o f irr a d ia. fed fa.e / o r mo vem ent ACTION: , ,e , _ g , , ,e ,. ;,. ,a g ; , y . r e J.
MODES 1, 2, 3 and 4:
fi/frasen With one Control Room Emergency ^*- Ch:ne; System inoperable, restore i the inoperable system to OPERABLE status within 7 days or,be in at least l HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .
MODES S andof '
6bihadsover f Gri f "airra ve md eian ted t o ffa.e/.1,%v/<a.s,.
im d ided h e / o< ~~ e"?
a. With one Control Room Emergency Mc-Cie nv[ System inoperable, restore the inoperable system to OPERABLE status within 7 days or initiate and maintain operation of the remaining OPERABLE Contrci f
Room Emergency R/WenAir Cleanup System*"'in* the p,m, *y*"r::irculation c?' acde.
. p, g ;,,,
b. With both Control Room Emergency A'r Ch,;nup Systems /i,no,perable, '
" " 'fedc/ "_or with the OPERABLE Control Room Emergency ^*- C!:fnup System of being powered by an OPERABLE emergency power source, suspend all operations involving CORC ALTERAHONS er pcsitive re;;tivity cheni,e..
m ren.i,& ef <rea.dided fu el e< imve<<ue,,+ c. / /oaJ.s ever irradia. red dt e /. ,
SURVEILLANCE REOUIREMENTS :
$ R/tra +ien
4. 7.9 Each Control Room Emergency Air Chanup System shall be demonstrated i OPERABLE:
a. At-4 east-ence-Ser 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by v:ri fying that-the-eentr:1 r00:
air-temperature-is-less-than er equal to [S07F; ,
ub. At least once per 31 days on a STAGGERED TEST BASIS by initiating, from the control room, flow through the HEPA filters and charcoal adsorbers and verifying that the system operatas for at least 10 continuous hours with the heatery Oper: ting;cen/rs/ cire cu / ener,ful.
1 0
,4 V0GTLE - UNIT 1 3/4 7-1+
J
. l I )
PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) b..es At least once per 18 months or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire, or chemical release in any ventilation zone communicating with the system by:
^ * /' ^ " ""# #
f.) V:r*#ying s
th:t th: ch:nup :y:ter ::th'h , th: h ph;;
penetreMoreand-byp :: h:kage tertiag rcc-a+'a"= "*4+="4= '-
af
',d"'* .: 5 M:: the t :t pro::dar; ;;id:n:: ":g h-
1. .y 7.,g , i re ,n
(' .... .th:n [*]*' :e.nde.....,
_ _ .. , . n ._. ..t.. .. :._.._ . ,. 20^
e _ __2 e
. . . . . e .. .a. ........__e.... : , _ ,
n__..,_
..., . ... e, , ,
..u_
e_.a,s
_, o
, _ . . . . , o. ,. .o , . .
u._,6 cfm 10%;
[ 9 %,
. 1!a. m.4 #. ,s ,4.
..,m-n .,ui. +.h. i n 'l l_ _ A_ _m,u _e_m f. +.m_ e.- - - s u a l. , .+hst
- - _ _ m_ i-nka _ _ _p - -s_ t_ e
_ m,s y", ) ^. iAp.,IC An d. w. N.w b uw w . .~ "
wA. $ier.,.O
^^
e .y, p m- _ -,i - a = --
e e <
bf 4 44,4.- ',*h J ___
f.L n__.,_.__.
nw, i .u ..y h__f.
..,....u w..,.a.,
.E i nmyw s a ww s y wuaun A..n.,
2.__2.
/Astrf O n_.2,t__ e u___L
, o7e ___.. .L.
.__..._.m.
_L___
...j
._,.t__
......3 .. . . . . .
3.+ r-/5. eie m <. emm.,m+m o om,4+4.mm e_ . e_ . m_ _., o ._,_
. . 3_. ., ._ __.__
c . . .u.. , eo o_mj.
r............-...n
,o,o ,__ _ __.m., ,_om . ..
,4mm , u.., . . , . . .
u- r**w. ma s _. _ , ,
3) Verifying a system flow rate of.2< ecccfm + 10% during system operation when tested in accordance with;XNSI N510- Wf&.
'-* ' W *
, se.cHen 1. o / <. .,
i n
m
,,e m_.__ _, _m ____, _2_ ___ _.
.....,..3, L /,/'/. u e, s y, (.
s it'* - 21 ___m__
day: 2*t .___,_
r r1
1,
_L._,__2 th:t
1:':r:t:
_____2____
y :::1y:
S :'_:_..
n__..,_.
r;pr:-
, ,, , y 9 _,n__.,.__.. e . 2 2. g n_. s_s__ e __L
,__si.
,e..,e,.., t
.......e,s, u. . . . . ,. o.7. 8, i m . f g ,% c, .
. ~ . . . . .
--++h.
4 4 e
i m# em,,,,,. _ 3msm.3t7 n . . u. ,7,tm e, t ; cy$tey . :< o ,o,o ;r ht: y "_::!!'
o . . , 4. ,. 4 - . 4, u. _. .__. u. . . ..., ... . . . . . . . , .
,_ ___m.., n C.S.a
\ Sdid ;: .:tr:ti: :' h:: th:n ;** *;
. da At least once per 18 months by:
1) Verifying that the pressure drop across the combined HEPA
filters and charcoal adsorber banks is less thanJ6 finches Water Gauge while operating the system at a flow rate of.,:y,f.,cc .
' cfm + 10%;
~
b ,, trol stoe ,m .Es o I. shen
. 2) Verifying that on a C:nt i-- nt "h::: "'" Schth; ; .d "igh
' D:h: " n:ity fest signal, the system automatically switches umegocv V into a Mr* : hti:n mode of operation with flow through the
' " HEPA filters and charcoal adsorberseenke.,
4 Verii'ying that the system maintains the control; roomjat a 4
l 3) positive pressure of greater than or equal to.T1/8 71nch Water l Gauge :t 1::: th:n :r : :1 t: : pr::: rb:ti:n '1 r :"
. c'- elit!'!: ** 2d*: r-t :rr : dur'n; t:: : ;r;t hn; ,
ye h t; t te rhe. own.de. atancspnere.d:"a:. rig +.>;yste,n ep varien.
. 4) Verifying that the heaters dissipate //e 4 kV when nd tested in caccordance .,r. .t a .a wiM7 AWNi e510-1M' 70 uetarin
5) Verifying that on a'High Chhrin

/ Toxic Gasdest signal, ,
m.m.
t ,.:. :y:t:r ::t-- 4.;tu..h;113..;. it;';; ic,tc, . .. .. J . ....
l 1
m u_ orn. 11,
..,,.4_.
_ m. ._m <_ _
._, r,-, ._
... .. . . H e e.,Jeal room is et s.rie n
. . . . . . . . .,...-........m
u. ,', e ,s e. e u w. m r u e .,is.
VCGTLE - UNIT 1 3/4 7- N IS i
as Insert to Page 3/4 7-15, item I ,
4 l
! 1. Verifying that the cleanup system satisfies the in-place testing i l acceptance criteria of greater than or equal to 99.5% filter '
efficiency while operating the system at a flow rate of 25,000 cfm + 10% and performing the following tests:
l I (a) A visual inspection of the control room emergency air cleanup j system shall be made before each 00P test or activated carbon
' adsorber section leak test in accordance with Section 5 of l ANSI N510-1980.
t (b) An in-place D0P test for the HEPA filters shall be performed in accordance with-Section 10 of ANSI N510-1980, l
i (c) A charcoal adsorber section leak test with a gaseous halogenated
! hydrocarbon refrigerant shall be performed in accordance with 1 Section 12 of ANSI N510-1980.
i.
i i
Insert to 3/4 7-15, item 2.
i O 2. Verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in accordance with :
1 Section 13 of ANSI N510-1980 meets the laboratory testing cri _ ,
) terion of greatec than or equal to 99.8% when tested with methyl i
> iodide at 800C and 70% relative humidity.
l Insert to 3/4 7-15, item c i
l c. After every 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of charcoal adsorber operation by verifying within
! 31 days after removal that a laboratory analysis of a representative
carbon sample obtained in accordance with Section 13 of ANSI N510-1980 i meets the laboratory testing criterion of greater than or equal to l 99.8% when tested with methyl iodide at 800C and 70% relative 1 humidity.
lO l -
r
s Pl. ANT SYSTEMS r SURVEILLANCE REQUIREMENTS (Continued) ,
g .<.,.. ,. . , - - .
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, . . . o m 4 n,4,,, un + u t . . . ,- .,,.m ,,,n ,4.,,- +u ., , , . . ........u.
.----- ,- -- r - - -- ...
. gf' ng.c g ;s!,11r -- - a - - - -
-,7
.a....u.+u,,,,...,.~.<u<, ,, r u , , +u- raw a ,
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g4y. - - - = - -- - - ---r----- -- -- - -- -
if.r m e n_ s ne a. :ag{ Q M T. M. C. i n ? u. 7..- C +.e. .1 A'7- l
_ ..<= .. . ,__. SPD
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t a. n e. n ce4._,1 . .,n t. ,_r . _ . , +-. e r. , s k . . s- a . '.
. . . _ <u,, , . .u . .__ ____ _ _,___ ...
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U l IQ l V0GTLE - UNIT 1 3/4 I -
l
_ . _ _ - . - . _ , - _ . - _,-- .._..___,,-,,c.., r , m.y - __ -.-,.--.m n-_.,__ , -
l
() Insert to Page 3/4 7-16 i
e. After each complete or partial replacement of a HEPA filter bank i by verifying that the HEPA filter banks remove greater than or equal to 99.5% of the 00P when they are tested in-place in accor-dance with Section 10 of ANSI N510-1980 while operating the system at a flow rate of 25,000 cfm +10%.
f. After each complete or partial replacement of a charcoal adsorber '
Dank by verify ng that the charcoal adscrbers remove greater than i
or equal to 99.5% of a halogenated hydrocarbcn refrigerant test gas when they are tested in-place in accordance with Section 12 of ANSI N510-1980 while operating the system at a flow rate of 25,000 cfm 110%.
O t
4 O
. . . , .- . _ - . _ . . _ . . _ , _ - . _ . _ _ , _ , . . . . . . _ _ . _ _ . _ _ . _ _ _ . . . . _ . _ . _ . - . . . ,---,._...._m. ._ ,. . ._.. . . ..
l
) PLANT SYSTEMS 3/4.7.8 ECCS PcM9 RCCM EXHAUST AIR CLEANUP SYSTEg
\s '
LIMITING CONDITION FOR OPf8iATION -' .
3.7.3 Tk incependent ECCS 9t.mp Room Exhaust Air Cleanup Systems s 11 be OPERABLE. \
\
APDLICA8ILIT g NODES 1, 2, 3, and 4.
ACTION:
\ s With one ECCS Purp Rcom Exnaust Air Cleanup Systemble, in.opeya/
restore tne incperable system to QPERA3LE status wf thin 7 days or t4 in at least HOT STANDBY within the nex' E hours and in COLD SHUTCOWN thin'the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .
l SURVEILLANCE REQUIREPENTS \
p 4.7.8 Each ECCS Pump Roca Exhaust rC anup System shall be demonstrated OPERABLE:
\
a. At least once per'31 days n STAGGERED TEST SASIS by initiating, -
from the centrol room, f sw thr gh the HEPA filters and charcoal adsorbers and verifyin that the bystem cperates for at lear,t 10 continuous hours w'un the hesters operatirg; ;
\
b. At least once per ' months or (1) after any structural maintenance on the HEPA filt or cnarcoal adsorber\hcusings, or (2) following painting, fire, or che'sical release in aby ventiletion zone communicating ith t!e system by:
'\
1) Verif ngthatthecleanupsystensatishiesthein-place pec ration and bypass leakage testing acdeptance criteria of le than [*]% and uses the test procedure ' guidance in Regula- '
. ry Positions C.5.a, C.5.c anc C.5.d of Regulatory Guide 1.52, flevision 2, March 1978, and the system flow rate is cfm i 10%; y -
\
) Verifying, within 31 days after removal, that a laboratory analysis of a representative carbon sample obtained in accor-dance with Regulatory Position C.S.b of Regulatory Cufde 1.52, Revision 2 March 1978, maets the lacaratory testing criteria of Regulatory Positi.on C.6.a of Regulatory Guide 1.52. Re\i-sien 2, March 1978, for a met.4y1 icdice penetration of lessq '
than [**]%; and \
VCGTLE - UNIT 1 3/4 7-18
/ - . . .
m PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
3) Verifying a system flow rate of cfm f 10% during syst operation when tested in accordance with ANSI N510-1975.
After every 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of charcoal adsorber operation, by ver fying, c\ within 31 days after removal, that a laboratory analysis o a epresentative carbon sample obtained in accordance with egulatory sition C.6.b of Regulatory Guide 1.52, Revision 2, Mar h 1978, me ts the laboratory testing criteria of Regulatory Po tion C.6.a of kegulatory Guide 1.52, Revision 2, March 1978, for a methyl iodid% penetration of less than [**]%; ,
d. At lees once per 18 months by:
1) Veri ing that the pressure drop across th combined HEPA filter. and charcoal adsorber banks is leafs than [6] inches Water Ga e while operating the system a flow rate of _
cfm f,10%,
2) Verifying th t the system starts on;I Safety Injection test signal,
/
3) Verifyingthatthesystemmaintaf6stheECCSpumproomata negativepressureNfgreatert n or equal to [1/8] inch Water Gauge relative to the outside tmosphere,
4) Verifying that the fi er c ling bypass valves can be manually O opened, and Verifying that the heat dissipate + kW when 5) tested in accordance w h SI N510-1 W -
e. After each complete or pa7 ti al rep cement of a HEPA filter bank, by verifying that the cl,eanup syste satisfies the in place pene-tration and bypass leakage testing ac ptance criteria of less than
[*]% in accorcance with ANSI N510-1975 or a DOP test aerosol while operating the system /at a flow rate of cfm i 10%; and
f. AftereachcompletIcrpartialreplacement f a charcoal adsorber i/ig that the cleanup system s isfies the in place bank, by verify per.etration and /b ypass leakage testing accepta ce criteria of less than [*]% in J cordance E with ANSI N510-1975 for halogenated hydrocarbonfefrigerant test gas while operating a system at a flow rate ,of cfm c 10%.
0.05% value a cable M HEPA filter or charcoal adsorber fficiency of 99% is assumed, or 1% when a HEPA filter or charcoal adsorber fficiency
\ of 95% or le/s is assumed in the NRC staff's safety evaluation. se the value assumed for the charcoal adsorber efficiency if the value for he HEPAfilt/risdifferantfromthecharcoaladsorberefficiencyinthe RC staff's tafety j evaluation.)
) **Value,spplicable will be determined by the following equation:
P=[0.-E, when P equals the value to be used in the test requirement j (ji),EicefficiencyassumedintheSERformethyliodideremoval(%),
, and SF is the safety factor to account for charcoal degradation between
. / tests (5 for systems with heaters and 7 for systems without heaters). j OGTLE - UNIT 1 3/4 7-M k
PLANT SYSTEMS 3/4.7.7 PIPING PENETRATION AREA FILTRATION AND EXHAUST SYSTEM LIMITING CONDITION FOR OPERATION
~
3.7.7 Two independent ECCS Piping Penetration Area Filtration and Exhaust Systems shall be OPERABLE.
APPLICABILITY: MODES 1, 2, 3 and 4 ACTION:
With one ECCS Piping Penetration Area Filtration and Exhaust Systen inoperable, restore the incpirable system to OPERABLE status within 7 days or be in at least H0T STAtlDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the j folicwing 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .
SURVEILLANCE RE0VIREl1ENTS 4.7.7 Each ECCS Piping Penetration Area Filtration and Exhaust System shall be demonstrated OPERABLE:
a. At least once per 31 days on a STAGGERED TEST BASIS by initiating, tron the control room, flow through the HEPA filters and charcoal adsorbers and verifying that the system operates for at least 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months with the heater control circuit energized.
b. At least once per 18 months or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire or chemical release in any ventilation zone communi-cating with the system by:
1. Verifying that the cleanup system satisfies the in-place test-ing acceptance criteria of greater than or equal to 99.5%
filter efficiency while operating the system at a flow rate of 16,000 cfm +10% and performing the following tests:
(a) A visual inspection of the piping penetration area filtra-tion and exhaust system shall be made before each DOP test or activated carbon adsorber section leak test in accor-dance with Section 5 of ANSI H510-1980.
(b) An in-place 00P test for the HEPA filters shall be per-formed in accordance with Section 10 of ANSI N510-1930, (c) A charcoal adsorber section leak test with a ga!eous bal-ogenated hydrocarbon refrigerant shall be performed in accordance with Section 12 of ANSI N510-1980.
2. Verifying within 31 days af ter removal that a laboratory analysis of a representative carbon sample obtained in accordance with Section 13 of ANSI N510-1980 meets the laboratory testing cri-terion of greater than or equal to 99.8% efficiency when tested with methyl iodice at 50cc and 70% relative humidity.
VeG7;P UtNO 5/4 7-/7
~ x e
PLANT SYSTEMS (T
b SURVEILLANCE REQUIREMENTS (Continued)
3. Verifying a system flow rate of 16,000 cfm +10% during system operation when tested in accordance with Section 8 of ANSI N510-1980.
c. After every 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of charcoal adsorber operation by verifying within 31 days after removal that a laboratory analysis of a repre-sentative carbon sample obtained in accordance with Section 13 of ANSI N510-1980 meets the laboratory testing criterion of greater than or equal to 99.8% efficiency when tested with methyl iodide at 80cc and 70% relative humidity.
d. At least once per 18 months by:
1. Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks of less than 6 inches Water Gauge while operating the system at a flow rate of 16,000 cfm 110%.
2. Verifying that the system starts on a Containment Ventilation Isolation test signal.
3. Verifying that the heaters dissipate 8014 kw when tested in accordance with Section 14 of ANSI N510-1980.
e. After each complete or partial replacement of a HEPA filter bank by verifying that the HEPA filter banks remove greater than or equal to 99.5% of the DOP when they are tested in-place in accor-dance with Section 10 of ANSI N510-1980 while operating the system at a flow rate of 16,000 cfm 110%.
f. After each complete or partial replacement of a charcoal adsorber bank by verifying that the charcoal adsorbers remove greater than or equal to 99.5% of a halogenated hydrocarbon refrigerant test gas when they are tested in-place in accordance with Section 12 of ANSI N510-1980 while operating the system at a flow rate of 16,000 cfm 110%.
a 4 ves rLE- usi1.4 5/4 7-16 1
i-i PLANT SYSTEMS
3/4.7.h SNUBBERS LIMITING CONDITION FOR OPERATION
! 3.7.9 8All snubbers shall be OPERABLE. The only snubbers excluded from the
! requirements are those installed on nonsafety-related systems and then only
) if their failure of failure of the system on which they are installed would have no adverse effect on any safety-related system.
APPLICABILITY: MODES 1, 2, 3, and 4. MODES 5 and 6 for snubbers located on
systems required OPERABLE in those MODES.
ACTION: .-
With one or more snubbers inoperable on any system, within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months replace or re-store the inoperable snubber (s) to OPERABLE status and perform an engineering eval-uation per Specification 4.7.9g. on the attachea component or declare the attached system inoperable and follow the appropriate ACTION statement for that system.
SURVEILLANCE REOUIREMENTS l 4.7.# Each snubber shall be demonstrated OPERABLE by performance of the following augmented inservice inspection program in addition to the require- ,
ments of Specification 4.0.5.
, a. Inspection Types tva r #66 r a i>
O a d $" ta'= a ci<ic ti of the same design and manufacturer, irrespective of capacity.
66 r-i b. Visual Inspections l Snubbers are categorized as inaccessible or accessible during reactor i operation. Each of these groups (inaccessible and accessible) may j be inspected independently according to the schedule below. The t first inservice visual inspection of each type of snupber shall te performed after 4 months but within 10 months of concencing POWER OPERATION and shall include all snubbers. If all snubbers of each
! typegon any system 7are found OPERABLE during the first inservice l
visualinspection,thesecondinservicevisualinspectionf6fthat systemyshallbeperformedatthefirstrefuelingoctage. Otherwise, j subsequent visual inspections [of a given system] shall be performed i
in accordance with the following schedule:
} No. of Ineperable Snubbers of Each Type Subsequent Visual
[on Any System] per Inspection Period Inspection period * **
l 18 mor.ths 25%
l 0
! 1 12 months i 25%
! 2 6 months
25%
! 3,4 124 days i 25%
l 5,6,7 62 days.* 25%
8 or more 31 days 1 25%
s
The inspection interval for each type of snubber n a given system shall not i
be lengthened more than one step at a time unless a generic problem has been i
~ identified and corrected; in that event the inspection interval may be
! lengthened one step the first time and two stepsJ hereafter if no inoperabla snubbersofthattypearefoundfonthatsystemJ.
I **The provisions of Specification 4.0.2 are not applicable.
V0GTLE - UNIT 1 3/4 7-f6 / f -
- . - - - - _ - . - , - _ - . - - - . - . , - . - , _ - - . - -,,-,z---.
i PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) e I
c. Visual Inspection Acceptance Criteria '
l Visual inspections shall verify that: (1) there are no visible ;
indications of damage or impaired OPERABILITY, (2) attachments to ;
the foundation or supporting structure are functional, and (3) fasten- l ers for attachment of the snubber to the component and to the snubber anchorage are functional. Snubbers which appear inoperable as a l result of visual inspections may be determined OPERABLE for the j purpose of establishing the next visual inspection' interval, provided l 2 that: (1) the cause of the rejection is clearly established and l remedied for at particular spubber and for'other snubbers irrespec-tive of typ on that system ythat may be generically susceptible; and (2) the affected snubber is functionally tested in the as-found condition and determined OPERABLE per Specification 4.7.4f?8 All
! snubbers connected to an inoperable common fydraulic fluid reservoir shall be counted as inoperable snubbers. TFor those snubbers common j to more than one system, the OPERABILITY of such snubbers shall be j considered in assepsing the surveillance schedule for each of the j relatedsystems.f l d. Transient Event Inspection An inspection shall be performed of all snubbers attached to sections
! of systems that have experienced unexpected, potentially damaging 4 transients as determined from a review of operational data and a visual inspection of the systems within 6 months following such an event. In addition to satisfying the visual inspection acceptance criteria, freedom-of-motion of mechanical snubbers shall be verified using at least one of the following: (1) manually induced snubber j movement; or (2) evaluation of in place snubber piston setting; or
(3) stroking the mechanical snubber through its full range of travel.
i
e. Functional Tests During the first refueling shutdown and at least once per 18 months
thereafter during shutdown, a representative sample of snubbers of j ;, each type shall be tested using one of the following sample plans.
The sample plan for each type shall be selected prior to the test period and cannot be changed during the test period. The NRC Regional Administrator shall be notified in writing of the sample plan selected for each snubber type prior to the test period or the sample plan used in the prior test period shall be implemented:
1) At least 10% of the total of each type of snubber shall be
! functionally tested either in place or in a bench test. For each snubber of a type that does not meet the functional test acceptancecriteriaofSpecification4.7.(f.,anadditional10%
of that type of snubber shall be functionkily tested until no more failures are found or until all snubmers of that type have 1
l been functionally tested; or j 1
i V0GTLE - UNIT 1 3/4 7-41
". hpeannd f
PLANT SYSTEMS
. SURVEILLANCE REQUIREMENTS (Continued)
e. Functional Tests (Continued)
A
2) A representative sample of each type of snubber shall be func-tionally tested in accordance with Figu ' 4. 7-1. "C" is the total number of snubbers of a type foun not meeting the accept-ance requirements of Specification 4.7. f. The cumulative number of snubbers of a type tested is denoted by "N". At the end of each day's testing, the new values of "N" and "C" (pre-i vious day's total plus current day's increments) shall be
+
plotted on Figure 4.7-1. If at any time the point plotted
falls in the " Reject" region, all snubb'ers of that type shall
be functionally tested. If at any time the point plotted falls i in the " Accept" region, testing of snubbers of that type may be i terminated. When the point plotted lies in the " Continue Testing" region, additional snubbers of that type shall be tested until the point falls in the " Accept" region or the
" Reject" region, or all the snubbers of that type have been tested; or l
3) An initial representative sample of 55 snubbers shall be func-
tionally tested. For each snubber type which does not meet the functional test acceptance criteria, another sample of at least one-half the size of the initial sample shall be tested until j
O the total number tested is equal to the initial sample size multiplied by the factor,1 + C/2, where "C" .is the number of J snubbers found which do not meet the functional test acceptance
! criteria. The results from this sample plan shall be plotted
using an " Accept" line which follows the equation N = 55(1 l + C/2). Each snubber point should be plotted as soon as the
! snubber is tested. If the point plotted falls on or below the
" Accept" line, testing of that type of snubber may be terminated.
If the point plotted falls above the " Accept" line, testing
must continue until the point falls in the " Accept" region or
! all the snubbers of that type have been tested.
i Testing equipment failure during functional testing may invalidate l that day's testing and allow that day's testing to resume anew at a later time provided all snubbers tested with the failed equipment during the day of equipment failure are ratested. The representative sample selected for the functional test sample plans shall be randomly selected
, from the snubbers of each type and reviewed before beginning the testing.
i- The review shall ensure, as far as practicable, that they are represen-i tative of the various configurations, operating environments, range of size, and capacity of snubbers of each type. Snubbers placed in the same location as snubbers which failed the previous functional test shall be ratested at the time of the next functional test but shall not be included in the sample plan. If during the functional testing, .
additional sampling is required due to failure of only one type of I O snubber, the functional test results shall be reviewed at that time to determine if additional samples should be limited to the type of snubber which has failed the functional testing.
\
y .
V0GTLE - UNIT 1 3/4 7-it i
...e-_- _ _ .- -- , . .-._. . - _ _ _.--- --_ _ - - _ -,.....-.---,-._ .- _ _,,- ... .
PLANT' SYSTEMS
' l l
SURVEILLANCE REQUIREMENTS (Continued) l if
f. Functional Test Acceptance Criteria I
The snubber functional test shall verify that:
1) Activation (restraining action) is achieved within the i specified range in both tension and compression;
2) Snubber bleed, or release rate where required, is present in both tension and compression, within the specified range;
3) For mechanical snubbers, the force required to initiate or maintain motion of the snubber is within the specified range
! in both directions of travel; and
4) For snubbers specifically required not to displace under
! continuous load, the ability of the snubber to withstand load without displacement.
4 Testing methods may be used to measure parameters indirectly or parameters other than those specified if those results can be correlated to the specified parameters through established methods.
1 g. Functional Test Failure Analysis An engineering evaluation shall be made of each failure to meet the-functional test acceptance criteria to determine the cause of the failure. The results of this evaluation shall be used, if applicable,
in selecting snubbers to be tested in an effort to determine the OPERABILITY of other snubbers irrespective of type which may be subject to the same failure mode.
For the snubbers found inoperable, an engineering evaluation shall be performed on the components to which the inoperable snubbers are attached. The purpose of this engineering evaluation shall be to determine if the components to which the inoperable snubbers are attached were adversely affected by the inoperability of the snubbers
~
in order to ensure that the component remains capable of meeting the j
designed service.
If any snubber selected for functional testing either fails to icck up or fails to move, i.e., fro 2.en-in place, the cause will be evaluated and, if caused by manufacturer or design deficiency, all
snubbers of the same type subject to the same defect shall be func-
! tionally tested. This testing requirement shall be independent of j the..requirementsstatedinSpecification4.7.ke.forsnubbersnot c>esting the functional test acceptance criteriah 8
i
.c O -
22.
V0GTLE - UNIT 1 3/4 7- H
PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
h. Functional Testino of Reoaired and Replaced Snubbers Snubbers which fail the visual inspection or the functional test acceptance criteria shall be repaired or replaced. Replacement snubbers and snubbers which have repairs whith might affect the functional test results shall be tested to meet the functional test criteria before installation in the unit. Mechanical snubbers shall have met the acceptance criteria subsequent to their most recent service, and the freedom-of-motion test must have been performed witnin 12 months before being installed in the unit.
i. Snubber Service Life Program The service life of hydraulic and mechanical snubbers shall be monitored to ensure that the service life is not exceeded between surveillance inspections. The maximum expected service life for various seals, springs, and other critical parts shall be deter-mined and established based on engineering information and shall be extended or shortened based on monitored test results and failure history. Critical parts shall be replaced so that the maximum service life will not be exceeded during a period when the snubber is required to be OPERABLE. The parts replacements shall be docu-mented and the documentation shall be retained in accordance with j Specification 6.10.3. '
~
i l
O 23 V0GTLE - UNIT 1 3/4 7-E+
O 1 l
10 ,
l 9
8:
1 REJECT 1
[
6 1
O G b**
4 ,7 , ,
CONTINUE TESTIN G V go %f 3
2- , *}g@
G+ ACCEPT i
1 .f O 10 20 30 40 50 60 70 80 90 100 N
O FIGURE 4.7-1 I
SAMPLE PLAtt 2) FOR SilUSBER FUtiCTI0tlAL TEST V0GTLE-U. LIT 1 3/4 7-24
J PLANT SYSTEMS 9 '
3/4.7.46 SEALED SOURCE CONTAMINATION J
LIMITING CONDITION FOR OPERATION 9
3.7.10 Each sealed source containing radioactive material either in excess of 100 microCuries of beta and/or gamma emitting material or 5 microCuries of alpha emitting material shall be free of greater than or equal to 0.005' microcurie of rerovable contamination.
APPLICABILITY: At all times. .
ACTION:
a. With a sealed source having remova'le u contamination in excess of the above limits, immediately withdraw the sealed source from use and either:
. 1. Decontaminate and repair the sealed source, or 4
4 2. Dispose of the sealed source in.accordance with Commission
> Regulations.
b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS
< 9
! 4.7.40.1 Test Requirements - Each sealed source shall be tested for leakage and/or contamination by:
a. The licensee, or 4
b. Other persons specifically authorized by the Commission or an Agreement State.
The test method shall have a detection sensitivity of at least 0.005 microcurie per test sample.
/
4.7.10.2 Test Frequencies - Each category of sealed sources (excluding
startup sources and fission detectors previously subjected to core flux) shall be tested'at the frequency described below,
a. Sources in use - At least onc.e per 6 months for all sealed sources containing radioactive materials:
2
1) With a half-life greater than 30 days (excluding Hydrogen 3),
and
2) In any form other than gas.
{}
VCGTLE - UNIT 1 3/4 7-36
3 PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
b. Stored sources not in use - Each sealed source and fission detector shall be tested prior to use or transfer to another licensee unless tested within the previous 6 months. Sealed sources and fission detectors transferred without a certificate indicating.the last test date shall be tested prior to being placed into use; and
c. Startup sources and fission detectors - Each sealed startup source and fission detector shall be tested within 31 days prior to being subjected to core flux or installed in the co,re and following repair or maintenance to the source.
2 '.10.2 Report: - ? 7: pert ; hell b: p pa. d ar.d ,oLm iisu Lv w.. Cwmm . a . . vi.
n :n :nnu:1 b :i: ::: led :: rc: r #iccier detect:- 12 E; t :t: :v:21 th: pr ;:n:: :P grc:ter th n :r :; :' t: 0.005 -i:r:Curi: :f :::v:b!:
-t: i n:tfr .
O
=.
e O_-
N ,
IG V0GTLE - UNIT 1 3/4 7-21 7
~ -r-- -- .my-g p - , ,,, _ _
4
\ \
PLANT SYSTEMS
)
3/4.7.11 FIRE SUPPRESSION SYSTEMS ,
FIRE SUPPRESSION WATER SYSTEM ,
LIMITING CONDITION FOR OPERATION N
3.7.11.1 The Fire Suppression Water System shall be OPERABLE with
a. At least [two] fire suppression pumps, each withacity a cap of /
[2500] gpm, with their discharge aligned to the fire / suppression '
heade r ,-s_ .
b. Separate ter supplies, each with a minimum cont.ained volume of gallon,s, and
c. An OPERABLE low path capable of taking suc ion from the tank and the tank and transferring the wat'er through distribution piping witn OPERABLE sectionalizing con)foi or isolation valves to the yard hydrant curb valves, the last/ valve ahead of the water flow alarm device on each' sprinkler or hose standpipe, and the last valve ahead of the deluge valve on each Q41uge or Spray System required to be OPERABLE per Specifications 3. ,.11.2, 3.7.11.5, and 3.7.11.6.
APPLICABILITY: At all times.
N
, ACTION:
a. With one pump and/or one ater s ply inoperable, restore the inoper-able equipment to OPER LE status ythin 7 days or provide an alter-nate backup pump or s pply. The pr isions of Specifications 3.0.3
, and3.0.4arenotaplicable.
/
1 b. With the Fire Suppression Water System otherwise inoperable, establish a backup Fire Suppression Water System within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .
/
\
\ '
,/ i
/ s
/
/
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! V0GTLE - UNIT 1 3/4 7-28
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i 3 PLANT SYSTEMS
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v \ SURVEILLANCE REQUIREMENTS ,
's 4.7.'11.1.1 The Fire Suppression Water System shall be demonstrated OPER,ABLE:
\
a.s At least once per 7 days by verifying the contained water supply
/
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b. At least once per 31 days on a STAGGERED TEST BASIS by, starting each electric motor-driven pump and operating it for at least 15 minutes on recirculation flow, /
c. At leas'ts once per 31 days by verifying that'each valve (manual, power-operated,'Nor automatic) in the flow path is in its correct position, N /
d. [At least orice per 6 months by performance of a system flush,]
\ /
e. At least once p'er 12 months by cycling eacb testable valve in the flow path throughsat least one complete, cycle of full travel,
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f. At least once per 18xmonths by perfopning a system functional test which includes simulated automatic actuation of the system througho6t its operating sequence, and: /
/
1) Verifying that each a tomat'.ic valve in the flow path actuates to its correct positioh /
2) Verifying that each pomp develops at least [2500] gpm at a systemheadof[2507 feet,
/
3) Cyclingeachvalv$intheflowpaththatisnottestableduring plant operatiov through at leas one complete cycle of full travel, and ,'
4) Verifying that each fire suppression \ pump starts [ sequentially]
to maintain the Fire Suppression WaterNSystem pressure greater than orfequal to psig. N
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g. At least o'nce per 3 years by performing a flow test of the system in accordancu with Chapter 5, Section 11 of the Fire' Protection Handbook, 14th Edition, published by the National Fire Protection Association.
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/ V0GTLE - UNIT 1 3/4 7-29 \ '
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"t& e"x PLANT SYSTEMS- ,
SURVEILLANCE REQUIREMENTS (Continued) ',
4./ 11.1.2 The fire pump diesel engine shall be demonstrated OPERABLE:
a At least once per 31 days by verifying:
) The fuel storage tank contains at least gallons of fuel, and
2) The diesel starts from ambient conditions and operates for at
. east 30 minutes on recirculation flow.. ,
b. At leas once per 92 days by verifying that a sample of diesel fuel from the heel storage ' tank, obtained in accor#nce with ASTM-D270-1975 is within t acceptable limits specified in able 1 of ASTM D975-1977 when checked or viscosity and water and s iment; and
c. At least once pe g18 months, during shu own, by subjecting the diesel to an inspection ig accordance with pr cedures prepared in conjunction with its manufacture ('s s
recommendati s for the class of service.
i 4.7.11.1.3 The fire pump diesel \ tarting 2 volt battery bank and charger shall be demonstrated OPERABLE:
a. At least once per 7 days by v rifying that:
1) The electrolyte level of ch battery is above the plates, and
2) The overall batter voltage greater than or equal to 24 volts.
! b. At least once per 9 days by verifyi that the specific gravity is-1 appropriate for co inued service of t e battery, and i c. At least once p 18 months by verifying at:
1 1) The bat eries, cell plates, and battery acks show no visual <
indic ion of physical damage or abnorma deterioration, and
! 2) Th battery-to-battery and terminal connecti ns are clean,
tirght, free of corrosion, and coated with ant) orrosion material.
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V0GTLE - UNIT 1 3/4 7-30 l 4 s
u+ eda Na -
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!. PLANT SYSTEMS SPRAY AND/OR SPRINKLER SYSTEMS ..
f' LI TING CONDITION FOR OPERATION
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3.7.11. The follosing Spray and/or Sprinkler Systems shall be OPERABLE:
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a. lantdependent-tobelistedbynameandlocation.][
b.
t j c.
APPLICABILITY: When ver equipment protected by the Spr / Sprinkler System is
< required to be OPERAB .
~
i ACTION:
a. With one or more of the above require Spray and/or Sprinkler Systems
} inoperable, within hour establish continuous fire watch with backupfiresuppres(sinequipment r those areas in which redundant j systems or components uld be d aged; for other areas, establish an.
i hourly fire watch patrol.
b. The provisions of Specifica ns 3.0.3 and 3.0.4 are not applicable.
r SURVEILLANCE REQUIREMENTS i
i_ 4.7.11.2 Each of the above quired Spray and Sprinkler Systems shall be 1 demonstrated OPERABLE:
1 i a. At least once er 31 days by verifying tha each valve (manual, power-operated, or futomatic) in the flow path is n its correct position,
! b. At least once per 12 months by cycling each te able valve in the j flow patl1I through at least one complete cycle of ull travel,
c. At least once per 18 months:
i 11j/ By performing a system functional test which inci es simulated
j / automatic actuation of the system, and:
/ Verifying that the automatic valves in the flow path
/ a) i f actuate to their correct positions on a \ test >
> / signal, and i /
f b) Cycling each valve in the flow path that is not testabl 1 during plant operation through at least one complete cyc g j of full travel.
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t j V0GTLE - UNIT 1 3/4 7-31 1
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PLANT SYSTEMS 1
SURVEILLANCE REQUIREMENTS (Continued) - '
2) By a visual inspection of the dry pipe spray and sprinkler /
headers to verify their integrity; and
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3) By a visual inspection of each nozzle's spray area, to verify the spray pattern is not obstructed.
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d. At\least once per 3 years by performing an air flow-test through i
eachsopen head spray / sprinkler header and verifying each open head sprayqprinkler nozzle is unobstructed. "
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PLANT SYSTEMS CO., SYSTEMS I
LIMITING CONDITION FOR OPERATION *
3. 11.3 The following High Pressure and Low Pressure CO j OPE BLE: 2 Systems hall be j
a. [ Plant dependent - to be listed by name and location b.
i c.
APPLICABILITY Whenever equipment protected by the CO2 ystems is required to i be OPERABLE.
1 l ACTION:
a. With one or re of the above requiredf 2 f0Systems inoperable, l within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months tablish a continuousj fire watch with backup fire
suppression equipment for those areas in which redundant systems or components could fire watch patrol, damaged; for oAher areas, establish an hourly
b. The provisions of Spe icati s 3.0.3 and 3.0.4 are not applicable.
1 -
SURVEILLANCE REQUIREMENTS N/
/
,X 4.7.11.3.1 Each of the above quired C0 Systems shall be demonstrated OPERABLE operated, or at least once automatic) in per 3 days by verifytog that each valve (manual, power-e flow path is in ts correct position.
j 4.7.11.3.2 Each of the ove required Low Pressu' CO Systems shall be demonstrated OPERABLE:
1 1 a. At least o
! be greater,Mcethan per 7 and days by verifying the CD storage tank level to
, pressure to be greate than psig, and '
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b. At leas / per 18 months by verifying:
t once 1
1) The system, including valves, associated ventil ion system 1
i fire dampers, and fire door release mechanisms, a\tuates manually and automatically upon receipt of a simulated actu fon signal,
and
2) Flow from each nozzle during a " Puff Test." l V0GTLE - UNIT 1 3/4 7-33 4
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PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
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4.h11.3.3 Each of the above required High Pressure CO2 Systems shall be demohstrated OPERABLE:
i
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a.\ At least once per 6 months by verifying the CO2 storage tank '
weight
ito be at least 90% of full charge weight, and ,.
b. At'least once per 18 months by: ,
1) Verifying the system, including associated ventilation system fire dampers and fire door release mechanisms, actuates manually i andtautomatically upon receipt of a simulated actuation signal, and '. ,
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2) Performance of a flow test through headers and nozzles to I assure no blockage. .
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, V0GTLE - UNIT 1 3/4 7-34 l
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PLANT SfSTEMS .
\HALON SYSTEMS LINTING CONDITION FOR OPERATION 3.7.11. The following Halon Systems shall be OPERABLE:
a. Plant dependent - to be listed by name and location.]
b.
c.
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APPLICABILITY: Whenever equipment protected by the Halon Sy tem is required to be OPERASLE. .
ACTION:
a. With one or m e of the above required Halo Systems inoperable, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months e tablish a continuous fire, watch with backup fire suppression equi' ent for those areas in,which redundant systems or components could b damaged; for other a'reas, establish an hourly fire watch patrol.
b. The provisions of Spec ications 3<0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS s ,
O x 4.7.'11.4 Each of the above required al Systems shall be demonstrated OPERABLE:
a. At least once per 31 d s by veri ng that each valve (manual, power-operated, or adomatic) in th flow path is in its correct position,
b. At least once per 6 months by verifying lon storage tank weight to be at leastf95% of full charge weight (or level) and pressure to be at least 90% 'f full charge pressure, and
c. At least once per 18 months by:
1) Ver'ifying the system, including associated V tilation System ire dampers and fire door release mechanisms ctuates manually and automatically, upon receipt of a simulated tuation signal, and \
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2) Performance of a flow test through headers and nozzlis to assure no blockage.
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V0GTLE - UNIT 1 3/4 7-35
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PLANT SYSTEMS q
/ FIRE HOSE STATIONS LIMITING CONDITION FOR OPERATION
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3. 7.1'Is 5 The fire hose stations given in Table 3.7-4 shall be OPERABLE.
APPLICABILITY: Whenever equipment in the areas protected by the fir hose stations is required to be OPERABLE.
ACTION:
a. With one or more of the fire hose stations given i Table 3.7-4 inoperakle,providegatedwye(s)onthenearestOffRABLEhose station (t . One outlet of the wye shall be conpected toThe the second standard 1 gth of hose provided for the hose ptation.
outlet of t wye shall be connected to a lepgth of hose sufficient to provide co rage for the area left unprofected by the inoperable hose station. ere it can be demonstrat that the physical routing of the f te hose would result i a recognizable hazard to operating technictgns, plant equipment or the hose itself, the fire hose shall be storeAin a roll at th outlet of the OPERABLE hose station. Signs shall\be mounted ab e the gated wye (s) to identify the proper hose to use N The abov ACTION requirement shall be accomplished within I hour if th inoperable fire hose is the p primary means of fire sup) ess dn; otherwise route the additional y) hose within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .
b. The provisions of Specific i s 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REOUIREMENTS 4.7.11.5 Eachofthefirehos!stationsgiveninTable3.7-4shallbe de.m enstrated OPERABLE:
a.
/
At least once per 31 days, by a visual i spection of the fire hose stations accessible during plant operatioh to assure all required it the station, equipmentisp
b. At least once per 18 months, by:
1) Vis, 1 inspection of the stations not accetgible during plant operations to assure all required equipment is at the station, k and
2) j emoving the hose for inspection and re-rackin 3)/ Inspecting all gaskets and replacing any degraded \ askets in the couplings,
c. /At least once per 3 years, by:
p
,' 1) Partially opening each hose station valve to verify val'Ve
/ OPERABILITY and no flow blockage, and
/ 2) Conducting a hose hydrostatic test at a pressure of 150 ps or p / atleast50psigabovemaximumfiremainoperatingpressure,\
d / whichever is greater.
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V0GTLE - UNIT 1 3/4 7-36 s
TABLE 3.7-4 FIRE HOSE STATIONS ELEVATION HOSE RACK NWMBER LOC TION *
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J(quipment.ist e all fire hose stations required to ensure the OPERABILITY o
/ V0GTLE - UNIT 1 3/4 7-37
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PLANT SYSTEMS A
U YARD FIRE HYDRANTS AND HYDRANT HOSE HOUSES MITING CONDITION FOR OPERATION
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3.7.11g6 The yard fire hydrants and associated hydrant hose housesingiven/
Table 3.7-5 shall be OPERABLE.
APPLICABIL TY: Whenever equipment in the areas protected by the yard fire hydrants is required to be OPERABLE.
/
ACTION:
With onexo r more of the yard fire hydrants or as
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iated hydrant a.
hose houses given in Table 3.7-5 inoperable, wi n 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months have sufficient' additional lengths of 2 1/2 inch dipmeter hose located in
~
an adjacent 0RERABLE hydrant hose house to pr vide service to the unprotected area (s) if the inoperable fire drant or associated hydrant hose ho'ose is the primary means of fire suporession; otherwise, provide the additional hose w hin 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .
b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
N SURVEILLANCE REOUIREMENTS 's 4.7.11.6 Each of the yard fire hydra nd associated hydrant hose houses given in Table 3.7-5 shall be demonstr ed OPERABLE:
a. At least once per 31 day , by v s<ual inspection of the hydrant hose house to assure all req red equi nt is at the hose house, p
b. At least once per 6 months (once dur ng March, April, or May and once during Septembyf, October, or No mber), by visually inspecting each yard fire hydrant and verifying th t the hydrant barrel is dry and that the hy/drant is not damaged, and
c. At least once er 12 months by:
1) ng a hose hydrostatic test at a ressure of 150 psig or Conduc)t50psigabovemaximumfiremainoperatingpressure, at leas whicfiever is greater,
2) Inspecting all the gaskets and replacing any degraded gaskets in the couplings, and
3) Performing a flow check of each hydrant to verify ts OPERABILITY.
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l V0GTLE - UNIT 1 3/4 7-38 sI
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} TABLE 3.7-5
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YARD FIRE HYDRANTS AND ASSOCIATED HYDRANT HOSE HOUSES LOCATION
HYDR 4:1T NUMBER
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/PERABILITYofsafety-relatedequipment.
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PL1NT SYSTEMS n
(v) \ 3/4.7.12 FIRE RATED ASSEM0 LIES 3
L1HITING CONDITICN 10R OPERATION .
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3.7.1d\All fire rated assemblies (walls, flocr/ ceilings, cable tray ef[ closures, and cthe'r, fire barriers) separating safety-related fire areas or sepprating portions of redundant systems important to safe shutcown within a ffre area and all seating devices in fire rated assembly penetrations (firej eoors, fire windows, firesdaepersi cable, piping, ar,d wentilation duct penet, ration seals shall be OPERARLE.
AFPLICABILITY: At all times.
\ '
' ACTION:
a. With one cr m' e of the aDove required fir ated assemblies and/or sealing cevicesginoperable, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months ther estaolish a continuous fire watch cn at leest one si e of the affecta.d assembly, or verify the OPERASILITY of fire dete ors on at least one side of the inoperable asse't(ly and establish an hourly fire watch patrol,
~
b. TneprovisionsofSpect{ cations 3 .3 and 3.0.4 are not applicable.
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SURVEILLANCE REQUIREMENTS 4.7.12.1 At least once per 18 mon he above required fire rated assemblies ard penetration sealing devices sh 1 be v rified OPERABLE by performing a
' l visual inspection of:
a. Tf.e expesed surf acer of each fire rated assembly, 3
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b. Each fite window / fire daner and associated hardware, and
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c. At least 10% of each type of sealed penetration. If apparent changes in appearance or abnormal degradatio'n6 are found, a visual irrspection/of an acditional 10% of each type o'fssealed penetration snall be,eace. This inspection process shall continue until a 10% '
sample ,with no apparent changes in appearance or ' abnormal degradation is found. Samples shall ce selected such that each enetration will be inspected every 15 years.
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- 1 PLANT SYSTEMS
\SURVEItLANCEREQUIREMENTS(Continued) '
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4. D12. 2 Each of the above required fire doors shall be verified OPERABL by .
inspsqting the automatic hold-open, release and closing mechanism and la ches
. at least once per 6 months, and by verifying: i i a. The OPERABILITY of the fire door supervision system for each electricallysupervisedfiredoorbyperformingaTRIPfUATING d
l DEVICE OPERATIONAL TEST at least once per 31 days, I
\
Thateachlockedclosedfiredoorisclosedatleas/onceper J b.
j 7 days,
! c. 'That doors \yith automatic hold-open and releasy mechanisms are free l of obstructiqns at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , a d a functional test is performed it least on e per 18 months, a
] h l c. That each unlockM fire doo. without ele rical supervision _ ;
is closed at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . ;
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FLANT SYSTEMS to (3 j 3/4.7.H AREA TEMPERATURE MON!T0 RING LIMITING CONDITJON F0k OPERATION - ,
do 3
3. 7. B The te7.perature of each area shown in Table 3.7-4 shall ro? De exceedec for more thar; 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or by mere than 20*F.
APPLICABILITY: 'Whenever the equipment in an affected area is required to be OPERASLE.
1 ACTION: g
- a. With one or adre areas ex:erding the terrperature %mit(s) shown in Table 3.7-3 for trere than 8 roars, prepare and submit to the Commission within 30 days, pursuant to Specificetion 6.9.2, a Special Report that proviacs a record of the ctculative tine and the amount by wnich the terperature in the affected arca(s) exceeded the limit (s) and an enalysis to deo:nstrate the continued OPERABILITY of the affected equipment. The crovisions of Specifi-cations 3.0.3 and 3.0.4 are not applicable.
d
b. With one or/rore areas exceeding the tegetature linit(s) shown in Table 3.7-f by more than 30*F, prepare. and submit a Special Report O
b as required by ACTION a. above and within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months either restore the area (s) to within the terperature limit (s) or det!are the equip-ment in the affected area (s) inoperaole.
SURVEILLANCE RE0VIREMENTS _
/0
1. 7. H The temperature in each of the arets Ahcwn in Table 3.7-f shaU be determined to t,e within its limit at least coce per 32 hours3.703704e-4 days 0.00889 hours 5.291005e-5 weeks 1.2176e-5 months .
27 V0GTLE - UNIT I 3/4 7-4e l
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TG5LE 3.7-6 AREA TEMPERATURE MONITORING l
MEA TENPERATURE LIMfT.(*E) i.
2 3.
4.
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i 3/4 7.11 ENGINEEREO SAFETY FEATURES (ESF) ROCM COOLER add SAFETY-RELATED_
CHILLER SYSTEM I
i LIMITING CONDITION FOR OPERATION ,
i e 3
j 3.7.11 Two independent ESF room cooler and safety-related chiller trains shall be OPERABLE.
A?PLICASILITY: MODES 1. 2, 3 and 4. i
ACTG1,:
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L'th only one ESF roon cooler and safety-related chiller train GPERABLE, restore two trains to DERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months cr be in at least riot STANDBY within tne next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SH0100VN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .
SURVEILLANCEP.EQUIqEMENTS 1 - .
. .-~
4.7.li Two EST ,um cooler ar.d safety-related chiller trains shall be demonstrated j OPERA 2/ E:
l' a. At least 0.ce oer 31 days by verif ir.3 f that each valve (manual, power operated or au'tomatic? servicing safety-related equipment that is not l locked, sealed or etherwise secured in position, is in its correct
- posf tion.
1 b At least once per '.8 months by verify 1rg that each automatic valve j servidna safety-rolated equipment ar.teates to its correct position
cn a str ety 1d oction test signal.
f c. At least ov.e per la months by verf fying that each ESF room cooler fan i ard each train of saf0ty-related chfilers (pro and chiller) start
, Outr.u:atically on a safety injectioc, test sigW..
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3/4.7.12 REACTOR C00LANT PUMP THERMAL BARRIER COOLING WATER ISOLATION 1
i i _ LIMIT:NG C0?dITION FOR OPERATION j 3.7.12 The reactor coolant pump thermal barrier isolation function shall be j OPERABLE.
~
APPLICABILITY: Modes 1, 2, 3, and 4.
ACTION:
' With the reactor conlant pump thermal barrier isolation function inoperable, restore to OPERABLE status within 7 days or be in at least HOT STANDBY '
l within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .
i i SURVEILLANCE RE0DIREMENTS 1
f 4.7.12 The reactor coolant pump thermal barrier isolation functica shall bo <
demonstrated OPERABLE by:
l a. Verifying that valve HV-2041 autom'atically closes on thermal barrier j outlet high header pressure and high header flow test signals at least i once per 36 months on a STAGGERED TEST BASIS.
J
! b. Verifying that valves HV-19051,19053,19055, ar.d 19057 automatically close Jn thermal barrier outlet high flow (FT-19052,19054,19056, and 19058) test signals at least once per 18 months.
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\,,j JUSTIFICATIONS FOR DEVIATIONS FROM STS -
SECTION 3/4.7 t
3/4.7:
Instrument tag numbers were added as appropriate to assist the operator in
' ensuring compliance with these Technical Specifications.
3.7.1.1:
4 The deletion of "of an unisolated reactor coolant 1ccp" in the LC0 reflects the fact that RCS loop isolation valves are not included in the VEGP design
. and the deletion of STS Table 3.7-2 is based on the fact that N-1 operation will not be permitted.
Similarly, Action Statement b was deleted to reflect the intent of the plant i not to operate in N-1 loop operation.
The revised action statements also allow operation in Maces 2 and 3 provided ,
at least one safety valve is OPERABLE without resetting reactor trip i
setpoints. Resetting trip setpoints in these modes is not necessary since the low power setpoint will be active. Also, forcing the plant to Cold Shutdown is unnecessary since the specification is applicable in Modes 1, 2, O and 3 only. The revised specification correctly replaces Cold Shutdown with Hot Shutdown.
j 1 !
This revision to Specification 3/4.7.1.1 1s sponsored to the NRC by the Westinghouse Ovners Group in letters OG-152 (June 1985) and 0G-161 (October ,
1985). ;
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] 4.7.1.2.1.4.4:
This surveillance requirement was revised for clarf fication. The words :
I i " automatic control" imply that the AFWS is controlled automatically as is i the main feedwater s) stem. When the AFV pump controls are in the automati.
position, the pumps will auto-star en on initiation signal. Once the pumps have started, AFW flow is maf ually rogulated to maintain the desired stcae t generator water level. Se4 tubsection 10.4 94 of the VEGP FSAR.
s i
4.7.1.2.2:
t
This surveillance requirement was delettrd on the bacis that the aux 111ary feodwater system is used for startup and shutdown operation. See paragraph
< 10.4.9.2.3 of the VEGP FSAR. ,
I i()
l l
3/4.7.1.3:
See subsection 9.2.6 of the VEGP FSAR for a description of the condensate storage tanks.
4.7.1.4: Table 4.7-1; Footnote *:
See the justification for revisions to Definition 1.10.
3/4.7.1.5:
This specification has been revised to reflect plant design. The VEGP is equipped with redundant main steam isolation valves and associated bypass isolation valves in each steam line. In the case where both MSIVs in a single steam line are inoperable, an outage time of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months (as opposed to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ) was chosen on the strength of an analysis which indicated that the increase in probability of core damage due to a main steam line break was negligible (1.01 X 10- 5 for operable MSIVs and 1.02 X 10 - 8 for two inoperable MSIVs in a single steam line for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months per year). In the case where one MSIV is inoperable for 7 days in a year, the probability of core damage in the event of a main steam line break was found to be 1.02 X 10 - * . For comparison purposes, the probability of core damage in the case of both MSIVs in a single line inoperable for a full year was found to be
2. 51 X 10 ~ * . The action statement for Mode 1 was revised to reduce power to less than or equal to S percent of rated thermal power within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to -
allow the operator to go to the lesser mode of startup which is addressed in the action associated with Modes 2 or 3. See paragraph 10.3.2.2.4 of the VEGP FSAR for a description of the main steam isolation valves.
3/4.7.3:
The word " loop" was replaced with " train" to be consistent with VEGP specific terminology used in the FSAR. Also, each train includes three 50-percent-capacity pumps with only two of the three pumps required for an operable train.
Thw surveillance requirements were modified to reflect the fact that the design of the component cooling water system does not include any automatic valees servicing safety-related equipment. See subsection 9.2.2 of the VEGP W R.
3/4.7.4; This soecification was revised to 'ifer to " trains" instead of loops to be consistent with terminology used in the FSAR and to reflect that the nuclear service cooling water (NSCW) system includes three 50 percent-capacity pumps per train. Therefore, only two pumps are required for an operable train.
See subsection 9.2.1 of the VEGP FSAR.
O
Ph*
3/4.7.5:
The terminology was revised to reflect the VEGP design, which consists of two NSCW tower basins, the combined inventory of which comprises the VEGP ultimate heat sink. Also, the water level is specified in plant elevation which does not necessarily correspond to feet above mean sea level. See subsection 9.2.5 of the VEGP FSAR.
The specified maximum water temperature is not an average temperature but rather that which is read in the common pump discharge header, and thus is a true measure of the temperature of the water being supplied to the cooled components.
Operability as a function of available NSCW tower fans and ambient wet bulb temperature is based on calculations which demonstrate that, given the specified ambient wet bulb terperature limits, the specified number of fan and spray cells are the minimum required to ensure sufficient cooling capacity for safe shutdown. This allows operational flexibility such that, should tower maintenance be required (fill replacement, fan maintenance, etc.) operation could continue.
3/4.7.6 (STS):
This specification was deleted in its entirety since the VEGP design includes adequate passive flood control protection in conformance with O Regulatory Guide 1.59. See paragraph 1.9.59.2 of the VEGP FSAR.
3.7.6:
Applicability and Actions:
Our draft involves two deviations from the STS: (1) The applicability for Modes 5 and 6 was revised such that the control room emergency filtration system is required to be OPERABLE only during movement of irradiated fuel or movement of loads over irradiated fuel; (2) The action for Modes 1, 2, 3, and 4 was revised to allow the option of initiating and maintaining the other flitration train in the emergency mode. Action b under Modes 5 and 6 was also revised to require suspension of operations involving movement of irradiated fuel and loads over irradiated fuel, to be consistent with the applicability as revised.
The justification for the revision to the action for Modes 1, 2, 3 and 4 has its basis in the fact that if the OPERABLE filtration train is placed in the emergency mode and maintained in that mode, the operators will be protected in the event of an accident or until the inoperable train can be restored to operable status. .
The justification for the revision to the applicability during Modes 5 and 6 has its basis in the fact that in these modes the filtration system is only necessary to protect the operators against a release resulting from a fuel
'j '
\
( handling accident. Therefore, as long as there is no movement of irradiated fuel or loads over irradiated fuel, the system should not be required to be .
operable. Similarly, Action b was revised to be consistent with the revised applicability.
Tr:e modifications to the applicability statement have been approved and 4
implemented in the Farley Technical Specifications.
4.7.6:
These surveillance requirements have been revised to incorporate plant-specific features and improvements which have already been approved by the NRC and implemented in the Farley Nuclear Plant Technical Specifications. See subsecticn 6.4.2 of the VEGP FSAR for a description of the control room emergency filtration system.
i 4.7.6.a (STS):
Surveillance Requirement 4.7.6.a was deleted from this specification and moved to 4.7.12. As discussed in the bases, the operability requirement ensures that the ambient air temperature does not exceed the allowable temperature for continuous-duty ratirg for the equipment and instrumentation cooled by this system. Because t.he control room emergency filtration system is normally not in operation it does not ensure tha:, the ambient air O temperature does not exceed the allowable. In adcition, as written, if the control room temperature exceeded the allowable, the emergency system must be declared inoperable, while it may be the means to reinstate proper temperature.
4. 7 . 6_. a :
Filtration unit heaters are cedgrad to maintain the moisture content of the air en.ering the charcoal adsorption oed t, clew 70 percent. The heaters are enabled by a contact in the fan starting circuit and heater control is governed by a moisture sensor in the heater outle . duct. Thers are no indicating lights associated with the heater.
The surveillance requirement states that the heaters must remain energized during the entire surveillance test period. With no indication of heater status this has necessitated the addition of steps in the operations surveillance procedures which require the measurement of voltage to the heaters. This is impractical from a safety standpoint and also from the standpoint that the heater is actually under control of the moisture sensor and will cycle on and off during the test. If the heater is cperating correctly, it should na remain energized for the 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months required in the Technical Specifications. Additionally, during periods of Icw relative humidity the heaters may not energize at all. Therefore, the surveillance requirement has been revised to require that the heater control circuit be
! energized during the 10-hour period. This meets the intent of the specification and is consistent with plant design.
i l
(]
\
,/
4.7.6.b:
The revisions to this specification are as follows:
1. The surveillance requirement to verify a maximum total bypass flow of 1 percent was deleted on the basis that the VEGP design does not include divert valves or the capability for bypass.
2. References to positions c.5.a, c.5.c, c.5.d, and c.6.b of Regulatory Guide 1.52 were replaced with references to Sections 5, 10, 12, and 13 of ANSI N510-1980. This was done for the purpose of clarification in an effort to reduce the potential for a misunderstanding of the requirements. The 1980 version of ANSI NS10 was used as stated in subsection 1.9.52 of the VEGP FSAR.
3 The laboratory testing criterion of 99.8 percent is based on the following equation for methyl iodide penetration which is found in draft Rev. 5 to NUREG-0452:
p ,100% - E SF where P = maximum permissible methyl iodide penetration, percent E = efficiency assumed in the SER for methyl iodide removal, percent p SF = safety factor to account for charcoal degradation v between tests (5 for systems with heaters and 7 for systems without heaters).
If E = 99 percent (Table 15.5 of the VEGP SER) and SF = 5 since VEGP is equippad with heaters, then:
P = 0.2%
Therefore, the laboratory testing criterion for the efficiency for methyl iodide removal should be 99.8 percent.
The filter efficiency of 99.95 percent is based on the assumption of bypass leakage. The VEGP control room emergency filtration system does not include the capability for bypass and, therefore, a filter efficiency of 99.95 percent is overly restrictive. A less restrictive value of 99.5 percent, when multiplied by the laboratory criterion of 99.8 percent, yields an overall iodine removal efficiency of 99.3 parcent which is still greater than the efficiency assumed in the SER.
The limiting criterion for the appropriate filter testing efficiency should be to ensure that at the completion of a surveillance interval, the filters will remove at least as much iodine as was assumed in the accident analysis. Therefore, the value of 99.5 percent should be sufficient to ensure that this criterion is met.
Reference to Section 8 of ANSI N510-1980 was added to the surveillance on system flowrate for the purpose of clarification.
()
s' f) 4.7.6.c:
Reference to position c.6.b of Regulatory Guide 1.52 was replaced with reference to Section 13 of ANSI N510-1980 for the purpose of clarification. i The laboratory testing criteria are listed in the surveillance requirement and the efficiency for methyl iodide removal of 99.8 percent is discussed under 4.7.6.b above.
i.7.6.d:
Item 2 was revised to be plant specific. The signals which result in control room isolation are shown on figure 7.2.1-1 (sheet 8) of the VEGP FSAR.
Item 3 was revised to reflect that the positive pressure for VEGP is measured relative to atmosphere rather than the adjacent areas. The VEGP system provides the required pressurization flow to maintain 1/8 in. Water Guage. This flow may vary depending on such conditions as ambient wind '
~
velocities, direction, and temperature. This is consistent with the-toxic gas analysis for VEGP.
Item 4 was revised to reference Section 14 of ANSI N510-1980 for clari'ication.
Item 5 was revised to reflect plant design with regard to toxic gas - ^
() isolation.
4.7.6.e:
t A filter efficiency of 99.5 percent was specified based on the argument !
presented above in 4.7.6.b and reference to Section 10 of ANSI N510-1980 was inserted for clarification.
4.7.6.f:
g v A filter efficiency of 59.5 percent was specified based on the argument presented above in 4.7.6.c an.d reference to Section 12 of ANSI N510-1980 was inserted for clarification.
3/4.7.8, ECCS Pump Room Exhaust Air Clcanup System (STS):
This specification was deleted since the VEGP design does not include such a system. See FSAR paragraph 6.5.1.1. for those systems which are taken credit for. The ESF room coolers are covered by Specification 3/4.7.13.
) 4-i P'% '
(ms) ,
g~
a-. . - , . , . . , - . .
~
) 3/4.7.7. Piping Penetration Area Filtration and Exhaust System:
This specification was added to reflect that credit is taken for the operation of this system to maintain offsite radiation exposures associated with post-accident recirculation outside containment within the guideline valves of 10 CFR 100. See FSAR paragraphs 6.5.1.1 and 9.4.3.2.2.2.
4.7.7.a:
The justification for the revision to this specification is the same as previously presented for 4.7.6.a.
4.7.7.b:
The justification for the revisions to this specification is the same as .
previously presented for 4.7.6.b.
4.7.7.c:
The justification for the revision to this specification is the same as previously presented for 4.7.6.c.
4.7.7.d:
The requirement that filter cooling bypass valves be vbrified capable of being manually opened was deleted because filter cooling is not part of the VEGP design.
Reference to Section 15 of ANSI N510-1980 was inserted for clarification.
4.7.7.e:
See the justification provided for 4.7.6.e.
4.7.7.f:
See justification provided for 4.7.6.f.
4.7.10.3 (STS):
This reporting requirement was moved to Section 6.9 in an effort to get the annual reporting requirement in one location within the Technical' Specifications. This revision has already been approved by the NRC and implemented in the Farley Nuclear Plant Technical Specifications.
l
em 3/4.7.11, Fire Suppression Systems:
3/4.7.12, Fire Rated Assemblies:
See the justification provided for the deletion of Specification 3/4.3.3.8.
3/4.7.11:
This specification was added on the basis that the operation of the ESF room cooler and safety-related chiller system ensures that the ambient air temperature does not not exceed the allowable temperature for continuous duty rating for the equipment cooled by the system. See paragraph 9.2.9.1 of the VEGP FSAR.
3/4.7.12:
This specification was added to ensure the operability of the isolation provisions that prevent a loss of reactor coolant through the nonsafety portions of the ACCW system. See paragraph 7.6.6.4 of the VEGP FSAR.
N
{
0050v
,st M
3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1 A.C. SOURCES d
OPERATING LIMITING CONDITION FOR OPERATION 3.8.1.1 As a minimum, the following A.C. electrical power sources shall be OPERABLE:
a. Two physically. independent circuits between the offsite transmission network and the onsite Class 1E Distribution System, and t
b. Two separate and independent diesel generators, each with:
1) 5

; r:t: day :nd engin; :ented fuel tank containing a minimum volum( of 75o gallons of fuel,
2) A separate Fuel Storage System containing a minimum volume of (A coo gallons of fuel,
3) A separate fuel transfer pump,
4) Lubricating ci' :terage cent:i ing : =i-ir r t:t:1 vele:: ef g;llen:; cf lubri : ting oil, :nd
5) CapabF Mty to tran:fer lubri : ting cil fre; :t:r;;; to th; di ::1 g:n:r:t:r unit.
APPLICABILITY: MODES 1, 2, 3, and 4.
ACTION: 4
a. With one offsite circuit of the/above-required A/C. electrical power sources inoperable, demonstrate the OPERABILITY /of the remaining A.C. sources by performing Surveillance Requirement 4.8.1.1.1.a within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter. If either diesel generator has not been successfully tested withir. T.he past u ,7 4 f f & 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , demonstrate its, OPERABILITY by performing Su.veillance dsse/ -rencane Requirements 4.8.1.1.2.a.4-and 4.8.1.1.2.a.-d- for ech such diesel
/s a..'r_/.z d y , generator, separately, within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Restore the offsite circuit oper.a.Ne#o '00=
'toOPERABLEstatuswithin72hoursorbeinatleastHOT0;;"Nbwing within the next-36 hours and in COLD SHUTDOWN within the fo y
-Q4 hours. [ \
frANDBy
b. WitheitherdieselgeneratorinoperablefdemonstratetheOPERABILITY of the above required A.C. offsite sources by performing Surveillance Requirement 4.8.1.1.1.a within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter. If the diesel generator became inoperable due to any cause other than preplanned preventive maintenance or testing, demonstrate the OPERABILITY of the remaining OPERABLE diesel gener-ator by performing Surveillance Requirements 4.8.1.1.2.a.-5 and
'SA diese!jener.zhr shah be ansiderel 6 he inoperd/e km h?ha,,e af
&ii- auas.sisas ac ,c ai, w ,. u - .ia - s .d..i-.. <
f. /. / 2 . 4. 5 .
V0GTLE - UNIT 1 3/4 8-1 l
l _ _ _ _ _.__ _ ._..,__ _ - . _._ _._ _ _ ,. .._._.. ~ - _ _ _ _._ ,._ _ _ . - _ _ _ _ . _
.u-,-, - g a. -- .. w -- - a+-u a-a . .
. - l ELECTRICAL POWER SYSTEMS LIMITING CONDITION FOR OPERATION ACTION (Continued) 5 4.8.1.1.2.a.Iwithin24 hours *.*Restoretheinoperablediesel generator to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or ne in at least HOT SrAA/DBy SH&TGOWN within the next ib2-hours and in COLD SHUTDOWN within the following p hours. '/, -
30 6 c
c. With one diesel generator inoperable in addition to ACTION e. or +.
above, verify that:
l
1. All required systems, subsystems, trains, components, and devices that depend on the remaining OPERABLE diesel generator as a j
source of emergency power are also OPERABLE, and j 2. When in MODE 1, 2, or 3, the steam-driven auxiliary feedwater
pump is OPERABLE.
2 If these conditions are not satisfied within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the i following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . 4
d. WithtwooftheahoverequiredoffsiteA.C.circuitsinopeable, demonstrate t OPERABILITY of two diesel generators sepa ately by performing t requirements of Specification 4.8.1.1.2.a. Vand j 4.8.1.1.2.a. within '. 5: r :nd :t !:::t : :: per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter, unless the diesel generators are already operating; restore at least one of the inoperable offsite sources to OPERABLE status within 24
, hours or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .4 W44h rgp2 Only :n: Off:it: :: :: r::ter:d, re:ter: :t !:::t te effrite
^
circuit: t: OPERABLE : tate: within 72 h: r: fr;; ti : Of initi:1 100:
Or b: in at 10::t HOT STANOSY within th; n;xt 0 h: r: :nd in COLO
, 0l-lL'T00"" within th f;11.;ing 20 5:27:.
. With two of the above required diesel generators inoperable, demonstrate the OPERABILITY of two offsite A.C. circuits by performing the require-ments of Specification 4.8.1.1.la. within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter; restore at least one of the inoperable diesel generators to OPERABLE status within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the
, following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .+ Rester: t 1:::t tw di:::1 ;;n:r:ter: t CuarfJ s e g nA;;; 3;;;u; witnin 72 h ;7: f7; ; ti : ef initi:1 12:: : be in 1:::t "0T STANDSY within the next,-S h: r: :nd S COLD SHUTDOW" within the-following 30 heu-s.
SURVEILLANCE REQUIREMENTS f
1 4.8.1.1.1 Each of the above required independent circuits between the offsite
transmission network and the Onsite Class 1E Distribution System shall be

i "This test is required to be completed regardless of when the inoperable diesel generator is restored to OPERABILITY.
V0GTLE - UNIT 1 3/4 8-2 N yvrymryMw+ v-wwr-'wv 9-ur 'w w urwy W --e tv*w--9 -9==y--verew---, v' w a- ew-- y-v -e Ce wf---+-ma. r=we- ev -- e eeiw e ---.r---em--w eenye e t-r-w-ww ,-
_m Insert 1
c. With one offsite circuit and one diesel generator of the above required A.C. electrical power sources inoperable, demonstrate the OPERABILITY of the remaining A.C. offsite source by performing Surveillance Requirement 4.8.1.1.2.a.4 within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter, and, if the diesel generator became inoperable due to any cause other than preplanned preventative maintenance or testing, demonstrate the OPERABILITY of the remaining OPERABLE diesel generator by performing Surveillance Requirement 4.8.1.1.2.a.5 within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , unless the OPERABLE diesel generator is already operating.
Restore at least one of the inoperable sources to OPERABLE status within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . Restore the other A.C. power source (offsite circuit or diesel generator) to
< OPERABLE status in accordance with the provisions of 3.8.1.1, Action Statement a or b, as appropriate, with the time requirement of that Action Statement based on the time of initial loss of the remaining inoperable A.C. power source. A successful test of diesel generator OPERABILITY per Surveillance Requirement 4.8.1.1.2.a.5 nerformed under the Action Statement for an GPERABLE diesel generato. or a restored to OPERABu. diesel generator satisfies the diesel p'arator test requirement of Action Statement a or b.
Insert 2 Following restoration of one offsite source, follow Action Statement a with the time requirement of that Action Statement based on the time of initial loss of the remaining inoperable offsite a.c. circuit. A successful test (s) of diesel OPERABILITY per Surveillance Requirement 4.8.1.1.2.a.5 performed under this Action Statement for the OPERABLE diesels satisfies the diesel generator test requirement for Action Statement a.
Insert 3 4
Following restoration of one diesel generator unit, follow Action Statement b with the time requirement of that Action Statement based on the time of initial loss of the remaining inoperable diesel generator. A successful test of diesel OPERABILITY per Surveillance Requirement 4.8.1.1.2.a.5 performed under this Action Statement for a restored to OPERABLE diesel satisfies the diesel generator test requirements of Action Statement b.
O
~
. .. .. .. - - - - -- .. ~.
ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) !
i
a. Determined OPERABLE at least once per 7 days by verifying correct i
breaker alignments,Ai ndicated power availabilitygand-atL
b. 0.menst.ceted OPC"AOLE at least once per 10 ;; nth; during :hutd:un by
tventferring (=:nually and-automatic 11y) unit p;wer -:upply fr:: th:
neemal-streutt-to-the-altern:t circuit.
4.8.1.1.2 Each diesel generator shall be demonstrated OPERABLE:
a. In accordance with the frequency specified in Table 4.8-1 on a
, STAGGERED TEST BASIS by:
i 1) Verifying the fuel level in the day :nd :ngin: ::ent:d fuel tank, i 2) Verifying the fuel level in the fuel storage tank, j
3) Verifying the fuel transfer pump starts and transfers fuel from the storage system to the day :nd engine ;;;;ted tank, t
- v a-e v (4
^
7:rifying the lubricating sil inventery in steiego, Mce2 W e
[
f -5) Verifying the diesel starts fr;; ;;bi:nt : ndit4en and/h 01:r:t:0 I
te at ! east [900]-rpr ia less than er equal t: [10] f r nd .*a.ed._
,The generator vol.tage and frequency ;h:11 b; U100] 2'D20] volts
- and,[60Y+fl.2THz within E40-] secondsT after the start signal.
The die ~ set generator shall be started for this test by using one of the following signals: g h a) Manual, or b) Simulated loss-of-offsite power by itself, or c) Simulated loss-of-offsite power in conjunction with an ESF Actuation test signal, or d) An ESF Actuation test signal by itself.
an inBWel- W-5-6) ,, Verifying the generator is synchronized, loaded to grc:t:r th:n
~ccoider equal t: [ continuous -rating] kW 'n le:: th:n er :qu:1 t: [50]
. ::: nd:*, and operates with : lead greater th:n er equal te i [centinuess rating] for at least 60 minutes, and 4 -?) Verifying the diesel generator is aligned to provide standby y,,,,,.r g power to the associated emergency busses.
7) Yerif/ ing h,e pressure. ir, a+ /easterre. diese/per.i4r niedart j receiter to beyreafer -t%n ce qu./ to af_g p,y .
l*These-diesel generator-starts-frc: :=bient-conditions-shal' be perf rd Only i once-per-184-days-in-these-surveMlance-tests-and--aH-ether engine--starts-fer the-purpose-of-this-surveillance te: ting :hal' be preceded by an engine pre!ube period-and/or-ethe- warmup precedure: curb :: gr:du:1 10: ding D150 :: ) :s l ! recommended-by the e nef::turer :: th:t th :::h:nic:1 :tr::: :nd :::r er the l
. diesebengine 1: =ir -f::d.
i V0GTLE - UNIT 1 .
3/4 8-3
,4 a ra ll ,,a+i,,i ,,e,/Ji.,, ,,c ,u
% a 74 is Sa.,2 :s mea.,+ as n 6,e.e Cr7 9/vs e . D&d.s s *n tsc e g3 o f 0/s is boe r,d. ef me ove los /se e.s Wetria.fio nJ Oug. fa Yet.or sp iney bot s lestds J/st.// nor ir,ndida.fc _M,t. &3 f, s , .
l t
i !
l 1
Insert 3A 1
All engine starts for the purpose of surveillance testing as required I by 4.8.1.1.2 may be preceded by an engine prelube period as recommended .
l by the manufacturer to minimize mechanical stress and wear on the diesel !
engig.
8 s
t i 4
i i i I
i i
i I I 'l t
I 4
O '
l !
i ,
i I i' i
i i
e i
l I l
l l
1, l i
l O
I I
- -- ,=.. __ -_ _ _ _ _ - - - . . __.,.v..-.--4,m.._ _ _-__. _. _ _ _ - - .
ELECTRICAL POWER SYSTEMS w SURVEILLANCE REQUIREMENTS (Continued)
b. At least once per 31 days and after each operation of the diesel where the period of operation was greater than or equal to I hour by checking for and removing accumulated water from the day and_eagiae-
.m....a <..-l tankg",
f'i c. At l, east once,per u...- 1. . , .. .. . ..
__ .- 92 days [:r On:: p;r
. .. . .u. . .u. .__ .. + u. . . . u.
u_..-_ 31 d:y:
., . u.(!' the
< . . . , gra"nd"et-r
.... . ,.4.,. ....._.
tanb)] by checking for and removing accumulated water from the fuel oil storage tanks;
.,+ s. . 9 29 __2-_ A- JA- Jari-
6. 9..J... y. .u.J
/A.
. . .A +. 1....,i. . . . . . , . .y.. .. . s _. ..u . .s w awua wa, ye awe ww 4 wa awuawawu t-o>..'.e stGr0g; tOnks by 'h;',. ,0. yle Obtained in GCC0rder.Ce
. . '.J:rifying O
. 3er 9,
r. e. . u.. . ms 4 .u._ n,4. s .n _ ..
, n, , .e, _- .
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r . .. . n_ f. n. o_u f. u_._1 n_ 41.. .
n_ n.
a.-s Atleastonceper18 months (,y)duringshutdown,by:
1) Subjecting the diesel to an inspection in accordance with procedures prepared in conjunction with it: e:nef::: rer':
recommendations die.s t forMthis class Mp? i of standby tr-driven service;Laderoswsn) auxi/4r f
2) Verifying thesg! ene ato'./
capa 6/(bility to reject a'/ 'e. load of grFeatef j than or equal to [.argest rin;'" r:rgency 10:d] kW while maintaining voltage at ['150] g ['20] volts and 'r:;tency :t.s ed of 4 s f id F,tt [50] [1.2] t less than or e 1 to 75% of the difference betweennominal(speedandtheOver ce oi ntf'ee--18
..u.,.,... ...-_4., . . u. . u. . . . ... 4.,. ,.,A.,.,
.. . . . . . -. .. .... N Trip nomina 4%c V.240,-410 Setp/ sPeeg ()tus pw.suant to Wse. save V!uce rentremen+s '
For any start of a diesef, the diesel must be operated with a load in accordance with the manufacturerk recommendations.
V0GTLE - UNIT 1 3/4 8-4
. _ . _ _ _ _ _ _ . . _ . _ ~ _ _ _ . . _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ - _ . _
- - --. - .- = . .-. . . . .- . - _ . _ -- - .
i L
I' Insert to Page 3/4 8-4 EO . ,
d. By sampling new fuel oil in accordance with ASTM-D4057 prior to addition to storage tanks and:
1) By verifying in accordance with the tests specified in ASTM-0975-81 prior to addition to the storage tanks that the sample has

a) An API' Gravity of within 0.3 degrees at 600 F, or a specific J
gravity of within 0.0016 at 60/600F, when compared to the sup-1 plier's certificate or an absolute specific gravity' at 60/600F t'
of greater than or equal to 0.83 but less than or equal to 0.89, or an API gravity of greater than or equal to 27 degrees but-
. less than or equal to 39 degrees:
b) A kinematic viscosity at 40 0C of greater than or equal to 1.9 cen-tistokes, but less than or equal to 4.1 centistokes, if gravity was not determined by comparison with the supplier's certification; c) A flash point equal to or greater than 1250F; and -
d) A clear and bright appearance with proper color when tested in i accordance with ASTM-D4176-82.
j 2) By verifying within 30 days of obtaining the sample that the other pro-perties specified in Table 1 of ASTM-D975-81 are met when tested in j accordance with ASTM-0975-81 except that the analysis for sulfur may
)
be performed in accordance with ASTM-D1552-79 or - ASTM-D2622-82.
1
! e. At least once every 31 days by obtaining a sample of fuel oil in accor-j dance with ASTM-02276-78, and verifying that total particulate contamina-tien is less than 10 mg/ liter when checked in accordance with ASTM-D2276-78,
. Method A; t
f. At least once per 184 days by:
1) Verifying the diesel starts
and the generator voltage and frequency
? are 4160 +240, -410 volts and 60 +1.2 Hz within 11.4 seconds after the start signal. The diesel generator shall be started for this test by using one of the signals listed in Surveillance Requirement 4.8.1.1.2.a.4.
This test, if it is performed so it coincides with the testing required by Surveillance Requirement 4.8.1.1.2.a.4, may also serve to concur-rently meet those requirements as well.
2) Verifying the generator is synchronized, loaded to an indicated value of
6100 to 7000 kW** in less than or equal to 60 seconds, and operates for
at least 60 minutes. This test, if it is performed so 'it coincides with.
the testing required by Surveillance Requirement 4.8.1.1.2.a.5, may also serve to concurrently meet those requirements as well, i
j "
All engine starts for tne. purpose of surveillance. testing as required by-i 4.8.1.1.2 may be preceded by an engine prelube period as recommended by the manufacturer to minimize mechanical stress on the diesel engine.
l ** This band _is meant as guidance to avoid routine overloading of the engine.
[. Loads in excess of- this band or momentary variations due to changing bus -
i loads shall not invalidate the test. l' l
I I
,-. , - - - -,--a ,w>y -y m,-- -e --._,,n.- .e.--en--g,-m,v.,p. , , , , , ,,g,, m.,-. >r- -wa --
nn-,..--r. .mm,,,.
ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 700
3) Verifying the4.eset enerator capability to reject a load of [::nt h;;;; 0 ating] kW without tripping. The generator voltage shall not exceed f47843 volts during and following the load rejection;
476C
4) Simulating a loss-of-offsite power by itself, and:
a) Verifying deenergization of the emergency busses and load shedding from the emergency busses, and n.S Verifyingthe$ieselstartsontheauto-startsignal, 4 b) energizes theiemergency busses with permanently connected loads within E493 seconds, energizes the auto-connected -
shutdown loads through the load se'quencer and operates for greater than or equal to 5 minutes while its generator is loaded with the shutdown loads. After energization, the
, steady-state voltage and frequency of the emergency busses i shal be maintained at raigg; i r "
I d60f]f fl.27Hz during I.his test.N] volts gogogo and
5) Verifying that on an ESF Actuation test signal, without loss-of-offsite power, the diesel generator startsW'on the auto-start signal and operates on standby for greater than or equal to 5
minutes. The generator voltage and frequency shall be n. .;
+
i 440+ va, po ? [?100] after; [?20] volts and.f603' the auto-start signal;liet %1.2fHz within generator steady-state {403' seconds voltage and frequency shall be maintained within these Ifmits during this test;
6) Simulating a loss-of-offsite power in conjunction with an ESF Actuation test signal, and:
a) Verifying deenergization of the emergency busses and load shedding from the emergency busses;
//.5 ..
b) Verifyingtheldieselstarts#ontheauto-startsignal, energizes the, emergency busses with permanently connected loads within [40-1 seconds, energizes the auto-connected i
emergency (accident) loads through the load sequencer and operates for greater than or equal to 5 minutes while its i generator is loaded with the emergency loads. After energization, the steady-state voltage and frequency of L0+ ,-d'o the emergency busses shall be maintained at f4160J +d' 2
E4f63voltsand,T6071fl.27Hzduringthistest;aY
/ew /uhe oil pressure-, hof je' kt f c Atter fe* Pod"'s c) Verifying that all automitic diesel generator trips, exceptengineoverspeed/andgeneratordifferential,are automatically bypassed upe- ! css ef ' felt ;: :n the emergency bus conc 9erent "ith a Safety Injection Actuati m 9"1
+ ,s ' e, ain e s r s. , .-: A< 1% n es e s / sa,ve;//.u.ee fe.snu as re. mire..L. Av
} u . 2. J.~/. 2. m .t hd. ffe ce dkd_ d a n '*M,4Ent. ,t976 b4bt. f eriod.'$5 ft C$ *** e nede d b=y
".i m v, .
. % ..y m i ~ a- n m ;,.; ,,,, u . ,n.riA ,n ic,s./ ,y , n , .2 n / m .,e s,., .% d4.se/
en v e V0GTLE - UNIT 1 3/4 8-5
=
....s s-y- 9 y e-- r , w--.,,.,,. .-w--.-.-- - - - ,
, ,_,-y .4v-g, g,,myy mw9--
3, ,y,.,,,_7,,.y-,.,9,,, ,,%,7,,,,%y. y,,--,.y,, 9,., ,7 , 3-
ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 4
j. 7) Verifying the diesel generator operates 'for at least 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .
During the first 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months of this test, the diesel generator shall be loaded to3 prest:r th:n er equ:! t [2-hour r: ting] W and during the remaining 22 hours2.546296e-4 days 0.00611 hours 3.637566e-5 weeks 8.371e-6 months of this test, the diesel generator shall be loaded toigr;;t:r th:n er :;ual t [ er-
~ir /od/u:4.4 c,ico- tinueu: Fitiiig] W.~The generator voltage and fraquency shall
,. c e m. r < r4150]
be$rth:
f :t:rt[120]
ign:1; voltsandj607+j{1.2fHIwithi-[10]::::nd; the :te:dy :tste g:ncr:ter v:lt g: :nd
're;uency :h:11 be sintain:d withi- th::: "-i t: during this 4 %c f 2.q .-g test! % ithin 5 minutes after completing this 24-hour test, perform Specification 4.8.1.1.2e.6)b);.,.
8) Verifying that the auto-connected loads to each diesel generator do not exceed th; 2000-h;ur r: ting Of cern kW;
9) Verifying the diesel generator's capability to:
a) Synchronize with the offsite power source while tha generator is loaded with its emergency loads upon a simulated restoration of offsite power, b) Transfer its loads to the offsite power source, and c) Be restored to its standby status.
10) Verifying that with the diesel generator operating in a test mode, connected to its bus, a simulated Safety Injection signal overrides the test mode by: (1) returning the diesel generator to standby operation, and (2) automatically energizing the emergency loads with offsite power;
11) Verifying that the fuel transfer pump transfers fuel from each fuel storage tank to the day :nd engin; m ted tank of each diesel via the installed cross-connection lines;
12) Verifying that the automatic load sequence timer is OPERABLE with the interval between each load block within 110% of its design interval;
10) Veei fyi ng-tha t-th e-foi l owi ng- di e s el- gene rato r-1 o c ko u t-feature s
> r:; nt di:::1 - generator-starting-only-when-required:
3) [ Turning gear-engaged], er b) [E: rgency-stop-1
> % .? ;,, w t 1 f>r .' es+r~o k + , k , u,1 x.+.+.
f + o
" fIf Specification 4.8.1.1.2fr.6)b) is not satisfactorily completed, ~it is not
! necessary to repeat the preceding 24-hour test. Instead, the diesel generator for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or until operating temper-i may be operated at [::n;in;;;;-ti \ ng] '
ature has stabilized. '
3/4%",6He ical speciWed. A., wet //anee l V0GTLE - UNIT 1 8-6 pff-emen f 4. 8. /. /. 2' . 4 . 5 l
w +- *-
N *
T~ , . - . . .
4
, - - , , - - - . . , , , , . _ , , , , , , , - - -,--,,~,.-.4 ,
N i INSERT 5 an indicated target value of 7650 kW (between 7600 and 7700 kW)**
INSERT 6
All engine starts for the purpose of surveillance testing as required by 4.8.1.1.2 may be preceded by an engine prelube period as recommended by the manufacturer to minimize mechanical stress and wear on the diesel engine.
This band is meant as guidance to avoid routine overloading of the engine. Loads in excess of this band or momentary variations due to changing bus loads shall not invalidate the test.
Failure to maintain voltage and frequency requirements due to grid ,
disturbances does not render a 24-hour test as a failure.
f 6
a l
. i
u - -
1 1
1 1
4 1
ELECTRICAL POWER SYSTEMS s
l SURVEILLANCE REQUIREMENTS (Continued) l l
'a4)- Venifyina, +h.=+. w i. +. h. . = ' . '
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. a. +. 1. a _. c. +. n. .n. , a.
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. M. +. a. P.A_ 7 nn a. n. A_ n_ n.e.n.
. . .h,. , .e .+ m , .+. . 4. ,.m. n h. a. t.h. . ,...
sanacata.r. e . e. 4. =. . "_1. +_ =_ a. a. a. "_ .e d_
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+,
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. .-.....a.,. . . . .
h.1n At least once per 10 years by:
1) Draining each fuel oil storage tank, removing the accumulated sediment and cleaning the tank using a sodium hypochlorite solution, and
2) Performing a pressure test of those portions of the diesel fuel oil system designed to Section III, subsection ND of the ASME Code at a test pressure equal to 110% of the system design pressure.
A . .o . 7. 1. .1 D. a. n. n. .+ e. . . - All A l.. a. .e a. l. n, a. .n. a. c. s + n. . e. 41. m_ m. a_ .eu_, _m 14. A_
f. .m n. e. n. .a. n u. m 1 4. A_ , c h. a_l l in a . + n +a d .. . .w. .D, a, na. . +. n n .e . . . n. +. .+ a. .e n a,4. # 4. ,+ .. 4. a =. c. . a o.9
.k m. ., a. n, .. ... . . . e. n. = 4. .e. e t. a n.
.. .s .c, n a. 4. .m 1 , . r-.---. r.. . .
nana.m+a. #miin.n, ekult fa, iud = +ka !=#---..
u i. +.h. . i n .i n. A_ _s n. e. . D.a.n r---n e+ e af Aineal s-----'
+. .4.....
3. . . .._...-.A ... A. 4. . n n ,- . .1.
6. .- _. g.
n-
.wei 2st.u- e e 9 L w.w.u
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uv'us A.Avo, i
nu v i .'"-
.c i n. .n 1
_, A. n. ,e n .c +. 1_.Q77
_ .T #. +.. h.. a n. n_.e_h n. e._a #. #. _s 41.-. o ca..e
- 4n + k. a. i_m .e +. , n. n u. _m 14.. A. +ae+e
.. . . . ...a .n.
.s n. . .n . n.u. r l. a. .s .
. . n_ n. 4. +. h _s .e 4. e.4. e a, c..a.s
. . + .a w + h. _s n. n. e.
... .. an, n_ s_1 + n. 7, +hn e. nnn et- - ehm11 r-- ---- --ha
$h99 l......
+n I. n..e l. n A n, D. in a,,-n n-.1. s + n. c. y o
. a.a n.'.a A .. . . -- ..
h. a, .s. A A 4. +. 4. n. n._s 1 ( n. #. n. n. e...s + 4.
-- . . . . e. n.
m.ne. n. .=. =. . a .n. A_ aA_
A.
O n. .= 4. +. 4. .m P. .- 1. .h
. a .#.. .
O. n n.11. .s... + A .,. n O..n . 4. A. .a 1. . .i n..O , OA.4.
m . . ... 4... 1. , o . .v.. . . a. 9. a,i, 7.
1 I
i l
l V0GTLE - UNIT 1 3/4 8-7 -
~ ...
TABLE 4.8-1 i w DIESEL GENERATOR TEST SCHEDULE i """er of F:f'"r:: ;
Number of Failures in '- L::t 100 i' lid Last 20 Valid Tests
Tee 4ea Test Frequency 51 --1 Once per 31 days 9
] > 2** -t Once per 7 days I !
i lO 1
Criteria for determining number of failures and number of valid tests shall be in accordance with Regulatory Position C.2.e of Regulatory Guide 1.108, but determined on a per diesel generator basis.
! For the purposes of determining the required test frequency, the previous
test failure count may be reduced to zero if a complete diesel overhaul to like-new condition is completed.,provided that the overhaul, including appro-i and testing, is specifically approved by (
priate post-maintenance the manufacturer and if accepta operat@,ble reliability has been demonstrated. The l
! reliability criterion shall be the successful' completion of 14 consecutive.
' tests in a single series. Ten of these tests shall be in accordance with the routine Surveillance Requirements 4.8.1.1.2.a.5 and 4.8.1.1.2.a.6 and four
tasts in accordance with the 184-day testing requirement of Surveillance
! .Ar.u 2/ _ Requirement / M.0.1.1.2.a.; er.d 4.0.1.1.2.a.C. If this criterion is not I satisfied during the first series of tests, any alternate criterion to be i
used to transvalue the failure count to zero requires NRC approval.
i
The associated test frequency shall be maintained until seven consecutive
' failure free demands have been performed and the number of failures in the last 20 valid demands has been reduced to one.
V0GTLE - UNIT 1 3/4 S-8 ;
ELECTRICAL POWER SYSTEMS A.C. SOURCES J
SHUTDOWN LIMITING CONDITION FOR OPERATION 3.8.1.2 As a minimum, the following A.C. electrical power sources shall be OPERABLE:
4
a. One circuit between the offsite transmission network and the Onsite Class 1E Distribution Syste:a, and
0. One diesel generator with; d fa.e/ da.y J
1) Oey :nd eng n ::ented fuel tanks containing a minimum volume of ag,g_ gallons of fuel,
2) A fuel storage system containing a minimum volume of 44,ooo gallons of fuel,
3) A fuel transfer pump, i) Lubri:: ting ci' :ter:ge cent:ining : -iWu- t:t:1 -veler of '
g:ll:n: cf lubrie: ting ci?, : s O
V 3) Cepebility te-transfer lubrieding cil fre;;; et r:g: te the dit::1 gener:ter unit.
APPLICABILITY: MODES 5 and 6.
in mo nLuree with SpeciReann 3.4 4..s.
a ACTION:
With less than the above minimun required A.C. electrical power sources OPERABLE, immediately suspend alp operations involving CORE ALTERATIONS, ,
positive reactivity changes, movement of irradiated fuel,ior crane cperation >
' with loads over the fuel storagej pool, and within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , de;;r:::urize 2nd vent the Reactor Coolant System through : greater th:n :r ::;u:1 te squr-e inch vent. In addition, when in MODE 5 with the reactor coolant loops not '
filled, or in MODE 6 with the water level less than 23 feet above the reactor vessel flange, immediately initiate corrective action to restore the required sources to OPERABLE status as soon as possible.
provide relief c@i/iG Av SURVEILLANCE REOUIREMENTS 4.8.1.2 The above required A.C. electrical power sources shall be demonstrated OPERABLE by the performance of each of the requirements of Specific-ations 4.8.1.1.1, 4.8.1.1.2 (except for Specification 4.8.1.1.2a.f))/ :nd '. 9. L L ?, '
O i
V0GTLE - UNIT 1 3/4 8-9 y-- -- ; ~,. _
4 3/4.8.2 D.C. SOURCES f%
d CFIRATING LIMITING CONDITION FOR OPERATION 3.e.2.1 As a minipum, the (cllowirrg D.C. electrical sources shall be OPERABLE: ,
a- [250/435-}= volt-Satteey hnt Nc.1, :nd it: t:: cf at:d full espe:ity] ;
G % :r, and (Se e ;,,u,r :
i
b. [250/1& vet-Battery Bank N . 2, and it: FP ebarger. ::cciatedfulicapacity)'##
3/4 t-10. .
APPLICABILITY: MODES 1, 2, 3, and 4. gg, 3
-4
, ACTION: /
bok Withoneoftherequiredbatterybanksand/lorbil::p::itychargersinoperable, !
l restore the inoperable battery bank and/or%ii-ce ecity r charger to OPERABLE status within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months anc in COLD SHUT 00WN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . >
I- ts-epeci ficatica is int:nced-fewus; ca-pherts .;ith wc di .isi;n; cf- 0.C. -
powsn ?y. "edi'ic

ticc: ::y be-cece:::ry,-on : plant uniqu: b; sis, to !
accomodne di'#erent de:igns.
4 4
e SURVEILLANCE REOUIREMENTS ene a.sscc.ia}ed ~
, 4.8.2.1 Each [9E+/1253-volt battery bank and,(charger shall be demonstrated .
1 0 ERASLE:
a. At least once per 7 days by verifying that: j
1) The pararneters in Table 4.8-2 meet the Category A Ifolts, and P
2) The total battery terminal voltage is greater than er equal to
[2:0/120] volts on float charge. j l.2. G, 1
3 VCGTLE - UNIT 1 3/4 8-10 I
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i
6+n 4
i Ir,ser t to Page 3/4 S-16
a. J.25 V*de Batiery bank 1.4D13 and one of its asto:i3ted full-capacity chargers. -
b. 125 V-cc Battery banx ISDJB, And 6ne of its associated fulT-capacity chargers.
c. .125 V-dc Eat *et'y bank 1C013, and one of its associated full-ccpacity chargers. .
d .- IM V-cc Bntory bar.k 10DIB, eno one of its ass 6ciated full-capacity cnargers. {
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SURVEILLANCE RIQUIRENEh75 (Certinued) _.
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b. At least or.ce per 02 days and within 7 days /after a battery disch&rge with battery termi.nal voltage belo.< E2.?0FJtF] volts, or battery overcharge with battery terminal voltage above (%57't] vo3ts, by verifying that: MC
1) The parametsrs in Table 4.8-2 maet the Category B limits,
2) There is ro visible corrosion at either terminals er connectors, or the cor.ne.-tion resistance af these (tems ig les.s it'en ,
, [1EC x 10J]Mr-ec6 zo % o a < t.h e c. ween 3L. as www<A-k ri,,9 H.c y/w t pre.cpstaMens/ t e.s.es < n L.
3) The average electtplyte teinperatu e of E4--Fopres+s444*e-wameee]
+4 conne:ted cells is above E60}' F. A c e j v e.
S6
c. At least once per 1$ months hy verif91ng that:
4
1) The cells, cell plates, and battery racks show no visual indication of physical damage or abnormal deterioratien, i 2) Tne cell-te-et11 and temir.a1 connecticas are clean, tight, and .
coated with anticorrosion material, -
3) The resistance of each call-to-cell and terminal connection is less than or equal to (450 x 10 83 A , w4 zo % ove.< /ua.vuaje<
me.a.s w <4 du<iog en pla.n fpespua.fio mJ
4) The battery charger will suoply at Itastf400}jen,s ag eresand.
at f125,E0] volts;for at least Efri tcurs.
t-ne m M c //y 11.
1
d. At least once par 18 months, during shutdown, by verifying that the 1 battery capacity is acequate to supply and maintain in CPGABLE status al? of tne actual or simulated emergency loads for the design duty cycle when the battery i.s subjected to a battery service test; I :
e. At least cecs per 6] nonths, during shutdown, by vedfying that the battery tapecity is at least 30% of the manufacturer's rating when stabjected to a performance discharge test. Once per 60-month interval this perforcants discharge tett may be cerformed in lieu cf the battery service test requirect by !pecification 4.8.2.1d. ; and j I
f. At least once per 18 months, dering shutdown, by giving performance l discharge tests of battery cao3 city to any battery that shows signs '
of degradation or hss reached 0% of the ser'<1ce life expectad for th., app 3 icatio.n. Degradatica is ir.dicated when the battery capacity drops scrs than 10% of r3tod cepacity. from (f.s average .on previous performance tests, or is pelcw 90% of the nanufacturer's rating.
,4e sy.sM, d <mL 3, soo ampeas log s ys te m C , s e..L 2 c c . sap s ro nr S ,i s re m D VCG7LE - UN!T 1 3/4 8-11
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_TA*LE 4. 8-_2 i
- d BATTERY $URVEILLANC5 REQUIKIMENTS CATEGORY A U) CATEGORY 3(2) j PARAMETER LIMITS FOR EACH LIMITS FOR EACM AttCWASLE(3)
OESIGNATED PILOT COMICTID CEtt VALUE FOR EACH CELL CCMECTED CII.t 1 '
Electrolyte >Hfnimun level >Minimur leve? Above top of ,
indication mark, indication r: ark, level 'pla ta s ,
and i h" .above and i k" abcve and not
maxiaue level auteum level overflowing indication mark indication mark
^
Flort Voltage 3 2.13 volts > 2.13 Volts (6) -- > N volts Nct mee thac i
0,C20 below the ,
~
! /. m average of an
! Specific > 1.-290(5) > h+M connected calls
^
sGravity (#)
Average of all Aserege of all
ccnnected cells connected cells ,
> b-295
"" y Q~ '~ S)
O TABLE NOTATIONS i~
(1) Fcr any Category A parameter (s) cutside the Mnit(s) shown, the batten '
may be considered OPERASE provided that within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months all tne Category B i measurer.ents ara taken and fcund to be wit.hin their allowacle values, end !
provided all Categorf A and B parameter (s) are restored to within finits <
, within the next 6 dcys.
(2) For any Category 3 parameter (s) outside the limit (s) shown, the catte.~j ,
may be censbered OPERAoLE previdea that the Category 8 parameters are within their allowable values and provided the Category B parameter (s) are ,
restered t,o within limits within 7 days.
. (3) Any Category B parameter cot withfr. Its allcwable value indicates an inoperable batte.~/.
(4) Corrected for electrolyte temperature and level he,.t.t '
(5) Crbatterychargingcurrentislessthanj2/ampswhenonj-harge.
I (6) Corrected for average electrolyte temperature. (
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V0GTLE - UNIT 1 3/4 G-12 l
. _ _ _ _ _ -.._,..._-__.1,
cnc huiei re b %d. ca ir ef I: 5 - V i.'2 de <y ha.r,ks (G < +ae r-1.:: 5 V 75a 6 ,v i,.za,tr tA 2 ir3 ,rn) (cgig er D.C. SOURCES !.: S - V ht Hr ry. l>a .,n: ' l dn ' e . t -J ! =0 .9 ' a ccL en e
z. s s o c i xted r a.u- cay L U .'y c l a oye r ,re r h1 derf b ank.
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/ SH'JTDOWN LIMI7ING CN OTTIC4 FOR; OPERATION ,
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} 3. 8. c. 2 As a minire,Jera [250/125NAt-battery 4enk-eehs n;;c?;t;d fell-eepedty-1,$ r;n shall be OPERABLE.
APPLICABILITY: HJOES 5 and 6.
dU'Y'c j A_CTION:
be?h f
! With the required battery bank and/or! full cogdby chdrgersinoperable,
1. mediately suspend gli operations involving C0AE ALT TIONS, positive i reactivity enanges, or sev;: ment of irradiated fuel; f itiate corrective action to restore the required battery back and firH-cepedtj charger to j
i meOPERABLE the Reactor status as soon Coolant System as possible, trievic, a and within square 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months sf.i P n.%e,a.ni:: c,r4
/o scor/ance, i ca. f., E G: c /Ec :thp a J. -!. L J I
s
.SUTtVE11 LANCE REGUIREMEt!TS ."
M* #8M*f '*FO##+7 !* '
L - -- _
f j r 4.8.2.2 The above required EEW125]-volt battery bank and full-capacity charer shat) be demonstrated OPERABLE in accordance with Specificat. ion
4. 5. 2.1.
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s VOSTLE - UNIT 1 3/4 3-13
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3/4.8.3 \0NSITE POWER DISTRT30 TION .
OPERATIfl0 LIMITINr; COND TION 102 GPDATION ,/
3.8.3.1 The fell ving electrical busses shall be enarg.'ied in t. spectff ed manner with tie br kers apan (both3 Detween reduur? ant busses wi in the unit (and between units the sace statien]:
a, Division el .C. Emergency Susses ::onsisting of:
1) [4150] 'I. It Emergency Bus # , and
2) [420]-Vol. Emergancy Sus # _ .
b. Division #2 A.C. tergency Basses consisting of' (416CJ-Volt Et'rgency Bus # , and '
1)
2) [480]-Volt Eme ency Bus # .
c. (120]-Volt A.C. Vltal us # energizedpromitsarsociated inverter connected to thC. BIisT ,
d. [E03-Volt A.C, Vitai Buh# anergizyd from its abociated inverter come::ed to D.C.'y3us e _a,/
a. [120]-Volt A.C. Vital Bus # \ ~ energfhedfromistassociated invertarconnectedtoB.C.Boge f, f, [1203-Volt A.C. Vital Bus # \ e, zec from its associated
inverter connected to C.C. B G'
g. [250/125]-Volt 0.C. Bus R energi.IdJ from 3attery Bank #1, and
h. .(250/1253-VoltD.C.hs#2energ)IekfromBatterySanx#2. ' '
APPLICA3!LITY: KODES 1, 2, 3, and 4. /
ACTIcy:
a. Vith are of tre required dipisions of A.C.\ergency basses not fully ecergized, reenergize tne divisica wit in 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or be in at leaJt M0T STANDBY within t.he nex1; E hturs and 'n COLD St!uTccw?i wi thin the folicwicq 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months , /
b V5thoneA.C.vitalbus/eithernctenergizedfrc. its assoc {ated inverter, cr with the /nverter not connected to i assada ;ec 0.C.
b'as: (1) reenergize ,the A.C. viul Bus ylthin 2 ho rs or be in at least h0T STANC,3Y wyhin tba next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLA Sf!UrD0VN within the fatlewing 10 nours; and (2) reenergize the'$n.C. vital bus frcen its asser. fated inverter connected to its asso ated 0.C.
bus wittin 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or ce in at 7 east HOT STANDBY within\the text 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in GLD SliUTDOWN within the folicwing 30 heart
Twoinvertersmayoedisconnectedfroatheir0,0.basforupto24habsas necessary, far the pur;.ase of perforcing an eqt.alizing charge on their bsociatad battery bank provided: (1) their vital busses are energized, and (2) the vital
' busses associatao wD.h tr.e ot: er batt.ary bank are energized from their associsted inverters and conrected to their associated D.C. bus. i
/
y0GTLE - UNIT 1 - 3/4 8-14 -
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LIMITING CONDITION FOR 09ERATIDX ---
1 pCTION (Cc:1tinued) l b?nk,
p. With one O.C.' bus not energhed from its associated batta ithin 2 hcurs t, reenergize the D.C. bus from its associated battery bank fa CO'D
~
or be in at least HOT STAN0BY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months a SHUTDCWN within ti,a follcwing 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .
AURVEILLANCE REQUIREMENTS
/
i 4 4.3.3.1 The specitfed bessss shall 3a determined energized in the recuired annner at least once per 7 days by verifying correct brepker alignw.ent and
' in:ticated voltage on the busses.
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- 3/4 8- 15 I V0GTL! - UNIT 3 l
. b ELECTRICAL POER SYSTEMS ,
3/4.8.3 ONSITF F3WER DISTRIBUTION i
GPERATING LIMITING CC%DITION F0R.0PE8AT1GH T.T.~3.1 Tne followin[electrica) Busses snall be energized in the soccified i manner.
a. A.C. Emergency Sussec consisting of:
1. Train A ,
a) 4160 volt saitch] ear 1AA02 b) 480 volt switcogear 1AB04
1) MCC 1AEE c) 480 volt switchgear IAB05 '
1) MCC 1ABA *
2) MCC iSBC
3) VCC 1ABF d) 480 volt swiCcrgear IAB15 i 1) MCC 1ABB 1 t
2) MCC 1 ADD >
4 2. Train B j e) 4160 volt switcrgear JEA03 ,
1 b) 480 vclt swit h g r I M 06 i
1) MCC 183E :
, c) 480 voit switchgear 18007 .
1) MCC 1BBA
. 2) MCC 1BBC i 3) MCC 1BBF
d) 480 volt swittbgear 18E1B
1) MCC IBBB ;
1 2) MCC 1880
b. 120 voit A.C. vital Busset
1. Associated with Trait A a) Channel I t
1) Panel 1AY1A energized fren inverter 1AD111 connected to l switchgear 1A01* l
2) Panel 1AY2A energized from inverter 1AD1111 connected to *
switchgear 1ADi*
b) Channel III J
1) Panel ICY 1A energized frcm inverter 1CDlI3 connected to switchgear ICD 1'
2. Associated with Train S l a) Channel II ^
l 1 1) Parel 1BY1B ene gized from inverter 1BD112 connected to switchgesr 180l*
2) Panel 1BY25 energized f cm invert.er 18D1I12 connected to i
, switchgear 1001* !'
b) Channel IV
1) Panel 10Y13 energized frorn irterter 100114 connected to Tsitchgear 1001*
' c. 125 volt D.C. Busses consisting 6f: ;
1. Associated with Train A t a) system n ;
1) Ssitchgear 1AD1 energized frcm battery 1AD13
2) M~C 1AD1M energi:ed frcm switchgear 1AD1 -
3 V0GTLE-UNIT 1 3/4 8-14
.-s - .-, - -- -c % w .,.-,r.,--.--.m%w . %-.,y,,e -...-,.--.------%,,r .- ,,. . . , %%- , ,-,,_.v-r,- e- ,*ws,..--ww.= y- --ws,.-+ .. -v%.-g ..< rc-.
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c 3) Distribution panel 1AD11 energized from switchgear 1AD1
- 4) Distribution panel 1AD12 energized from switchgear 1AD1 b) System C
1) Switchgear ICD 1 energized from battery ICD 1B
2) MCC ICD 1M energized from switchgear 1C01
3) Distribution panel ICD 11 energized from switchgear 1CD1
2. Associated with Train B a) System B
1) Switchgear 1811 energized from battery 1801B
. 2) MCC IBDIM energized from switchgear 1801
3) Distribution panel IBD11 energized from switchgear 1801 1
4) Distribution panel 1BD12 energized from switchgear 1801 b) System D 4 1) Switchgear 1001 energized from battery IDD1B
2) Distribution panel 10D11 energized from switchgear 1001 APFLICASILITY: MODES 1, 2, 3, and 4 j ACTION:
3
a. With one of the required Trains of A.C. Emergency Busses not fully i energized, re-energize the Train within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or be in at least i
H0T STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hcurs.
! b. With one 120 volt A.C. Vf tal Bus either not erergized from its associated inverter or with the inverter not ccnnected to its associated Switchgear: (1) Re-energize the Panel within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months I
' O. or be in at least HST STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTD0hN withir. tre following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months ; and (2) Re-energize the I Panel feori its associated inverter within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in at i least HOT STANDBY uithin f hours and in COLD SHUTD0in! within the folicwing 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .
. c. '4f th one 125 vcit D.C. Bus not energized in the specified manner, re- i energize the Switchgear or Distribution Panel in the specified manner within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or be in at le.ast HCT STANDBY in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD '
SNUTDOWN in the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .
} SURVEILLANCE REQUIREMENTS ,
a
! 4.8.3.1 The specified busses- shall ce aetermined energized in the specified manner at least acce per 7 days by verifying correct breaker alignment and *
indicated voltage.
1
Up to *.we inverters in a single train may be disconnected from their associated Switchgear for up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months as necessary, for the purpose of
! Battery Cank or inverter maintenance provided: (1) Their associated Panels are ens gized from their regulated transformers and (2) The panels i associated witn the other Battery Bank powered from that AC Train are energiz9c in tna specified manner.
f
! V0GTLE-VMIT 1 3/4 8-15 }
I 4
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O ONSITE POWER DISTRIBUTION O SHUT 00WN LIMITING CONDITION FOR OPERATION L+)
3.8.3.2 As a minimum, the following electrical bussesAshall be energized in the specified manner: , !
tr.an )
a. One divisien of A.C. emergency busics consisting of one f4160 volt sw/My%
ruecen4-ene f480pvolt A.C. emergency be:,ani six + 7c vot/ AC. A etu Gnfrel Cmters.
s win.he A w-r Cn e resin c r
b. -Twof120pvolt A.C. vital busses energized from their associated inverters connected to their respective D.C.' busses, and One. reatn o/ swtMyw wL a.s s o dated- dis +nLHon egwpment
c. Cn: [250/125P, volt D.C. bus energized from its associated battery bank.
APPLICABILITY MODES 5 and 6.
ACTION: /70 Vide f*lI'l W "bilIV l*f With any of the above required electrical ausses not energized in the required j manner, immediately suspend all operations involving CORE ALTERATIONS, positive
- reactivity changes, or movement of irradia ted fuel, initiate corrective action to energize the required electrical busses in the specified manner as soon as possible, and within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , d:pr:::tri:: : d ::t the RCS thr;;gh at ':::t i s-- :qu:r: in:h ':: t:in accord.wce. wah SpexsWeaken 3. + 9 3.
4 4
l SURVEILLANCE REOUIREMENTS i
i 4.8.3.2 The specified busses shall be determined energized in the required manner at least once per 7 days by verifying correct breaker alignment and indicated voltage on the busses.
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i 4
i s 4ll e h r< h.! bus es sha.tl be frsm the sm train .
VCGTLE - UNIT 1 3/4 8-16
. .. - = _ . . .- - - . - _ ...
i 3/4.8 4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES
! A.C. CIRCUITS INSIDE PRIMARY CONTAINMENT 1
LINTING CONDITION FOR OPERATION ,
3.8.4 At least the following A.C. circuits inside primary containee shall be deeneggized:
\ ] in panel [
a. ,- and J.
0(rcuitnumbers[ ,
b. Circuit numbers [ , , _ and 2 in panel [ 3
\
APPLICABILITY: MO0ES 1, 2, and 3.
ACTION:
N
With any of the above required circuits energized, tr' the associated circuit breaker (s) in the specified panel (s) within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .
I /-
SURVEILLANCE REQUIREMENTS s. /
! 4.8.4.1 Each of the above require A.C. rcuits shall be determined to be deenergized at least once per 24 ho s* y verifying that the associated O' circuit breakers are in the tripped c ition.
/
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Except at least once per 31 days if locked, sealed, or otherwise se ured in the, tripped condition.
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V0GTLE - UNIT 1 3/4 8 !
i
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.s I
l ELECTRICAL EOUIPMENT PROTECTIVE DEVICES CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES juo pggo gg
&erAxfMS rs / set A77ar/ r/2AA/S for2M S/2.5 8 5-r w f EA/ 4 9n V r/ETM /2 ~
LIMITING CONDITION FOR OPERATION duccr ANO e/a>-am / c soropurm-1.
3.8.4.E- All containment penetration conductor overcurrent protective devices v,44 given in T b h 3.S-1 shall be OPERABLE.
Weeder breders to isola. Hon APPLICABILITY: MODES 1, 2, 3, and 4. fraodormers be Aun veo V c/us /E busses and non-c/*ss /f ACTION: efu./pment With one or more of the containment penetration condue.:r overcurrent protective device (s) ,civer in Tab!: 3. S-1 inocerable: , and /e eder 4reders ** s'o /4*/on transformers bekeen 4so V c/ ass /E busses and non-C/ Ass / f epoymenf
. {eetere the protective device (s) te OPEPABLE statur er deenergize th: circuit ( ) by tripping the ::: :f:ted b::kup circuit br::ker See ,sserr er reckin;; cut er -e-e"ing the 49eperable ci-cuit 5-ee'e- uithia de 3Af-/7 72 Scurc, de !:re the :ffected :y:ter er cc pene-t daaperab!e, and ver fy the beck"a circuit breaker te be tripped er the inepe&
i able circuit breaker r eked cut er re Oved :t 1:::t enc: p:r 7 d:y:
there:fter; the previcien: Of Speci#ic:tien 3.0.A 3pg ggg yp;}jgyg}g te evercurrent device 4- Circuit: which h:v: their b::kup circuit bre ker: tripped, thei i oper:ble n circuit breaker: r: k:d Out, Or krer:ved,er e
c.4. Be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD .
SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .
SURVEILLANCE RE0VIREMENTS 1
4.8.4.-e All containment penetration conductor overcurrent protective devices and gi/en in T:b h 3.3-1 shall be demonstrated OPERABLE: / seder 4reder.s to /se/a.7/en fransformers bekeen 480 V
a. At least once per 18 months: (j,,,, / r fusses and non-(/4ss
/ 3' B / E e9 u.i
1) By verifying that the x dium v lt:g y W15 kV] circuit fmenfbreakers are OPERABLE by selecting, on a rotating basis, at least M the circuit breakers of each voltage level, and performing the following:
a) A CHANNEL CALIBRATION of the associated protective relays, b) An integrated system functional test which includes simulated automatic actuation of the system and verifying that each relay and associated circuit breakers and control circuits function as designed, and O _
I7 V0GTLE - UNIT 1 3/4 8 . - -. _. -. .-. - - _ . . - - - _
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(_,) Insert to Page 3/4 8-17
a. Restore the protective device or feeder breaker to OPERABLE status or deenergize the circuits by racking ~ out or removing the inoperable cTr.cuit breaker or protective device and tripping the associated backup circuit breaker within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ,1 declare the affected system or compo-nent inoperable, and verify the inoperable circuit breaker or protective deviterracked out or removed at least once per 31 days thereafter; the provis' ions of Specification 3.0.4 are not applicable to overcurrent
~
tevices or feeder breakers in circuits which have their backup circuit
~
~ breakers tripped, their inoperable circuit breaker racked out or removed, or
b. Deenergize the circuits by racking out or removing the inoperable cir-cuit breaker or protective device within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , declare the affected system or component inoperable, and verify the inoperable circuit breaker or protective device racked out or removed at least once per 7 days thereafter; the provisions of Specificatior 3.0.4 are not applicable to overcurrent devices or feeder breakers in circuits which have their inoperable circuit breaker racked out, or removed, or 1
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ELECTRICAL EQUIPMENT PROTECTIVE DEVICES SURVEILLANCE REQUIREMENTS (Continued)
O n 4*
c) For each c fcuit breaker found inoperable during these functional tests, an additional representative sample of at least of 44 the circuit breakers of the inoperable type shall also be functionally tested until no more failures are found or all circuit breakers of that type have been functionally tested.
2) By selecting and functionally testing a representative sample of at least 10% of each type of lower vo.itage circuit breakers.
Circuit breakers selected for functional testing shall be selected on a rotating basis. Testing of these circuit breakers shall consist of injecting a current # fth : ::h:
equa4-to 300% of the pickup Of th: kng-ti : d; hy trip element and 150% of the pickup cf the :hcrt-ti : de hy trip e-lementr-and verifying that the circuit bre:k:r Op:r t::
w&tMn-the-time-dehy b:nd etath for th:t curr:nt :p::ified by-the-manufacturer.---The-4estanten :us ch;;nt sh:11 b tested by injecting a current equal to ML" cf the pickup vake-of the eh::nt and-veef fying that th: circuit br::ker "
tFip: instant:neeu:1y-with : intention:1 tim dehy. .:1d:d e::: circuit breaker testing chel' she fell = t'f: pree:dur:
( except-that-generaMy-no-more-than-two-teip k;;nt:. tie:
\ d- hy and instant:necer, ef'1 be 4 vehed. Circuit breakers found inoperable during functional testing shall be restored to OPERABLE status prior to resuming operation. For each circuit breaker found inoperable during these functional tests, an additional representative sample of at least 10% of all the circuit breakers of the inoperable type shall also be function-ally tested until no more failures are found or all circuit breakers of that type have been functionally tested; and L. At least once per 60 months by subjecting each circuit breaker to an inspection and preventive maintenance in accordance with procedures prepared in conjunction with its manufacturer's recommendations.
1 lM 2X C dS 5 e! Sh L brea 5(. erb /1 C W1ir1A ! $4 ffoin? eta me a. 3 u r;y rh e. r?spense dim e - 7~he eraea s u red.-
'respcnrc Nme wi// be. cs m p xre L h //,e. man.<.hefuer}
da f.t h ensa.re %F if h wi+hin H,e tolera nces
\. -rp e.c.;.Re 1. u y th c. ma.n aAt- c+a.rer-O
/7 l V0GTLE - UNIT 1 3/4 8-19 i
pg TABLE 3.8-1 ,
V CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES DEVICE RUMSER SYSTEM AND LOCATION POWERED
\ /
1. 6900 VAC Reactor Coolant pump (Primar9sbreaker) 1 (Backup b'neaker) 2
. 3 4
2. 480 VAC from AD Centers List all; prima'ry breakers Backup breakers Backup breakers
3. 480 VAC from MCC List all; primary breakers Backup breakers Backup breakers
4. 125V DC Lighting O- List all; primary breakers Backup breakers Backup breakers 440 VAC CRDM Power
/
5.
\
Primary breakers / \
Backup breakers /
/
Backup breakers 7
/
\
/ \
/
/
o /
V0GTLE - UNIT 1 ,
3/4 8-20 l
~
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) ELECTRICAL E0VIPMENT PROTECTIVE DEVICES J SA Fe~ry- Ret.4 RED
,5 MOTOR-OPERATED VALVES THERMAL OVERLOAD PROTECTION wig BYPASS DEVICES LIMITING CONDITION FOR OPERATION Sde+y -re / del me to r operAfrA 3.8.4.hThethermalovrioadprotectionandbypassdevices[integr:1withth:
actor : tarter of each valve If:t:d '- T:ble 2. S 2 shall be OPERABLE.
APPLICABILITY: Whenever the motor-operated valve is required to be OPERABLE.
ACTION:
With --- -- -- : fthethermaloverloadprotectionanMeebypassdevice[
inocerable, declare the affected valve (s) inoperable and apply the appropriate
% l ACTION Statement (s) of the affected valve (s). 4 hr a.ny one er more. snfe+y -feMEL mehr-opernM '
SURVEILLANCE REOUIREMENTS 2.8.4.3 The above-required E_.
ther=:1 cverload-protect-ica and bypass devices f* S F **
M f 5,k~ .1,1 L.
. a m - m.
a. A.t least once per 19 months, by the performance of a TMP-AGTUAHNG BEVICE OPERATIONAL TEST cf-the byp::s-etrettitry-for-these ther al O' everload-dev4ces-which are either:
1. Continuous 4y-bypassed :nd temporarily placed-indorce only th:n
-the-valve-motors-are-undergoing periodic-oe-maintenance-testing, or-l 2. HermaHy-in-force--during-plant-operation and bypassed under acc4 dent-conditiens.
b. At least-ence-per la caths by the performance df a CHANNEL-CAMBRAHON
-of-a-representatrive s:mple-of at 1c::t 25% of:
1. ^1' thermal-over4oad-dev4ces-which are net bypassedy-such that-each-non-bypassed device is calibrated et least-once-per-6-year 4,
2. A14-thermal-oveeload-dev4ces-which are continuously-bypassed-and temporar44y-placed-in-force only when the v:ive ;tcrs are stedergoing-periodic or m:intenance testing, :nd thermal. overlead dev4ces-nor=:lly in force and bypassed under ::c4 dent-condf# fens such-that-each-thermal ever! cad is cal 4brated and :ch-valve f:
cycled-through-at-least-one-complete cyc!: Of fcil travel with the-motor-operatcr when the therma! Overicad i: OPERASLE :nd net bypassed, at least ence pe" 6 years --
%e ins er t h pa.ge. 3p t- 19 I
V0GTLE - UNIT 1 3/4 8 I
, _ , . _ . , , _ _ . _ , _ _ _, _,m.._~ _ , _ _ _ . . . _ _ , . . _ - - - - , . - . , _ , . . . _ . ~ - , - . . , - , . . _ _ . . _ , . _ . . - . _
4- w o *64 O INSERT TO PAGE 3/4 8-19 4.8.4.2 The above recuired thermal overload protection bypass devices shall be verified to be OPERABLE.
a. Following maintenance on the valve motor starter, and
b. Following any' periodic testing during which the thermal overload device was temporarily placed in force.
c. At least once per 18 months, during shutdown.
d 1
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- _ - _ . . _ _ _ _ . , _ . - . ~ , . _ _ - . . _ . - . _ - .
1 i
i i \
j TABLE 3.8-2 MOTOR-OPERATED VALVES THERMAL OVERLOAD PROTECTION BYPASS DEVICE SYSTEM (S)
AFFECTED VALVE NUMBER IC tinuous)(Accident Condi ons)(No)
\ /
\
7
/
. z
/
/
/
O e
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V0GTLE - UNIT.1 3/4 8-22
.-m., .,v,_,._._,4_.____m,__., ,,._., ,,.c, _.,-,-,m, .. ,,,....,.,.~_,,.._,,,.,-m.,.,,, ..._,_,,,.,,.,v_,,,.,,_,__.__,,
JUSTIFICATIONS FOR DEVIATIONS FROM STS SECTION 3/4.8 3
3.8.1.1.b:
i The VEGP diesel engines are not equipped with engine-mounted fuel tanks.
l The requirement for lubricating oil storage minimum volume and transfer i capability was deleted on the basis that the VEGP diesel generator lobe-oil system is integral to the diesel package and will be covered under the definition of operability of the diesel itself. See figure 9.5.4-1 and
response to Question 430.35 in the VEGP FSAR.
3.8.1.1, Action a:
, This action statement was revised to clarify that, given the loss of an
{-
offsite circuit, the untested diesel generator may already be operatinj; in which case, testing would be unnecessary. In addition, the action statement was revised to specify Hot Standby in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and Cold Shutdown within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months for the sake of consistency.
1 3.8.1.1, Action b:
This action statement was revised to clarify the point at which a diesel generator is considered to be inoperable. The requirement to be in Hot Shutdown within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and Cold Shutdown within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months was 3
revised to require Hot Standby in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and Cold Shutdown within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months for the sake of consistency.
I i 3.8.1.1, Action c:
This action statement was inserted to provide for the case where an offsite circuit and a diesel generator may be inoperable. As part of the equipment is restored, action is transferred to an earlier action statement. Explicit language has been included to highlight this transfer and clarify the i starting time for the requirements of the action statement being entered.
! The transfer statement also clarifies that if a diesel generator has already been successfully tested as part of the present action statement, it does j not have to be retested.
!' 3.8.1.1, Action e:
i This action statement has been revised to in'clude explicit language which transfers action to a previous action statement when one offsite source has been restored and one remains inoperable. In addition, the requirement to a perform surveillance on the diesel generators within I hour and once per 8
!O I
i i
f
l l
l I
hours thereafter was revised to within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . This was done to reduce operating the diesels in parallel with a degraded system; reduce demands on the operators' time for testing; and reduce the number of tests required of the diesel generators. As presently worded, the Standard Technical Specifications could result in as many as nine tests of each diesel generator (a total of 18 tests for two diesels) over a 72-hour period. We believe this kind of testing to be excessive, especially in view of the fact that current industry experience indicates that excessive testing can be detrimental and could cause loss of the EDG when it is needed most.
Further, since this type of testing involves operating the EDG in parallel with a degraded system, the possibility of losing the EDG during such testing is increased. The EDGs are designed and intended to be standby power sources. Therefore, the negative implications of such a test are not trivial.
There is also the consideration for operator actions during plant abnormalities. Operator capabilities to diagnose and track plant conditions have been increased and priority of ooerator actions has been given further attention. Actions to assure adequate reactor core cooling may continue over hours and have priority over the testing of redundant equipment.
Operator actions that could distract the operator, require his time to be spent on lower priority actions, or have the potential to unnecessarily generate doubts or confusion, need to be minimized.
3.3.1.1, Action f:
O' The action statement has been revised to transfer action to a previous action statement as equipment is restored to operability. The transfer statement also clarifies that if a diesel generator has already been successfully tested as part of the present action statement, it does not have to be retested.
4.8.1.1.1.b:
This surveillance requirement is not applicable to VEGP on the basis that 4 the Class 1E distribution system has two independent ac power trains, each fed from an independent Class 1E bus with immediate access to the offsite transmission network through a physically independent circuit. A failure of a single active component in the onsite electric distribution system will not prevent the safety-related systems from performing their function. The VEGP design of the offsite transmission network to the onsite safety-related loads, with power from the offsite transmission network for each circuit available immediately, provides the independence and redundancy required to effect a safe shutdown of the plant assuming a single failure. This is acccmplished without taking credit fu alternate feed capability.
4.3.1.1.2.a.1 &
4.8.1.1.2.a.3:
O G
There is no engine-mounted fuel tank.
J-f() 4.S.I.1.2.a.4 (STS):
See the justification provided for 3.8.1.1.b.
i a
j 4.8.1.1.2.a.4:
i
$ This surveillance has been revised to require that'the generator voltage and i frequency be at acceptable levels within 11.4 seconds after the start signal. The specification on engine speed is redundant to the requirement on frequency and is therefore not necessary. From the standpoint of the
safety-related loads, the proper voltage and frequency arr the important parameters. This surveillance has also been revised by deleting the words "from ambient conditions" since each start for the purpose of.this surveillance will be preceded by a prelube period in accordance with manufacturer's recommendations. The primary purpose of routine (monthly) testing should be to implement the recommended testing and to verify 1 starting and load handling capability, rather than to simulate the design I
basis accident conditions.
j 4.8.1.1.2.a.5 and Footnote *:
1 This surveillance has been revised to eliminate the fast loading requirement on the basis that routine (monthly) testing should demonstrate starting and
load handling capability rather than simulating design basis accident 1 conditions. The fast loading will be done every 184 days' as per new
. Surveillance 4.8.1.1.2.f. Also, a loading band has been specified rather than a target load. This has its basis in the fact that some operating plants have been forced to routinely overload their diesels during testing man effort to ensure that minimum test requirements are met. When ,
instrument uncertainties are factored in, the overloading could be significant. The monthly test should exercise the diesel generator, confirm its operability, and detect degradation or a failure before a second diesel j generator failure is likely to occur. During the 18-month testing, the test
loads envelope the calculated accident loads. It should not be necessary or
! desirable to envelope the design basis accident loads, which might occur in
! 10,000 years, by a test that is repeated 12 times each year. The. band specified (6100 - 7000 kW) is based on a maximum auto-connected load of 6032 kW and the continuous duty rating of 7000 kW. Footnote
has been revised since 4.8.1.1.2.a.5 no longer requires fast loading.
4.8.1.1.2.a.7:
This surveillance requirement was added to replace the requirement to start i the diesels at least 5 times given a certain pressure in the air start i receivers. Since the volume of the air start ~ receivers is fixed and the capability to start the diesels at least 5 times without compressor assistance is demonstrated during preoperational testing, verifying that adequate pressure is available should be sufficient to demonstrate the j capability to start the diesels. In addition, diesel generator start time
! is verified on at least monthly intervals. This revision reduces the
! required testing while maintaining the intent of the specification.
4 i
i
. _ . - . _ _ _ . - - - - . , _ .. __ .-_.__.__,__,.__...-.,__...r... ,_..-...,_-,..,,m,,, _.
O's ,, 4.8.1.1.2.b:
There is no engine-mounted fuel tank.
4.8.1.1.2.c through 4.8.1.1.2.e:
The fuel oil sampling program has been revised based on a commitment to the NRC as stated in the response to Question 430.9 of the VEGP FSAR. This program is consistent with what has already been implemented at the McGuire Nuclear Station.
4.8.1.1.2.f:
See the justification provided for 4.8.1.1.2.a.5.
4.8.1.1.2.q, Footnote #:
This footnote was revised for clarification.
d.8.1.1.2.g.1:
The words "its manufacturer's" were deleted on the basis that GPC is
() committed to the TDI Owners Group Program rather than relying solely on the manufacturer's maintenance recommendations.
4.8.1.1.2.g.2:
The word " diesel" was inserted in the first line to clarify that this test is intended to verify that the engine governor is operating properly.
The value of 484 rpm is based on the TDI Owners Group Design Review and Quality Revalidation Report for VEGP. An analysis of engine operation indicated a need to limit engine speed to less than or equal to 484 rpm to avoid critical speed resonance at 496 rpm.
4.8.1.1.2.g.3:
Again, the word " diesel" was inserted for clarification.
4.8.1.1.2.c.6.c:
The words "upon loss of voltage on the emergency bus concurrent with a safety injection actuation signal" were deleted on the basis that they are redundant to the first line of 4.8.1.1.2.g.6.
O rv - - -+v-
- ei - - - ,-'* * -
< -y -
N M dsty 4.8.1.1.2.g.7:
This surveillance has been revised to reflect a band for loading based on
.the reasoning presented in the discussion of revisions to Surveillance Requirement 4.8.1.1.2.a.5. In this case, however,.it is desirable to test as close to the 2-hour rating as possible without overloading the diesel.
For this reason, a target value of 7650 kW has been proposed for the 2-hour portion of this test.
The words "within ( ) seconds after the start signal; the steady-state 1 generator voltage and frequency shall be maintained within these limits" ;
were deleted on the basis that there is no requirement to have a separate
' start to 'cegin the 24-hour test. This test could be perfcrmed in conjunction with another test which has already started the diesel and required that start time, voltage, etc. be verified. The footnote ** was added since the diesel generator is paralleled with the grid. Therefore, i disturbances attributable to the grid should not invalidate the 24-hour ,
test.
j The reference surveillance to be performed within 5 minutes of the 24-hour test was revised to 4.8.1.1.2.g.4.b on the basis that this test requires-a
, higher loading for the diesel generator and has less impact on the plant than repeating the ESF actuatier, test signal and loss of offsite power start i test. The intent of the surveillance remains unchanged, ie., the diesel is j capable of emergency start after the 24-hour test. ,
iO j 4.8.1.1.2.g.8:
4 The 2000-hour rating for the VEGP diesels is 7000 kW. However, the maximum
) auto-connected load is 6032 kW. Therefore, a test value of 6100 kW will ensure that the auto-connected loads are within-design limits.
1 4.8.1.1.2.g.11:
i There is no engine-mounted fuel tank.
4.8.1.1.2.g.13 (STS):
4
! The VEGP decign is such that the diesel is prevented from starting any time l the turning gear (barring device) is engaged or the emergency stop is actuated.
! d,g,1,1,2,g,14:
)
j This item was deleted as previously discussed under 4.8.1.1.2.a.7. l l0 i
4.8.1.1.2.h (STS):
This surveillance requirement was deleted on the basis that starting both diesel generators simultaneously is not a test designed to demonstrate freedom from interdependencies. Design controls are in place to prevent interdependencies at the design and construction stage and controls will continue to be enforced during the operations phase to preclude modifications which could result in interdependencies.
4.8.1.1.3:
This reporting requirement has been revised to an annual diesel generator reliability data report. The information proposed for this report is based on Generic Letter 84-15 and is contained in Section 6.9 of the Technical Specifications. This revision is also consistent with the intent of Generic Letter 83-43.
Table 4.8-1:
The proposed test schedule bases accelerated test requirements only on the number of failures in the last 20 valid tests. Basing the test frequency on failures during the previous 100 valid tests is deleted because it is not an accurate measure of the reliability of the diesel. Nearly 2 years are required to complete 100 tests at the accelerated test frequency. Thus, an isolated problem which caused several test failures before being corrected
~
could result in accelerated testing for a 2 year period. This could be excessive testing which is counter to vendor and industry recommendations for optimizing diesel generator reliability. The proposed test schedule is consistent with the intent of Generic Letter 84-15. In addition, footnote
was revised to be consistent with the fact that the 184-day testing requirements are listed under 4.8.1.1.2.f in our markup.
3.8.1.2.b:
There is no engine-mounted fuel tank.
The action statement was revised to reference Specification 3.4.9.3 for pressure relief capability for the RCS. The words "depressurize and vent" were deleted to allow credit for the RHR suction relief valves. The RHR suction isolation valves can be manually opened in the event of a loss of power.
4.8.1.2: .
Surveillance Requirement 4.8.1.1.3 was deleted to be consistent with the fact that 4.8.1.1.3 was deleted from Specification 3/4.8.1.1. See the justification provided for the deletion of 4.8.1.1.3.
O l
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(- 3.8.2.1 a & b and Action Statements:
This limiting condition for operation and the action statements were revised to reflect the plant-specific design for the 125 V-dc system. See subsection 8.3.2 of the VEGP FSAR for a system description.
4.8.2.1.b.2 & 4.8.2.1.c.3:
The limit on connection resistance was revised to "20 percent over the average as measured during the plant preoperational tests" on the basis that this is consistent with the recommendations of IEEE Standard 450.
4.8.2.1.C.4:
This surveillance has been revised to reflect the VEGP design.
Table 4.S-2:
The limit on float voltage was revised from 2.07 V to 2.10 V on vendor recommendations. The specific gravity was revised to reflect the VEGP specific manufacturer's full charge specific gravity of 1.210.
3/4.8.2.2:
This specification has been revised so that it is applicable to the VEGP specific design for the 125 V-dc system. See subsection 8.3.2 of the VEGP FSAR for a system description.
The action statement was revised to reference Specification 3.4.9.3 for pressure relief capability for the RCS as discussed in the justification for 3.8.1.2.b.
3/1.8.3.1 and 3/4.8.3.2:
These specifications were revised to reflect the VEGP specific design in as clear and straightforward a manner as possible. See section 8.3 of the VEGP FSAR for a system description.
The action statement of 3/4.8.3.2 was revised to reference Specification 3.4.9.3 for pressure relief capability for the RCS as discussed in the justi fication for 3.8.1.2.b.
3/4.8.4.1 (STS):
This section is not applicable to VEGP based on tie fact that all the containment electrical penetration and penetration conductors will be O protected by demonstrating the operability of prinary and backup overcurrent protection circuit breakers during periodic surveellance, as required by Specification 3/4.8.4.1.
3.8.4.1:
This specification has been revised to include-the feeder breakers to the isolation transformers between 480 V Class 1E busses and non-Class 1E equipment as discussed in Section 8.4.5 of the VEGP SER.
! Table 3.8-1 was deleted on the basis that such a lengthy listing does not provide useful information to the operator in ensuring Technical Specification compliance. This listing can be maintained in plant procedures without affecting the intent of the Technical Specifications.
Action Statement a. has been revised to reflect the fact that the inoperable circuit breaker or protective device will be racked out in addition to tripping the backup breaker. Therefore, a 31-day surveillance interval should be sufficient to ensure that the affected circuit (s) remain deenergized until the protective device (s) are restored to operable status.
This is consistent with other action statements or LCO's which require that equipment be rendered inoperable during certain phases of plant operation j
(reference Surveillance Requirement 4.5.3.2, for example).
action Statement b has been added to reflect the STS action of 7-day surveillance for those circuit breakers or protective devices where it may not be desirable to trip the backup breaker in addition to racking out the circuit breaker or protective device. For example, if the backup circuit breaker feeds more than one circuit breaker or protective device, it would 4 not be desirable to deenergize all the circuits associated with the backup
. . breaker.
The above revisions provide additional flexibility without altering the intent of the specification.
4.3.4.1:
There are only four circuit breakers which fall into the category of 4-15 kV 1
breakers. These are the 13.8 kV breakers for the reactor coolant pumps.
Therefore, 4.8.4.2.a nas been revised to reflect this. See figure 8.3.1-1
~
sheet 1, of the VEGP FSAR. In addition, the feeder breakers to isolation i transforners between 480 V Class 1E busses and non-Class 1E equipment have i
been included as discussed in Section 8.4.5 of the VEGP SER. Surveillance Requirement 4.8.4.2.a.2 has been revised to be less prescriptive on the basis that different breakers are tested by different methods. Therefore, such a prescriptive statement as contained in the Standard Technical l Specifications is not practical for all low voltage circuit breakers. In general, the lower voltage circuit breakers will be tested in a more conservative fashion than called for in the Standard Technical Specifications.
O
La_.-
D.
'Q 3/4.8.4.2:
This specification was revised to reflect the fact that prior to core
loading and during plant operation, the thermal overload relay trip contacts for all of the Class 1E valves are permanently bypassed with jumpers, in accordance with Regulatory Guide 1.106. See paragraph 8.3.1.1.2.k.5 of the VEGP FSAR.
) Table 3.8-2 was deleted on the basis that this listing does not provide information which is essential for the operator in ensuring compliance with Tecnnical Specifications. This list can be maintained in procedures without affecting the intent of the specification.
i 1
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l 3
0051v t
.%g L 6+ .
!- e 3/4.9 REFUELING OPERATIONS t
i 1 3/4.9.1 BORON CONCENTRATION .
+
i
/ LIMITING CONDITION FOR OPERATION i
3.9.1 The boron concentration of all filled portions of the Reactor Coolant i System and the refueling canal shall be maintained uniform and sufficient tc
'- ensure that the more restrictive of the following reactivity conditions is met; i either:
a. A K,ff of 0.95 or less, or
b. A boron concentration of greater than or equal to,[2000[ppe.
APPLICABILITY: MODE 6.
j ACTION:
i With the requirements of the above specification not satisfied, immediately suspend all operations involving CORE ALTERATIONS or positive reactivity '
3 changes and initiate and continue boration at greater than or equal to & gpm
! of a solution containing greater than or equal to70oo ppm boron or its !
j' equivalent until K is reduced to less than or equal to 0.95 or.the boron - l l
concentration is rINored to greater than or equal to [2000] ppe, whichever is !
j the more restrictive. ;
a ,
i I SURVEILLANCE REQUIREMENTS I
k*
l ' 9.1.1 The e e r::trictive of the deve tr ce rt!vity c rditie- thM '
j dete + ed ;-fe- te-l 3. %-"i eg e r u-bel ti g t% ----te- "- ee' k--t e-d j
b. utggg7:g7 ef 37 y;7s_ L ,y ert.37 7 3 4. e.,g ef 3 frg <72 ,
it; fully it:rted p=f tfr eith*- tM rerte r :-'
I 4.9.1.2 The boron concentration of the Reactor Coolant System and the refueling canal shall be determined by chemical analysis at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .
2 n Ac t -u +-rrr,, IM1 % .m, a l-uq -ItJ, a nd iW-0% -i% t 1 4.9.1.+ Valves {-1501446 vi interated ;ter ; w n

] shall be verified closed and secured in position by mechanical stops er by - r:r! ef M er e! r t-f r' ,
powee at least once per 31 days.
l
'% r-te- e M , be --!- -t =e <- m E s 2:= = f=, i: er ee ==tr
' frcM uth th= ueese! Med c!ceu e belte !ece th e fuM y t = ' n d r rith r u.a ___..;
. u_
V0GTLE - UNIT 1 3/4 9-1 l r;
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() REFUELING OPERATIONS 3/4.9.2 INSTRUMENTATION LIMITING CONDITION FOR OPERATICN 3.9.2 As a minimum, two Source Range Neutron Flux Monitors shall be OPERA 3LE, each with conticuous visual indication in the control room and cce with audibic indication in the co,tainment and control roco.
AFPLICA3ILITY: MODE 6.
ACTION:
a. With one of the above reoujred ronit ?s insterable or not operating, f errediately suspatrJ all cperations involving CORE ALTERATIIXIS or pcsitive retctivity changes. ,
b. With both of the above required conitors inoperable ce nat operating, '
deternine the boron cer.contration of the Resctor Coolant System at least or.ce per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .
SURVEILLANCE FEQUIREMENTS 4.9.2 Each Scurce Range Neutron Flux Monitor shall be demonstrated OPERABLE cy perforcance of:
a. A CHANNEL CHECK at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ,
b. An ANALCG CHANNEL OPERATIONAL TEST within 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months prior to the initial start cf CORE ALTERATIONS, and
c. A3 ANALOG CHANNEL GFERATIONAL TEST at least once per 7 days.
()
VCGTLE - UNIT 1 3/4 9-2 i
P REFUELING O#EEATIC#5 3/4.9.3 DECAY T1PE LIMIT i.%. '>3ITION FOR CFERATION 3,9.3 The reactor shall be sue:ritical for at least'100 bcurs.
APPLICABILITY: Ouring movement of irradiated fuel in the reactor vessel.
ACTION:
With the taacter subtritical for less than 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months . .suscend all operations involving movement of f rradiated fuel in the rasctor vessel.
i
(~)
s_-
SURVEILLANCE REOUIREMENTS 1
4.9.3 The reactor shall be determined to have been subcritical for at least 100 hcurs by verification of the date ar.d time of subtriticality prior to ;
movement of irradiated fuel in the reactor vessel. ,
i P
l
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VCGTLE - UNIT 1 3/4 9-3
c i
_ _ _ __ .,_,_____m ._ -_ . _ _ _ . . . , _ _ _ . . . . . _ _ ~ , _ _ . _ _ _ , _ _ . - . , _ _ . - . _ _ _ , _ _ _ , _ _ . _ . , _ _ . , . . . . _ . - . _ _ . - , _ _ _ _ . . . _ _ . .
s
[s_ j REFUELING OPEAATIONS 3/c.9.4 CONT A I NMENTJU It 31 *G F EN ETPA'*? C N S L'MITING_t,0f0iTION FOR CPSRATION _ _ _ _ _ _ ,
?,9. 4 The cor.tainnert building penetratiens shall be in the followfog status:
a. The equipaar.t ccer .closec and held in placa by a minimum cf four co;ts,
c. A air. imam cf one doct 10 each airlock is closec, and
c. Eacn peneteatfun previding direct accaas from the contair. ment atmosphere to the cutsids etnespher.e shal? be either:
1) Ciosed cy en isolation valve, olind flange, or manual valve, or
2) De capacle of beirg cicsed by an GPERAELE automatic containment purgr r+eetrtret isolaticn valve. <
r? .rr . x. ,..,
AoPLICA$!LITY: Caring CORE ALTERATI0h5 cr movement of irradiated fuel within the cce tainr.er.t.
ACTION:
With the recuirenents of the abcve specification ,1st satisfied, immediately suscer.d all cperatter.s involving CCRI ALTEFATIONS er novament of irradiated fuel in ite contair.sent building.
l SURVE:LLANCE RECUIREMEN'TS
~ ~ '
varn O 4 E e n 4.9.4 Each of tha above reautred ccattircent bui7 ding penetrations shall be ceta.-mired to ce eitner in its closec/ isolated cor.ditten or captble of being closed by an OPiMELE autonatic certa'rme.9t pwge anr. exhestt isolation ualve within 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months prior to the start of anc at least once per 7 days during CORE ALT 5RATICNS or r.ovement of irradiated fuel in tre containnent builcirg ty:
a. Verifyira the penstrations are 1.1 trtle clcsed/fsciated conditior.,
or ,
sass 1**.:n
b. Testing the ccntainnent une n-. td-e=haust isolatien valsas per the a;olicable portices of Specification 4.6.4,2.
3 VG3TLE - LHIT 1 -
3/4 9-4 ,
,Q '
! i _,/ REFUELING OPERATIONS 3/4.9.6 CCMHJNIC4flCNS LIMI~ING CONCITION FOR OPERATICH 3.9,5 Direct conmunicxtions snall ha r.aintained bst etn the contr.ol room and '
psesonnel at the refueling station.
APPLICABILITV: Duri..g 10.iE ALTERATIONS.
i ACTION: .-
i hhan direct commun#citions between 1.5e control room and personnal at the refueling station cacnot be maintaired, suspend all CORE ALTERATION 3. ,
i t
Sth'.*11LL ANT.E REQUIREMENTS _ ,
i 4.S.5 Direct ccmcuaicatiens between the control room and perscnne? at the !
refuelirg station stall be demonstrated within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months prior to tne start of and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during CORE ALTER 4TIch5.
6 V0GTLE 'JNI? I 3/4 9-5
1. 1
/
l REFUELING OPERATICh5 _
j [
,\ /
$ 3/4.9.6 MANIPULATOR CRANE /
i i J
LIMITING C093ITION FOR OPERATI21 s
1 ,
i 3.9.6 The franipulator crane and auxiliary doist sna11 be .used for snovenent of i crive rcos er fuel assetablies and shall te OPERA 3tE with: ,
I a. The canipulator crane Jsed for movement :f fuel assamtlies taving:
1
1) ' A minimum canacity of (2750] pcuoes, and O An'cVerload cutcff limit less *,han or equal to [27002 pounds.
1
. t. The smilfary toist used for latching am: unlatching drive rods hawir.g:
j
1) A minixcm capacity of (610] pounds,-and i I 2) A load indicator whien shall be used to prevent lif ting loads ,
j in excess of (6003 pounds.
i APPLICA*ILI1Y: During movenent of drhe rods or fuei assemblies within !
tne reacter vessel. \ !
I ACTIDf: .
i r c
l Mith the requirements fcr crane and/or' hoist OPERABILI71 not satisfied, suspend ;
j use of any inoperaala canipulator crane ec/or auxiliary toist frcn operatiens
invciving the a

ovament of crive rots anc fuel assenultes witnin tne reactac !
vessel, t
0
.l SURVEILLANCE REOUIREEDeTS j l _
)
v s t
i 4. 9. 6.1 Eac'1 manip:Jiator crane used for ecvement of fim1 assemb'Ifes withio
! the reactor vessel shalt te cemonstratad OPEF.AELE within 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months prior to !
l the start af such eperations by cerf:rming a load test cf at tesst (17EC] pounds 4
.and demcnstrating an autoestic Isad cutoff when thra crane load exceeds '
(2.'33] po.;nds. I i (
i 4.9.6.2 Each auiliary Moist .and associated lead inafcatar used for.naverent i i of drive reds within the reactor vessel shall de eneanstrated OPERASLE within i l 130 hours0.0015 days 0.0361 hours 2.149471e-4 weeks 4.9465e-5 months prier to the start of such operations by perfo win 0 a load test of ,
j at least (610] pour.ds. ;
i 1
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} V0GTLE - UNIT 1 3/49-6 l 4
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!,,,_. _ _ _ __.._ ,-.._ _ _. _ ,.._._,. _ _. _ _,_ _ _ ~. . . - . - , _ . . . _ , . . . _ . . . _ , . . . . . . _ .
a>W-N REFUEL 1?tG OPERATION 5 p
v 1/ J .1. 6 EUl'!Li!JG MACHINF LIN_ITlhG C01)lTICV FOR OPERATION _
3.9.6 The refueling eachine and auxiliary hoist shall be used for movenent of drive rods or fuel assetDlies and shall be OPERABLEwitn:
a. The refueling machir.e used for novement of fuel assem31ies having:
2. A design rated load of 3165 pounds on the hoist (19ef pour.ds an the gripper),
2. Automatic cutoffs Wit 5 the following setpoints relative to the suspende.1 weight of the assen>ly in water:
a. Pricary overload is.
1. Plus 250 ;ounds fer wet conditions.
7. Flcs 350 pounds for dry conditions.
9. Secondary overload is:
1. Plus 150 pounds above primary overloed.
c. Load POduction 15:
1. Minus 250 pounds for both wet and dry conditions.
() b. ine oJxiliary hoist uted for latchir.g and unlatching drive r0ds having:
A 5.inimun capacity of 2000 pounds, and 1.
2. A 1000 pov d (minimun} load indicator which will be used to acritor lift 199 inads for ttese oseratices.
A 7 P'.1C A B 'L I TY : QJefnq movement of drive rods or fuel assemblies within the.re3CL1r pressure vessei.
ACTl's -
'nith the re ;J. . ementi for tt,e refieling 63 chine and/or auxiliary bois; CPER/5!LITY t ot titisfied, suspend use of cny vnopersole refueling machine and/or auxiliary hoist from c;e*Jtions inv 1ving the movement of fuel assecolies and/cr drive reds w i t *.T , the reacter cres:ure vessel, The provisions of Specification 3.0.3 are not inchcrble. ,
SURVEILLANLC FECalEEMENTS __ _,_
4.9.5.1 Each refueline r.s-htca used fcr movement of (nel assemblies within the
+: tor oresture vessel shall be demonstrated CPERAGit within 103 hours0.00119 days 0.0286 hours 1.703042e-4 weeks 3.91915e-5 months prior :o start of such operation; by terforming a Toad tett of it least 125 of the design rit+] Itad and Li demons'.rarleg an eutcnatic icod catcff when tFe refueling nachino lovi er teecs the setpoints of SCAct fication 3.9.6.a.2.
4.9.6 2 Eacn auxMlary hoist ar.d asasciated load indicatcr used for movement of L
drive rsd .ithtn the eattor cressure vessel shall be demcr.strated OPERABLE within Mc hni.rs c riar to thf start of such operaticos by perfornirg a load test of at lea o Izte counds, -
'Joirit-lWIP 1 3/4 9-6
j w . 41 I
j 1 L !
1
} REFUELIM OPERATIONS
! 2 /4. 9.7 CRANE TRAVEL - SPENT FUEL STORAGE AREA 5 i
1 L
LIMIT

NG_ C00IT10N FOR OPERATION _ _
=
j l 3.5.7 Loads e excess of/146) pounds shall be prohibited frert travel over 4
fuel usemolics in the stoHge pool. ,
I APOLICA5ILITY: f 1
kith,vfuel assembifes in the storage pool.
i i
ACTTON:
irrA h'A *dd- .
?
! , a. With the requirements of the above specification not satisfied, place :
I the cra e load in a sMe condition. !
i ,
! b. The provisiens of Specifications 3.0.3 and 3.0.4 are not applicable. i
! i i
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! r lO
,1 SURVEILL4NCE REQUIREMENTS i
e i,
4., 9. 7 Crane intericeks and physical stops which prevent crane travel with j loads ir. ixcess of/Lsrerlpounds over fuel assemblies shall be denonstrated '
OPERABLE W.hin 7 day

n price to crane use and at leas: once per 7 days i thereafter during crane operation. j i
1 1
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4 i V0GTs.1 - 11 NIT 1 3/4 9-7 I
i._......__.___-_._.__.__._ _ _._._. _ ___ _
J O
(s-) REFUELING OPERATIONS 3/4.9.8 FESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION 1
hISH 'sATER LEVEL LIMITI!iG CONDITICN FCa OPERATISH
< e s. n 3.9.3.1 At least one residual beat removal (RHR) Toep shall be OPERASLE and la operation."
ApoLICASILITY: MCDE 6, when the water lesel above the top of the reactor vessel fiaige is greater than or equal to 23 feet.
ACTICN:
+ -+ n Wit 5 no RHR lees OPERABLE anc in cperation, suspeed all operations involving an increase in the reactot deca / heat load or a reduction in boron concentration of the Reactor Coolant System and imnediately inf tiate corrective action to ritorn tr.e re.1uired RHR ic+p to 0?ERABLE and operating status as soon as possib1e. Cicst all certainment penetrations p-oviding direct access from
tre containment atmospneret to the cutside atmosphere within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .
(
+ .< x . t.
SURVEILLANCE REGUIFEMENTS 1
tr.ur, 4.9.8.1 At least one RHR Fece shall be verified in ope *ation and circulating >
reactor coolant at a flow rate of greater than or equal to (2800] gpm at least once per 12 hcurs.
i 9
i
< . . .s . n "The RPR loop may be removed from operation for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months per 8-hour period ,
during tna performance of CORE ALTERATICNS in the vicinity of the reactor vessel not legs.
{}
V0GTLE - UNIT 1 3/4 9-8 1
I f t
= !
REFUELIf*O CPEAATIONS i
i I LOW WATER LEVEL 1 ;
I i
i LIMITING CONDITION FOR OPERATION :
f
+ as.i n s 3.9.8.2 Two independent residual heat removal (RHR) toops shall be OPERABLE, t
! and at least one RHR leep shall be in speration.* {
re .e
APDLICABILITY: MOCE 6, wnen the water level above the top of the reactor 6 l vessei flange is less than 23 feet. j ACTION: '.p g i e i & ns i a. With less than the required RNR leoes OPERABLE, ,immediately initiate [
l corrective action to return the required RHR loops to OPERABLE <
) status, or to establish greater than or equal to 23 feet of water +
j above the rea:: tor vessel flange, as soon as possible. I i
w ;n 3 2 b. With no RHR loop in operation, suspend all operations involving a i l
reduction in boron concentration of the Reactor Coolant System and j
! imediately initiate corrective action to return the required RHR i
} + 4 leep to operation. Close all containment penetrations providing ;
direct access from the containment atmosphere to the outside !
atmosphere within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . l t !
I SURVEILLANCE REOUIREMENTS l l
h . tin I 4.9.8.2 At least one RHR leep shall be verified in cperation and circulating l reactor coolant at a ficw rate of greater than or equal to [20003 ppm .st {
l
-j least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . (44&J ;.
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q
Priortoinitial[ criticality,theRHRloopmayberemovedfromoperationfor i
)
op to I hour per+ hour period during the performance of CCRE ALTERATIONS in ;
i the vicinity of the reactor vessel hot legs. [
j '
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V0GTLE - UNIT 1 3/4 9-9 l i
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i
REFUELING OPERATIONS VdMnLAno "m
d 3/4.9.9 CONTAINMENT PU^0" evumcT ISOLATION SYSTEM LIMITING CONDITION FOR OPERATION i '/a s i m'en 3.9.9 The Containment Pbewe anu Eahesit Isolation System shall be OPERABLE.
APPLICABILITY: During CORE ALTERATIONS or movement of irradiated fuel within tne containment. i ACTION:
' lad; i. uric r.
a. With the Containment P rge ;.M C,,hesst Isolation System inoperable, i
close each of the pur;;: 3-d exhattet penetrations providing direct access from the containment atmosphere to the outside atmosphere.
terillulon i b. The provisions of Specificaticns 3.0.3 and 3.0.4 are not applicable.
4
(
'O SURVEILLANCE REQUIREMENTS i */.2&',1. uric e,
! 4.9.9 The Containment PL.ge =d Ed aust Isolation System shall be demonstrated CPERABLE within 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months prior to the start of and at least once per 7 days during CORE ALTERATIONS by verifying that containment puiv. .ua -ahewit isolation
' occurs on-nanueMattiet4mutnd on a High Radiation test signa 1Tfrom each of
the containment radiation monitoring instrumentation channels. 4, .; 7, l
i i
O
! V0GTLE - UNIT 1 3/4 9-10
REFUELING OPERATIONS 3/4.9.10 WATER LEVEL - REACTOR VESSEL L MITING CONDITION FOR OPERATION -
\
\
3.9.10 \At least 23 feet of water shall be maintained over the to of the reactor' vessel flange.
i APPLICABILITY: During movement of fuel assemblies or control ods within the containment when either the fuel assemblies being moved or e fuel assemblies seated within the reactor vessel are irradiated while.in E 6.
ACTION: '
N, With the requirement's,of the abovt specification not satisfied, suspend all operationsinvolvingmovementoffuelassembliesor,!controlrodswithinthe reactor vessel. x <
s O SURVEILLANCE REQUIREMENTS \//
4.9.10 The water level shall be/ determined to be at least its minimum required depth within 2 tours prior to the start of and at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter during movement offfuel assembifes or control rods.
\-s N'\
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.V0GTLE - UNIT 1 3/4 9-11
REFUELING OPERATIONS 3/4.9.10 WATER LEVEL - REACTOR VESSEL FUEL ASSEMBLIES LIMITING CONDITION FOR OPERATION 3.9.10.1 At least 23 feet of water shall be maintained over the top of the reactor vessel flange.
APPLICABILITY: During movement of fuel assemblies within the containment when the fuel assemblies being moved are irradiated.
ACTION With the requirements of the above specification not satisfied, suspend all operations involving movement of fuel assemblies within the reactor vessel.
SURVEILLANCE REQUIREMENTS f 4.9.10.1 The water level shall be determined to be at least its minimum required depth witnin 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months prior to the start of and at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months there-after during movement of fuel assemblies.
V0GTLE - UNIT 1 3/4 9-11
REFUELING OPERATIONS WATER LEVEL- REACTOR VESSEL ,
CONTROL RODS LIMITING CONDITION FOR OPERATION :
3.9.10.2 At least 23 feet of water shall be maintained over the top of the irradiated f0el assemblies within the reactor pressure vessel.
APPLICABILITY: During movement of control rods within the reactor pressure vessel while in MODE 6.
4 ACTION: . .
With the requirement of the above specification not satisfied, suspend all operations involving movement of control rods within the pressure vessel.
SURVEILLANCE REQUIREMENTS O-4.9.10.2 The water level shall be determined to be at least its minimum required depth within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months prior to the start of and at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter during movement of control rods within the reactor vessel.
k lO 1
V0GTLE - UNIT 1 3/4 9-12 1
() REFUELING OPERATIONS 3/4.9.11 WATER LEVEL - STORAGE POOL LIMITING CONDITION FOR OPERATION 3.9.11 At least 23 feet of water shall be maintained over the top of irradiated fuel assemblies seated in the storage racks.
APPLICABILITY: When6ver irradiated fuel assemblies are in the storage pool.
ACTION:
f a. With the requirements of the above specification not satisfied,
'j suspend all movement of fuel assemblies and crane operations w+th l 8
Ef "0 P3e! yes.spc.c,#Fivods ir th; ';;i sterage eree; and restore the water level to within ?
5' its limit within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .
b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
O SURVEILLANCE REQUIREMENTS 4.9.11 The water level in the storage pool shall be determined to be at least its minimum required depth at least once per 7 days when irradiated fuel
~
assemblies are in the fuel storage pool.
i 7
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V0GTLE - UNIT 1 3/4 9-it
l REFUELING OPERATIONS i 3/4.9.12 FUEL STORAGE POOL AIR CLEANUP SYSTEM LIMITING CONDITION FOR OPERATION ,
i \
3.9.12 Tw(independentFuelStoragePoolAirCleanupSystemsshallbeOPERABLE.
APPLICABILITY) Whenever irradiated fuel is in the storage pooI.
ACTION:
' \
- a. With one Fuel Storage Pool Air Cleanup Syste inoperable, fuel
- movement within the storage pool or crane operation with loads over j the storage pool may proceed provided the,0PERABLE Fuel Storage Pool Air Cleanup Sys' tem is capable of being powered from an OPERABLE emergency power source and is in operation and discharging through at least one train of HEPA filters and' charcoal adsorbers.
b. With no Fuel Storage Pool Air Cleanu/p System OPERABLE, suspend all operations involving movement of , fuel within the storage pool or crane operation with loads over the storage pool until at least one '
Fuel Storage Pool Air Cleanup, System is restored to OPERABLE status.
N ,'
! c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
l \
/ \
SURVEILLANCE REQUIREMENTS /
4.9.12 TheaboverequiredFueIStoragePoolA'irCleanupSystemsshallbe demonstrated OPERABLE:
\
a. At least once per 31 days on a STAGGERE TEST BASIS by initiating, j frem the control room, flow through the HEPA filters and charcoal adsorbers and verifying that the system operates for at least 10 continuo'us hours with the heaters operating;
,- g
b. At least'once per 18 months or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings,'or (2) folicwing painting, fire, or chemical release in any var.tilation zone communicating with the system by: N p 'N
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O l V0GTLE - UNIT 1 3/4 9-13 ,
a
~
REFUELING OPERATIONS
$URVEILLANCE REQUIREMENTS (Continued) 1
\
\
\ 1) s Verifying that the cleanup system satisfies the in plac penetration and bypass leakage testing acceptance cri,t'eria of less than [*]% and uses the test procedure guidance in
\ Regulatory Positions C.S.a. C.5.c, and C.S.d of Reg'ulatory Guide 1.52, Revision 2, March 1978, and the system flow rate is cfm 10%; ,-
2) Verifying, within 31 days after removal; that a laboratory analysis of a representative carbon sample,,obtained in accor-dance with Regulatory Position C.6.b of Regulatory Guide 1.52,
- Revision 2, March 1978, meets the laboratory testing criteria
.of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978, for a methyl iodide penetration of less than [**]%;
and
3) Verifying a system flow rate of cfm i 10% during system operation when tested in accordance with ANSI N510-1975.
c. After every 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of charcoal adsorber operation by verifying, within 31 days after removal, that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory O Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978, for a methyl iodida penetration of less than [**]%.
d. At least once ;er 18 months by:
1) Verifying that the pressu e drop across the combined HEPA
~
filters and charcoal adsorber banksiis less than [6] inches Water Gauge while operating the system at a flow rate of cfm i 10%,
2) Verifying that on a High Radiation test signal, the system automatically starts (unless already operating) and directs its exhaust flow through the HEPA filters and charcoal adsorber banks, N
\
. 'N s
!O V0GTLE - UNIT 1 3/4 9-14 L _ __ ___- - - . - --- . - - - . . . . - - . . - -._. - -. _ , . . . . - -
.. a REFUELING OPERATIONS S VEILLANCE REQUIREMENTS (Continued)
,i
3) Verifying that the system maintains the spent fuel stora'ge pool area at a negative pressure of greater than or equal to [1/4]
inch Water Gauge relative to the outside atmosphere during system J
operation,
4) Verifying that the filter cooling bypass valves.can be manually opened, and ,
~.
Veri'fying that the heaters dissipate
5) kW when tested,in accordance with ANSI N510-1975. -
e. After each complete or partial replacement of a HEPA filter bank, by verifying that 'the cleanup system sathfies the in place penetration and bypass leakage testing acceptance criteria of less than [*]% in accordance with ANSI N510-1975 for a 00P test aerosol while operating the system at a flow. rate of cfm i 10%.
f. Aftereachcompleteobpartialreplacementofacharcoaladsorber bank, by verifying that'the cleanup system satisfies the in place penetration and bypass leakage testing acceptance criteria of less O
V than [*]% in accordance with ANSI N510-1975 for a halogenated hydrocarbon refrigerant testsgas while operating the system at a flow rate of cfm i 10%.N \
s
,i
0.05% value applicable when a HEPA filter or charcoal adsorber efficiency of 99% is assumed, or 1% when a HEPA filter or charcoal adsorber efficiency of 95% of less is assumed in the NRC staff's safety evaluation. (Use the i
value assumed for the charcoal adsorber efficiency if.the value for the HEPA filter is different from the charcoal adsorber efficiency in the NRC staff's safety evaluation).
Value applicable will be determined by the following equation:
P= 00 5 E , when P equals the value to be used in the test requirement
(%), E is efficiency assumed in the SER for methyl iodide removal (%),
and SF is the safety factor to account for charcoal degradation between tests (5 for systems with heaters and 7 for systems without heaters).
l 'N l \
x 7
\
O V0GTLE - UNIT 1 3/4 9-15 .
w qbp b
() JUSTIFICATIONS FOR DEVIATIONS FROM STS
SECTION 3/4.9 i
3/4.9.1:
\'
Footnote
was deleted as redundant since it met ely repeats the definition i for Mode 6 Surveillance Requirement 4.9.1.1 was deleted since this requirement is
' provided within the definition of Mode 6 which must be met prior to entry into Mode 6 as required by Specification 3.0.4.
This is consisteat with the WOG "Short-Term Tech. Spec. Improvements" which were prioritized as a " Staff Action" item during the January 29, 1986 i meeting between the WOG Tech. Spec. Subcommittee and the NRC Staff.
4.9.1.2:
Reference to air and electrical power was deleted since the valves in question are manual valves.
i
! 3/4.9.4 & 3/4.9.9:
(
I
" Purge and Exhaust" was replaced with " Ventilation" to reflect plant-specific terminology. See subsection 6.2.4 of the VEGP FSAR.
! 3/4.9.6: .
i t The Generic STS Specification 3/4.9.6 (Manipulator Crane) was deleted in its l entirety. A new Specification 3/4.9.6 (Refueling Machine) sas inserted in its place. The new specification reflects the installed equipment in Plant Vogtle. See subsection 9.1.4 of the VEGP FSAR.
I 3.9.7:
The addition of the word " irradiated" reflects the intent of this specification.
3.9.8.1:
The change reflects the VEGP specific terminology " Residual Heat Removal
! Train" instead of " Residual Heat Removal Loop." See subsection 5.4.7 of the l VEGP FSAR.
O
_ ~ . . . . . . _ . . . . _ . _ , _ _ _ _ _ _ _ . _ _ _ . . . . _ _ _ . _ , , _ _ _ _ . , , , _ , _ , . . _ _ , _ _ , . _
~
1 f 3.9.8.2:
1 4 The change reflects the VEGP specific terminology " Residual Heat Removal Train" instead of " Residual Heat Removal Loop." See subsection 5.4.7 of the j VEGP FSAR.
l This revision has been approved by the NRC Staff at a recent operating plant i and is consistent with the WOG short-term recommendations as discussed in the Jauuary 29, 1986 meeting between the WOG Tech. Spec. Subcommittee and
l. . the NRC Staff. This item was prioritized as a " Staff Action" item at that meeting.
The revision of footnote
allowing the RHR train to be out of operation for-
> 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months out of every 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months (initial core load) was ma 6 on the basis that, since there is ne decay heat involved, there should be no problem with boron l stratification or precipitation. Also, there is no likelihood or a boron dilution transient as discussed in 15.4.6 of the VEGP FSAR. .
l!
3/4.9.9:
This specification was revised to reflect VEGP specific terminology and the fact that the design does not include a specific hand-switch for manual initiation. See subsection 6.2.4 of the VEGP FSAR.
3/4.9.10:
T
{ This specification was split into two specifications td facilitate the' i connection of the drive rods to the control rods. When fuel assemblies are ;
being moved, at least 23 feet of water is required above the vessel flange to provide the necessary shielding. However, when fuel assemblies are not
4. being moved but the drive rods are being connected to the control rods, 23 feet of water above the vessel flange is not necessary for proper shielding. By requiring 23 feet of water above the top of the irradiated. . t f
fuel assemblies, the necessary shielding is maintained, and the operation of l connecting the drive rods to the control rods is facilitated. This has been approved by the NRC and implemented in the Farley Nuclear Plant Technical Specifications.
{
I 3/4.9.11:
The replacement of the phrase "with loads in the fuel storage areas" with ,
"over the spent fuel pool" in this specification's action statement provides a more defined action and improves the consistency between the action and
. the assumptions used in the accident analysis.
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3/4.9.12: a j
TheVEGPisequippedwithafuelfiandlingbuildingpostaccidentexhaust . '
!. system. However,_ Chapter 15 of the SER spe~cifi'cally states that no credit-was assumed for the operation of these filters in mitigating the I consequences of a fuel handling accident in the fuel handling building.
1 Consequently, this specification has been deleted since it is not required l
to validate an assumption made 'in the safety analysis.
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$ 3/4.10 SPECIAL TEST EXCEPTIONS w
3/4.10.1 SHUTDOWN MARGIN LIMITING CONDITION FOR OPERATION 3.10.1 The SHUTDOWN MARGIN requirement of Specification 3.1.1.1 may be suspended for measurement of control rod worth and SHUTDOWN MARGIN provided reactivity equivalent to at least the highest estimated control rod worth is available for trip insertion from OPERABLE control rod (s).
APPLICABILITY: MODE 2.
ACTION:
a. With any fu"-?:ngth control rod not fully inserted and with less than the above reactivity equivalent available for trip insertion, immediately initiate and continue boration at greater than or equal to g gpm of a solution containing greater than or equal to vooo ppm boron or its equivalent until the SHUTDOWN MARGIN required by Specification 3.1.1.1 is restored.
b. With all full-lengt.5 control rods fully inserted and the reactor subcritical by less than the above reactivity equivalent, immedi-ately initiate and continue baration at greater than or equal to So gpm of a solution containing greater than or equal to oco ppm boron or its equivalent until the SHUTDOWN MARGIN required by Specification 3.1.1.1 is restored.
SURVEILLANCE REOUIREMENTS 4.10.1.1 The position of each-full-length control rod either partially or fully withdrawn shall be determined at least once per 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .
4.10.1.2 Each full-length control rod not fully inserted shall be demonstrated capable of full insertion when tripped from at least the 50% withdrawn position withinp24--heues prior to reducing the SHUTDOWN MARGIN to less than the limits of Specification 3.1.1.1.
\'7 da.y.s O
V0GTLE - UNIT 1 3/4 10-1
4 SPECIAL TEST EXCEPTIONS 3/4.10.2 GROUP HEIGHT, INSERTION, AND POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION 3.10.2 The group height, insertion, and power distribution limits of Specifications 3.1.3.1, 3.1.3.5, 3.1.3.6, 3.2.1, and 3.2.4 may be suspended during the performance of PHYSICS TESTS provided:
a. The THERMAL POWER is maintained less than or equal to 85% of RATED THERMAL POWER, and ,
b. The limits of Specifications 3.2.2 and 3.2.3 are maintained ,,
and determined at the frequencies specified in Specification 4.10.2.2 below.
APPLICABILITY: MODE 1.
ACTION:
With any of the limits of Specification 3.2.2 or 3.2.3 being exceeded while the requirements of Specifications 3.1.3.1, 3.1.3.5, 3.1.3.6, 3.2.1, and 3.2.4 are suspended, either:
Reduce THEPMAL POWER sufficient to satisfy the ACTION requirements O a.
of Specifications 3.2.2 and 3.2.3, or
b. Be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .
SURVEILLANCE-REQUIREMENTS 4.10.2.1 The THERMAL POWER shall be determined to be less than or equal to 85% of RATED THERMAL POWER at least once per hour during PHYSICS TESTS.
4.10.2.2 The requirements of the below listed specifications shall be performed at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during PHYSICS TESTS:
a. Specifications 4.2.2.2 and 4.2.2.3, and
b. Specification 4.2.3.}s.
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O V0GTLE - UNIT 1 3/4 10-2 i
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SPECIAL TEST EXCEPTIONS 3/4.10.3 PHYSICS TESTS LIMITING CONDITION FOR OPERATION 3.10.3 The limitations of Specifications 3.1.1.3, 3.1.1.4, 3.1.3.1, 3.1.3.5, and 3.1.3.6 may be suspended during the performance of PHYSICS TESTS provided:
a. The THERMAL POWER does not exceed 5% of RATED THERMAL POWER,
! b. The Reactor Trip Setpoints on the OPERABLE Intermediate and Power 25% cf P^
Range channels are set at le : th:r. Or :w:1 t:
7 a/c 2 >/ T'lEP"AL POWER, and
c. The Reactor Coolant System lowest operating loop temperature (T**9) is greater than or equal to F.
APPLICABILITY: MODE 2.
ACTION:
e. With the THERMAL POWER greater than 5% of RATED THERMAL POWER, immediately open the Reactor trip breakers.
S+l
b. With a Reactor / Coolant System operating loop temperature (T,y9)
O. less than E+ H3*F, restore T a to within its limit within
15 minutes or be in at least MT STANDBY within the next
15 minutes.
f SURVEILLANCE REOUIREMENTS 4.10.3.1 The THERMAL POWER shall be determined to be less than or equal to 5%
of RATED THERMAL POWER at least once per hour during PHYSICS TESTS.
4.10.3.2 Each Intermediate and Power Range channel shall be subjected to an ANALOG CHANNEL OPERATIONAL TEST within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months prior to initiating PHYSICS TESTS.
4.10.3.3 The Reactor Coolant System temperature (T,yg) caH be deteMned to be greater than or equal to E+ M3*F at least once per 30 minutes during PHYSICS TESTS. 6.H O
V0GTLE - UNIT 1 3/4 10-3
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SPECIAL TEST EXCEPTIONS 3/4.10.4 RhACTORCOOLANTLOOPS l
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LIMITING CONDITION FOR OPERATION
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3.10.4 The limitations of Specification 3.4.1.1 may be suspended during the '
performance of STARTUP, and PHYSICS TESTS provided: j
a. The THERMAL POWER does not exceed the P-7 Interlock'Setpoint,
, and ,
\
b. The Reactor Trip Setpoints on the OPERABLE Intermediate and Power Range channels are\ set less than or equal to 25% of RATED THERMAL
- POWER.
APPLICABILITY: During operation below the P-7 Inter}cck Setpoint.
/
l ACTION: /
WiththeTHERMALPOWERgreaterthanteP-7InterkockSetpoint,immediately open the Reactor trip breakers.
j
/
SURVEILLANCE REQUIREMENTS j \
4.10.4.1 The. THERMAL POWER shall be/ determined t Setpointatleastonceperhourdur,ingSTARTUPan%belessthanP-7 d HYSICS TESTS. Interlock
~
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4.10.4.2 Each Intermediate and P6wer Range channel, and P-7 Interlock shall be subjected to an ANALOG CHANNEL OPERATIONAL TEST within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months prior to initiating STARTUP and PHYSICS / TESTS. \\
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V0GTLE - UNIT 1 3/4 10-4
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\.,_/ SPECIAL TEST EXCEPTIONS i
3/4.10.4 REACTOR COOLANT LOOPS 4 LIMITING CONDITION FOR OPERATION i
r 3.10.4 The limitations of the following requirements may be suspended: ,
a. Specification 3.4.1.1 - During the performance of startup and PHYSICS TESTS in MODE 1 or 2 provided: l f
1) The THERMAL POWER does not exceed the P-7 Interlock Setpoint, and 3
2) The Reactor Trip Setpoints on the OPERABLE I'ntermediate and .
Power Range channels are set in accordance with Table 2.2-1. ,
b. Specification 3.4.1.2 - During the performance of hot rod drop time measurements in MODE 3 provided at least two reactor coolant .
i loops as listed in Specification 3.4.1.2 are OPERABLE. l f' APPLICABILITY: Ouring operation below the P-7 Interlock Setpoint or performance of hot rod drop time measurements.
ACTION:
f a. With the THERMAL POWER greater than the P-7 Interlock Setpoint
! during the performance of startup and PHYSICS TESTS, immediat.ely '
open the Reactor trip breakers.
b. With less than the above required reactor coolant loops OPERABLE during performance of hot rod drop time measurements, immediately open the reactor trip breakers and comply with the provisions of the ACTION statements of Specification 3.4.1.2.
SURVEILLANCE REQUIREMENTS 4.10.4.1 The THERMAL POWER shall be determined to be less than P-7 Interlock
Setpoint at least once per hour during startup and PHYSICS TESTS.
I 4.10.4.2 Each Intermediate and Power Range channel, and P-7 Interlock shall i be subjected to an ANALOG CHANNEL OPERATIONAL TEST within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months prior to
! initiating startup and PHYSICS TESTS.
4.10.4.3 At least the above required reactor coolant loops shall be deter-
! mined CPERABLE.within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months prior to initiation of the hot rod drop time ,
l measurements and at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months during the hot rod drop time i .
measurements by verifying correct breaker alignments and indicated power availability and by verifying secondary side wide range water level to be greater than or equal to 17"..
Vo & 7 L G - c.)Ah 7 j 3/4 so - 4
WW-SPECIAL TEST EXCEPTIONS 3/4.10.5 POSITION INDICATION SYSTEM - SHUTDOWN 1
LIMITING CONDITION FOR OPERATION 3.10.5 The limitations of Specification 3.1.3.3 may be suspended during the performance of individual Nil 6.6.. shutdown and control rod drop time i measurements provided;
a. Only one shutdown or control bank is withdrawn from the fully inserted position at a time, and ,.
b. The rod position indicator is OPERABLE during the withdrawal of the 2 rods.*
APPLICABILITY: MODES 3, 4, and 5 during performance of rod drop time measurements.
ACTION:
With the Position Indication Systems inoperable or with more than one bank of rods withdrawn, immediately open the Reactor trip breakers.
i SURVEILLANCE REQUIREMENTS 4.10.5 The above required Position Indication Systems shall be determined l
i to be OPERABLE within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to the start of and at least once per
- 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter during rod drop time measurements by verifying the Demand
' Position Indication System and the Digital Rod Position Indication System agree:
a. Within 12 steps when the rods are stationary, and l
> b. Within 24 steps during red motion.
f
This requirement is not applicable during the in i it al calibration of the Digital Rod Position Indication System provided: (1) K f is maintained less than or equal to 0.95, and (2) only one shutdown of fontrol rod bank is withdrawn from the fully inserted position at one time.
O V0GTLE - UNIT 1 3/4 10-5 7_,yn 9 y ,,- -.---...y*g- ,y. -- -y.. -
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() JUSTIFICATIONS FOR DEVIATIONS FROM STS ,
SECTION 3/4.10 3/4.10:
Since the plant has only full-length control rods, the words " full-length" have been deleted as redundant. See paragraph 4.2.2.3 of the VEGP FSAR.
4.10.1.2:
The purpose of Surveillance Requirement 4.10.1.2 is to assure the reliability of the reactor control rod insertion capability prior to reducing shutdown margin below specified levels during low power (less than 5 percent) physics tests. The revision from 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to 7 days will allow a ,
more expeditious startup following initial criticality and subsequent refuelings without significantly increasing the probability that a stuck rod will occur in conjunction with a positive reactivity addition transient.
This has been implemented at San Onofre.
3.10.3:
LCO item b was deemed redundant to Table 2.2-1 (which sets both the IR and
() PR low setpoint at s 25 percent RTP) and was modified accordingly.
3.10.4:
LCO item a.2 was deemed redundant to Table 2.2-1 (which sets both the IR and PR low setpoint at s 25 cercent RTP) and was modified accordingly.
I Special Test Exception 3.10.4(b) allows the suspension of 3.4.1.2 during the no flow hot rod drop time test in Mode 3 provided at least two reactor ccolant loops are OPERABLE. The two reactor coolant loops would provide redundant decay heat removal paths. An uncontrolled RCCA bank withdrawal during performance of the subject testing in Mode 3 is not anticipated since the operator would be actively monitoring the rod control system and would be able to detect and correct any problems arising with the red control system. This special test exception has already been approved and implemented in the Byron Unit 1 Technical Specifications.
4.10.4.3:
This surveillance requirement was added to provide a means to determine OPERABILITY of the two reactor coolant loops.
i l
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1 2-1 3.10.5:
4 The words " full-length" are unnecessary since the VEGP does not use
! part-length rods. See paragraph 4.2.2.3 of the VEGP rSAR. i i
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_ i 3/4.11 RADI0 ACTIVE' EFFLUENTS 3/4.11.1 LICUID EFFLUENTS CONCENTRATION LIMITING CONDITION FOR OPERATION 1
3.11.1.1 The concentration of radioact/ive material released in liquist effluents to UNRESTRICTED AREAS (see Figure 5.1-M shall be limitcc to the concentrations specified in 10 CFR Part 20, Appendix B, Table II, Column 2 for radiensc11 des otner than dissolved or entrained noble gases. Fct dissolved or entrained neole gases, the concentration shall be limited to 2 x 10 4 microcurfe/mi total activity. ,.
APPLICABILITY: At all times.
ACTION:
2 With the concentration of radioactive saterial released in ligeid effluents to UNRESTRICTED AREAS exceeding the above limits, inmediately restore the concen-tration to within the above Ifmits.
e. Th pro va'sto ns es .9ee Wies +ian s r e. 3 a rod I c 4 **
not ary/icable -
SURVEILLANCE REQUIREMENTS 4.11.1.1.1 Radioactive liquid wastes shall be sampled and analy::ed accceding O to the sampling and analysis program of Table 4.11-1.
4.11.1.1.2 The results of the radioactivity analyses shall be used in acccrdance with the cethodology and parameters in the 00CM to assure that the concentrations at the point of release are maintained within the limits of Specification '
3.11.1.1. .
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! V0GTLI - UNIT 1 3/4 11-1 l
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i TABLE 4.11-1 i RADI0 ACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM
~
LOWER LIMIT 1 OF DETECTION MINIMUM I
SAMPLING ANALYSIS TYPE OF ACTIVITY (LLD)(1)
} LIQUID RELEASE ANALYSIS (pCi/ml) 1 TYPE FREQUENCY FREQUENCY P
! 1. Batch Vaste P 5x10 7 Release EachJ3atch Each Batch Principal Gamma Tanks (2) Emitters (3) ,
4 vm3#c A%ise' T31 1x10 * ,
ra ,, x ,
a. s 9 e r- % ~cc4 M Dissolved and 1x10 5 i
- P '
One Batch /M Entrained Gases ;
, D',7 ^' ' ",* ' (Gamma Emitters)
b. i w rs_-cio i P M H-3 1x10 8 Each Batch Composite (#) [
Gross Alpha 1x10 7
pr < ,,,y e -
1 oe
c. sa w s P Q Sr-89. Sr-90 5x10.s Each Batch Composite (4) ,
1x10.s Fe-55 ,
! i l
W Principal Gamma 5x10 7 !
i 2. Continuous Releases (5) Centinuous(6) Composite (6) Emitters I3)
' ve wde " I-131 1x10.s l
/d + t n + i e >* f
a. b,4-l M Dissolved and 1x10 5 !
M Grab Sample Entrained Gases !
j
! (Gamma Emitters) {
! -b ,, !
4 M H-3 1x10 5 Continucus(6) Composite (6)
Gross Alpha 1x10 7 4
-en v Q Sr-89, Sr-90 5x10.s ,
Continuous (6) CompositeIO) !
1 1x10 8 l
Fe-55 t
1, lO i
.; , V0GTLE - UNIT 1 3/4 11-2 i i
1 l - . . . . _ . . . -
m l _ g,.
q(j TABLE NGTATIONS (1)The LLD is daffr.ed, for purposes of these specifications, as the se.allest concentration cf radioactive material in a sample that will yield 3 net count, above system background, that will be detected with 95% prcbability '
with only 5% probacility of falsely concluding that a bitak observaticn represents a "real'8 signal.
For a particular censure.eent system, which may include radiecnenicel separation:
4.66 s b E V 2.22 x 10'S Y exp ( .ut) ,
Wnere:
LLD = the "a priori" lower limit of detection (microcurie per unit cass cr volume),
s = the standard deviation of the background ccunting rate or of b
the counting rata cf a black sample as apprcpriate (ccunts per minute),
E = the counting efficiar.cy (counts per disintegration),
p' v V = the sample $12e (units of r: ass or volume),
2.22 x 108 = the number of disintegrations per minute per microcurie.
Y = the fractional radicchemical yield, when appifcable, .
A = the radioactive decay constant for the particular radicauclide (sec 1), and at = the elapsed time between the midpcint of sample collection and the time of counting (sec).
Typical values of E, V, Y, and At sh:uld be used in the calculatfen.
It should be recognizea that the LLD is definad as an a priori (before the fact) limit representing the capability of a neasurement system and not as an a pesteriori (after the fact), limit for a particular measurement.
(2)A batch release is the discharge of liquid wastes of a discrete volume.
Prior to sampling for analyses, each tatch shall be isolated, and then thorougnly mixed by a method described in tne ODOM to assure .
representative sampling.
O t
V0GTLE - UNIT 1 3/4 11-3 i
t
TABLE 4.11-).(Continueol
(_,) TABLENOT&T10NS(Cqn,tinue.jl .
(3)The principel gamma entiters fer which the LLD .specificatico appli,es include the fcilowing radionuclides: Mc-54, Fe-59, Co-58, .C0-60, Zn-65, Ho-99, Cs-134, ts-137, a.1d Ce-141. Ce-144 thall also be measured, but 4 witfi an LLD of 5 x 10 8 ThIs list docs not mean that.only thase nuclides '
are to be cons'idered. Other gamme pecks that are identifiable, together with those of tha above natlides, chall alto .be analyzed and reported 'In the Sem! annual Radioective Effluent RalEEse RGport pursuant te SpaCTfica- ,
tion 6.9.1.4 in the format . outlined in Regulatory Guide 2.21, Appt.idix B,, i Revision 2, June 1974.
I (4)A ceraposite samole is one in which the quantity of Tiquid sarpled is prcportielai to' the quantity of. liquid waste discharged and in which the :
tethcd of sampling empic3ed results in a speciman that is representatiive cf the liquids released.
4
( )A continuous release i,s the discharge of liquid wastes of a nondiscrete '
volare, e.g. , from a voltee cf a system that hss an input flow during the continuous release. ,
4 (5)To be repre.sentative of the cuantiti6s and concentrations .of radioictive materials in licuid affluents, samples shall be collected cor.ticuausly in ,
proportion to the eate of flow of the effluent strean. Prior to ar.alyses,
()
til samples taken for the composite shall La thoroughly mixed in order for the ecmposite sample to be representative of the effluent release, f
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I V0GTLE - UNIT 1 3/4 11-4 l .
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Q RADICACTIVE EFFLDENTS E92 LIMITING CONNTION FOR OPEFATION _
3.1L 1.2 The dese or dose c:mitment to a MEM3ER OF THE PUBLIC free radioactiv'a materials in liquid affluents releasea, fron each unit, to ONRESTRICTED AREAS (see Fir;cre 5.1-t) shalt be limited:
. L
a. During try 'cahndar quarter to less than er equel to LO .sreas to the whoh body anc to less than or equal to 5 mre::s to a:y orgar, l and L . ;
i c.
0uring any calendar year to Iess than or equal to 3 mreas to the ;
i ,
whole body and te less than or equal to 10 crus to arry organ.
, APPLICABILITY: At all tir.es.
A*TIcte
.% With the calculated dose from the rehase of radienctive eterials in licpto effleents exceeding any cf tr.s above limits, prcare
and submit to the Ccetssion within 30 Cays, pursuant to 5pecification 4 6.5.2, a 5:ecial Report that identifies the cause(s) for exceeding i
the limit (s) and defines the cerrective actions that have been taken ,
to refJCe the relet $Cs and the prCOosed Correct-IVe SCtiers to be 8 taken to assure that subsequent release.s will be in compliance with !
the above 1imits. his-SpecieMep+et-she4'f-e4+see4*,de --G-)--the ;
cesults of radioloqual-analyses of :the4rinMag =ter set-carand l (2)--th(-easicAegieAimpate-Mnkhedde4A4eg .;;tcr :gpcih: w4h regad-t+-t#e-eequi remts-+f-40-EFMert-141, Sife ki,Wicrii '!ater ,
t
) Ack*
I
- d. The provisions of f;ecifications 3.0.3 and 3.0.4 are not applicable.
SL'R1EILLf.hCE REOUIEEMENT5 4.11. 1 2 Cunulative dose contributions frca lignid effluents for the current calendar quarter and tha current calendar year snail be catermiced in accordance with the utbodeloqy and parameters in the GCCM at hast once per 31 days.
l .
, 'The--requirements-c?-ACTICM-*Ghand-G4-are-wpMc2h -44y--if dHeb; =ter '
supply-is -takan-from-P.he--receiving +atee-bedr*Mhin 3 d h: af t',e ph it ;
oiscnargs--4-thuase-of- r4vemited-plants-tnis-is-3 d k: t,c- a tr;= = Iy. i V0GTLE - UNIT 1 3/4 11-5 .
- - , - - - - - . , - - - - , , p. e, -,- , , , , - . ---wr.,m..
.. ,e, , . , . , , , , - , , , . , - , . . ~ , . _ ,
-g RASICACTIVE EFFLUENTS U
!LIOUIC RA3h'ASTE TREATMENT SYSTEM ilMITItG CONDITIGN FOR OPERATION _
3. 11.123 The Liquid Rad [aste Treatment System shall he OfERABL2 and .tppropriate porticas of the system shall be used to redece releases of racioactivf ty when the projected aesas due/t a the liquid effluent, from ea:b unit, to U! RESTRICTED AREAS (see figure 5.L-F) would ex:eed Sve6-arem to the whole body or 0.2 mrem to ar;y ergar + 212xhy ;--M. C. .rB i
pu c.uu4.trpar-fer
AoPLICA2ILITY

At all times. .
4 9 A.CTIO,N.:
$ a. With radioactive liouid waste being discharged without +.reittent ant in excess of tte above limits and any portion of the liquid Radwaste T* sat.T.ent System r.ot in operatica, prapar,t acd sub2it tc tte Commis-1 sion within 30 days, cursuant to Spa ffication 6.9.2, a .Special Report i that incluces the felicwing informaticc:
l 1. Explanation of vny liquid rad aste was beinC discharge: without ',
treatment, identification of aay incperaale equipnant or
, subsystems, ana the reason for tne inoperability,
2. Action (s) talen to restore the ihoparable equipment to CPERABLE i status, and i
i 3. Summary description of action (s) taken to prevert a recurrence. L i
b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicabis.
SEVIEILLANCE P1001REH9fTS _
4.11.1.3.f'D:ses due to ifquid releases froc each unit to UWREEIRICTED AREAS h shall be projected at least or.ca per 31 days in accordance with the rethodology and oartreters in the ODCM when Liquid Raasasta Treatr.ent Systems ere not being ;
fully Ltilized. t c
. . . I e dTI84~ t*!b lh d EifDFe8 ebad Sysbd I T 2 i
sc.m cci G u-uu uy metirrSpeeHicati . . . . .
t t
l
$(]
j YCGTLE - UNIT 1 3/4 11-6
_ __ _ ._ . . .. _. . ~ . _ - __ _ - _ . _ _ _ _. _. ______. _ _ _ _ _ _ _ _
i; 4
. j L
i ;
RADI0 ACTIVE EFFLUENTS
?
LIQUID _NOLDUP TANKS * ,
1 s
LIMITING CONDITION: FCR OPERATI_9N
~
ev+ side +emporar
\
3.11.1.4 The quant $ty of radioactive material contained in oech s' the felic kg y l wgretect:d :vedeer tankf/ shall be lictited to less than or equal to fo Curies, ex:Toding t9f ttura and dissolved or entrained noble gases:
4
+ _w i A s 3
- r
.s. Out:id: t=.p: ~; t & .
i +
AfDLICABILITY: At all times.
ACT:0N: e ,.44 t< :
I a. With the quantity of rJdfoactIVe CatMial 3R sny ofTtha aboVe listUd tanks exceeding the above limit, imediately St.sper,d all additices i of radioactive material to the tank, within 49 hour5.671296e-4 days 0.0136 hours 8.101852e-5 weeks 1.86445e-5 months s} rtdece the tank t T ccDtents to witnin the limit, and describe the events leading tn ,
this condition in the next Semiannaal Radioactive-Effluent Release Report, pursuant to Spect f' ation 6.9,1.4jo r p ro rMe ,,d/4e4#o, 4 Me. ;
Corv,m:s3:en pursesr,+ to Specifiedoon 4 ~ 9.1 wv/,s 34 day:, in lieu of any -
t
b. The provisions of Speelfic.ation.s 3.0.3 and 3.0.4 are not applicable. 8l^*r, , f.
SURVEILLAKME R_EQUIREMEXTS_
. 4 Hh << ,
l 4.11.1.4 The cuantity offradiaactive material cont.ained in each of the above !
f
( listed tanks sr.all be dei' trained to be within the abova {lmit by analyzing a ,
representative sample of the tank's contents at least once per 7 days wnen radioactive materials are being added to the tank.ror ea-c.A base /, # radc4cs'Ne.
/>,xtertri prie r- +o i+s .s sdifie r. fe di> e A n ti. ,
i 1
i i
) .
Tanks included in this specification are those outdoor tanks thatare rot c
surrounced by liners, dikes, or walls capaele of holding the tank contents and that do not have tank overflows and surrounding aree drains connectied !
l i
to tne Liquid Padweste Treatment System. (
c I
r VCGTt.E 'JIIT 1 3/4 11-7 :
i l
l r .
l .m . , - . _ _ _ _ _ - _ _ . . _ . _ _ , _ _ _ _ _ _ _ _ _ _
-ti l RADIOACTIVE EFFLUENTS
! 3/4.11.2 GASEOUS _ EFFLUENTS 00SE RATE LIMITING CONDITION FOR 0PERATION I
3.11.2.1 The dose rate due to radioactive materials released in gaseous
effluents from the site to areas at and beyond the SITE NUNDARY (see Figure :
5.lshallbelieftadtothefollowing:
a. For ncble gases: Less than or equal to 500 mrems/yr to the whole body and less than or equal to 3000 arems/yr, to the skin, and
'; b. For lodine-131, for Iodine-133, for tritium, and for all radio-nuclices in particulate form with half-lives greater than 8 days: 4 Less than or equal to 1500 mrems/yr to any organ. , AFDLICA8ILITY: At all tioes. i ACTION: l >*4ith the dose rate (s) exceeding tne above linits, immediately restore the release rate to within the above limit (s). , h ri,e praisiens o f &eeis4ca+ ions 3 o 3 and 3 c.1 are no t *?Yhenb!e - , ' SDRVEILLANCE REQUIREMENTS I 4.1L 2.1.1 The dose rate due to noble gases in gaseous effluents shall be determined! to be within the above limits in accordance with the methodology i and paras.eters in the ODCM.
.se m n sn g ny
] 4.11.2.1.2 The dose rate due to Iodine-131, Icdine-133, tritium, anc all
- radienuclides in particulate form with half-lives greater than 8 days in ,
ssseous affluents stall be determined to be within the above limits in , accordanca with the methodology and parameters in the 00CM by obtaining representative sacles and performing analyses in accordanca with the sa:pling and analysis progrse specified in Table 4.11-2. l p l l i 4 h i s a \O l 4 l 7 V0GTLE - UNIT 1 3/4 11-8 , i i \
O O D 9
- TABLE 4.11-2 / 'Q1CE. < RADIDAC11VE CASE 005 WASTE ShT4iTiilf AND ANALYSIS PROGRAM '
o ' ci -- HINIMUM LOVER LlHIT OF 5AMPLING ANALYSIS TYPE OF DETECTION (LLD)II) ACTIVITY ANALYSIS (pCi/ml)
$ GA5E005 RELEASE TYPE FREQUENCY FREQUENCY
1. Waste Gas Storage P P Tar.k N ca-y Each Tank Each Tank Principal Gama Emitters (2) 1x10 4 i ;
Grab Sample
2. Containment Purge ? P or-Vent 24 "or #4 " Each PURGE C) Each FIIP.GE I3I Principal Gamma Emitters (2) 1x10 4 Grab Sample - ,
H H-3 (oxide) 1x10 8
3. a. Plant Vent H(3)'(4)i * \ l Principal Ganna Emitters (2) 1x10 4 Grab $ ample M(3)
, <* H-3 (oxide) _ lx10 8__ __ f N(5)w ) Principal Gassa Emitters (2) 1x10 4 w b. #"uMS$ rah T
efe c kr

Area m * --' Grab Sample /\/\ _ _
8 i 4 I' g jf, Ventilation H-3 (oxide) _ _ 13107
c. Auxiliary H Principal Gama freitters I) 1x10 4 Bldg, Radwaste Grab Sample M
%Its;/ve.f.'*a --> Area, 50.". Vent t Bansing ggg,n _
All Reluse Types Continuous (6) y(7) y_333 lxl0,l,
- 4. ,
as listed in-1., 2.r Charcoal
,.s.,a 3 3 and 3.Pabove. Sample :
! A ho 3 6 1,er, s - - y(7) Principal Gamma Emitters (2) 1x10 11 N 1S/2 $'e,I,'eY N/. Continuous (6) Particulate Samp1H (19c 2 - VG - 001,190 4- V6 -00: , 19e2-vs-vos,nch Continuous (6) M Gross Alpha 1x10-(1 '
n o z - v<,-o os, noc va ' V:,
-oo-c.co , 4, Composite Par-efe r - vo-oo t, nos- ve- o o f, ticulate Sample l
A '9 2 V6 - o'o ) _ Continuous (6) Q Sr-89, Sr-90 1x10 11 Composite Par-
ticulate Sample ,.
$, d.U- ) TABLE 4.11-2 (Continued)
TABLE NOTATIONS (1)The LLO is defined, for purposes of these specifications, as the smallest concentration of radioactive material in a sample that will yield a net count, above system background, that will be detected with 95% probability ' with only 5% probability of falsely concluding that a blank observation represents a "real" signal. For a particular measurement system, which may include radiochemical separation: 4.66 s g , < E V 2.22 x 10s . y . exp (. g t) Where: , LLO = the "a priori" lower limit of detection (microcurie per unit mass or volcma), l s = the standerd deviation of the background counting rate or of - b the counting rate of a blank sample as appropriate. (counts per < minute), E = the counting efficiency (counts per disintegration), V = the sample size (units of mass or volume), , 2.22 x 108 = the number of disintegrations per minute per microcurie, Y = the fractional radiochemical yield, when applicable, A = the radicactive decay constant for the particular radionuclide , (sec.2),and ' at = the elapsed time between the midpoint of sample collection and the timo of countir.g (sec). Typical values of E, V, Y, and at should be used in the calculation. It should be recognizac that the LLO is defined as an a criori (before the fact) Ifmit representing the capability of a measurement system and not as an a posteriori (af ter the fact) limit for a particular measurement. f O i V9GTLE - L' NIT 1 2/4 11-10 . i
. _ _ , _ . - _ - . . _ - - , _ _ . _- -~ - , . , . . _ - - - ._ _ _ - . , ,-
^
TABLE 4.11-2 (Continued) TABLE NOTATIONS (Continued) (2)The principal gamma emitters for which the LLD specification applies include the following radionuclides: Kr-87, Kr-88, Xe-133, Xe-133m, Xe-135, and Xe-138 in noble gas releases and Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, I-131, Cs-134, Cs-137, Ce-141 and Ce-114 in Iodine and particulate releases. This list does not mean that only these nuclides are to be considered. Other gamma peaks that are identifiable, together with those of the above nuclides, shall also be analyzed and reported in the , Semiannual Radioactive Effluent Release Report pursuant to Specification 6.9.1.4 in the format outlined in Regulatory Guide 1.21, Appendix B, Revision 1, June 1974. On issuous increase or decre a se in I3)Samp/ or a THERMAL POWER chrge exceeding 15% of RAT 1-hour period.+ ! (4) Tritium grab samples shall be taken at least once per 24 hours when the
refueling canal is flooded.
(5)Tritiu= grab samples shall be taken at least once per 7 days from the ventilation exnaust from the spent fuel pool area, whenever spent fuel is in the spent fuel pool. I ! vf) i (6)The ratio of the sample flow rate to the sampled stream flow rate shall be kncwn for the time period covered by each dose or dose rate calculation ! ! made in accordance with Specifications 3.11.2.1, 3.11.2.2, and 3.11.2.3. ,
' ""*""'**d'**"
(7) Samples shall be changed at least once per 7' ays and analyses shall be completed within 48 hours after changing, or after removal from sampler. Sampling shall also be performed at least opce per 24 hours for at least , 7 days following each shutdcwn, startup, or* THERMAL POWER chcag; exceeding
- 15% of RATED THERMAL POWER within a 1-hour period and analyses shall be ccmpleted within 48 hours of changing. When samples collected for 24 hours are analyzed, the correspor. ding LLDs may be increased by a factor of 10. - ! This requirement does not apply if (1) analysis shows that the DOSE -
i EQUIVALENT I-131 concentration in the reactor coolant has not increased j more than a factor of 3; and (2) the noble gas monitor _shows that i ; ef fluent activity has not increased more than a factor of 3. 9' %mpiin; ee < rep i <<d un los a. primary - te - sec,nda.ry ha e u i
' ca ,, /irm e d i,
t re s airement b es no t an,/v i / u) .e n a/,ci, s;,w, psar ti, a p., e. jl , 67L , a aisn.le n t Im i e ,,e ,, rra +ien i., th e p</m a ry c.. /a n t As s ,,.t in crea., a. :
~
n ~ s es enan a. tie for e f 3 ; la i tsse. n o d It' ea s er,on/for ~shes.n tao f effluers Y '
?ne+ar of Jj an d O } a ,oeuy e ^ .tcHviki tras no t in crease /- ins re ih n.n n r i.s in nem<ess. .s O
YOGTLE - UNIT 1 3/4 11-11 j i I
- , , - - - - , . . . ..-,,,.---.,-..-.c- --,-r,. w,- , -,,v,,, . ,,,-w, --n, - _ . . . , , . w,. .-c-~ . . - , ,-_.. . , , , .
ay&-
+ m.
4
() RADIOACTIVE EFFLUENTS DOSE - NOBLE GASES LIMITING CONDITION FOR OPERATION 3.11.2.2 The air dose due to noble gases released in gaseous effluents, from
+ each unit, to areas at and beyond the SITE BOUNDARY (see Figure 5.1-4) I shall be limited to the following:
a. During any calendar quarter: Less than or equal to 5 mrads for gamma radiation and less than or equal to 10 mrads for beta radiation, and During any calendar year: Less than or equa-l' to 10 mrads for gamma b.
radiation and less than or equal to 20 mrads for beta radiation. APPLICABILITY: At all times. ACTION
a. With the calculated air dose from radioactive noble gases in gaseous l effluents exceeding any of the above limits, prepare and submit to i
i the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report that identifies the cause(s) for exceeding the limit (s) and defines the corrective actions that have been taken to reduce () the releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits.
b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REOUIREMENTS i
~
4.11.2.2 Cumulative dose contributions for the current calendar quarter and current calendar year for noble gases shall be determined in accordance with the methodology and parameters in the ODCM at least once per 31 days. f i 4 \. i VCGTLE - UNIT 1 3/4 11-12
+
l 1
.. -,.__...___.-.__m._..__._-__ . . _ . . _ . _ . . . . _ . . _ . . _ _ _ - . _ . _ . _ . _ . _ _ . , _ _ _ _ . _ _ .
~ ~ . - -
l i RADI0 ACTIVE EFFLUENTS DOSE - 10 DINE-131, 10 DINE-133, TRITIUM, AND RADIOACTIVE MATERIAL IN PARTICULATE FORM LIMITING CONDITION FOR OPERATION 3.11.2.3 The dose to a NEMBER OF THE PUBLIC from Iodine-131, Iodine-133, tritium, and all radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents released, from each unit, to areas at and beyond the SITE BOUMDARY (see Figure 5.1-J) shall be limited to the following: 1
a. During any calendar quarter: Less than or equal to 7.5 mrems to any organ and, ..

b. During $ny calendar year: Less than or equal to'15 mrems to any organ.
APPLICABILITY: At all times. ACTION: s
a. With the calculated dose from the release of Iodine-131, Iodine-133, tritium, and radionuclides in particulate form with half-lives greater than 8 days,'in gaseous effluents exceeding any of the above limits, prepare and-submit to the Commission within 30 days, pursuant O to Specification 6.9.2, a Special Report that identifies the cause(s) for exceeding the limit (s) and defines the corrective actions that have been taken to' reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits.

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE RE0VIREMENTS j 4.11.2.3 Cumulative dose contributions for $le c "rrect calendar quarter and current calendar year for Iodine-131, Iof P'c< ;$2 tritium and radionuclides in particulate form with half-lives grette t)[ > days shall be determined in accordance with the methodology and parameters in the ODCM at least once < per 31 days. lO V0GTLE - UNIT 1 3/4 11-13
- - _ _ . _ , _ - - . _ - _ ___ _ __ . _ _ _ _ . . . ~ .- _ _ _ _ _ _ _ _ _ - _ _ . .
3 1 RADI0 ACTIVE EFFLUENTS GASEOUS RADWASTE TREATMENT SYSTEM LIMITING CONDITION FOR OPERATION Gv9sfous WAsrE PhefCss/A 3.11.2.4 The VENTILATION EXHAUST TREATMENT SYSTEM and theWACTE C",5 "0LOL'? - SYSTEM shall be OPERABLE and appropriate portions of these systems shall be
used to reduce releases'of radioactivity when the projected doses in 31 days j due to gaseous e.ffluent releases, from each unit, to areas at and beyond the SITE BOUNDARY (see Figure 5.1-p would exceed f .e./Niert
a. 0.2 mrad to air from gamma radiation, or

b. 0.4 mrad to air from beta radiation, or
' c. 0.3 mrem to any organ of a MEMBER OF THE PUBLIC. i APPLICABILITY: At all times. ACTION:
a. With radioactive gaseous waste being discharged without treatment and in excess of the above limits, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a
, Special Report that includes the following information:
1. Identification of any inoperable equipment or subsystems, and the reason for the inoperability,

2. Action (s) taken to restore the inoperable equipment to OPERABLE status, and

3. ~ Summary description of action (s) taken to prevent a recurrence.
~
l
b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
i SURVEILLANCE REOUIREMENTS i d 4.11.2.4.2' Doses due to gaseous releases from each unit to areas at and beyond the SITE BOUNDARY shall be projected at least once per 31 days in i accordance with the methodology and parameters in the ODCM when Gaseous Radwaste Treatment Systems are not being fully utilized. 4.11.2.4.2 Th: in:talled VENTILATION EXMAUST TP,CATMENT SYSTCM and L'ACTC GAS-HOLDUP--SYSTEM-shall-be-cengider:d OPER",SLE by .n ; ting Sp;;ific;ti:n; s....... oou s. .&.& v. s. .&.s.
~
O l V0GTLE - UNIT 1 3/4 11-14
RADI0 ACTIVE EFFLUENTS EXPLOSIVE GAS MIXTURE [Syst a; n;t d;;igned t; withstand a hyd ;;;n : p!;;ien] 1 LIMITING CONDITION FOR OPERATION
~
I das cous wn.srE PNOW sW6-3.11.2.5 The concentration of oxygen in the WAST: OAS ;;0 LOUP SYSTEM shall be limited to less than or equal to 2% by volume whenever the hydrogen concentration exceeds 4% by volume. APPLICABILITY: At all times. ACTION: Gasscus wasre fmcssstab-
a. With the concentration of oxygen in the STE C.*S "^LDUP SYSTEM greater than 2% by volume but less than or equal to 4% by volume, reduce the oxygen concentration to the above limits within 48 hours.
&ascous wnsre pnocsssists-a b. With the concentration of oxygen in the WASTE-GMHieE99P SYSTEM greater than 4% by volume and the hydrogen concentration greater than 4% by volume, immediately suspend all additions of waste gases to the system and reduce the concentration of oxygen to less than or equal to 4% by volume, then take ACTION a., above.
l c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. SURVEILLANCE REQUIREMENTS GA.sGous Wstsrs fnocEssie 4.11.2.5 The concentrations of hydrogen and oxygen in the WASTE CAS ll0 LOUP SYSTEM shall be determined to be within the above limits by continuously monitoring the waste gases in the WASTC CAS !!0 LOUP SYSTEM with the hydrcgen and oxygen monitors required OPERABLE by Table 3.3-13 of Specification 3.3.3.11. G-A:Gous wASrf PROCE:55W6 l O V0GTLE - UNIT 1 3/4 11-15 .
~. ~ . - - . . .. . - .- .
M%e eWA RADIOACTIVE EFFLUENTS

EXPLOSIVE GAS MIXTURE [ Systems designed to withstand a hydrogen explosion]
\ / /
LIMITNG CONDITION FOR OPERATION 3.11 . 5 The concentration of hydrogen or oxygen in the WASTE GAS HOLDUP SYST hall be limited to less than or equal to 4% by volume. APPLICABILITY: At all times. ACTION:
a. Witb the concentration of hydrogen or oxygen.in th'e WASTE GAS HOLDUP SYSTEM exceeding the limit, restore the concentration to within the I
' limit within 48 hours. ' /
i ~
b. The pro isions of Specifications 3.0.3 and 3'.'0.4 are not applicable.
/
SURVEILLANCE REQUIR : ENTS S
/ \
4.11.2.5 The concentra ion of hydrogen or oxyge'n in the WASTE GAS HOLDUP SYSTEM shall be determined to within the above limits by continuously monitoring the waste gases in the WASTE HOLDUP SYSTEM wjth the hydrogen or oxygen monitors required OPERABLE by Table 1.3-13 of Speciff' cation 3.3.3.11. O
\ \ \
_ / \ . / \s
\ \~ ^
r. \
\ / \ / \, / /
e j
/ 3/4 11-15b \
V0plE - UNIT 1
-- -=-mp rt.- we, e, ,--*e r-m ve r v+++s--e*vv-r--ve-t-r ++at=- - -w-d-w t w- -~nwar,**-w-*N'T 7 w " ' - - * * ^v
s RADIOACTIVE EFFLUENTS D EcA>' GAS STC SCE TANKS i LIMITING CONDITION FOR OPERATION de ca-3.11.2.6 The quantity of radioactivity contained in each gas ster:yg: tank shall be limited to less than or equal to Curies of noble gase's (considered as Xe-133 equivalent). 4 g A 0 W te APPLICABILITY: At all times. ACTION: decay ;
a. With the quantity of radioactive material in, any gas 34eeese tank exceeding the above limit, immediately suspend all additions of radioactive material to the tank, within 48 hours reduce the tank contents to within the limit, and describe the events leading to this condition in the next Semiannual Radioactive Effluent Release Report, pursuant to Specification 6.9.1.4.

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.11.2.6.1The quantity of radioactive material contained in each gas tank shall be determined to be within the above limit at least once per 24 7 days O hours when radioactive materials :r: bei~; added to the tank /dariny /^e. previous 1 d 2 ys. ha.ve.been 4 !, 2 6 2. :n ?A e e v e n t .1 c ,, /;,,,, ed. majo r fu e / y% ;/are. ( > / % ) each wasfe rne, ,y aa.n fify of r.a..licac fi/e ma feria./ es,, tain ed in 2.2 s J e es y ?.2 n A sh t // Je d e seem ined to b e. w/ Noin /4e a.6ove 4* moi
'.t ' le.e ., " enee pe, : - s,ovrs wnen rad;ca efive ma terials Aase h e e. ,, +. fAa aa in ene pre vo' sus z9 hours . . .x i a 4
l l O V0GTLE - UNIT 1 3/4 11-16 _ _ _ . _ . . . _ _ _ _ - ~ _ , ~ . _ - _ _ . , _ . . . _ _ . - _ . . . . _ . _ . _ . , , . _ _ . . . _ , _ , . _ . . _ , _
~
l RADIOACTIVE EFFLUENTS O v 3/4.11.3 SOLID RADIOACTIVE WASTES LIMITING CONDITION FOR OPERATION 3.11.3 Radioactive wastes shall be solidified or dewatered in accordance with the PROCESS CONTROL PROGRAM to meet shipping and transportation requirements during transit, and disposal site requirements when received at the dis,posal site. APPLICABILITY: At all times. ACTION:
a. With SOLIDIFICATION or dewatering not meeting disposal site and chipping and transportation requirements, suspend shipment of the inadequately processed wastes and correct the PROCESS CONTROL PROGRAM, the procedures, and/or the Solid Waste System as necessary to prevent recurrence.

b. With SOLIDIFICATION or dewatering not performed in a'ccordance with 4
the PROCESS CONTROL PROGRAM, test the improperly processed waste in each container to ensure that it meets burial ground and shipping requirements and take appropriate administrative action to prevent recurrence.
c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.11.3 SOLIDIFICATION of at least one representative test specimen from at least every tenth batch of each type of wet radioactive wastes (e.g., filter sludges, spent resins, evaporator bottoms, boric acid solutions, and sodium sulfate solutions) shall be verified in accordance with the PROCESS CONTROL j PROGRAM:
a. If-any test specimen fails to verify SOLIDIFICATION, the SOLIDIFICATION of the batch under test shall be suspended until such time as additional 2
~
test specimens can be obtained, alternative SOLIDIFICATION parameters can be determined in accordance with the PROCESS C.ONTROL PROGRAM, and a subsequent test verifies SOLIDIFICATION. SOLIDIFICATION of the batch may then be resumed using the alternative SOLIDIFICATION parameters determined by the PROCESS CONTROL PROGRAM;
b. If the initial test specimen from a batch of waste fails to verify SOLIDIFICATION, the PROCESS CONTROL PROGRAM shall provide for the collection and testing of representative test specimens from each consecutive batch of the same type of wet waste until at least three consecutive initial test specimens demonstrate SOLIDIFICATION.
The PROCESS CONTROL PROGRAM shall be modified as required, as provided in Specification 6.13, to assure SOLIDIFICATION of subsequent batches of waste; and
c. With the installed equipment incapable of meeting Specification 3.11.3 or declared inoperable, restore the equipment to OPERABLE status or provide for contract capability to process wastes ~as necessary to satisfy all applicable transportation and disposal requirements.
V0GTLE - UNIT 1 3/4 11-17
~,
4
RADI0 ACTIVE EFFLUENTS 3/4.11.4 TOTAL DOSE LIMITING CONDITION FOR OPERATION 3.11.'4 The annual (calendar year) dos 2 or dose commitment to any MEMBER OF 1
THE PUBLIC due to releases of radioactivity and to radiation from uranium fuel cycle sources shall be limited to less than or equal to 25 mrems to the whole body or any organ, except the thyroid, which shall be limited to less than or equal to 75 mrems. APPLICABILITY: At all times. ACTION: ,.
a. With the calculated doses from the release of radioactive materials in liquid or gaseous affluents exceeding twice the limits of Speciff-cation 3.11.1.2a., 3.11.1.2b., 3.11.2.2a., 3.11.2.2b., 3.11.2.3a., or C
3.11.2.3b., calculations shall be made including direct radiation contributions from the units (including outside storage tanks etc.) i to determine whether the above limits of Specification 3.11.4 have ! been exceeded. If such is the case, prepare and submit to the Commis-sion within 30 days, pursuant to Specification 6.9.2, a Special Report that defines the corrective action to be taken to reduce l subsequent releases to prevent recurrence of exceeding the above , limits and includes the schedule for achieving conformance with the above limits. This Special Report, as defined in 10 CFR 20.405(c), ,O i shall include an analysis that estimates the radiation exposure (dose) to a MEMBER OF THE PU8LIC from uranium fuel cycle sources, including all effluent pathways and direct radiation, for the calendar year that x includes the release (s) covered by this report. It shall also describe levels of radiation and concentrations of radioactive material involved, and the cause of the exposure levels or concentrations. If the estimated dose (s) exceeds the above limits, and if the release condition
, resulting in violation of 40 CFR Part 190 has_not already been corrected, the Special Report shall include a request for a variance in accor-dance with the provisions of 40 CFR Part 190. Submittal of the report is considered a timely request, and a variance is granted until staff
action on the request is complete.

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.11.4.1 Cumulative dose contributions from liquid and gaseous effluents shall be determined in accordance with Specifications 4.11.1.2, 4.11.2.2, and 4.11.2.3, and in accordance with the methodology and parameters in the ODCM. 4.11.4.2 Cumulative dose contributions from direct radiation from the units (including outside storage tanks etc.) shall be determined in accordance with the methodology and parameters in the ODCH. This requirement is applicable only uncer conditions set forth in ACTION a. of Specification 3.11.4. i O V0GTLE - UNIT 1 3/4 11-18
(b/ JUSTIFICATIONS FOR DEVIATIONS FROM STS SECTION 3/4.11 3/4.11: Instrument tag numbers were added as appropriate to assist the operator in , ensuring compliance with these Technical Specifications. 3.11.1.1, Action b: 10 CFR 20.106(a) states "For purposes of this section concentrations may be averaged over a period not greater than one year." Therefore,,it is not ' reasonable to require shutdown of the reactor within I hour when 10 CFR 20, Appendix B, Table II, Column 2 concentrations are exceeded. In addition, requiring shutdown of the reactor within 1 hour when Appendix B concentrations are exceeded is not consistent with 10 CFR 50.73 (B) (viii) B which allows levels .of twice Appendix B values when averaged over 1 hour as a criterion for determining whether the event is reportable. Further, releases from radwaste systems may not be related to the operational made of the reactor; therefore, reactor shutdown may not be appropriate as a means of restoring concentrations to within Appendix B O limits. Therefore, the provisions of Specifications 3.0.3 and 3.0.4 should not apply. T 3.11.1.2, Action a: The deletion of the additional special report information is justified on the basis that there is no drinking water taken from the river within 3 miles of plant discharge as noted by footnote *. 3.11.1.3: The change that revises the dose projection from a 31-day period to a quarterly basis is for coerttional convenience and does not alter the intent of the specification in any way since the corresponding dose limit has been adjusted for a quarterly projection. 4.11 L ;2 (STS): This surveillance requirement was deleted'on the basis that is redundant to the requirements of 3.11.1.1 and 3.11.1.2. , O
1 l' 3.11.1.4: The outdoor tanks which contain radioactive material at VEGP are Seismic Category I and are missile protected. Consequently, this specification should apply only to temporary tanks which are brought onsite.for additional ! storage capacity. See paragraph 2.4.13.2 of the VEGP FSAR The referenced 10-curie limit is justified based on NUREG-0133. 3.11.1.4. Action a: 4 Action . Statement a was modified to. permit greater than 48 hours for reduction of tank contents so long as immediate action is taken (i.e., suspend additions to the tank)land an increased level of reporting is complied with. This would allow a more realistic time frame for the implementation of actions necessary to reduce the tank contents to within
; the limit. For a temporary outside tank, this could involve delivery of an j
additional tank or modification of the existing tank to drain or pump the i contents. t i
4.11.1.4:
i 1 The revision recognizes that a temporary tank may not allow a representative i sample to be obtained. Analyzing each batch prior to addition to the tank will be the same as analyzing the total tank contents. Thus, the intent of the specification is preserved. i 3.11.2.1, Action b; ) According to Technical Specification BASES the annual dose limits are the doses associated with the concentrations of 10 CFR 20, Appendix 3, Table II, Column 1. 10 CFR 20.106 (a) states ".For purposes of this section ! concentrations may be averaged over a period not greater than one year." Therefore, it is not reasonable to require shutdown of the reactor within 1 hour when the dose rate (s) associated with Appendix B is(are) exceeded, k In addition, requiring shutdown of the reactor within 1 hour when a dose rate associated with Appendix B concentrations is exceeded is not consistent with 10 CFR 50.73 (b) (viii) A which allows levels of twice Appendix B . j values when averaged over 1 hour as a criterion for determining whether the event is reportable. j. I Further, releases from radwaste systems may not be related to the i operational mode of the reactor; therefore, reactor shutdown may not be appropriate as a means of restoring concentrations to within Appendix B ! limits. Tnerefore, the provisions of Specifications 3.0.3 and 3.0.4 should not apply. !O 1 I 4
4.11.2.1.1: The word " semiannually" was inserted on the basis that the determination
' will be made semiannually for the semiannual report. ' Table 4.11-2:
See section 11.3 of the VEGP FSAR. Item 1:
" Decay" was added to reflect plant-specific nomenclature.
Item 2:
" Vent" was deleted to reflect plant-specific design. Footnote 3 was revised to incorporate the same conditions used in footnote f on the basis that both footnotes are applied to sampling for principal gamma emitters. Footnote 3 was also revised to require sampling only when a purge is in progress.
Item 3a: Footnote 5 was added since the fuel storage area ventilation exhausts through the plant vent. Item 3b:
" Condenser Air Ejector and Steam Packing Exhauster" reflect plant-specific design. " Fuel Storage Area Ventilation" exhausts through the plant vent.
Footnote 8 was added to clarify that sampling need not be performed unless a primary-to-secondary leak exists. See the justificatiun provided for item 4 below. Item 3c:
" Auxiliary Building" exhausts through the plant vent. The "Radwaste Solidification Building" identifies a plant-specific feature. The steam generator blowdown vent exhaust through the condenser air ejector and steam packing exhaust and was therefore deleted.
Item 4: Types 1 and 2 (decay tanks and containment purge) need not be included since both are released via the plant vent and sampling at the plant vent is sufficient. Gaseous release type 3b need not be included unless a primary-to-secondary leak has been confirmed on the basis that unless such a leak exists the condenser air ejector and steam packing exhauster does not represent a release pathway. Specification 3/4.7.1.4 will ensure that primary-to-secondary leaks are identified in a timely manner via 'a gross radioactivity determination at least once per 72 hours. O
I 1 l O V Footnotes 3 and 7 were revised by the addition of the words " continuous increase or decrease in" for clarification. Without the revision, there was concern that a 15 percent change in power in a 1-hour period could be interpreted as, for example, the sum of a 5 percent decrease, a subsequent 5 percent increase, and another 5 percent decrease. 3.11.2.4: The system terminology change from waste gas holdup system to gaseous waste processing system reflects plant-specific nomenclature. See section 11.3 of the VEGP FSAR. 4.11.2.4.2 (STS): This surveillance requirement was deleted on the basis that it is redundant to the requirements of 3.11.2.1, 3.11.2.2, and 3.11.2.3. 3.11.2.5, Action Statements a and b, and 4.11.2.5: The change in terminology (from waste gas holdup system to gaseous waste processing system) was made to reflect plant-specific nomenclature. See section 11.3 of the VEGP FSAR. 3/4.11.2.5, Explosive Gas Mixture (Systems designed to withstand a hydrogen explosion): This specification was deleted since VEGP is not equipped with a gaseous waste processing system designed to withstand a hydrogen explosion. 3.11.2.6: The change in terminology (from " Storage" to " Decay") reflects plant-specific nomenclature. See section 11.3 of the VEGP FSAR. The basis for the 2.0 x 10' Ci for Xe-133 is outlined in NUREG-0133 and complies with the criteria in Branch Technical Position ESTB 11-5 as outlined in NUREG-0800. 4.11.2.6.1 & 4.11.2.6.2: The proposed revision relaxes the surveillance requirements for quantifying the gas decay tank inventory from once per 24 hours when adding radioactive j material to the tank to once per 7 days when activity has been adde'd during the previous 7 days - in the absence of confirmed major fuel failure (> 1 percent). b
(Vh a) The total noble gas inventory associated with 1 percent defects (i.e., approximately 90,000 Ci) is significantly below the activity inventory which could result in the total body exposure limit of 0.5 rem (300,000 Ci). Further, plant operation with fuel failures approaching or exceeding the equivalent of 1-percent fuel defects is not likely based on previous plant experience and other Technical Specification limitations (i.e., iodine RCS activity). l l b) Based on the above, a daily sampling and analysis schedule is ! considered to be too frequent and inconsistent with the ALARA l philosophy; i.e., it does not minimize occupational radiation exposure to the operators who obtain and analyze the samples. O 0052v
._- - . . . _ . . _ ~. _.
1 i i I 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.1 MONITORING PROGRAM , LIMITING CONDITION FOR OPERATION ta) 3.12.1 The Radiological Environmental Monitoring Program shall be conducted as specified in Table 3.12-1. APPLICABILITY: At all times. I t ACTION:
a. With the Radiological Environmental Monitoring Program not being conducted as specified in Table 3.12-1, prepare and submit to u,,, t //me the Commission, in the Annual Radi_ologic_a1 Environmental 0; r:t h Report required by Specification [6._9_.1.3Da description of the reasons for not conducting the program as required and the plans for preventing a recurrence.
w firmed.g
b. With theslavel of radioactivity as the result of plant effluents in rem rhe e4 d s an environmental sampling medium at a specified location exceeding Me arWW 3.12-2 when averaged over any calendar c d e w f e rreJ the reporting levels of Tablequarter, prepare and submit to the Commis e r a./ a r re w r m - to Specification 6.9.2, a Special Report that identifies the cause(s) l Won, sA /co ne r for exceeding the limit (s) and defines the corrective actions to be

/<t W taken to reduce radioactive' effluents so that the potential annual dose
to a MEMBER OF THE PUBLIC is less than the calendar year limits t
of Specifications 3.11.1.2, 3.11.2.2, or 3.11.2.3. When more than j one of the radionuclides in Table 3.12-2 are detected in the sampling j medium, this report shall be submitted if: concentration (1) concentration (2) + ***> 1.0
' reporting level (1) , reporting level (2) -
' When radionuclides other than those in Table 3.12-2 are detected and are the result of plant effluents, this report shall be submitted if the potential annual dose
to a MEMBER OF THE PUBLIC from all radio-nuclides is equal to or greater than the calendar year limits of Specification 3.11.1.2, 3.11.2.2, or 3.11.2.3. 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 Radiolgd cal Environmental i Opeceting Report required by Specification 6.9.1.3./
%.en ; i Iane.e.
i
*The methodology and parameters used to estimate the potential annual dose to a MEMBER OF THE PUBLIC shall be indicated in this report.
O 4
V0GTLE - UNIT 1 3/4 12-1
i 1 1 RADIOLOGICAL ENVIRONMENTAL MONITORING O LIMITING CONDITION FOR OPERATION ACTION (Continued)
'#" E "'"" " N " #'' " I' adeo uale sa,>>plesW '*]
t
c. With/ milk oncr;;n leafy ":getatien ,aN unavailable from one or more of the sample locations required by Table 3.12-1, i t atify 4
c**3 ,. d ,',;' jp:cf #ic locations for obtaining replacement samples and edd-@en//3 access /.d,// within 30 days to the Radiological Environmental Monitoring Program #^ese 5**
'u b t'. /,4 given in the 00CM. The specific locations from which samples were " ###"#
unavailable' may 4 hen be deleted from the monitoring program. Pursuant to Specification 6.14, submit-in the next Semiannual Radioactive Effluent Release Report documentation for a change in the 00CM i including a revised figure (s) and table for the 00CM reflecting the
; new location (s)twith supporting information identifying the cause of the unavailabil ity of samples and justifying the selection of the' new location (s) for obtaining samples,or /Ac. turave//44///7 a7 " if<ury suHab/c new locajior,s.
3
d. The provisions of Spec'ifications 3.0.3 and 3.0.4 are not applicable.
i T SURVEILLANCE REQUIREMENTS I 4.12.1 The radiological environmental monitoring samples shall be collected pursuant to Table 3.12-1 from the specific locations given in the table and figure (s) in the ODCM, and shall be analyzed pursuant to the requirements of Table 3.12-1 and the detection capabilities required by Table 4.12-1. 4 h o 1
.t . ; h e re au in in er f 5 or 'd d i* Nj ie s/ r' vir o ru rrr en ts/ r*'* *'i /i r-in) a re &re sarese
\ f
u. do n . ,, , es t? #he scie. TA ., ,, ,1 s/ng/e y*rs;,a.,, in e ta,t;,,, ms n cfo,iny, l !a ,. i a t. w e a y , a -. :( f a s.//ly .e e.r v vst er c e seeve.s bo th ver Ms .
! t i s: . D !. *n r 4 1s s c e n llrarr.s.h e y e e n au A /y3 ois o / !h e. e to 9i n a !, .*. dy o'a 'e a.fe , . e, .t ru a e a. m p ie as x y p n y r i. d e.. 7'A t te n. /fs e f /h e cors k s & ,,,. t
. * * . A ly 2 *f
sl1 A // h& C s e & v'y lE ! & *$ s9 f !b e- A !' / t' t.3 ,# fitr1C C d 94 5i% !W. rt ? Lse d'!$1 l .-
3 ,,a :,..s a . l 7
O V0GTLE - UNIT 1 3/4 12-2
A A V U U TABLE 3.12-1 l 5 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM *,'
~
g r-NUMBER OF REPRESENTATIVE l E EXPOSURE PATHWAY SAMPLES AND SAMPLING AND TYPE AND FREQUENCY Z AN0/0R SAMPLE SAMPLE LOCATIONS I COLLECTION FREQUENCY OF ANALYSIS
" b ^ - 5 6'
1. Direct Radiation (2) for[y routine monitoring stations Quarterly. Gamma dose quarterly.
(URl-DR40) either with two or more dosimeters or with one instrument for measuring and recording dose rate continuously, placed as
. follows:
An inner ring of stations, one in each meteorological sector in the general area of the SITE BOUNDARY M
+
(DR1-DR16); M An outer ring of stations, one in
,O each meteorological sector in apro x6ndet/7 j" ' the 6-to-8-km range from the .A' ~ site (DR17-DR32); and The balance of the stations -{DR33-DR40) to be placed in special interest areas such as population centers, nearby residences, schools, and in one or two areas to serve as control stations. "The number,- media,- frequency, and location-of-samples-may-vary-from-site-to-site This-table-presents-an acceptable-minimum-program-for a-site-at-which-each-entry-is-applicable--Local site characterist-fes must be examined-to determine if-pathways-not-covered by this table-may-significantly re-tr! bete te an individual's dose and should be-included in the-sample-program The ce* !stters in parentheses, e-1, DR1,- A1,-previde-ene-way-of-defining-sample-locations--in-this-specif4catien that can be used te identify the-specific locations in the map (s)-and-table-in-the-00CN. .
.-.- _.- - - . - . - ._ _ - .. - . ~ ~ .... - . . - _- _- .--- - ._-. -...- - - - .- . .. O , O O TABLE 3.12-1 (Continue.1) h RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM , NUMBER OF REPRESENTATIVE SAMPLING AND ! E EXPOSURE PATHWAY SAMPLES AND gg) TYPE AND FREQUENCY Q AND/OR SAMPLE SAMPLE LOCATIONS COLLECTION FREQUENCY OF ANALYSIS
" 2. Airborne :
1 i Radiciodine and Samples from five locations Continuous sampler oper- Radiolodine Cannister: Particulates (Al=AS): ation with sample collec- 1-131 analysis weekly. tion weekly, or more Three samples (Al-A3) from/oc hans frequently if required by close to the three-SITE dust loading. Particulate Sampler: Gross beta radioactivity BOUNDARY locations, in dif ferent sectors,-of-the- analysis following highest-calculated ;. .;;I filter cha I3) and t'
***#*9*~8# I* % gamma isoto c analysis O) '
of composite (by M
One sample (-A4} from the location) quarterly. A vicinity of a community having the highest calcu-lated annual average ground-level 0/Q; and One sample (AS) from a control
location, as-foc-example-15-to- .
34-ka-distant-and-in-th: ?::M preva l ent- wi nd - d i rec t ienca n' 44, g t to node s Jr.s ta a,f er leyond .
3. Waterborne C- isotopic analysis I4)

a. Surface (5) One sample upstream (Wal). Composite sample over 1-month period.(6) t . C si r One sample downstream (Wa3h t , , .,iy_
m b.-Ground - - -Samples from one-or- two-sources Quarterly. E.r*:. isotopic',' and t- .. < --- w, i ...w g g y }7 ;y_jga jk,3o . _ a gg aa.e aas M 5
. . _ _ _ - - . - . . . - . _ _ . _ _ - _ . - . . - - _ . - . - - . _ - _ . _ _ . ~ _ - - . . - . - - - . _ . . . - - . O O - O c, / r a u r u.a /er m,n ,- in ha c d' " '" # " ' ""'# TABLE 3.12-1 (Continued) \ p, L. ,, r 8 RADIOLOGICAL ENVIRONMENTAL MONIl0 RING PROGRAM /
/ '" NUMBER OF 4 REPRESENTATIVE !
E EXPOSURE PATHWAY SAMPLES AND SAMPLING Aft 0 / TYPE AND FREQUENCY SAMPLE LOCATIONS I3) COLLECTION FREQUENCY OF ANALYSIS Z AN0/0R SAMPLE
3. Waterborne (Continued)
Tuo af Composite sample over I-131 analysis on each sae//c b y". Drinking -One- samplesef each of one to three Wcl-- Wc3) of the nearest composite when the dose 2-week period when calculated for the con-
/ I-131 analysis is per-I'{eal, par, water affectedupplies that could be by its discharge, sumptinn of the water :
formed; monthly com- is greater than 1 mrem Go af posite otherwiseg'a,g One samples free a control ,f r,6 u ,,,fae ,, f. p ,y,, y )f Composite location (Wc4). ;r, ,,, , /, , s . , v " d ' for gross beta and gamma
. .. c n ua / e r /, e , / ,,,,, e ~
d isotopic analyses (4) u k ka,.tewry.t.,.e.e,esor,,t),1,, asmonthly. sp> prey Composite
</a te . for g
tritium analysis quarterly, a c .g. Sediment One sample from downstream area Semiannually. Gamma isotopic analysis from with existing or potential semiannually. Shoreline recreational value (Wdl). , Ingestion
4. '
(s)
a. Milk Samples from milking animals Semimonthly
n- Gammaisotopig($emi-4) a in three locations (Ial --la3) animals-are-en-pasturet 1431 analysid monthly when-animals monthly-at-ether-timesr.
a s ,,, /e. ' within -&-km distance having the highest dose potential. If are-en-pastum,-monthly there are none, then one strother-timew , sample from milking animals in each of three areas (Ial-la3) between 4-te3 /4 aboa/Sn></e.s 8-km distant where doses , are calculated to b than 1 arem per yr,489reater 7 One sample from milking animals M_ at a centro 1 location tiety, sJ..a/ w en;/es ,/astard or ley-aula dprefe><d/7i,e a-15-to-30 km-distant-andin-tte least-prevalent wind directiosy'o/ /csse, pre /.s/ence.
-J-
(
\
p O u) TABLE 3.12-1 (Continued) l 1
'=
l 8 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM d e
"' NUMBER OF REPRESENTATIVE l E EXPOSURE PATIMAY SAMPLES AND SAMPLING AND TYPE AND FREQUENCY gg) OF ANALYSIS l Q AND/0R SAMPLE SAMPLE LOCATIONS COLLECTION FREQUENCY e
4. Ingestion (Continued)

b. Fish and '<.,s t.dne sample of e$'d commercially Sample in season, or Gamma isotopic analysis I4)
Inverte- o, and recreationally important semiannually if they on edible portions. l
-4 rates species in vicinity of plant are not seasonal.
discharge area. -(Ib1 li,-) . dsan ' ". .fe e" r' a, y . _ . _
' - ,+ dne sample of-sden_. species in' .r o v ca,,,,ne re as tly er a~al*w"'.a...4 s ,e c ;< s i, . >
f Ag ! N ' areas not influenced by plant v e ir < < /ta..,a st, , ,,,,,a r ts ,J o/f......r //s c L, /' ' discharge (-Ibl0- -Ib-). u pu,;,,, ,p ri,,At-time-ofda>pw'IN'1seau,,. Gamma-isotopic-analyses
} c. food One-sasiple-of-each-peincipal rvest .
g P.m & cts class of-food-products,-from on-edible-portion. 7 d,'w " or any area that--is-irrigated
'I ' ' 'y 'f d d,% . b'y water-in which-liquid plant-wastes-have-been J e /ca y r e,.c s. .. ,,,4 ,' discharged--(Icl =-Ic-). /-,,,, wo ,,.,fe /, ,,r,,,,, , < , , //,, )7 Samples-of three different Monthly during Gamma isotopic (4) and I-131 kinds of-broad-leaf-vegeta- growing season. analysis.
sff f xe t, vo m v ;,, i
./. / r,< , < ,. / s e < /< < s . ) t, ion-grown nearest exh of two-di f ferent-of fsite-loca-tions-of-highest-predicted .J/ /e.,s / c.,e s,..f>/ h 7 'a' nnual-average- ground-level l . , , , , , , .s c...ere/ ) -
D/Q-if-milk-sampling is-not
,,.,,/,,,,...e...'vr f performed-(Ic10-1c13).
ia .. .9 . ./ .s /.s . ? ' Monthly-during Gamma -isotopicg-and 131
, ; One sample-of-each-of-the e"- -<r*-j- similambroad-leaf-vegeta- growing-season. analysis.
tion-grown to-30-ka-dis-tant-in-the--least-prevalent l wind-d i rection-i f-milk-sam-l pling is-not-performed-(Ic20 -
-Ic23):
i
- . . - - - ~- . . _ _ - _ _ _
M^*' N A*/. lae h sa my/e h e.a Ks , w A fe e h sh.Y M (A Y A w. n* m e. N ***"*
' { Dr*y e r inc!emers.' wea.s'isw TABLE 3.12-1 (Continued)'
I TABLE NOTATIONS , ,.,, f , j na e a .. s e-a/>?e,n.m A (1) SpecificparametersofdistanceanddirectionsectorfromtM$$$t$-We Of :ne reactocr, and additional description where pertinent, shall be pro- :
. vided for each and every sample location in Table 3.12-1 in a table and ;
figure (s) in the @TM7 Refer to NUREG-0133, " Preparation of Radiological , Ef fluent Technical Specifications for Nuclear Power Plants," Octcber 197.8, i and to Radiological Assessment Branch Technical Position, Revision 1, , NoveTber 1979. Deviations are permitted from the required sampling schedule if specimens are uncbtainable due to circumstances such as e
" 9f ,' hazardous conditions, seasonal unavailanilitydaad malfunction of ewee- #
Q C, 7 ~ ' ~ ~ me44TiiEpHng equipment,K If specimens are unobtainable due to sampiing equipment malfunction, effort shall be maca to co'aplete corrective action l prior to the end of the next sampling pericd. All caviations frcm the sampling schedule shall be documented in the AnnuaLJ4diologict1 Environ- , h ; /a .,,, ~ mentar WM4*g Report pursuant to Specificationf6.9.1.3 s It is recog-ni:ed that, at times, it may not be possit:1e or prRt'icanle to continue ,
" " " ' " ' to cbtainisartples of the media of choice at the most desired location or w ,.. A r time. In these instances, suitable alternative media and locations may-be ' ,c_ , w.
h. i 4ho4+a made within for30thedays particular pathway in the Radtological in question Environmental and appropriate Monitoring Program. subs '
given-k. tM 006M. Pursuant to Specification 6.14, submit in the next Semiannual Radicactive Effluent Release Report documentation for a change in the CDCM including a revised figure (s) and table for the ODCM reflect- L ing the new location (s),with supporting infermation identifying the cause O of the unavailability o ' samples for the pathway and justifying the selec-tion of the new locatio )(s) for obtaining samples., or /Ac u,us.m/M//dy e/ t
, Weny, suiM/e new loediens (2) One er mere instruments, such as a pressurized ion chamber, for maasuring f and recording dose rate continuously may be used in place of, or in addi- i tion to, integrating cosi:neters. For the purposes of this table, a tnernoluminescent dcsimeter (TL3) is considered to te one phosphor; two i or mere phosphors in a packet art considered as two or more dosimeters. 5 Film badges shall not be used as dosimeters for measuring direct radiation.
FTM 40 .itationt-h-cot-en-ebsolute-numbert-The .d:r ef-4frest444twien t moni tor 4 r.g- sta ti ons-may-to- reduc e d-acc o rd ir.g-to-geograpbkal - 1 1 m i tet4ew, a g. , et en ^'ean-s4te,_some-sectors-wi44-be ev:r = tee-5,-thet-tM =Mr I of- do s.izete r4-may -be-reduc ed ac c o rd i ng ly,--Tae-# : p: ncy-e f--eea ty+ 4-ee reacout.for_.Tt0-systaas-will depend gea th=-cnaracter4stics of the red-fIc-system-used-and-sbould-be-selected-to-obtain-optie deze "for-2t N l within4tnimal-feding,-) (3) Airborne particulate sample filters shall be analyzad for gross beta l radioactivity 24 hours or more after sampling to allow for Madon and (
~!horon daughter decay. If gross beta activity in air particulate samples i is greater than 10 tia.es the yearly mean of control samples, gases ;
isotopic analysis shall be performed on the individual samole's, f 5 O i ! V0GTLE - UNIT 1 3/4 12-7 l l ! t
. ~ .
Te.BLc 3.12-l__(Cortinued) TABLE h0TATI0h5_LCo,r.tinued) . (4) Gamma isotopic analysis seaas the identification and gaantification of gamma-emitting radi.onuelidas that tray be att.-ibutable to the affluents
from the facility.
' (5) The " upstream sa' epic" shall t i taken at a distance 11eyond significant f,1 fluence o.' tne discharge, The "dewnstream" sa.mple stall be taken in ar. , area reyond but,near the mixing zote, u9psteeep--samp12- it:-eveeessey 5 't ,l muet-be-tekee-far en:ughWPGam-u4e-beyewf the p?ad-ir'! r . wat.sr :ha!! te-sampluf an?y 9en-tl.e -rece4Ar.g-nter-is-et414 ate.4*r recreation 4L-act.tvi t iat ,. (6) A<empoe4te-+ ample-45-one !r A,4ch-the quant 4ty-(*14Wt) Of tigeM-sempted is-propertieral to the ptit;' ef flowing 1!q"14-ar.d-ina? 5 tne asthed :f
h i of-: ling -employed-results 4A-a -spec'<aea-4ha*.-is- r';;r;;:nt::h:
Rwi4-f b:. S th44-swiras.s*cuposite sag,le alicto..s
sha;1 be co?lected at time intervals that are very sh.rt (e.g., hourly) raiative to tFs cer;ositing period (e.g. , r.onth2y) in order to assure obtain!ng a r.epresentative sample,
(+)-Grounenete-t2:aples she4-ee-tam wr-;r this tre,et ee-4+ *,*wM-rco-dr4aMeg ; 4 eM rr igat44.Epurpcserr-44-ares: "here-th4.-hydrew14c # g?adieat-or-cocher+t < pcopeWee-arEw14able-foc-seta.r.'n:t :n i trt - (G) The dose shall t,e calculated for the maxima.n organ and age group, u;;ing the metnedology and parameters in the ODCM. {+)--4f- harvest-eceves-vere-than-ence-a-yeney-eampMeg-ohali t:4:efersed j eve 4rg-tech-diocrett-haryest,---If-barvest-gccurr,-continuwelf, r.pling r444e-wanthly.---Attention-shal14eSei6te -f r.cluding-sample +4( ; Me ras-end -root--food-prod. cts,
~ '
e' - no.D.;,,,, .sa ',v as ,r oa e. ject / p r.~/u o e.y ma'.t .br , Y s>-o.t,, o n f u m p+ , t r:
k
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O TABLE 4.22-1 (Continued) 7ABLE NOTATIONS (1)This Itst does r,st e.ean t!nt only'/ as piant these nuclides areeffLen As to be considered.
; other peaks that at e icenti/f able, together with those of the above nuclides, shs11 aisc De analyzed and reported in the Annugl Radio 1cgical
Environmental 4ta ee pursuant to Specification @.9.1.3.)
m 9.vo%g 8ieport i (2) Required detection capabilities for thermo7urinescent dosimeters used for environmental measurenents shall be in accordance with tht recommenda-tions of Regulatory Guide 4.13. (3)The LLO is defined, for purposes of these specifications, as the smallest concentration of radioactive materf at in a carple that will yield a not count, above systeo backgrouca, that wil1 be detected with 95% probability with only 5% probability of falsely concluding tttat a blank observation represents a "real" signal. for a particular measurement syster, which may include ratlochemical , separation: 1 4.66 s D LLD : 3 E - V - 2.22 - Y - exp(-Mt) Where: LLD = the "a priori" Tower limit of detection (pics 0uries per unit mass or volume), s b
= the standard deviation of the background counting rate or of the ,
counting rate of a blank sample as appropriata (counts per ninute),
~
, E = the counting efficiency (counts per disintegration), t l V = the sample sizo (units of mass or volume), ! 2.22 = the netbe' of disintegrations, per minute per picocurie, ! Y
the fractional radiochemical yield, when applicable, A = the radioactive decay constant for the particular radionuclide (sec-2),and I
At =theelapsedtimebetweeneN tal collection, or end of l the sample collection period, and time of counting (sec). Typical values of E, V, Y, and at should be used in the calculation. i I O l ! V0GTLE - INT I , 3/4 22-11 , I
a_.' TASLE 4.12-1.(Continued) TABLE NOTATICNS (Continued) It shoul.d be recognized th-t the LLD is defined as an a priori (before the fact) limit repressnting the capability of a measureme'nt syst4m and not as an e posteriori (aft 6r the fart) limit fer a particciar seAsufement, ActTyses shali be perfctmed in such a manner that the stated LL55 will be acifeved under routine conditiont. Occasionally backg.osnd fluctuations, j er.avotdable small sample cizes, the presence of foterfering cuclides, or otner uncontrollable cirecestances may render tr.ese LLDs unacnf evable. ; in such cases, the contributirag factcts sha11 be -ident111ed and described 10 tree M. nut! Radiciogical Environsentsi 4pe et4fMy Report pursuant to , Spec 1ticaticnltT,.C% s -
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4 i 4 RADJOLOG7 CAL ENVIRONMENTAL MONITORING
~~ L a r,d w H Jro*n +1, e Sa..m n ,a A ym 9'Her Mast ivry he e vc /&d )'
} %) 3/4.12.2 LAND USE GEN 6BS f "~' M'I 5" "d Y - i I i . ' 1 .e l' ' " 'W,e cw , LIMITING CONDITION FOR OPERATION , 1. i
$. 3.12.2 A Land Use~s uev4ensus shall be conducted and shall identify within a i distance of 6-kaf5 miles /the location in each of the 16 meteorological l j
i sectcrs cf the nearest milk animal, the nearest residence, and the nearest i 2
! qardanMIofgreaterthan50m (500 ftz) producing broad leaf vegetation.'
l J H Teveted-releerts :: defhed-4n " guleterj Cuid: 1r111.-2 ;hien 1, ; i My1977,--the-1.an&Use-Census-shall-+1-se-identHy with,in ; dht:.n Of 5 '- c (-hnles)-the-locations-in- each-4f-the 16 =^ tee-c!c;4ca? 5=r+^~ af =11 milk i animals-and- all g:rdens-of-greater-than-50 2 ; retch.; brer M f rr; tat 4aa ] - i ' I APPL!CABILITY: At all times. i r j acTICV. smv r With a Land Use Geaews identifying a location (s) that yields a
~
i a. d calculated dose or dose commitment greater than the values currently ' j beingy,Altdated in Specification 4.11.2.3, pursuant to Specifica-ticQ9 ]d.Dide7tify 4 the new location (s) in the next Semiannual
. i i Radioacttve Effluent Release Report.4 i V .u ,/'les are un > lao, fe .
smy t, Witn a Land Use Genews identifying allocation (s) that yields a ! l t esiculateddcseordosecommitment(viathesameexposurepathway) ! 2C% greater tr.an at a location from/which samples are currently i being obtained in accordance with Specification 3.12.1, add the new l 4 Iccation(s) 'dthin 30 dhthe Padiological Environmental Moni- l tering ProgracKin_the ODCW The sampling location (s), extlud- ! l ing tM control station location, having the lowest calculated dose or dese coraf tment(s), via the same exposure pathway, may be deleted , i fror. this nocitoring program.a&te*44ctobet43-of-the-year 4n W" i ems-ta24;V;tt;fensus;weew:onducte_e,r/ Pursuant to SpecificatM.lb i- ~ f~ submit in tr.e next Semiannual Radioactive Effluent Release Report
~
( documentation fer a change in the 00CM including a revised figure (s) \ , j and table (s) for the OCCM reflecting the new location (s) with ~ informa / i i , tien su; porting the change - -- in sampling locations. l
,n
! c. The previsions of $pecifications 3.0.3 and 3.0.4 are not applicable. I t i i I 1 l
*Scond leaf wjdation-sampl4ng-of-Mr4 east-three dyrcr;;t k';d ef ;;-- +-* 4 --
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4 hy v. .e nni sn e ve v l',*" > i RADIOLCGICAL ENVIRONMENTAL MONITORING ,. , y j ,. f., c ,. j , ,.c 4 /f SURVEILLANCE REQUIREMENTS i . A rvey I - 4.12.2 The Land Use Ceae.es shall be conducted during the growing season at least once per 12 months using that information that will provice the L;;facA 4 results, such as by a door-te-door survey, ;; riel ::.:r;:j, se by consulting ~
"1 local agriculture authoritier,3,The results of the Land Use Gewus shall be inclucao in the, Annual Radiological Environmental 'gr:th; Report pursuant to SpecificationC6. fit 37J N Sar vei//aric e-4 $f b y M C Wi d- CO M l' fl h" f C A C Y w >e enerAeds as r's rinde.
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3/4.12.3 INTERLABORATORY COMPARICON PROGRAM
., r $ LIMITING CONDITION FOR OPERATION . 3.12.3 Analyses shall be performed on all radioactive materials, ' supplied a part of an Interlaboratory Comparison Program that has been approved by the . . Commission,thatcorrespondtosamplesjrequiredbyTable3.12-1. ' APPLICABILITY: At all timas.
# # #" V# " .
I ACTION: l
a. With analyses not being performed as required above, report tne corrective actions taken to prevent a recurrence to the Comission
~
l in the Annual Radiolo ical Environmental Op nti g-Report pursuant , J to SpecificationT S a r " * /M " * * ;
t. The previsions of Specifications 3.0.3 and 3.0.4 are not applicable.
] I SURVEILLANCE REQUIREMENTS s 4.12,3 The Interlaboratory Comparison Program shall be described in the 00CM. F 4 ? A summary of the results obtained as part of the above required Interlaboratory
Ccmparison Frogram shall be included in the Annual Radiological Environmental '~
j ype n ti g Report pursuant to Specification @
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' 's b m ,/ JUSTIFICATIONS FOR DEVIATIONS FROM STS SECTION 3/4.12 . i
3.12.1, Fcotnote a:
' inis footnote was added to clarify that, whereas VEGP is a two-unit site, tha-e will be one site-related radiciogical environT. ental monitoring program. 3.12.1, Action a: j Tn's annual report, which documents radiological envircnmental monitoring
! a:tivities for the calendar year, might more appropriately be entitled the Annusi Radiological Environmental Surveillance Report; this title is used at
our other nuclear plant. Inis change in terminology has been mad
*broughout the remainder of the draft of Section 3/4.12. i 4 \ , 3.:2.{, Action 0:
} Cgnfirmation provides assurance from making unnecessary reports. The woro i
"cenfirmed" is added to ce consistent with the Radiological Assessment .
Branch Technical Position (BTP), Rev. 1 "ovember 1979. !
\
The insertion of the words "frcm the end of the af fected calendar quarter or af ter confirmation, whichever is later" clarifies when a report must be
made. .
l 3.12.1, Actior c:
*t needs to be clarified that this paragraph is not acplicable to cases of '
i teeparary unavailability. It may not he possible to find suitable l replacement locations for milk or vegetation (or other samples). Difficulty
! is being experienced in finding milt animals within 5 miles. Adequate
' vegetation sa. pies may not be available during periods of hot, cold, or dry weather. Replacement locations c.sn be added only when and if adequate samples are available. When samples become permanently or persistently cravailable at a g'ven location, that sampling location is de facto e ceieted. ;
3.12.1, Footnote b

1 ; ! This footrote was added to define the term " confirmed." The definition is i consistert with tne BTP referenced above. i , I 1
--_._--_______--__________._-______-___._______-_._-__--_--_-____--__-_-_-_l
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Table _3.12-1: General: o Designators for sample location? are provided in the DDCM and have been deleted from this table. Distances are specified in miles as opposed to kilometers. The unit of miles is preferred since maps of the area around the plant are based on that unit. Wherever a range is specified for a sample location, the words "approximately" or "at about" have been inserted to achieve some needed flexibility. Specific: i Footnote *: i This footnote was deleted on the basis that it simply provides guidance as to how this specification should be written. As such, it should net'be a part of these Technical Specifications. Item 1: This footnote was deleted on the basis that it simply provides guidance as to how this specification should be written. As such, it should not be a O part of these Technical Specifications. I Item 2, " Airborne"- The words "of the highest calculated annual average ground level D/Q" were deleted on the basis that there is not a large variation in the calculated values of the annual average D/Q at the site boundary among the azimuthal
sectors. Practical considerations may prescribe the most suitable location as one other than that with the highest or next highest D/Q.
The words "as for example 15-30 Km distant and in the least prevalent wind direction" were deleted on the basis that they are instructional and should 4 not be part of these Technical Specifications. The insertion of the words "at about 10 miles distant or beyond" ensures i that the D/0 value at the control station is significantly lower than at the i indicator location. Item 3, " Waterborne"- Groundwater samp' ling is not required on the basis that there are two distinct water zones underlying the plant site - an unconfined water table which is replenished by natural precipitation and a deep aquifer which is isolated from the water table by a thick aquiclude. The area on which the
) plant is situated is bounded by stream channels that act as drains for the
0) A _/ s water table, thereby intercepting groundwater that moves laterally and prevents inflow or outflow to adjacent areas. (See VEGP Draft Environmental Statement Section 4.3.1.2.) Drinking water onsite is taken from the confined aquifer. Groundwater is not used for irrigation purposes within a few miles of tne piant si.te. At each downstream water treatment plant a composite sample of river water is taken near the intake and a grab sample of finished water is collected. We have been unable to obtain composite samples of finished water. Item 4, "Ingastion"-
a. Milk:
The insertion of the words "about 10 miles distant or beyond" will ensure that the D/Q value at the control station is significantly lower than that at the indicator location. The words "when animals are on pasture, monthly at other times" were deleted since milk animals are on pasture essentially year round. Tne addition of new note 9 eliminates the need for the I-131 analysis when the gamma isotopic analysis is sufficiently sensitive to meat the LLD for I-131. Otherwise, a separate analysis of I-131 will be performed. () The addition of new note 8 specifically defines a milk animal. The deletion of a specific range for the control location for milking animals is based on the fact that milk animals are scarce and it is not practical to place unnecessary restrictions on control locations.
b. Fish:
The revisions to this portion of Table 3.12-1 are based on the following: Only fish species are collected because sufficient quantities of invertebrates suitable for human consumption, such as oysters or clams, have not been found near the plant. Not infrequently, it is difficult to obtain adequate samples of any commercially or recreationally important species of fish. Sampling of seasonal fish is listed separately since sampling at the control station is not meaningful and the sampling periods may be indicated.
c. Food Products:
This was deleted on the basis that there is no downstream use of river water for irrigation purposes. r~~ i O
O (m_/ Grass may be substituted for leafy vegetation cn the basis that grass has been found to be an effective sampling medium. Grass is available during a larger portion of the yeer than broad-leaf vegetation. Tne basis for two onsite sampie loca-1:ns near tha site boundary in different sectors is as follows: The BTP suggests the collection of vegetation at the site boundary is preferable to collections at an actual nearby offsite garden. There is not a large variation in the calculated values of the annual average D/Q at the site boundary amongst the azimuthal sectors. Practical considerations may prescribe the most suitable location as one other than that with the highest or next highest D/Q. The main advantages of maintaining an onsite grass or vegetation plot are:
1) One is not subject to the good will of a private citizen for obtaining adequate samples; and

2) The calculated annual average ground level 0/Q is likely higher than any of the offsite gardens located in the annual land survey.
Concerning the sample from a control location at about 10 miles distant or sg beyond, the important thing is that the D/Q value at the control station should be significantly lower than at the indicator location. This may be assured by the 10-mile distance. Table 3.12-1, Table Notation: Note 1: The addicion of the words "a point midway between the center of the two" is based on the fact that all area maps have been constructed with the midway point as the origin. The centers of the two reactors are a few hundred feet apart. The addition of the words "Each sample location will be design &ted by a number, name or some other label" was added to clarify that each location will be designated with a label in the ODCM. The words " dry or inclement weather" were added to indicate that deviations may be necessitated as a result of the weather. The words "or other justifiable reasons" were added to allow for deviations resulting from unforeseen circumstances other than those explicitly called out. Note 2: The words deleted from this note were instructional in nature and as such
, should not be part of these Technical Specifications.
%Aw Note 5:
The deleted words are not applicable to VEGP. Note a: It is impractical to obtain liquid samoles which can be demonstrated to be proportional to flowrate in the Savannah River due to the complex nature of flaw throughout a cross-section of the river under various flow conditions. True proportionality could only be achieved by relating the size of each alliquot to the integrated flowrate across the river cross-section at the time each aliquot is taken; this informaton cannot be obtained at reasonable cost and effort. However, because the Savannah River is protected from extreme flowrates by dams, representative sampling can reasonably be achieved by obtaining samples with automatic sampling equipment set to obtain aliquots at short intervals (on the order of hourly) with the sampling interval adjusted periodically in accordance with river flow conditions. Note 7 (STS): This note was deleted for the reasons already discussed under Item 3,
" Waterborne."
Note 9 (STS): p(_/ This note was deleted on the basis that there is not downstream use of river water for irrigation purposes. Table 3.12-2: The separation of Zr-95 and Nb-95 and Cs-137 and Ba-140 removes any uncertainty regarding the reporting level or LLD.* The reporting levels for the specific radionuclides in the separated chains were derived on the same casis as the other radionuclides in Table 3.12-2. The reporting level is the concentration at which anhJal Appendix I dose limit to the most sensitive organ and age group would be acquired with 1 year's expusure. Table 4.12-1: The values in Table 4.12-1 are the same as those found in the BTP rounded to one significant figure. Laboratories usually report such results to one significant figure.
Members of the decay chain may be determined separately by gamma isotopic s analysis.
g.- s/ Note 1: The addition of the words "as plant effluents" eliminates the need for reporting naturally occurring radionuclides which are not part of the plant effluents. 3/4.12.2: The term " survey" is believed to be more appropria e than the word
" census."
The statement pertaining to elevated releases was deleted on the basis that VEGP does not make elevated releases. The addition of the statement excluding land within the Savannah River Plant from the survey is based on the fact that this land does not constitute a pathway to the public since access is controlled by SRP. 3/4.12.2, Action b: Tne addition of the words "if samples are available" is based on the fact that, although milk and vegetation samples have generally been available upon request, it cannot be assumed that this will always be the case. For the VEGP site, no significance is recognized in delaying deletion until after October 31 since there is year round pasturing, and vegetation samples C, N / become scarce in hot, dry, and cold weather. 4.12.2: The additions to this surveillance were made to provide clarification as to the manner in which tne land use survey may be conducted. 3.12.2, Footnote *- This footnote was deleted on the basis that, although grass or leafy vegetation is to be sampled at two site boundary locations, the garden survey is to be performed for dose calculations due to plant effluents. 3/4.12.3: The words "and analyses" were inserted to clarify that only those analyses of aoplicable samples as required by Table 3.12-1 need be performed in the Interlaboratory Comparison P ogram. O v 0053v
a. ms 4
t BASES FOR SECTIONS 3.0 AND 4.0 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS O O l I
, - w----..-- --r- .w__. ,.
j -eM i i i 4 1 ! 4 g i j j NOTE The BASES contained in succeeding pages sumarize l the reasons for the Specifications in Sections 3.0 j^ and 4.0, but in accordance with 10 CFR 50.36 are not part of these Technical Specifications. I lO . ) I i 1 .. I, I 5 4 l 1
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W.i% ' 1 - l 3/4.0 APPLICABILITY BASES The specifications of this section provide the general requirements applicable to each of the Limiting Conditions for Operation and Surveillance Requirements within Section 3/4. In the event of a disagreement between the requirements stated in these Technical Specifications and those stated in an applicabl.e Federal Regula- - tion or Act, the requirements stated in the applicable Federal Regulation or Act shall take precedence and shall be met. 3.0.1 This specification defines the applicability of each specification in terms of defined OPERATIONAL MODES or other specified conditions and is provided to delineate specifically when each specification is applicable. 3.0.2 This specification defines those conditions necessary to constitute compli-ance with the terms of an individual Limiting Condition' for Operation and associated ACTION requirement. 3.0.3 , The specification delineates the measures to be taken for those circum-stances not directly provided for in the ACTION statements and whose occurrence would violate the intent of a specification. For example, Specification 3.5.2 requires two independent ECCS subsystems to be OPERABLE and provides explicit ACTION requirements if one ECCS subsystem is inoperable. Under the requirements of Specification 3.0.3, if both the required ECCS subsystems are inoperable, within 1 hour measures must be initi-ated to place the unit in at least HOT STANDBY within the next 6 hours, and in at least HOT SHUTDOWN within the following 6 hours. As a further example, Specification 3.6.2.1 requires two Containment Spray Systems to be OPERABLE and provides explicit ACTION requirements if one Spray System is inoperable. Under the requirements of Specifica-tion 3.0.3, if both the required Containment Spray Systems are inoperable, within 1 hour O measures must be initiated to place tM Unit in at least HOT STANOBY within the next 6 hours, in at least HOT SHUTDOWN within the following 6 hours, and in COLD SHUTDOWN within the subsequent 24 hours. It is acceptable to initiate and complete a reduction in OPERATIONAL MODES in a shorter time interval than required in the ACTION statement and to add the unused portion of this allowable out-of-service time to that provided for operation,in subsequent lower OPERATION MODE (S). Stated allowable out-of-service times are applicable regardless of the OPERATIONAL MODE (S) in which the inoperability is discovered but the times provided for achieving a mode reduction are not applicable if the ine'perability is discovered in a mode lower than the applicable mode. For exam-ple if the Containment Spray System was discovered to be inoperable while in STARTUP, the ACTION Statement would allow up to 156. hours to achieve COLD SHUTDOWN. If HOT STANDBY is attained in 16 hours rather than the allowed 78 hours, 140 hours would still be available before the plant would be required to be in COLD SHUTDOWN However, if this system was discovered to be inoperable while in HOT STANOBY, the 6 hours provided to achieve HOT STANDBY would not be additive to the tine available to achieve COLD SHUTDOWN so that the total allowable time is reduced from 156 hours to 150 hours. I 3.0.4 This specification provides that entry into an OPERATIONAL MODE or other specified applicability condition must be made with: (1) the full complement of required systems, equipment, or ccaponents OPERABLE and (2) all other paran.eters as specified in the Limiting Conditions for Operation being met without regard for allowable deviations and out-of-service provisions contained in the ACTION statements. The intent of this provision is to ensure that facility operation is not initiated l 4 with either required equipment or systems inoperable or other specified. limits being exceedec. Exceptions to this provision have been provided for a limited number of specifi- , O cations when startup with inoperable equipment would not affect plant safety. These excections are stated in the ACTION statements of the appropriate specifications. l VCGTLE - UNIT 1 B 3/4 0-1 l i N, ,.._ -.- ,,. -,-,- .- -- -...-,- - - --..-..---._ - -
l APPLICABILITY BASES 4.0.1 This specification provides that surveillance activities necessary i to ensure the Limiting Conditions for Operation are met and will be performed during the OPERATIONAL MODES or other conditions for which the Limiting Condi-tions for Operation are applicable. Provisions for additional surveillance activities to be performed without regard to the applicable OPERATIONAL MODES i or other conditions are provided in the individual Surveillance Requirements. ' Surveillance Requirements for Special Test Exceptions need only be performed when the Special Test Exception is being utilized as an exception to an individual specification. 4.0.2 The provisions of this specification provitie allowable ,olerances for performing surveillance activities beyond those specified in the nominal surveillance interval. These tolerances are necessary to provide operational flexibility because of scheduling and performance considerations. The phrase "at least" associated with a surveillance frequency does not negate this allowable tolerance value and permits the performance of more frequent surveillance activities. The tolerance values, taken either individually or consecutively over three test intervals, are sufficiently restrictive to ensure that the reliability associated with the surveillance activity is not significantly degraded beyond that obtained from the nominal specified interval.' 4.0.3 The provisions of this specification set forth the criteria for determination of compliance with the OPERABILITY requirements of the Limiting Conditions for Operation. Under these criteria, equipment, systems or components are assumed to be OPERABLE if the associated surveillance activities have been satisfactorily performed within the specified time interval. Nothing in this provision is to be construed as defining equipment, systems or components OPERABLE when"such items are found or known to be inoperable although still meeting the Surveillance Requirements. Iter: ::y 5: det: *n:d "n:per:ble-ith thi.; :p :ification. dud ~; ure, dur%g curve!'1:ne: t::t:, 0 * ::::rd: :: .; Ther:f:re, ACTICM state :nt: 0r: :ntered wh:n the Srv:ill:n:: R:quir ::nt: cheuld h:ve beer perfe :d r:ther th:n at the ti : it 1: Jt:::vered th:t th:
-tect: u:re n:t perf:rmed.
4.0.4 This specification ensures that tha surveillance activities associated with a Limiting Condition for Operation have been performed within
the specified time interval prior to entry into an OPERATIONAL MODE or other applicable condition. The intent of this provision is to ensure that surveil-I lance activities have been satisfactorily demonstrated on a current basis as required to meet the OPEPABILITY requirements of the Limiting Condition for Operation.
Under the terms of this specification, for example, during initial plant i STARTUP or following extended plant outages, the applicable surveillance l activities must be performed within the stated surveillance interval prior to l placing or returning the system or equipment into OPERABLE status. O VCGTLE - UNIT 1 B 3/4 0-2 l
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APPLICABILITY BASES 4.0.5 This specification ensures that inservice inspection of ASME Code Class 1, 2 and 3 components and inservice testing of ASME Code Class 1, 2 and 3 pumps and valves will be performed in accordance with a periodically updated version of Section XI of the ASME Boiler and Pressure Vessel Code and Addenda as required by 10 CFR 50.55a. Relief from any of the above requirements has been provided in writing by the Commission and is not a part of these Technical Specifications. This specification includes a clarification of t!)e frequencies for per-forming the inservice inspection and testing activities required by Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda. This clarification is provided to ensure consistency in surveillance intervals throughout these Technical Specifications and to remove any ambiguities relative to the frequencies for performing the required inservice inspection and testing activities. Under the terms of this specification, the more restrictive requirements of the Technical Specifications take precedence over the ASME Boiler and Pressure Vessel Code and applicable Addenda. For example, the requirements of Specification 4.0.4 to perform surveillance activities prior to entry into an s OPERATIONAL MODE or other specified applicability condition takes precedence over the ASME Boiler and Pressure Vessel Code provision which allows pumps to be tested up to 1 week after return to normal operation. And for example, the Technical Specification definition of OPERABLE does not grant a grace period before a device that is not capable of performing its specified function is declared inoperable and takes precedence over the ASME Boiler and Pressure Vessel Code provision which allows a valve to be incapable of performing its specified function for up to 24 hours before being declared inoperable. O VCGTLE - UNIT 1 B 3/4 0-3
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3/4.1 REACTIVITY CONTROL SYSTEMS BASES 3/4.1.1 BORATION CONTROL 3/4.1.1.1 and 3/4.1.1.2 SHUT 00WN MARGIN A sufficient SHUTDOWN MARGIN ensures that: (1) the reactor can be made suberitical from all operating conditions, (2) the reactivity transients asso-ciated with postulated accident conditions are controllable within acceptable limits, and (3) the reactor will be maintained sufficiently subcritical to preclude inadvertent criticality in the shutdown condition.
.:n 100 G'S / .2n.t' 2 SHUTDOWN MARGIN requirements vary throughout core life as a function of fuel depletion, RCS boron concentration, and RCS T,yg. 7he most restrictive condition occurs at E0L, with T,yg at no load operating temperature, and is associated with a postulated steam line break accident and resulting uncon , , , , trolled RCS cooldown. In the analysis of this accident, a minimum SHUTOOWN " MARGIN of B-@3 Ak/k is required to control the reactivity transient, j Accordingly, the SHUTDOWN MARGIN requirement is based upon this limiting condition and is consistent with FSAR safety analysis assumptions. 'dith T avg S less-than-2009Fr the-reactivity tr:n:fent: r::elting f n; : p::tel:t:d :t::=
line break co ld;wn cr =ini=:1 :nd : 1% ak/k SHUTDOW" "^3CI" pr:vid::- 9pt: pr:t::ti:n.
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! 3/4.1.1.3 MODERATOR' TEMPERATURE COEFFICIENT j
The limitations on moderator temperature coefficient (MTC) are provided i to ensure that the value of this coefficient remains within the limiting condition assumed in the FSAR accident and transient analyses. The MTC values of this specification are applicable te a specific set of plant conditions; accordingly, verification of MTC values at conditions other i than those explicitly stated will require extrapolation to those conditions in j order to permit an accurate comparison. The most negative MTC, value equivalent to the most positive stoderator ! density coefficient (MDC), was obtained by incrementally correcting the MDC used in tne FSAR analyses to nominal operating conditions. These corrections l 4 i l O V0GTLE - UNIT 1 8 3/4 1-1 l L _ _ _ .
Insert A to B 3/4 1-1 In MODES 3, 4, and 5 the most restrictive condition occurs at BOL, associated with a Boron Dilution Accident. In the analysis of this accident a minimum SHUTDOWN. MARGIN of (Later) delta-k/k (MODES 3 and 4) and (Later) delta-k/k (MODE 5) is required to allow the operator 15 minutes from the initiation of the Source Range High Flux at Shutdown Alarm to reactor criticality. Accordingly, the SHUTDOWN MARGIN require-ment is based upon this limiting requirement and is consistent with the FSAR accident analysis assumptions. O . O
REACTIVITY CONTROL SYSTEMS BASES MODERATOR TEMPERATURE COEFFICIENT (Continued) , involved subtracting the incremental change in the MDC associated with a core condition of all rods inserted (most positive MDC) to an all rods withdrawn condition and, a conversion for the rate of change of moderator density with temperature at RATED THERMAL POWER conditions j This value of the MDC was then transformed into the limiting MTC valuef-3.9J x 10 4 Ak/k/* F. The MTC value of;[-3.07 x 10 4 Ak/k/*F represents a conservative value (with corrections for burnup and soluble boron) at a core condition of 300 ppm equilibrium boron concentration and is obtained by making these correcticns to the limiting MTC value of [-3.9] x 10 4 Ak/k/*F. The Surveillance Requirements for measurement of the MTC at the beginning and near the end of the fuel cycle are adequate to confirm that the MTC remains within its limits since this coefficient changes slowly due principally to the reduction in RCS boron concentration associated with fuel burnup. 3/4.1.1.4 MINIMUM TEMPERATURE FOR CRITICALITY ggj p This specification ensures that the reactor will not be made[ critical v with the Reactor Coolant System average temperature less than E6413*F. This limitation is required to ensure: (1) the moderator temperature coefficient is within itsanalyzed temperature range, (2) the trip instrumentation is within
, J ts normal operating range, (3) the ? 12 interleck is ab;ve its setpcint, J (+) the pressurizer is capable of being in an OPERABLE status with a steam bubble, and ( ) the reactor vessel is above its minimum RTNDT temperature.
3/4.1.2 BORATION SYSTEMS
,and Tne Boron Injection System ensures that negative reactivity control is available during each mode of facility operation. The components required to perform this function include: (1) borated water sources, (2) charging pumps, (3) separate flow paths,4(4h boric acid transfer pumpss f8A-essociated ll cat Treeing-Systems, and-(5) anj;;rgency power supply frek OPCa0LE dittd 9*****t M - rn e.
With the RCS avarage temperature above 200*F, a mininum of two boron injection flow paths are required to ensure single functional capacility in
' the event an assumed failure renders one of the flow paths inoperable. The boration capability of either flow path is sufficient to provide a SHUTDOWN i
O l V0GTLE - UNIT 1- B 3/4 1-2 y +, ,e-- __- . , , , y _.-.
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\ ~' \ B0 RATION SYSTEMS (Continued) \ / \ MARGIN f. .. exp.cted spacetirg cenditian3 f 1.0*' M/h after xenon decay The maximum expected boration capability requirement \andcooldownto200*F. occurs at EOL from , full power equilibrium xenon conditio E-Ele &3 gallons of 7c00Tppm horated water from the boric acid storage tanks of 2000 ppm borated water from the refueling water storage 7MY. or tankP" 522] gal oneh je g, /,, e, (RWST).
With the RCS temperature below 200 F, one Boron Injection System is acceptable withcut single failure consideration on the' basis of the stable reactivity condition of the reactor and the additional restrictions prohibiting CORE ALTERATIONS and positive reactivity changes in the event the single Boron Injection System becomes Inoperable.
" '-itation for-a-aaxiace of cne centrifug:1 ch:rging pump to be OPERASLE-and-the-Sur+efMance-Requirem:nt te verify all charging p"=e except the-required-OPERABLE pump-to-be-inopecable Sciet [275]*c provid:: :::urance that : ::: addition-pressure-trans4ent-can4e-reHav d by the speration of a a~m omu "'*~ ~ *
(Laer) (Lder) s Thebaroncapadilityrequiredbelow200*Fissufficienttoprovidea SHUTDOWN MARGIN of -1!E M/k after xenon deca and cooldown froa 200*F to [v) 140*F. This condition requires either dg311ons of i7000}" ppm borated water from the boric acid storage tanks or(4dvJgallont.of 2000 ppm borated water from the RWST. Qu /, n /u, c The contained water volume lirr.its include allowance for water not available because of discharge line location and other physical characteristics. The limits en contained water volume and/m/C.Sboron ccncentration o also ensure a pH value of between X8.5Thnd [41-O} for the solution recirculated within containment after a LCCA. This pH band mininizes the evolution of iodine and minimizes the effect of chloride and caustic stress corrosion on mechanical systems and components. The OPERABILITY of one Boron Injection System during REFUELING ensures that this system is available for reactivity centrol while in MODE 6. 3/2.1.3 "GVABEE-CONTROL AISE"SLIES
-The-spec 4f4 cations-of-this- section-enswe-thatM1) Occeptabl+-pcwer dietri-bution4iefts-are-maintained -(M--the ninimur SHUTG9WN-MARGIM-4s m:intained, and 3
(3) te retential-e#ects-of-rod-misaMgnment-on-assoc-inted ::cid:nt :n:ly::: re Matted,-CP4RA8ikM cf the4ontrol-rod-position-tedicatcr: i: r;qu'r:d to determine control-fod-positisns--and-thereby-easwe re pliance 'ith the ecrtr4 rod-aMgr ent 2nd %section4imits.-Verif4 cation-taat-the4tgital--Red P;;ition indicator-agrees-with-the-demanded ;:sition uith4- t I? r* =ps =+ ?A, AS, I?n,
,V0GTLE - TWIT 1. B 3/4 1-3 i
\ REACTIVITY CONTROL SYSTEMS BASES i / \
MOVABLE CONTROL ASSEMBLIES (Continued) and 22 \ steps withdrawn for the Control Banks and 18, 210, and 228 s eps with-drawn fornthe Shutdown Banks provides assurances that the Digital d Position Indicator is operating correctly over the full range of indicatio . Since the Digital Rod Position Indication System does not indicate the ac}5al shutdown rod position between 18 steps and 210 steps, only points in the indicated ranges are picked for' verification of agreement with demanded positjdn. TheACTIONstatementswhichpermitlimitedvariations'/ from the basic requirements are accompanied by additional restrictions w'ich h ensure that the original design criteria are met. Misalignment of a rod requires measurement of peaking factors and'a restriction in THERMAL POWER / These restrictions pro-vide assurance of fuel rod integrity during continu operation. In addition, those safety analyses affected by a misaligned to are reevaluated to confirm that the results remain valid during future oper ion. 1 ., The maximum rod tirop timhrestriction is consistent with the assumed rod
drop time used in the safety arialyses. Mea rement with T,yg greater than or equal to [541]*F and with all reactor coo nt pumps operating ensures that the measured drop times will be representat# e of insertion times experienced during a Reactor trip at operating coo tions.
Control rod positions and OPER IllTY of the rod position indicators are required to be verified en a nomi 1 basil.of once per 12 hours with more fre-quent verifications required if n automatic monitoring channel is inocerable. These verification frequencies re adequate for assuring that tne applicable LCOs are satisfied. \N
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/m k,) s 3/4.1.3 MOVABLE CONTROL ASSEf1BLIES The specifications of this section are necessary to ensure that the follow-ing requirements are met at all times during normal operation. By observing that the RCCAs are positioned above their respective insertion limits during normal operation,
1. At any time in life for MODE 1 and 2 operation, the minimum SHUTDOWN MARGIN will be maintained. For operational '10 DES 3, 4, 5, and 6, the reactivity condition consistent with other specifications will be maintained with all RCCAs fully inserted by observing that the boron concentration is always greater than an appropriate minimum value.

2. During normal operation the enthalpy rise hot channel factor, FAH' Will be maintained withir acceptable limits.

3. The. consequences of an ejected RCCA accident will be restricted below the limiting consequences referred to in the ejected rod analysis.

4. The core can be made subcritical by the required SHUTDONN MARGIN with one RCCA stuck. In the event of an RCCA ejection, the core can be made subcritical with two RCCAs stuck,where one of the RCCAs is assumed to be the worst ejected rod control assembly.

5. The trip reactivity assumed in the accident analysis will be available.
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6. Dropping an RCCA into the core or statically misaligning an RCCA during normal operation will not violate the thermal design basis with respect to DNBR.

7. The uncontrolled withdrawal of an RCCA will result in consequences no more severe than presented in the accident analysis.

8. The uncontrolled withdrawal of a control assembly bank will not result in .a peak power density that exceeds the center line melting criterion.
OPERABILITY of the control rod position indicator channels (LC0 3.1.3.2) is required to determine control rod positions and thereby ensure compliance with the control rod alignment. SPERABILITY of the Demand Position Indication System (LC0 3.1.3.2) is required to d?termine bank demand positions and thereby ensure complian:e w:th the insertion limits. The ACTION statercents which permit limited variations from the basic requirements are accompanied by auditional restrictions which ensure that some of the original criteria are met. Misalignment of a rod requires measure-ment of peaking factors or a restriction in THERMAL POWER, either of these restrictions provide assurance of fuel rod integrity during continued opera-tion provided no further abnormal condition develops. s For Specification 3.1.3.1 ACTIONS b and c it is incunbent upon the plant to , g verify the trippability of the inoperable control rod (s). This may be by verification of a control system failure, usually electrical in nature, or that the failure is associated with the control rod stepping mechanism. In V0GTLE-UNIT 1 B 3/4 1-4
the event the plant is unable to verify the rod (s) trippab'ility, it must be assumed to be untrippable and thus . fall under the requirements of ACTION a . Assuming a controlled shutdown from-10J% RATED THERMAL POWER, this allows approximately four hours for this verification. The maximum rod drop time permitted by (LC0 3.1.3.4) is consistent with the assumed rod drop time used in the accident analyses. Measurement with T 551 degrees-F and with all reactor coolant pumps operating ensures t$aE the measured drop times will be representative of insertion times experienced during a reactor trip at operating conditions. Bank demand positions and OPERABILITY of the rod position indicators are required to be verified on a nominal basis of once per 12 hours with more frequent verifications required if an automatic monitoring channel is IMOPEMBLE. These verification frequencies are adequate for assuring that the applicable LC0's are satisfied. O V0GTLE-UNIT 1 B 3/4 1-5 l
b l 3/4.2 POWER DISTRIBUTION LIMITS i BASES The specifications of this section provide assurance of fuel integrity during Condition I (Normal Operation) and II (Incidents of Moderat~e Frequency) events by: (1) maintaining the minimum DNBR in the core greater than or equal to 1.30 during normal operation and in short-term transients, and (2) limiting the fission gas release, fuel pellet temperature, and cladding mechanical properties to within assumed design criteria. In addition, limiting the peak linear power density during Condition I events provides assurance that the initial conditions assumed for the LOCA analyses are met and the ECCS acceptance criteria limit of 2200'F is not exceeded. . The definitions of certain hot channel and peaking factors as used in these specifications are as follows: Fq(Z) Heat Flux Hot Channel Factor, is defined as the maximum local heat flux on the surface of a fuel rod at core elevation Z divided by the average fuel rod heat flux, allowing for manufacturing tolerances on fuel pellets and rods; Nuclear Enthalpy Rise Hot Channel Factor, is defined as the ratio of Fh the integral of linear power along the rod with the highest integrated 4 power to the average rod power; and F xy(Z) Radial Peaking Factor, is defined as the ratio of peak power density to average power density in the horizontal plane at core elevation Z. 3/4.2.1 AXIAL FLUX DIFFERENCE Thelimi5sonAXIALFLUXDIFFERENCE(AFD)assurethattheF(Z) q upper
- bound envelope of M 2 times the normalized axial peaking factor is not exceeded during either normal l operation or in the event of xenon redistribution following power changes. 2 3c Target flux difference is determined at equilibrium xenon conditions. - _ The '_ N : ;th rods may be pcsitioned within the core in accordance with their respective insertion limits and should be inserted near their normal position for steady-state operation at high power levels. The value of the target flux difference obtained under these conditions divided by the fraction of RATED THERMAL POWER is the target flux difference at RATED THERMAL POWER for the associated core burnup conditions. Target flux differences for other '
THERMAL POWER levels are obtained by multiplying the RATED THERMAL POWER value by the appropriate fractional THERMAL POWER level. The periodic updating of the target flux difference value is nec.essary to reflect core burnup considerations. O V0GTLE - UNIT 1 B 3/4 2-1 1
PCWER DISTRIBUTION LIMITS IG' BASES AXIAL FLUX DIFFERENCE (Continued) Although it is intended that the plant will be operated with the AFD-within the target band required by Specification 3.2.1 about the t'arget flux difference, during rapid plant THERMAL POWER reductions, control red motion - will cause the AFD to deviate outside of the target band at reduced THERMAL POWER levels. This deviation will not affect the xenon redistribution suffi-ciently to change the envelope of peaking factors which may be reached on a subsequent return to RATED THERMAL POWER (with the AFD within the target band) provided the time duration of the deviation is limited. Accordingly, a 1-hour , penalty deviation limit cumulative during the previous'24 hours is proyided foy operation outside of the target band but within the limits of FigureJ3.2-l'T For
- while at THERMAL POWER levels between 50% and 90% of RATED THERMAL POWER.
THERMAL POWER levels between 15% and 50% of RATED THERMAL POWER, deviations of tne AFD outside of the target band are less significant. The penalty of 2 hours actual time reflects this reduced significance. Provisions for monitoring the AFD on an automatic basis are derived from ' the plant process computer through the AFD Monitor Alarm. The computer deter-mines the 1-minute average of each of the OPERABLE excore detector cutputs and provides an alarm message immediately if the AFD for two or more OPERABLE excore channels are outside the target band and the THERMAL POWER is greater
than 90% of RATED THERMAL POWER. During operation at THERMAL POWER levels ,
s 7-s -) between 50% and 90% and between 15% and 50% RATED THERMAL POWER, the computer \'/ outputs an alarm message when the penalty deviation accumulates beyond the limits of 1 hour and 2 hours, respectively. Figure B 3/4 2-1 shows a typical monthly target band. 2 / ' . h2-an63/4. 0. 2 'EA' rLUX HOT CP":EL "'CTOR, xd RC: rLC',; RAT: AND EisM-ENMLP 5: MOT Cen:L L'CR
' - The4imits-on-heat,41ux-hot-channe4-factoMG5 fl= r:te, :nd axle:r enthalpy -ice- hot--channe4.-factor-ensure-that: (1) th; dc;i;n i iimit.; :.. peJ ' ::1
pcwe- density--andminimum-GNER-are-not-exceeded -and (2) in th: ;rt Of s-iOCA-the peak fuel clad-temper:ture "i not-exceed the 2200 F ECCS ::::pt:n :
C-itari: '#~it. Each-of thee: it measurab!c but v normally crly be detmeined i per# edi c2y es-soeci +4 ed-4 Sc.eci?ications 4,2. 2 and A. 2. 3. his peried'c serve !'l an:: u suffic4ent-to-ensure-that-the 'imit: Or: maint ined pr:vid:d: Control . rods _in-a-single -group move together with aa 4rdi"4d"=1 *^A 2. is,ser-tion-differing by =cre-than- 12 step:, i dicated, n # rem the group dem:nd petitien;
5. Centrol red-group: 2re : quenced 'ith cverlapping grcup: :: der -ibed Gn-5; ci'ic: tier 2.1.2. ;
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1%tCATE3 ulAL U.W Cl#9PCJCI
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[v; x 20WER CISTRIBUTIGN LIMITS
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HEAT'fLUV HOT CHANNEL FACTOL ar.d RCS FLOW RATE AND htJCLEAR ENTHALPY RISE
~
F CSAMiEL Ii QOJ (Continudd) c.NJhe contro! rod insertion limits of Specifications 3.1.3p and 3gl.3.6 are maintained; and j
d. The axial power distribution, expressed in terms of , AXIAL '
FLUX DIFFERENCE, is maintained within the ifr.its. N
~ /
cy vill be maintained within its lirdts providad. Coltditions a. through
d. aceva ara maintained. As noted on Figure 3.2-3 RCS/ fiow rate and Ff :nay be " traded off" against.or.e anather (i.e., a low vaasurad RCS ficw rate h a: rptacle if the measured F H is also low) to ensde t!'at the calculatec CNE'l wil! nct be belu the desig'rt CNBR vaha. The ra.r axation of Fh, as a function of THEPPAL POWER allows changhs in the er.dialpswer shape for all permissible
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red ir,sertion limits. f r'3 R as calculated in Specificati' 3.2.3 and used in Figure 3.2-3, accounts ( > 1s valt.a is used in the various accicent for F'N less tnaa or et;ual to 1.49. analy s where F influen;espera.e[atgothertaanONSR,e.g.,peakclad temperatura,andthusisthemaxi[m das hasurec" vaius allowed. f g
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Fuel rod.: wing recuces the value of DNB ratio. Credit is available to offset-this reduction in the generic margin. The teneric margins, totaHng
~ 9.1". DN33 comp'.etely of fset any rod bow penaltieb., This margin Mcluces the following: x 1
n. Design limit 0.!BR of (1.20 vs L28),

c. Grid Spacing (7. ) of Q.045 vs 0.059], -

c. Taersa) Diffucion Ccafficient of [0.038 vs 0.059],

d. DNBR Multiplier of [0.85 vs 0.28], ar.d

c. Pitch reduction, ,
The acplicable values of rod tow ;enalt.ies are refereoced in the FfAR. s 4.
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( ,/ \ l V0GTLE - UNIT 1 0 3/4 2-4 \ ,
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p0'.iER OISTRIBUTION Littifs
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BASES __ 3/4,2.2 and 3/4.2,3 HEAT FLUX HOT CHANNEL FACTOR, HUCLEAR ENTHALPY HOT CHANSEL FACTOR The li:ntts on heat flux bot channel. factor, and nuclehr enthalpy riso hot
channel factor ensure that 1) the cesign limits on peah local power d.9nsity and miniro EBR are not exceeded and 2) in the event of a LOCA the peak fuel clad temperature wfll not exceed ttie 2200 F ECCS acceptance criteria limit, Each of these is measureable but uill normally only be determined periodically as specified in Specifications 4.2.2 and 4.2.3. Thk perisdic surveillance is sufficient tu ensure that the linits are maintained provided

1. Control rods in a single group move together with no individual rod insertion differing by more than + 12 steps, indicated, frca the ,
group demand position.
6. Control rod banks are sequenced with overlapping groups at described in Specification 3.1.3.6;

c. The control rod insertion if ttits of Specification 3.1.3,5 anc' 3,1.3.6 are maintained; and :
O d. The axial power distribution, expressed in terms of AXIAL FLUX
\ DIFFERENCE, is m'aintained within the limits, f5g wil) be maintained within its limits provided conditions a. tnrough
d. ateve are maintained. TherelaxaticnofF2gasafunctlpnofTHE@ALF0WER allows changes in the radial shape for all permissible rod insertion limits.
'#,er. c,n Fa measurement is taken, dn allowance for both experimental error and marufectaring telGrance must be made. An allowance of 5*, is appropriate for a fW core mao 7.af m Ath the incore detector flux rapping system and a 3t 3ibaar.ce i3 a;propriate for canufacturing tolerance.
M:en F2p is ressured, (i .e. , inferred), measurement docertaintv li,e. , the acorocriste un';ertair,ty or, toe incere inferred hot rod peaking factor) rcust be Cicwed for ard G is the arpec?riate allowance for a full core map taken with the incore detectior systes
oci rod bairg reducet the value of C;lB ratic. Credit is available to o" set th!s reduc. tion in the generic margin. The generic design margins, t9taling 9.11 Om, conc *etely offat at:y red Scw penalties. This ^.argin inf.uds the folloaing:
D Dedor,1imit D3BR of L30 vs. L28,
2) Gridbacing (K.) of 0.0dE vs 0.059,

3) Ther. mal Di'fasidr' Ceefficient of 0.G3d vs. ').C59,

4) CHER yuitiplier of 0.B3 vs. 0.35,acd 7
( S' Pitch reduction Tr.e applica~ ole v61ues cf rod t,0w penaltias are referenced in the F5 Art. YO;rLE-DMIT I B 3/4 2-4
POWER DISTRIBUTION LIMITS v \ BASES HEAT, FLUX HOT CHANNEL FACTOR, and RCS FLOW RATE AND NUCLEAR ENTHALPY RI H0T CHANNEL FACTOR (Contipped) , Wheh an F measurement is taken. an allowance for both experimpntal error g and manufacturing tolerance must be made. An allowance of 5% is , appropriate for a full-core map taken with the Incore Detector Flux Mapping / $ystem, and a 2% allowance 'is appropriate for manufacturing tolerance. /
\ /
The Radial peaking Factor, Fxy(Z), is measured period,ilally to provide assurance that the' Hot Channel Factor, F (Z), remains within its limit. The n
) as provided the Radial Peaking F
y limitforRATED'TyRMALPOWER(F Factor t.imit Report per. Specification 6.9.1.6 was determined from expected peer control manuevers over the full range of burnup conditions in the core. N /
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2 WhenRCSflewrateandhareseasured,na'additionalallowancesare p necessary prior to comparisonwith s the liasits/of Figures 3.2-3 and 3.2-4. g
Measurement errors of [2.1]% fohRCS total j flow rate and 4% for Fg have been allowed for in determination of the desigs ONBR value.
The measurement error for RCS at flow rate is based upon performing a precision heat balance and using the ' sult to calibrate the ACS flow rate indicators. Potential fouling of efydwaterventuriwhichmightnotbe ,- detected could bias the result fr the pqecision heat balanca in 4 non-conservative manner. Therefore a penaltysof [0,1]% for undetected fouling of , the feedwater venturi is inclu d in Figure \3.2-3. Any fouling which might , bias the RCS flow rate ceasu ment greater than [0.1]% can be detected by monitoring and trending va us plant performahce parameters, If detected.
- action shall be taken bef e performing subsc-cutint precision heat balance measurenants, i.e., eithyr the effect of the foultng shall be quantified and ccrpensated for in the,RCS flow rate measurement orsthe venturf shall be cleahed to eliminate the fouling. / \
The 12-hour pfriodic surveillance of indicated RCS\(low is sufficient to detect only flowfdegracation which could lead to operation outside the accept-able region of cperation shown on Figure 3.2-3. 3/4.2,4 OUAQ2 NT p0WER TILT PATIO
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The AUADRANT POWER TILT RATIO limit assures that the radial power distribu-tion satisfies the design values used in the power capability analysis. RadiaVpower distribution measurements are mado during STARTUP testing and perio~dically during power operation, ' t
/ The limit of 1.02, at which corrective action is required, provides DHB and linear heat generation rate protection with x-y plane power tilts. A- / limit of 1.02 was selected to provide an allowance for the uncartainty J / associated with the indicated power tilt. ,/
VCGTLE - UNIT 1 5 5/4 2-5
W'I 90WER OISTRIBUTION LIMITS 8ASES The radial peaking factor, F# (Z), is measured periodically to provide . additional assurance that the hot channel factoryg(Z), remains within its limit. The Fxy limit for RATED THERMAL POWER (Fxy J as provided in the Radial Peaking Factor limit report per Specification 6.9.1.6 was determined from expected power control maneuvers over the full range of burnup condi-tions in the core. 3/4.2.4 QUADRANT POWER TILT RATIO The QUADRANT POWER TILT RATIO limit assures that the radial power dis-tribution satisfies the design values used in the power capability analysis. Radial power distribution measurements are made during startup testing and periodically during power operation. The limit of 1.02, at which corrective action is required, provides DNB and linear hoat generation rate protection with x-y plant power tilts. A limit of 1.02 was selected to provide an alicwance for the uncertainty associated with the indicated power tilt. . The two-hour time allowance foe operation with a tilt condition greater than 1.02 but less than 1.09 is provided to allow identification and cor-rection of a dropped or misaligned control rod. In the event such action () fm does not correct tilt, the margin for uncertainty on F0 is reinstated by reducing the maximum allowed power by 3 percent for each percent of tilt in excess of 1.0. For purposes of monitoring QUADRANT POWER TILT RATIO when one excore detector is inoperable, the movable incere detectors are used to confirrt that the normalized symmetric power distribution is consistent with the QUA0EANT POWER TILT RATIO. The incore detector monitoring is done with a fuii incore flux map or two sets of four symmetric thimbles. The two sets of four symmetric thimbles is a unique set of eight detector locations. These locations are H-3, H-13, C-8, N-8, E-5, E-11. L-5, and L-11. 3/4.2.5 DNB PARAMETERS The limits on tne DMB related parameters assure that each of the paraneters are maintained within the normal steady-state envelope of opera-tion assumed in the transient and accident analyses. The limits are consistent with the initia) FSAR assumptions and haye been analytically demonstrated adequate to Saintain a minimum DNBR of 1,30 throughout each analyzed tran-sient. The indicated Iy2 value of (Lator) and the indicated pressurizer pressure value of (LateF1 psig correspond to analytical limits of (Later) and (Later) psig, respectlyely, with allowance for meO5urement uncertainty. I . V0'1TLE-UNIT 1 8 J/4 2-5
, -1 .
POWER DISTRIBUTION LIMITS ' sj N N
= \ BASES _ _ . = -
QU.CRANT POWER TILT RATIO (Ccntinued)
\ . 'The 2-hcue time allowance 'for operaticn wit.h a tilt gondition gr9ater then 110.2 but less than 1.D9 is provided to allow identification and/ correction of a dropped or misaligned control rod. In the event such action adtien does not correc.t the tilt, the margin for uncertainty on Fg is rei.nsta49a by reducing the ca,vimum flowed tower by 3% for each percent of tilt in exC/ss of 1.
For purpdses of monitoring QUADRANT PCWER TILT RATIO n one excore wF,e/ , detector is inopgr$ble, the moveable incore detectors.are usea tp confirm that the Acrmalized symmetric power distribution is consistent'with the @ADRANT POWER TILT RATIO. 'The incere detector monitoring is gene 9th a full incore four synettric flux map or two sets'ef four syritetric thimbics. The t . thimoTes is a unique sut Of eight detector locatfoes.! These wo sets are locatfor.3 of C-8, E-5, E-11, H-3, H-13, L-5, L*L1, N-8. / 3/4.2.5 DNB PARAMET5!RS . TnelimitoontheDNB-rehtedparameters ssure that each of the parameters ' are maintained within the norma' steady-stat'e envelope of operation assumed in the transient and accident analy 3 Thefl imits are con:istent with the o initial FSAR assumptions and have een Kalytically demonstrated adequ4te to ' Q maintain a minimum DNBR of 1.30 thr c#out each analyzed transitot. The indicated T,yg value of [581]*F and ' e indicated pressurizer pressura valce of [2220] psig correspond to analytic 11 its of 595"F and 2205 psig respeca tively, with allowance for measur ment u ertainty. The 12-hour periodic su illance of se parameters through instrument readout is suVficient to ensu that the paraheters are rest 6ted within their .
- liciits following load changes and other expected transicot operation, -
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V0 GILE - UNIT 1 B 3/4 2-6 \ i
[} v PM R DI,STRIBUTION 1.IMIT5 BA$ES - - - - m._ ;- 3, , _.i k The 12 7.our periodic survaillence of these paraceter's throgh thstrument rendout is sufficie'.C to ertsun o thAt the parameters are rertored Within their limits following load cMngs ar.d Other exDected transient operatioP.. The 18 month periodic o.easuremmit of the RCS total fiow r4te is adequate to detect flou degrad3 tion and ensare correlation of the flow indicatic.) channels with treasured ficN such that the indicated percen- flow will prevfde sufficient verifi ation c of tr.e fbw rate on a 12 hour basji. A 1 O - 4 2 t I lO V0GTLE-UNIT 1 $ 3/4 2.E , a - ,_. -- . - - . .
,. - . . _ _ . _ _ _ _ . _ - .. . _ . . _ _ _ _ _ , _ . . _-~ - . -
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s e ensis +e n t wow,' miro Mi,.,;ber, P?aYrepeakte s + ie e, ! i /e a 1 e / re Nai /t b & th e fe ris*r e en 4// sci; a M Ey / n e e ru_. SA W /y h a A rc l 3/4.3 INSTRUMENTATION I BASES 3/4.3.1 and 3/4.3.2 REACTOR' TRIP SYSTEM and ENGINEERED SAFETY FEATURES M TTON SYSTEM INSTRUMENTATION TheCPERA8ILITYoftheReactorhSystemar.dtheEngineeredSafety Features Actuation System instrumentat1on and interlocks ensures that: (1) the
' associated ACTION and/or Reactor trip will be initiated when the parameter ,
{ nonitored by each channel or combination thereof reaches its Setooint (2) the a specified coincidence logic h :St:Scd, (3) sufficient redundancy is main-i tained to pemit a channel to be out-of-service for testing or maintenancey j and (4) sufficient system functional capability is available from dtverse
parameters.
The OPERABILITY of these systems is required to provide the overall f I reliability, recundancy, and diversity assumed available in the facility design fcr the prote.ction and mitigation of accident and transient conditions. The integrated operation of each of these systems is consistent with the assumptions used in the safety analyses. The Surveillance Requirements spect-j fied for these system 3 ensure that the overall system functiona'l capability is ; maintained comparable to thie original design standards. The periodic surve11-~ ' lance test, perfctmeo at the mi.nimum frequencies are sufficient to demonstrate this capability. % 3,sart .Jo, E 3M 3 /. 4 The E6pineered Safety Teatures Actuation System Instrumentation Trip ' l O Set. points specified i,n Tame 3.3-4 are the nominal values at which the bistablet are set for each func3fenal unit. A Setpoint is considered to be adjusted consistant with the nominal value when the "as measured 5etpoint is within i the band allowed for calioration accuracy. i To accemmodate the instn. ment crf ft assumed t.o occur between operational tests and tha. accuracy to which Setpoints can he me&sured and calItarated, Allo-able Valus for the Setpoints have bi:en .specified in Table 3,3-4. . Opera-tien with Setpoints less conservative than tne Trip Sets,oint but within tne l
Allwab1; Value is acceptable since an allowance has haen .1ade in the safety analysis to accommodate thts error. An optional provision has been included i
for datermining the 072RABILITY of a channel when its Trip 56tpoint is fodad to exceeG the AllowaH 2 VeltA .The methodology of 1. hip ost5.an utilizes the d as measured" c'edation from the 'specified calibration 74f nt for rack and i i sensor comoooents in conjuN: tion witb a stati.st? 11 comkioatish of the otner uncartaintiss of the instrumentation to eneasure the prccess vathible and the l urc6ctainties in calibrating the ir.struuntatica. In Equation 3.3-1, +
7. + R S < TA, t.he interactive effects of the errors in tha r.inck Ond tf.e i >
i sensor, and the "as measured" values of the errors are considsr.d. z, <ss
l specifit0 in Table 3.3-4, in percent span, 4 the statistical sutaation
( of errors assumed in the analysis excludiaq thosa associated with the sensor and ra<:k drift and the accutscy of their measurerent. TA br Total Allowance is the differenes, 'in percent spai A or Rack Errer is the "as measured i deviation, In the percent span, for the affected channel from the specified Trip 5etpoibit. E nr Sansor Error it either the *as measured" deviation of l i j V0GTi.E - UN!T J B 3/4 3-1 , l < u -_ j
C) Insert to Page B 3/4 3-1
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Specified surveillance intervals and surveillance and maintenance outage times have been determined in accordance with WCAP-10271, " Evaluation of Surveillance Frequencies and Out of Service Times for the Reactor Protection Instrumentation System", and supplements to that report. Surveillance intervals and out of service times were determined based upon maintaining an appropriate level of reliability of the Reactor Protection System and Engineered Safety Features instrumentation. t*\ b
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, INSTRUMENTATION BASES l REACTOR TRIP SYSTEM and ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION (Continued) the sensor from its calibration point or the value specified in Table 3.3-4, in percent span, from the analysis assumptions. Use of Equation 3.3-1 allows , for a sensor draft factor, an increased rack drift factor, and provides a threshold value for REPORTABLE EVENTS.
The methodology to derive the Trip Setpoints is based upon combining all of the uncertainties in the channels. Inherent to the determination of the Trip Setpoints are the magnitudes of these channel uncertainties. Sensor and rack instrumentation utilized in these channels are expected to be capable of operating within the allowances of these uncertainty magnitudes. Rack drift in excess of the Allowable Value exhibits the behavior that the rack has r.ot met its allowance. Being that there is a small statistical chance that this will happen, an infrequent excessive drift is expected. Rack or sensor drift, in excess of the allowance that is more than occasional, may be indicative of more serious problems and should warrant further investigation. The measurement of response time at the specified frequencies provides assurance that the Reactor trip and the Engineered Safety Features actuation associated with each channel is completed within the time limit ast,uaned in the safety analyses. No credit was taken in the analyses for those channels with d response times indicated as not applicable. Response time may be doct.strated by any series of sequential, overlapping, or total channel test measurements provided that such tests demonstrate the total channel response time as defined. Sensor response time verification may be demonstrated by either: (1) in place, onsite, or offsite test measurements, or (2) utilizing replacement sensors with certified response time. The Enginee:ed Safety Features Actuation System senses selected plant parameters and determines whether or not predetermined limits are being exceeded. If they are, the signals are combined into logic matrices sensitive to combina-tions indicative of various accidents events, and transients. Once the required logic combination is completed, the system sends actuation signals to those Engineered Safety Features components whose aggregate function best serves the requirements of the condition. As an example, the following actions may be initiated by the Engineered Safety Features Actuation System to mitigate the consequences of a steam line break or loss-of-coolant accident: (1) Safety Ecc 5 In3eetson pumps start and automatic valves position, (2) Reactor trip, (3) feed water isolation, (4) startup cf the emergency diesel generators, (5) containment spray pumps start and automatic valves position (6) containment isolation, (7) steam line isolation, (8) turbine trip, (9) auxiliary feedwater pumps start and automatic valves position, (10) containment :::lin;; fr: start and A cc e rs automatic valves position, (11) essential : mice water pumps start and auto- ; matic valves position, and (12)'Gentrol Roo= -Ise4ati:n =d ":ntihtien Systems start. l % no / 4 %NMen En qny k&&en l 1 ww aya-,,..,_ s.,, l O ! V0GTLE - UNIT 1 B 3/4 3-2 I l
INSTRUMENTATION BASES REACTOR TRIP SYSTEM and ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION (Continued) The Engineered Safety Features Actuation System interlocks pe'rform the following functions: P-4 Reactor tripped - Actuates Turbine trip, closes main feedwater valves on T,yg below Setpoint, prevents the opening of the main feedwater valves which were closed by a Safety Injection or high Steam Generator Water Level signal, allows Safety Injection block so that components can be reset or tripped. Reactor not tripped prevents manual block of Safety Injection. and /owsfeam /ine. pre.ssare. P-11 On increasing pressurizer pressure, P-11 automatically einstates Safety Injection actuation on low pressurizer pressur On decreasing pressure, P-11 allows the manual block of Safety Injection actuation on low pressurizer pressurefand low slessrr /ine pressure cada//ow.s
.s/wr /ine isola.fien on nega.+sk .s fenar /ine. press ve. rate to Jecome. -P-12 Cn incr:ning rc :t:r 00:10.. 10:p t::p:r:ture,+ .P. .12 : t;:: tic:lly acjf,"P' ..t....+.. c.<.>u ua. +u- . +...+u- .. u<-u < u~ . . : .. u . . .
R....':.i.'c...
.. . ..- : u.... i. - ' . i.X ... :. ;.7. . r,. C ..r..:, .:D.. ,. .: T. ...,".:..mnad eming signe! to the r 0" r Syst- On d::r:::ing r:::ter l
c el:nt 100p temperature, P-12 :llew: th: ::nt:1 bleek of S:f:ty 5/e w hae. essure Injection T cr 1:w::tuation en high :te:: steam-1-ine-presrsur: andfl oute:: w ::incid:nt tic:llywith rc.;= ;ith;r
th: 10: le y$2 .~

ing avg r4;n:1 #rce the Steam O' ? Syste . -
i l P-14 On' increasing steam generator water level, P-14 automatically trips all feedwater isolation valvesaand inhibits feedwater control valve i modu 1ation. , int +iades a thrbine frip, . 3/4.3.3 MONITORING INSTRUMENTATION l ! 3/4.3.3.1 RADIATION MONITORING FOR PLANT OPERATIONS l The OPERABILITY of the radiation monitoring instrumentation for plant I operations ensures that: (1) the associated action will be initiated when the radiation level monitored by each channel or combination thereof reaches its Setpoint, (2) the specified coincidence logic is maintained, and (3) suffi-cient redundancy is maintained to permit a channel to be out-of-service for testing or maintenance. The radiation monitors for plant operations senses radiation levels in selected plant systems and Tocations and determines whether or not predetermined limits are being exceeded. If they are, the signals are combined into logic matrices sensitive to combinations indicative of various accidents and abnormal conditions. Once the required logic combination is
~
completed, the system sends actuation signals to initiate al' arms or automatic isolation action and actuation of Emergency Exhaust or Ventilation Systems. V0GTLE - UNIT 1 B 3/4 3-3
l
INSTRUMENTATION BASES 3/4.3.3.2 MOVABLE INCORE DETECTORS 1 The OPERABILITY of the movable incore detectors with the specified minimum complement of equipment ensures that the measurements obtained fro'm use of this system accurately represent the spatial neutron flux distribution of the core. The OPERABILITY of this system is demonstrated by irradiating each detector used and determining the acceptability of its voltage curve.
ForthepurposeofmeasuringF(Z)orFhafullincorefluxmapisused. q Quarter-core flux maps, as defined in WCAP-8648, June 1976, may be used in recalibration .of the Excore Neutron Flux Detection Syt, tem, and full incore flux maps or symmetric incore thimbles may be 'used for monitoring the QUADRANT POWER TILT RATIO when one Power Range channel is inoperable. 3/4.3.3.3 SEISMIC INSTRUMENTATION The OPERABILITY of the seismic instrumentation ensures that sufficient-capability is available to promptly determine the magnitude of a seismic event and evaluate the response of those features important to safety. This capa-bility is required to permit comparison of the measured response to that used in the design basis for the facility to determine if plant shutdown is required pursuant to Appendix A of 10 CFR Part 100. The instrumentation is consistent
~
with the recommendations of Regulatory Guide 1.12, " Instrumentation for Earth-quakes," April 1974. 3/4.3.3.4 M'ETEOROLOGICAL INSTRUMENTATION The OPERABILITY of the meteorological instrumentation ensures that sufficient meteorological data are available for estimating potential radiation doses to the public as a result of routine or accidental release of radioactive materials to the atmosphere. This capability is required to evaluate the need i for initiating protective measures to protect the health and safety of the public and is consistent with the recommendations of Regulatory Guide 1.23, "Onsite Meteorological Programs," February 1972. 3/4.3.3.5 REMOTE SHUTDOWN SYSTEM The OPERABILITY of the Remote Shutdown System ensures that sufficient capability is available to permit safe shetdown of the facility from locations outside of the control room. This capability is required in the event control room habitability is lost and is consistent with General Design Criterion 19 of 10 CFR Part 50. The OPERABILITY of the Remote Shutdown System ensures that a fire will not preclude achieving safe shutdown. The Remote Shutdown System instrumentation, O V0GTLE - UNIT 1 B 3/4 3-4
l I
N (d INSTRUMENTATION BASES 1 REMOTE SHUTDOWN SYSTEM (Continued) g,3g, ; + '
control, and peu:r circuit:-and transfer switches ne'cessary to eliminate effects of the fire'and allow operation of instrumentation,! control :nd p:'.::r circuits required to achieve and maintain a safe shutdown condition are independent of areas where a fire could damage systems normally used to shut down the reactor. This capability is consistent with General Design Criterion 3 and App:ndix t: 10 Orri Part 50. C,W E B 9. S. /. 3/4.3.3.6 ACCIDENT MONITORING INSTRUMENTATION The OPERABILITY of the accident monitoring instrumentation ensures that sufficient information is available en selected plant parameters to monitor and assess these variables followirg an accident. This capability is consis-tent with the recommendations of Regulatory Guide 1.97, Revitica 3, "Instrumen-tation for Light-Water-Cooled Nuclear Power Plants to Assess Plant Conditions During and Following an Accident," " y 10^' and NUREG-0737, " Clarification of TMI Action Plan Requirements," November I b
/Ee n.twn 2, Deceder spe 3/4.3.3.7 CHLORINE DETECTION SYSTEMS The OPERABILITY of the Chlorine Detection Systems ensures that sufficient
() capability is available to promptly detect and initiate protective action in the event of an accidental chlorine release. This capability is required to protect control room personnel and is consistent with the recommendations of Regulatory Guide 1.95, Revisier 1, " Protection of Nuclear Power Plant Control room Operators Against an Accidental Chlorine Release," J nuary 1077. 3/4.3.3.8 FIRE-DETEETION-INSTRC"E"TATIO"
%e-OPER'dB-ILID' of-the4 ire-detection in:tru= ntation ensure; that both adequate waraing-capatality-is-availat>1e-for prompt detecti:n f fires and thet Eire-Suppression--Systemsy-that-are-ac-tuated-by-f-ire detector:, will discharg:
extinguishing -age n t s-i n -a-ti mely- manne r..-Prompt-detecti on-and - s up p re s s i o n - of
<4.. ,, . . . a. 3 , .m.a..... .. .+w.
mn.+.-+.4..,,.
. . , . <m.. A , . , n, o. + n. e s < n. +.,u. .
1. ., +. a. A a
n, . 4.c---. n m o n +. . n. A.
.. 4. e an integra' e' : nt 4a the evera faci'ity C4 e o retectier "regr::.
Fire-detector: that are u :d t: ::tuate ri re Supprea :ier Syste represent a-more-critically-important-component f pl:nt : Fir retection Pr:gra than detectors that :re in :tal'ed :olely for ::rly fir; warning and n:tifica-tier Centequently, the -inimum aumber of OPERABLE **re detector: meet be geester. The '::: cf detection capabi'ity fer tire Suppre::ien Sy:te::, ::tunted by # ire detector:, represent: 2 :ignificant degr:d:tien of #ir; protecti:n fer r l V0GTLE - UNIT 1 B 3/4 3-5
-d9 5
4-INSTRUMENTATION BASES , FIRE DETECTION INSTRUMENTATION (Continued)
.m .... 4, ......i+ + w. . , + . w n . w_. + m , . <<.. m.+,w .+. i ..... u. :_ .
i ti$tEh'5':A'5rIEE'5tehA'tb5a -^U5Ubh'e5Er5nteh'fSE~theIEss':[5EtEEtr5 that previt only ::rly fire warning. The ::t:bli:h: nt ;f fre gent fire patrols in the affected-areas _is-req"f red to provide detectien - rei'ity unti! the inoper:ble in:tr;;;nt:ti:n is r;;tered te OICTA"I;.!!". 8 - 3/4.3.3.+ LOOSE PART DETECTION SYSTEM The OPERABILITY of the Loose-Part Detection System ensures that sufficient capability is available to detect loose metallic parts in the Reactor System and avoid or mitigate damage to Reactor System components. The allowable out-of-service times and surveillance requirements are consistent with the recommendations of Regulatory Guide 1.133, " Loose-Part Detection Program for . the Primary System of Light-Water-Cooled Reactors," May 1981. i 9 l 3/4. 3. 3.M RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION j The radioactive liquid effluent instrumentation is provided to monitor
,and control, as applicable, the releases of radioactive materials in liquid effluents.during actual or potential releases of liquid effluents. The j
' O Alarm / Trip Setpoints for these instruments shall be calculated and adjusted in accordance with the methodology and parameters-in the ODCM to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20. The OPERABILITY'and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50.
to 3/4.3.3.+1 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION The radioactive gaseous effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous efflu-ents during actual or potential releases of gaseous effluents. The Alarm / Trip Setpoints for these instruments shall be calculated and adjusted in accordance j
with the methodology and parameters in the ODCM to ensure that the alarm / trip i will occur prior to exceeding the limits of 10 CFR Part 20. This instrumenta-tion also includes provisions for monitoring (and controlling) the concentrations of potentially explosive gas mixtures in the WASTC CA: ll^' 0"I SYSTEM. The OPERA-1 BILITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50. The sensitivity of any noble gas activity monitors used to show\ compliance with the gaseous effluent release requirements of Specification 3.11.2.2 shall be such that concentrations as low as 1 x 10 8 pCi/mlaremeasurable.{
. ., hurous t.4Sff MucswW6r 4/ 4. 3. 3. // - Se e fue ef lcr a 3/43 -4 O
V0GTLE - UNIT 1 8 3/4 3-6
l G J INSERT FOR B 3/4 3-6 3/4.3.3.12 HIGH ENERGY LINE BREAK ISOLATION SENS0RS The operability of the high energy line break isolation sensors ensures that the capability is available to promptly detect and initiate protective action in the event of a line break. This capability is required ta pre-vent damage to safety-related systems and structures in the auxiliary building. l O E f l l O l
INSTRUMENTATION BASES 3/4.3.4 TURBINE OVERSPEED PROTECTION This specification is provided to ensure that the turbine overspeed protection instrumentation and the turbine speed control valves arh OPERABLE and will protect the turbine from excessive overspeed. Protection from turbine excessive overspeed is required since excessive overspeed of the turbine could generate potentially damaging missiles which could impact and damage safety-related components, equipment or structures. i ) 8 4 J O P O V0GTLE - UNIT 1 B 3/4 3-7 - l
1
'O 3/4.4 REACTOR COOLANT SYSTEM V
BASES 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION The plant is designed to operate with all reactor coolant loops in operation and maintain DNBR above 1.30 during all normal operations and anticipated transients. In MODES 1 and 2 with one reactor coolant loop not in operation this specification requires that the plant be in at least HOT STANDBY within 6 hours. In MODE 3, two reactor coolant loops provide sufficient heat removal capability fcr removing core decay heat even in the event of a bank withdrawal accident; however, a single reactor coolant loop provides sufficient heat removal capacity if a bank withdrawal accident can be prevented, i.e., by opening the Reactor Trip System breakers. Single f ailure cencideratient r-equire-that-two-Toop be OPERABLE st 211 time:. fr.un In MODE 4, and in MODE 5/with reactor coolant loops filled, a single reactor coolant loop or RHR Wep provides sufficient heat removal capability for removing decay heat; but single failure considerations require that at (either RHR or RCS) be OPERABLE. Hai" least two p/coi In MODE 5 w th /reactor train.s coolant loops not filled, a single RHR 'r' esp provides sufficient heat removal capability for removing decay heat; but single failure considerations, and the unavailability of the steam generators as a heat (Q
/ removing component, require that at least two RHR 1 peps be OPERABLE.4 + rain r .;Gwert 4 8 3/ 4 &/ here.
The operation of one reactor coolant pump (RCP) or one RHR pump provides adequate flow to ensure mixing, prevent stratification and produce gradual reactivity changes during boron concentration reductions in the Reactor Coolant System. The reactivity change rate associated with boron reduction will, therefore, be.within the cacability of operator recognition and control. 3SC The restrictions /on starting an RCP with one or more RCS cold legs less than or equal to [t?S3*F are provided to prevent RCS pressure transients, caused by energy additions from the Secondary Coolant System, which could exceed the limits of Appendix G to 10 CFR Part 50. T! 9 RCS will be protected against overpressure transients and will not exceed the41imit: Of Appendix G by either-(1) r;;tric-t4ng-the-water-volume-in-the-presstn4zer and th;r:by pr:viding : v !ume fer the reactor-coolant-to-expand 4nte r -or--{2-) b.v restrictin? startinc of the RCPs to when the secondary water temperature of each steam generator is less : than 5a *F above each of the RCS cold leg temperatures. .
/
l rnnhnm,, ne mina/ to & Arc fiv't & // pot. der w ' ' ~m *.*n& G rene fer ye a rs Uf p y! Ay,f.T to5se) N O 7 /u'mo I Tho com4 amant +n m s 4 n t -m i n +ha hnenn ranscat.;tiga gf ga {gg!3ted Trep 9 seater thed er equal te the b' wen cencentratien Of th: :p;r: ting le:;;':n:;r:: that-no reacti"ity additi:n--t:- th: 04ee-could :: ;r during :t:rtup ;f :n
'::': tad 1:0p. Ver 'itation Of the b rcr ::n::ntr: tion ' :n idi; 1;;; pri;r 4
t+ opening th; :t p valve; pr vid;; ; r;;;;ur;n;; ;f th; ;dca ::y ;f th: M. ;n j
n::ntrati r '- the f Ol ted !:cp. Cper: ting the i:012t:d 10:p er reci cul ti g (q
'r at 1:::t 00 ' ut:: pri;r t: :;;r'n; it: :t:p :1;:: :n::r:: :d: ; :t f " " ' ^'; c ' t": coelant 4- thic leep and prevent: 2ny *eacti"ity ef %ct: 90 te uro.wcance ntration-Str a ti ' f ::t i :n: .
V0GTLE - UNIT 1 B 3/4 4-1
'6
( Insert 1 to B 3/4 4-1 The locking closed of the required valves in Mode 5 (with the loops not filled) precludes the possibility of uncontrolled boron dilution of the filled portion of the Reactor Coolant System. This action prevents flow to the RCS of unborated water by closing flowpaths from sources of unbora-ted water. These limitations are consistent with the initial conditions assumed for the boron dilution accident in the safety analysis. V(D l i f l l O
l REACTOR COOLANT SYSTEM (v~') BASES DE^C700 000L*"7 LOODS ^"O CCOL?"7 CIDCULATIC" (C:ntinued) [C? TIC"iL] St:-tup cf 2n idle-Joop-wi i nject-ce:1 water fr:m the 1 :p ic.t; tha
*. The reactivity-trans4:nt rc ulting ' rem thi: :::1 ;;ter 'r.j::ti r i; ~4"4=ized by de' ying i: !2ted ? :p t:rtup unti' it: temper:ture f: ith'-
20 F cf-the-operating 1;;;;. " h'n; th: rc::ter :ub:riti::! pri:r :: 12:p ct: +up p cut-t: 2ny per :pik: "ich :uid rc:uit 'r:m thi; ;;! w Re @ cc:1 re::t i it; tr:nci -t. 3/4.4.2 SAFETY VALVES The pressurizer Code safety valves operate to prevent the RCS from being pressurized above its Safety Limit of 2735 psig. Each safety valve is designed to relieves 2c.ccolbs per hour of saturated steam.:t the v:!ve 50tprint. The relief capacity of a single safety valve is adequate to relieve any overpressure condition which could occur during shutdown. In the event that no safety valves are OPERABLE, an operating RHR , connected to the RCS, provides-g-'s overpressure relief capability and will prevent RCS overpressurization. In (_,) addition, tyhe Overpressure Protection , ystem provides a diverse means of protection [against RCS overpressurization at low temperatures. GIL Vran During operation, all pressurizer Code safety valves must be OPERABLE to prevent the RCS from being pressurized above its Safety Limit of 2735 psig. The combined relief capacity of all of these valves is greater than the maximum surge rate resulting from a complete loss-of-load assuming no Reactor trip until the first Reactor Trip System Trip Sotpoint is reacned (i.e. , no credit is taken for a direct Reactor trip on the loss-of-load) and also assuming no operation of the power-operated relief valves or steam dump valves. Demonstration of the safety valves' lift settings will occur only during shutdown and will be performed in accordance with the provisions of Section XI of the ASME Boiler and Pressure Code. 3/4.4.3 PRESSUR1ZER The limit on the maximum water volume in the pressurizer assures that the parameter is maintained within the normal steady-state envelope of operation assumed in the SAR. The limit is consistent with the initial SAR assumptions. The 12-hour periodic surveillance is sufficient to ensure that the parameter is restored to within its limit following expected transient operation. The maximum water volume also ensures that a steam bubble is formed and thus the RCS is not a hydraulically solid system. The requirement that a minimum number of pressurizer heaters be OPERABLE enhances the capability of the plant to control Reactor Coolant System pressure and establish natural circulation. O V V0GTLE - UNIT 1 B 3/4 4-2 l I
i 4 REACTOR COOLANT SYSTEM BASES 3/4.4.4 RELIEF VALVES The power-operated relief valves (PORVs) and steam bubble function to relieve RCS pressure during all design transients up to and including the design step load decrease with steam dump. Ooeration of the PORVs minimizes the undesirable opening of the spring-loaded pressurizer Code safety valves. Each PORV has a remotely operated block valve to provide a positive shutoff capability should a relief valve become inoperable. 3/4.4.5 STEAM GENERATORS The Surveillance Requirements for inspection of the steam generator tubes ensure that the structural integrity of this portion of the RCS will be main-tained. The program for inservice inspection of steam generator tubes is based on a modification of Regulatory Guide 1.83, Rcvision 1. Inservice inspection of steam generator tubing is essential in order to maintain surveil-lance of the conditions of the tubes in the event that there is evidence of mechanical damage or progressive degradation due to design, manufacturing errors, or inservice conditions that lead to corrosion. Inservice inspection of steam generator tubing also provides a means of characterizing the nature j and cause of any tube degradation so that corrective measures can be taken. v The plant is expected to be operated in a manner such that the secondary coolant will be maintained within those chemistry limits found to result in negligible corrosion of the steam generator tubes. If the secondary coolant chemistry is not maintained within these limits, localized corrosion may Me4y result in stress corrosion cracking. The extent of cracking during plant operation would be limited by the limitation of steam generator tube leakage between the Reactor Coolant System and the Secondary Coolant System g ,,,m., TTeactor-to-secondary leakage = 500 gallons per day per steam generator). Cracks having a reactor-to-secondary leakage less than this limit during operation will have an adequate margin of safety to withstand the loads imposed during normal operation and by postulated accidents. Operating plants have demonstrated that reactor-to-secondary leakage of 500 gallons per day per steam generator can readily be detected by radiation monitors of steam generator blowdown. Leakage in excess of this limit will require plant shutdown and an unscheduled inspection, during which the leaking tubes will be located and plugged. Wastage-type defects are unlikely with proper chemistry treatment of the secondary coolant. However, even if a defect should develop in service, it yill be found during scheduled inservice steam generator tube examinations. m ,,,, .- , Plugging will be required for all tubes with imperfections exceeding the plugging limit of f40]% of the tube nominal wall thickness. Steam generator tube inspections of operating plants have demonstrated the capability to reliably detect degradation that has penetrated 20% of the original tube wall ! thickness. ( l V0GTLE - UNIT 1 B 3/4 4-3
I t i REACTOR COOLANT SYSTEM l . BASES i STEAM GENERATORS (Continued) g go, y Whenever the results of any steam generator tubing inservice" inspection fall into Category C-3, these r9sults will be promptly reported to the Commission
': Spe:f:1 " g;rt pursuant to Op u ffi s t ha C.0.2 within 20 dq: and prior to 4
resumption of plant operation. Such cases will be considered by the Commission on a case-by-case basis and may result in a requirement for analysis, laboratory examinations, tests, additional eddy-current inspection, and revision of the
' Technical Specifications, if necessary.
3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE a'
; 3/4.4.6.1 LEAKAGE DETECTION SYSTEMS 1
The RCS Leakage Detection Systems required by this specification are provided to monitor and detect leakage from the reactor coolant pressure boundary. These Detection Systems are consistent with the recommendations of f Regulatory Guide 1.45, " Reactor Coolant Pressure Boundary Leakage Detection i Systems," May 1973. 1 I 3/4.4.6.2 OPERATIONAL LEAKAGE i PRESSURE BOUNDARY LEAKAGE of any magnitude is unacceptable since it may be indicative of an impending gross failure of the pressure boundary. Therefore, i the presence of any PRESSURE BOUNDARY LEAKAGE requires the unit to be promptly i placed in COLD SHUTDOWN. i i Industry experience has shown that while a limited amount of leakage is expected from"the RCS, the unidentified portion of this leakage can be reduced to a threshold value of less than 1 gpm. This threshold value is sufficiently low to ensure early detection of additional leakage. ! The total steam generator tube leakage limit of 1 gpm for all steam l generators net i: hted f a; the--RGS ensures that the dosage contribution 4 from the tube leakage will be limited to a small fraction of 10 CFR Part 100 dose guideline values in the event of either a steam generator tube rupture or steam line break. The 1 gpm limit is consistent with the assumptions used
in the analysis of these accidents. The 500 gpd leakage limit per steam j generator ensures that steam generator tube integrity is maintained in the event of a main steam line rupture or under LOCA conditions, f The 10 gpm IDENTIFIED LEAKAGE limitation provides allowance for a limited l
amount of leakage from known sources whose presence will not interfere with ! the detection of UNIDENTIFIED LEAKAGE by the Leakage Detection Systems. I The-CONTROLLEO LEMf.CE li-itat4en-restr4ct: g:r tha h:n th: t:t:1 '?: I supp!!:d to-t,%-reactor-ccchnt pu p :::1-s-exc::d: 10 g ; =fth-th: ::d:hti ; !O l V0GTLE - UNIT 1 8 3/4 4-4 1
=
P (O
\ /
REACTOR COOLANT SYSTEM BASES OPERATIONAL LEAKAGE (Continued) va!ve % the supply Line-fully-open-at_.a--nominal-RCS-pre::ure Of 2235 ;;!;. 4- th: Overt Of : LOCA, th: ::f:ty 'nf :ti: .
"i: l'-itat4on :n ure: that "# "et be les: thin 2 u-^d 4a the safety an !y:::.
The leakage from any RCS pressure isolation valve is sufficiently low to ensure early detection of possible in-series valve failure. !t u .ppe.. a.t that when pre:: vee-4so4ation-44-peswided-by-twe 'rc::rie: v:iv:: :nd aber fai?cre Of en: valv: #r the p fr c:n g: undetected for d:t:nti ? ?: ;ih f ti e, ver"!:stier of valve St^gr!ty i: re ; !r:d. Since these valves are important in preventing overpressurization and rupture of the ECCS low pressure piping which could result in a LOCA, these valves should be tested periodically to ensure low probability of gross failure. The Surveillance Requirements for RCS pressure isolation valves provide added assurance of valve integrity thereby reducing the probability of gross valve failure and consequent intersystem LOCA. Leakage from the RCS pressure isolation valve is IDENTIFIED LEAKAGE and will be considered as a portion of , the allowed limit. 3/4.4.7 CHEMISTRY The limitations on Reactor Coolant System chemistry ensure that corrosion of the Reactor Coolant System is minimized and reduces the potential for Reactor Coolant System leakage or failure due to stress corrosion. Maintaining the chemistry within the Steady-State Limits provides adequate corrosion protection to ensure the structural integrity of the Reactor Coolant System over the life of the plant. The associated effects of exceeding the oxygen, chloride, and fluoride limits are time and temperature dependent. Corrosion studies show that operation may be continued with contaminant concentration levels in excess of the Steady-State Limits, up to the Transient Limits, for the specified limited time intervals without having a significant effect on the structural integrity of the Reactor Coolant System. The' time interval permitting continued operation within the restrictions of the Transient Limits provides time for taking corrective actions to restore the contaminant concen-trations to within the Steady-State Limits. l The Surveillance Requirements provide adequate assurance that concentrations in excess of the limits will be detected in sufficient time to take corrective , action. 3/4.4.8 SPECIFIC ACTIVITY The limitations on the specific activity of the reactor coolant ensure that the resulting 2-hour doses at the SITE BOUNDARY will not exceed an V0GTLE - UNIT 1 8 3/4 4-5
e . - . .- REACTOR COOLANT SYSTEM BASES SPECIFIC ACTIVITY (Continued) appropriately small fraction of 10 CFR Part 100 dose guideline values following g ,, _< _ a steam generator tube rupture state ^~+--to-secondary steamaccidentgenerator in conjunction with leakage rate ofan assumed 1 gpm. Thesteady-values for the limits on specific activity represent Ilmits based upon a parametric evaluation by the NRC of typical site locations. These values are conservative
! in that specific site parameters of the v% */e. site, such as SITE BOUNDARY
! location and meteorological conditions, were not considered in this evaluation. The ACTION statement permitting POWER OPERATION to continue for limited time periods with the reactor coolant's specific activity greater than I microcurfe/ gram DOSE EQUIVALENT I-131, but within the allowable limit shown on Figure 3.4-1, accommodates possible iodine spiking phenomenon which may occur following changes in THERMAL POWER. g ( The sample analysis for determining the ross specific activity and E can exclude the radioiodines because of the lo eactor coolant limit of 1 microcurie / I gram DOSE EQUIVALENT I-131, and because, i the limit is exceeded, the
radioiodine level is to be determined ever'y 4 hours. If the gross specific activity level and radiofodine level in the reactor coolant were at their limits,theradioiodinecontributionwod1dbeapproximately1%. In a release I
of reactor coolant with a typical mixtdre of radioactivity, the actual radio-iodine contribution would probably b 9/ about 20%. The exclusion of radio-nuclides with half-lives less than M minutes from these determinations has been made for several reasons. The first consideration is the difficulty to identify short-lived radionuclides in a sample that requires a significant time to collect, transport, and analyze. The second consideration is the j predictable delay time between the postulated release of radioactivity from
the reactor coolant to its release to the environment and transport to the i SITE BOUNDARY, which is relatable to at least 30 minutes decay time. The
'y , choice of -10 minutes for the half-life cutoff was made because of the nuclear characteristics of the typical reactor coolant radioactivity. The radionuclides in the typical reactor coolant have half-lives of less than 4 minutes or
! half-lives of greater than 14 minutes M h hh c1 hw: c d htinctha h tw: n the ! -idianuc!ide 2beve and below-a ha!f '4fe cf 10 inuter. For these reasons the radionuclides that are excluded from consideration are expected to decay l to very low levels before they could be transported from the reactor coolant to the SITE BOUNDARY under any accident condition. 1 I !O i l 3 i V0GTLE - UNIT 1 B 3/4 4-6 I -_ . _ - - _ - _ . - - . _ - - _ _ _ _ _ _ _
REACTOR COOLANT SYSTEM BASES SPECIFIC ACTIVITY (Continued)
, ~ _ - .. .-_ /~ # d ' - - -
( Based upon the above considerations for excluding certain radionuclides from the sample analysis, the allowable time of 2 hours between sample taking and completing the initial analysis is based upon a typical time necessary to ,
+
perform the sampling, transport the sample, and perform the analysis of about 90 minutes. After 90 minutes, the gross count should be made in a reproducible geometry of sample and counter having reproducible beta or gamma self-shielding properties. The counter should be reset to a reproducible efficiency versus energy. It is not necessary to identify specific nuclides. The radiochemical
- determination of nuclides should be based on multiple counting of the sample within typical counting basis following sampling of less than 1 hour, about
2 hours, about 1 day, about I week, and about 1 month.
! wx ,% < Reducing T to less than 500*F prevents the release of activity should asteamgenerat@9 tube rupture since the saturation pressure of the reactor
coolant is below the lift pressure of the atmospheric steam relief valves.
4 The Surveillance Requirements provide adequate assurance that excessive specific activity levels in theJeactor coolant will be detected in sufficient time to take corrective actiondA reduction in frequency of isotopic analyses following power changes may be permissible if justified by the data obtained. 3/4.4.9 PRESSURE / TEMPERATURE LIMITS The temperature and pressure changes during heatup and cooldown are limited to be consistent with the requirements given in the ASME Boiler and Pressure Vessel Code, Section III, Appendix G: 3
1. The react.or coolant temperature and pressure and system heatup and cooldown
~
rates (with the exception of the pressurizer) shall be limited in accordance l with Figures 3.4-2 and 3.4-3 for the service period specified thereon: i
a. Allowable combinations of pressure and temperature for specific temperature change rates are below and to the right of the limit lines shown. Limit lines for cooldown rates between those presented may be obtained by interpolation; and

b. Figures 3.4-2 and 3.4-3 define limits to assure prevention of non-ductile failure only. For normal operation, other inherent plant characteristics, e.g., pump heat addition and pressurizer heater capacity, may limit the heatup and cooldown rates that can be achieved over certain pressure-temperature ranges.
I i i O V0GTLE - dNIT 1 8 3/4 4-7
p v REACTOR COOLANT SYSTEM BASES PRESSURE / TEMPERATURE LIMITS (Continued)
2. These limit lines shall be calculated periodically using methods provided below,

3. The secondary side of the steam generator must not be pressurized above 200 psig if the temperature of the steam generator is below 70*F,

4. The pressurizer heatup and cooldown rates shall not exceed 100*F/hrand 200 F/hg respectively. The3 spray shall not be used if the temperature difference between the pressurizer and the,gspray fluid is greater than f(625] F, and unhy

5. System preservice hydrotests and inservice leak and hydrotests shall be performed at pressures in accordance with the requirements of ASME Boiler and Pressure Vessel Code, Section XI.
Peter to M /es/
+ The fracture toughness propertie.s of the ferritic materials in the reactor vessel are determined in accordance'with the NRC Standard Review Plan, ASTM E185-73, and in accordance with additional reactor vessel requirements. These properties are then evaluated in accordance with Appendix G of the Itpi <umme_r N o Addenda to Section III of the ASME Boiler and Pressure Vessel Code, r d the ,b Wbet -, ~92 ,,, !, N, ' - , . . % mort Ar 83/44-7 Heatup and cooldown limit curves are calculated using the most limiting value of the nil-ductility reference temperature, RTNDT, at the end of
[12] effective full power years (EFPY) of service life. The h EFPY service life period is chosen such that the limiting RTNDT at the 1/4T location in the core region is greater than the RTNDT f the limiting unirradiated material. The selection of such a limiting RT NDT assures that all components in the Reactor Coolant System will be operated conservatively in accordance witn applicable Code requirements. The reactor vessel materials have been tested to determineReactor their initial opera-RT NDT; the results of these tests are shown in Table B 3/4.4-1. tion and resultant fast neutron (E greater than 1 MeV) irradiation can cause an increase in the RTNDT. Therefore, an adjusted reference temperature, based upon the fluence, copper content, and phosphorus content of the material in question, can be predicted using Figure B 3/4.4-1 and the largest value of ART computed by either, Regulatory Guide 1.99, Revision 1, " Effects of NDT m ,,x ure m er Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials," or the Westinghouse Copper Trend Curves shown in Figure B 3/4.4-2. The heatup and cooldown limit curves of Figures 3.4-2 and 3.4-3 include predicted adjust-O ments for this shift in RT NDT at the end of g EFPY as well as adjustments U for possible errors in the pressure and temperature sensing instruments. I V0GTLE - yNIT 1 B 3/4 4-8 l , l l
NN Insert for B 3/4 4-8 O The heatup and cooldown limit curves in Figures 3.4-2 and 3.4-3 are applicable to'Vogtle-Unit 1 for up to 16 EFPY. The most limiting material for Vogtle-Unit I has an initial RTtlDT f 30 F0 and a copper content of 0.06 WT %, whereas the hehtup and cooldown curves in Figures 3.4-2 and 3.4-3 are based on ar: initial RT of 40 0F and a copper content of 0.10 WT 5. As a result, applicahke to Vogtle-Unit 1. I 1 i
O 1
2 3 i O
O O O
~
i- s , x i
'N TABLE B 3/4.4-1 i
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8
--4 'N s REACTOR VESSEL TOUGHNESS t V \ 'sN ASME 50 FT-LB/35 MIN. UPPER LF ,
l 7 COMP MATERIAL CU P NOTT HIL TEMP *f RT NOT FT-L ! t E COMPONENT CODE TYPE %_ %_ *F LONG 1RANS *F LONG TRANS
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4 4
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i V0GTLE-UNIT 1 8 3/4 4 11
/N j)
REACTOR --COOLANT SYSTEM -- 8ASES PRESSURE / TEMPERATURE LIMITS (Continued) determined in this mannor may be used until the results Values of aRTNDT from tne material surveillance program, evaluated according to ASTM E185, are avaib h Capsules will be removed in accor hetwsurveillance A the_r_e._uiremen spec men with-
#;-7stM E185-M and 10 CFR Part 50, Appendix t drawal scheaule it sh g ir Te.ble 4.4-5 4 lead factor represents the rela- ,Ti5hshfie'tweeh'ThefastndtIO ensity at the location of the capsule ) and the inner wall of the reketor vessel. Therefore, the results obtained from the surveillance specifrens can be used to predict future radiation damage to the reactor vessel material by,_psihthe lead factor and the withdrawal -
i time of the capsuleMolehtup and cooldown curves mustMuMhen treDTgT deledned from the surveillance capsule exceeds the calculated ART f r the equivalent capsule radiation exposure. NDT Allowable pressure-temperature relationships for various heatup and cooldown rates are calculated using methods derived from Appendix G in Sec-tion III of the ASME Boiler and Pressure Vessel Code as recuired by Appendix G to 10 CFR Part 50, and these nethods are discussed in detail in WC#-7924-A.
> ;, a so now,9 pmyap hs.
[d A The general method for calculating heltup and cooldown limit curves is based upon the principles of the linear elastic fracture methanics (LEFM) technology. In the calculation procedures a semielliptical . surface defect with a depth of one quarter of the wall t.hickness, T, and a length of 3/2T is assumed to exist at the inside of the vessel vall as well as at the outside of the vessel wall. The dimensions of this postulated crack, referred to in Appendix G of ASME Section III as the reference flaw, amply exceed the current capabilities of inservice inspection techniques.
- Therefore, the reactor operation limit curves developed for this reference crack are conservative and provide sufficient safety margins for protection against nonductile failure. To assure that the radiation embrittlement ef fects are accounted for in the calculation of the limit curves, the most limiting value of the nil-ductility reference temperature, RTNDT, is used and this includes the radiation-induced shift, ARTNDT, corresponding to the und of the period f or which heatup and cooldown curves are generated.
The ASME approach for calculating the allowable limit curves for various heatup and cooldown rates specifies that the total stress intensity factor, K , for the combined thermal and pressure stresses at any time during heatup 7 or cooldown cannot be greater than the reference stress intensity f actor, KIR' l for the metal temperature at that time. K 7p is obtained from the reference f racture *,nughness curve, defined in Appendix G to the ASME Code. The K IR curve is given by the equation: nv V0GTLE - UNIT 1 8 3/4 4-12 i
- ~ _
b) v REACTOR COOLANT SYSTEM BASES _ PRESSURE / TEMPERATURE LIMITS (Continued) K s 26.78 + 1.223 exp (0.0143(T-RTET + 360)] (1) 7g Where: K IR is the reference stress intensity fr. tor as a function of the tretal temperature T and the metal nil-duct'lity reference temperature RTNDT* IDUS' the governing equation for the heatup-cooldown ualysis is defined in Appendix G of the AS*E Code as follows: ( CK7g + XIt 1#IR Where: K the stress intensity factGr caused by membrane (pressure) stress, yg K 7g = the stress intensity factor caused by the therta) gradients, K IR = constant provided by the Code as a function of temperature of the material, relative to the RT NDT C = 2.0 for level A and B service iimits, and C = 1.5 for inservice hydrostatic and leah test operations. p b At any time during the hestup or cooldown transient, %g is determired by the metal temperature at the tip of the postulated flaw, the appropriate value for RTNOT, and the reference fracture toughness curve. The thermal stresses resulting from ter.;perature gradients through the vessel wall are calculated and then the corresponding thermal stress intensity factor, KIT, f r the reference flaw is cc:rputed. From Equation (2) the pressure stress intentity factors are obtained and, from these, the allowable pressures are calculated. C00LOOWN For the calculation of the allowable pressure tersus coolant tenperature during cooldown, the Code reference flaw is assumed to exist at the inside of the vessel wall. During cooldown , the controlling location of the flaw is always at the inside of the vall because tne thermal gradients produce tensile stresses at the inside, which increase with increasing cooldown rates. Allowable pressure-temperature relations aro generated for both steady-state and finite cooldown rate situations. Fron these relations, composite limit curves are constructed for each cooldown rate of interest. The use of the composite curve in the cooldown analysis is necessary because control of the cooldown procedure is based on measurement of reactor coolant temperature, whereas the limiting pressure is actually dependen' on the material temperature at the tip of the assumed flaw. During cooldown, the 1/4T vessel location is at a higher temperature than the fluid adjacent to the
)
vessel 10. This condition, of course, is not true for the steady-state situa-l l [V tion. It follows that at any given reactor coolant temperature, the 6T developed during cooldown results in a higher v ,lue of Kgg at the 1/4T location l VocTLE - UNIT 1 i 8 3/4 4-13 l l
b y REACTOR COOLANT SYSTEM BASES - PRESSURE / TEMPERATURE L!MliS Gontinued) for finite cooldown rates than for steady-state operation. Furthertiore, if cor.ditions exist such that the increase in K7p exceeds Kgg, the calcu bted dilowable pressure during cooldown will be greater than the steady-state value. The above procedures are needed because there is no direct, control on leftperature at the 1/4T location; therefore, allowable pressures sey cnknowingly be violated if the rata of cooling is decreased at various intervals along a ccoldown ramp. The use of the composite curve climinatas thip problem and assures conservat.ive operation of the system for the entfre cooldown period. 3ATUp Three separate calculations are required to determine the limit curves for finite heatup rat,es. As is do1e in the cooldown analysis, allowable pressure-temperature r@lationships arc developed for steady-state Conditions as well as finite hootup rata conditions assuming the presence of a 1/4T defect at the insid? cf the vessel waTI. The thermal gradiente during heatup O v produce compressive stresses at the insido of the Wall that allevlate the tensile stresses produced by internal pressure. The metal *.ct:perature at the crack tip lage the coolant temperatum; therefore, the Egg f or the 1/4T ct'a:#, during heatup is lower than the Kgg far the 1/4T crack during steady-state cond tions at the safre coolant temperature. During haatup, esi.ecially at the end of the transient, conditions may exist such that the effects of compressive i , thermal stresses and dif ferent Ag's for steady-state and finite heatup ratos do not offset each other and the pressure-tencerature curve based on steady-state ccnditions no longer represents a lower bound of all similar tunes for finite heatup ratas when the 1/4T flaw is considered. Therefore, both cases have to be analyted in order to assure that at any coolant temperature t% lower value of the allowabic pressure calculated for stency-stato end finite heatup rates is obtained. . The second portion of the heatup analysis concerns the calS 91ation of ortnsure-tersperature limitations for the case in whic'i a 1/4T deep outside surface flaw is a;sumed. Unlike the situation at the vtssel fodde surface, the thermai gradients established at the ct.itsiJe surface durIng heitup prociuce stresses which are tensilo in nature and thus tend to reinforce any pressure stresses presant. These thermal stresses, of course, are dependent on both the rate of Acatup and the time (or coolant temparature) along the heatup ramo. Further:nore, since the thermal stresses at the outside are tensile and increasc wita increasing heatup rate, a lower bound cLeve cannot be defined. l Rather, each hestuo rate of interest must be analyzed on an individual basis. i l
~
V0G7tE - UNIT 1 83/44-14 j l l 1 i- _ J
s. -- - ,
t
; REACTOR 000LANT SYSTEM %,)
3- - _ Pr7Eb50RE/TEMFERATURC LIMITS (Contiated) Folicwing the geaaration of pressue2-temperature citrvec for both the steMy* state dnd finite heatup rata situations, the final limit curvas are produccc as folicws. A ccuposite curve is conttructe6 based 0,7 a point-by-poirit comparison of the steady-ctate and finite heatup rate cata. i.t any given teg orature, the allowable pressure is taken to be the 16 ser of the three values taken from the curves under consideration. The use of the composite curve is necessary to set canservative hea'.up limitations because it is possible for conditions to exist such that over the cturse of thG he tup ramp the controlling condition Witches from the inside to the outside and the pressure limit most at all times be based on alaTysis of the
most critical criterion.
, /c".)( f FiftrHy, the cceposite curves for the heatup rate data and the cooldown rate data are hdjusted for possible errors in the pressure and temperature 5, e ,,n ur sensing ir.struents by the values indicetta on the respective curves. /er 6 i 4 r ,2 Althougn '.he pressurizer operates in terporature ranges above those for which there is reason for concern of nonductile failure, operating limits are provided to assure corrcatibility of operation with the fatigue analysis performed in accorr'ance with the A5ME Code requirements. JAe f//h'.stc& !;ubWMtNes .kFh E-me,m- OVERPRESSURE PROTECT [0N SYSTEM _ ) ~
TheOPERABILITYoftwoPORVsMranRCSventOpening-c-ot-leasrt c - - - - - square inches ensures that the RCS will be protected from pressure transients which could exceed the limits of Appendix G to 10 CFR Part 50 when one or more of the RC$ cold legs are less than or equal to [-NS-3 F. Either PORV has adequate relieving cap 6bility to protect ttle RCS f rom overpressurization when the trsnsient.is limited to either: (1) the start of an idle RCP with the secordary Water terperature of the steam generator less tnan " or equal to g*F stove the RCS cold log temperatures, or (2) the start off e-- MI p g om iu M injection into a water-solid RCS. a// M r4< e m m ', * * '"p> c a.s Ce M. c s as e m u m .* ~' The MaxirNm All' owed PORY Setpoint for thehperatore Overpressure _cO Protection SystemSTOPS) is derived by analysis which models the perfornance of thr605 assuming various mass input and heat input transients. Operation with a PORV Setpoir.t less than or equal to the maximum Setpoint ensures that Appe4% criteria will nct be violated with consideration for a maximum pressure overshoot j Oeyond the PORV Setpoint which can occur o a result of tima de.tcys in signal , prccessing anti valvo npening, instrument uncertainties, and single failur6, To , ensurts that mass and heat input transients morC severe than those assumed cannot l occur, T9chnical Specifications require lockout of all but-ene safety infection i pump; and414-but-one-centr-iAg A charg Mg- M =p while in MODES 4, 5, and 6 with the reactor vessel heac installed and disallow start of an RCP if secondary temp I erature is more than 50 F above primary temperature. The Maxinm Allowed p0RV Setpoint for the LN will be updated based on the I results of examinations of reactor vessel material irradiation surveillance l m specinens performed as required by 10 CFR Part 50, Appendix H, and in accordancn l [v ) with the schedule in Table 4.4-5. w r.w.,a/aeffumy .4,V , we r s ea ,.a 1.e.ny ! ,f . p ,,.t,, 5 VCGTLE - UNIT 1 8 3/4 4-15
i 1 i Insert for B 3/4 4-15 O 0 -- Finally, the new 10CFR50 Appendix G Rule which addresses the metal tempera-ture of the closure head flange and vessel flange regions is considered. This rule states that 'the minimum metal temperature of the closure flange ragions should be at least 1200F higher than the limiting RTflDT for these regions when the pressure exceeds 20 percent of the preservice hydrostatic test pressure (621 psig for Westinghouse Plants). For Vogtle Unit 1 the minimum temperature of the closure flgnge and vessel flange regions is 140 F, since the limiting RTNDT is 20 F (see Table B 3/4-4.1). The Vogtle Unit I heatup curve shown on Figure 3-4.2 is not impacted by the new 10CFR50 rule. However, the Vogtle Unit 1 cooldown curve shown in Figure 3-4.3 is impacted by the new 10CFR50 rule. O O
4
,s REACTOR COOLANT SYSTEM BASES 3/4.4.10 STRUCTURAL INTEGRITY The inservice inspection and testing programs for ASME Code Class 1, 2, and 3 components ensure that the structural integrity and operational readiness of these components will be maintained at an acceptable level throughout the life of the plant. These programs are in accordance with Section XI of the ASME Boiler 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)(i).
Ccmponents of the Reactor Coolant System were designed to provide access to permit inservice inspections in accordance with Section XI of the ASME A dded. Boiler and Pressure Vessel Code, Cdition and AJJwde through As #1oMr>Nf 3/4.4.11 REACTOR COOLANT SYSTEM VENTS Reactor Coolant System Vents are provided to exhaust noncondensible gases and/or steam from the Reactor Coolant System that could inhibit natural circulation core cooling. The OPERABILITY of 1;;st on Reactcr C0clant--Sy+tema. reu4r a sce/ Ac44 vent path from-th [rt cter .essel h::d], th: [R:::ter C001:nt Sy:ter 'igh point], the-Epre::urizer steam-spac ], and the [f;clation :;nden::r Sigh pd p+%*t3 ensures that the capability exists to pe O rm this function. The valve redundancy of the Reactor Coolant System vent paths serves to minimize the probability of inadvertent or irreversible actuation while ensuring - that a single failure of a vent valve, power supply, or control system does not prevent isolation of the vent path. The function, capabilities, and testing requirements of the Reactor Coolant
- System vents are consistent with the requirements of Item II.B.1 of NUREG-0737, "ClarificationofTMIActionPlan$ Requirements," November 1980.
9 O V0GTLE - UNIT 1 8 3/4 4-16 .
._m 3/4.5 EMERGENCY CORE COOLING SYSTEMS ,
l BASES 1 I I 3/4.5.1 ACCUMULATORS r 4 i The OPERABILITY of each Reactor Coolant System (RCS) accumulator ensures 1 that a sufficient volume of borated water will be immediately forced into the j reactor core through each of the cold legs in the event the RCS pressure falls below the pressure of the accumulators. This initial surge of water into the core provides the initial cooling mechanism during large RCS pipe ruptures. l The limits on accumulator volume, boron concentration and pressure ensure r that the assumptions used for accumulator injection in the safety analysis are met. The accumulator power operated isolation valves are considered to be l " operating bypasses" in the context of IEEE Std. 279-1971, which requires that I bypasses of a protective function be remoted automatically whenever permissive , l condition. are not met. In addition, as these accumulator isolation valves l fail to meet single failure criteria, removal of power to the valves is required. j The limits for operation with an accumulator inoperable for any reason i except an isolation valve closed minimizes the time exposure of the plant to i a LOCA event occurring concurrent with failure of an additional accumulator a which may result in unacceptable poak cladding temperatures. If a closed isolation valve cannot be immediately opened, the full capability of one l accumulator is not available and prompt action is required to place the reactor in a mode where this capability is not required. 3/4.5.2 and 3/4.5.3 ECCS SUBSYSTEMS 1 2 The OPERABILITY of two independent ECCS subsystems ensures that sufficient
- emergency core ccaling capability will be available in the event of a LOCA-r assuming the loss of one subsystem through any single failure consideration.
Either subsystem operating in conjunction with the accumulators is capable of i supplying sufficient core cooling to limit the peak cladding temperatures I within acceptable limits for all postulated break sizes ranging from the ~ double ended break of the largest RCS cold leg pipe downward. In addition, each ECCS subsystem provides long-term core cooling capability in the , recirculation mode during the accident recovery period. I With the RCS temperature below 350*F, one OPERABLE ECCS subsystem is acceptable without single failure consideration on the basis of the stable reactivity condition of the reactor and the limited core cooling requirements. > ~ i !O l V0GTLE - UNIT 1 8 3/4 5-1 .
- ----.-e + - ,e - -ew ee y--,w.-ww--+'ers t,-e,---- - =-ee- ,--r9,mw g-- , ,,y. - ,-w y v t ,7 + - -e p -m e r,- - - wm=y-+- , -g
EMERGENCY CORE COOLING SYSTEMS ha BASES ECCS SUBSYSTEMS (Continued) 2 // /:,, ey,e ra b /e The limitation forj 2 maximum of ene-centr 4 fugal-charging-ptmp :nd one safety injection pump.sto be 4DERADLE and the Survei'!ance "equire ent *e ver fy 21' charging pu ps and safety injectier p"=ps except the -equired i C" O SLE charging pump tc be 'n:per:ble below-[275] F.provides assurance that a mass addition pressure transient can be reliev'ed by the operation of a single PORV. ],,f,w ggo e p The Surveillance Requirements provided to ensure OPERABILITY of each component ensures that at a minimum, the assumptions used in the safety analyses are met and that subsystem OPERABILITY is maintained. Surveillance Requirements for throttle valve position stops and flow balance testing provide assurtnce that proper ECCS flows will be maintained in the event of a LOCA. Maintenance of proper flow resistance and pressure drop in the piping system to each injection point is necessary to: (1) prevent total pump flow from exceeding runout conditions when the system is in its minimum resistance configuration, (2) provide the proper flow split between injection points in accordance with the assumptions used in the ECCS-LOCA analyses, and (3) provide an acceptable level of total ECCS flow to all injection points O equal to or above that assumed in the ECCS-LOCA analyses /a,,d co 7o cesare /^a/a etio,,/r% w m .ren
& es ,,t r> q a i ei.a < y ,y. p a m p ia j, /5 /ess W e..o o.n. o. .n. u. . v.u. . 3 e n.r. t. n.u. , ev. ereu. r n. o. r. t.n.u. . n. ] _' "/,d,,,,9 c ',io a pa ',*, , , ,,, a ,, .,4, The OPERABILITY of thefScron A'e fu Injeet4cn eli y M+eSy r Starage tem as part n,,,eofcaw.s r) ensures the ECCS
__ that sufficient negative reactivity is injected into the core to counteract
, any positive increase in reactivity caused by RCS cooldown. RCS cooldown can be caused by. inadvertent depressurization, a loss-of-coolant accident, or a ssteam line rupture.
The-limits- on-4njection tank minimum contained volume and beren concentration-ensure-that the : umption: used #^ th: :tcam lin: break analysis are-metAontained-water-vclum: Mmit includes sa slig & ace for water- not usable-because of tank discharge-line-locatien er ether physical characterictics. Phe-CPERABILITY-of--the redundant heat t"acia; channels - associated wi+h e^1"+4^^
^# +ha h^*^"
the berer i jectier n syst:- ensures that the se'ub4'4ty wA-be-mainta4ned-above-the-solubi'ity ' 4-i t cf 135 e :t 22,5CC pp- 50 0 .] J 3/4.5.5 REFUELING WATER STORAGE TANK Tee-CPERABILIT' cf the refueling water storage tank (DUST) as pa"+ af tha ECCS ensure that : cufficient supply .f c berated water is available fe" 4nject4 0" by the ECCS Jn the event of a LOCA The limits on RWST minimum volume and boron
concentration ensure that

(1) sufficient water is available within containment I ,
to permit recirculation cooling flow to the core, and (2) the reactor will remain subcritical in the cold condition following mixing of the RWST and the (J V0GTLE - UNIT 1 8 3/4 5-2 t w -,
M A!4 4 I 1 1 () EMERGENCY CORE COOLING SYSTEMS BASES REFUELING WATER STORAGE TANK (Continued) . RCS water volumes with all control rods inserted except for the most reactive control assembly. These assumptions are consistent with the LOCA analyses. The contained water volume limit includes an allowance for water not usable because of tank discharge line location or other physical characteristics. 10 5 The limits on contained water volume and[ boron concentration of the RWST 2 and M. for the solution recirculated alsoensureapHvalueofbetween;{8.5p'Hbandminimizestheevolutionof within containment after a LOCA. This p - iodine and minimizes the effect of chloride and caustic stress corrosion on
! mechanical systems and components.
i O 4 s i i O V0GTLE - UNIT 1 8 3/4 5-3
3/4.6 CONTAINMENT SYSTEMS BASES 3/4.6.1 PRIMARY CONTAINMENT 3/4.6.1.1 CONTAINMENT INTEGRITY Primary CONTAINMENT INTEGRITY ensures that the release of radioactive materials from the containment atmosphere will be restricted to those leakage paths and associated leak rates assumed in the safety analyses. This restriction, in conjunction with the leakage rate limitation, will limit the SITE BOUNDARY radiation coses to within the dose guideline values of 10 CFR Part 100 during accident conditions. 3/4.6.1.2 CCNTAINMENT LEAXAGE The limitations on containment leakage rates ensure that the total containment leakage volume will not exceed the value assumed in the safety analyses at the peak accident pressure, P3 . As an added conservatism, the measured overall integrated leakage rate is further limited to less than or equal to 0.75 L, Or 0.75 Lg , :: :pp?ic2!c, during performance of the periodic (~' test to account for possible degradation of the containment leakage barriers I d between leakage tests. The surveillance testing for measuring leakage rates is consistent with the requirements of Appendix J of 10 CFR Part 50. 3/4.6.1.3 CONTAINMENT AIR LOCKS The limitations on closure and leak rate for the containment air locks are required to meet the restrictions on CONTAINMENT INTEGRITY and containment leak rate. Surveillance testing of the air lock seals prcvides assurance that l the overall air lock leakage will not become excessive due to seal damage l during the intervals between air lock leakage tests.
. .C. .4 CONTAIN"CNT ISOLATION VAL'!: ANO C"ANNCL WELO PRES "RIZATION T T 1"5 ;C?TI h ]
W-CRESSILIP cf the hel:tien Yahe and C nt:f ent Charn:1 U:ld
' :::uri:stier Syst::: is required t: :::t-the r::tricti:n: : :;;r:ll i
l l ecnt:f cent h k :te :cuced '" the :sfety n !y:::. Th: Eurvei'hn::
':;ui :::nt: f:r deter '^ing CPERTSILIT' cre cen ictent with ';;;ndi, J f 1: "7 ';rt 50.
1 O ! O V0GTLE - UNIT 1 8 3/4 6-1
.~.
CONTAINMENT SYSTEMS BASES 4 3 /4. 6.1. 5' INTERNAL PRESSURE The limitations on containment internal pressure ensure that: (1) the containment structure is prevented from exceeding its design negative pressure differential with respect to the outside atmosphere ofj3Ypsig, and (2) the containment peak pressure does not exceed the design pressure of [.54] psig during [LO:3 or steam line break conditions]. 52 41 9 a.sswariarg an irsi+ial cor, trio,r,,ent pre ssm do.s ps y The maximum peak \ pressure expected to be obtained from a [LOCA e. steam line break] event is @45-]-psig The limit of containment pressure will limit the totalsure presto/37'psig forwhich Efe-] psig, initialis positive less thandesignpressureandisconsistentwiththesafety/ analyses. S 45 3 /4. 6.1. 9 AIR TEMPERATURE The limitations on containment average air temperature ensure that the over-all containment average air temperature does not exceed the initial temperature condition assumed in the safety analysis for a [LOCA e, steam line break accident]. Measurements shall be made at all listed locations, whether by fixed or portable instruments, prior to determining the average air temperature. 3/4.6.1.hCONTAINMENTSTRUCTURALINTEGRITY P rc:tre:::d concret: cent:1 7: t "ith ungreeted tenden:] 41 9 Thislimitationensuresthat/thestructuralintegrityofthecontainmentwill be maintained comparable to the riginal design standards for the life of the facility. Structural integrity s required to ensure that the containment will with-stand the maximum pressure of psig in the event of a [ LOC? ;r steam line break acciden . The measurement of containment tendon lift-off force, the tensile tests of the t ndon wires or strands, the visual examination of tendons, anchorages and exposed interior and exterior surfaces of the containment, and the Type A leakage test are sufficient to demonstrate this capability. (The tendon wise-ar strand samples will also be subjected to stress etciing tests and to accelerated corro-sian tests to simulate the tendon's operating conditions and environment.) Ruts;on z. of The Survcillance Requirements for demonstrating the containment's structural integrity are in compliance with the recommendations of png:::d Regu atory Guide 1.35, " Inservice Surveillance of Un routed Tendons in Presthrencrete sed _ Containment Structure < " igr" 1979. an proposed'ReguTatory Guide 1.35.1, " Deter-
~
S Tng Prestressing Forces for Inspection of Prestressed Concrete Containments," i April 1979. j i The required Special Reports from any engineering evaluation of containment l abnormalities shall include a description of the tendon condition, the condition of the concrete (especially at tendon anchorages), the inspection procedures, the tolerances on cracking, the results of the engineering evaluation, and the correc-tive actio _ V0GTLE - UNIT 1 B 3/4 6-2
(Sea /ed c/oseL isola Wen valves are isola / ton va./ves ander CONTAINMENT SYSTEMS administra.Hve. co&rol fo as sure tha-/ Hey cannot be.
) inadver/enH BASES ) mechanicaly epened.
devices ddminMrs.+we
/o sealer cm/rol lock /he Va./ve c/csed, /Aeinclud L use of b/ind r%
CONTAINMENT STRUCTURAL INTEGRITY (Continue'ger,or removal ofpr fa ne valve. d) ,gg gf, y., [-Reinfcrced concrctc containment] Thi: li:-itation ensure: that-the--Structural integrity of the centainment will be4aintained cmarable to the original.-des 4gn-standards for the 'if: ef the-fac414ty. Structural 4-tegrity i: required te en;ure that the containmcat
> i'l wit-h:tand the maxi mum pre :ure Of [48] psig i- the event of a [LOCA er i steam line break accident]. A " ice 1 4nspection 4- conjunction with the Type ^ !cakag0 t00t: i cuf#icient te demonstrate this capability.
7 3/4.6.l h CONTAINMENT VENTILATION SYSTEM a . The,[42-inch 7 containment purge supply and exhaust isolation valves are required to be sealed closed during plant operations since these valves have not been demonstrated capable of closing during agLOCA or steam line break accidentK Maintaining these valves sealed closed during plant operation ensures that exces-sive quantities of radioactive materials will not be released via the Containment Purge System. To provide assurance that these containment valves cannot be inad-vertently opened, the valves are sealed closed in accordance with Standard Review Plan 6.2.4 which includes mechanical devices to seal or lock the valve closed, or prevents power from being supplied to the valve operator. + The use of the containment purge lines is restricted to the inch purge j supply and exhaust isolation valves cinec, unlike the [42-inch] valvc:, [.h;
,g valvc arc capabic Of cic ing during [LOCA cr :te = line brcok accident]. There-eeme &rfW the SITE BOUNDARY dose guideline of 10 CFR Part 100 would not be exceeded in the event of an accident during containment PURGING operation. Operation with one pai cf the:: valve: Oper "i be 'i-ited te [1000] heur; dur ing a calendar year.
N: tet:! time the cent 2i" Tent purge (vent) syste- i 012 tier v2!ve "'2y be oper dur ng "00ES 1, 2, 3, and
in : :!cador year i ; function of anticipated acid i
and operating experience. Or!y esfety-related rescent; e.g., contai m:nt.prc: cure centro! cr the reductica cf airborne radioac44vity to facilitata perscanci : cess fcr :urvcillance and m:intenance asti.'itics, may bc used to support the additional time requests. Only :sfetyvelated-ressons-shecid be used te justifj the wco,ug of these isc! tion valves-during-MODES 1, 2, 3, and i in :ny calendar yccr regardles cf the allee b!c hourt. Leakage integrity tests with a maximum allowable leakage rate for containment purge supply and exhaust supply valves will provide early indication of resilient material seal degradation and will allow opportunity for repair before gross leak-age failures could develop. The 0.60 L leakage limit of Specification 3.6.1.2b. shallnotbeexceededwhentheleakagefatesdeterminedbytheleakageintegrity tests of these valves are added to the previously determined total for all valves and penetrations subject to Type B and C tests. 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS 3/4.6.2.1 CONTAINMENT SPRAY SYSTEM The OPERABILITY of the Containment Spray System. ensures that containment C' depressurizationandcoolingcapabilitywillbeavailableintheeventofaILOCA or steam line trealQ'. The pressure reduction and resultant lower containment leakage rate are co'nsistent with the assumptions used in the safety analyses. V0GTLE - UNIT 1 B 3/4 6-3
m
) CONTAINMENT SYSTEMS s
BASES CONTAINMENT SPRAY SYSTEM (Continued) [Cr:dit tak:n for i din r;;;vol] gg, gfg The Containment Spray System and the Containment Cooling Systemj:r r:dund:nt t 22 ' :ther ' prc'/iding post-accident cooling of the containment atmosphere. However, the Containment Spray System also provides a mechanism for removing iodine from the containment atmosphere and therefore the time requirements for restoring an inoperable Spray System to OPERABLE status have been maintained consistent with that assigned other inoperable ESF equipment. [N; cr:dit tak n f:r iodin; r;;;;;1] The Centa-inment-Spray-Sy ter and the-Gontain;;nt-Cooling System a~s redundant ta each-cther-wprovMing post accident eccling of th; ;catain;;nt stac phere. Stace-no-credit-has-beca taken for icdine reacv 1 by the Contain;;nt Spray Sy: tem, the sila ble cut f :crvice ti : require::nt: for the Cent:ia crt Spray Sys M-
nd C ntairment C Oli ng Syster have beer '-terrelated 2nd adjusted to reflect t@acditional redundancy ' : Ol'ng ::p:bi'ity, 3/4.6.2.2 SPRAY ADDITIVE SYSTEM B TTICNAL]
The OPERABILITY of the Spray Additive System ensures that sufficient NaOH is added to the containment spray in the event of a LOC,A. he limits on Na0H volumeandconcentrationensureapHvalueofbetweenf8.5 and Est-@ for the g*g solution recirculated within containment after a LOCA. Th s pH band minimizes O w/ the evolution of iodine and minimizes the effect of chloride and caustic stress corrosion on mechanical systems and components. The contained solution volume limit includes an allowance for solution not usable because of tank discharge line location or other physical characteristics. These assumptions are consistent with the fodine removal efficiency assumed in the safety analyses. 3/4.6.2.3 CONTAINMENT COOLING SYSTEM [CPTICNAL] The OPERABILITY of the Containment Cooling System ensures that: (1) the containment air temperature will be maintained within limits during normal operation, and (2) adequate heat removal capacity is available when operated in conjunction with the Containment Spray Systems during post-LOCA conditions. [Cr:dit taken f r iodine rezcv 1og by,spray lypt Xst;;;] both pro vo.de The Containment Cooling System end theIC ontainment Spray System ere redm-d:nt to each other 4a prefidia; post-accident cooling of the containment atmos-phere. As a result of this r;dundedcy 'n cooling capability, the allowable out-of-service time requirements for the Containment Cooling System have been appropriately adjusted. However, the allowable out-of-service time requirements for the Containment Spray System have been maintained consistent with that assigned other inoperable ESF equipment since the Containment Spray System also provides a mechanism for removing iodine from the containment atmosphere. [Ne ;recit tsken far isdine removal by : pray syst ::] The Contair... nt Occling Sy:t;; :nd th: C;ntair;;.t Spr:y Sy:t:r cr: r;dundant to-eech other in providing pc t :::ident eccling of th: : ntai ::nt atmosonere. Sine n credit 5:0 beer taker for iodine re-e'!:1 by the Coataf"- nen: Spray-Systea, the alicweble cut of service time re;;uir;;;nt; f;r the Con; air an; 0; cling Syst:= :nd Centain::nt Spray Sy:t: h:v: been '-t:rr:1:ted ad-ad-jwed-to reflect thi Oddi-t4enM--redundancy-in-ccoling ceps;itys V0GTLE - UNIT 1 B 3/4 6-4 l -
() CONTAINMENT SYSTEMS BASES f Pt E A uf tn 7 9,/,9..1 T..A M .Um P .mmm~. .e V. e T.m ru. s ..Af PA.NT.T.AMm ; Th CPE"ASILITY cf th: cont *inment-iediee-f4+ tee-train; ensures that
uf#icient-icdine removal-capabi'ity "4 bc : vail:b!: 4- th: :=nt of : LCCA.
The reductier ia contai-ment 1 dine 4 ventary reder:: the re:ulting SITE SOUNDAov radiation-dese ::::ci:ted "ith centai ment le:k:ge. Oper:ti^ of th: cy:ter with the h::ter: Operating for at ? ::t 10 heur: 4-31-day peri:d i: suf'ic4ent-to-reduce-the-bui-idup-of =0f stura on th: adscrbers and "CPA
'i'ters. 'he operation Of thi: cyster and resultant iodine r:mov:1 ::p;;ity re :0n:ictent ith the :cumption used 4- the LOC? On:ly:::. "!S! "510-1975 and Centric Letter 83-13, "Clarificati:n of Surveillon e ";qu r ;;nts fec "CPA Tiiters and Chacecal-Adsorbee-Units in Standard T: hnical Sp::ification :n ESI Cleanup-Systems" cra used as procedural guid;- for surveillance testing 3 3/4.6.-)- CONTAINMENT ISOLATION VALVES The OPERABILITY of the containment isolation valves ensures that the containment atmosphere will be isolated from tha outside environment in the event of a release of radioactive material to thi containment atmosphere or pressurization of the containment.and is consistant with the requirements of General Design Criteria 54 through 57 of Appendix A to 10 CFR Part 50. Contain-ment isolation within the time limits specified for those isolation valves
("')s s ,, designed to close automatically ensures that the release of radioactive material to the environment will be consistent with the assumptions used in the analyses for a LOCA. 4 3/4.6.-E COMBUSTIBLE GAS CONTROL The OPERABILITY of the equipment and systems required for the detection , and control of hydrogen gas ensures that this equipment will be available to maintain the hydrogen concentration within containment below its flammable limit during post-LOCA conditions. Either recombiner unit [or the Purge System] is capable of controlling the expected hydrogen generation associated with: (1) zirconium-water reactions, (2) radiolytic decomposition of water, ~ and (3) corrosion of metals within containment. [Cumuistiv; :p;rati n of th pege-Gystem-wi-th the h:: tars cpar; ting for 10 entinu-5 heur5 in a 01 dei period is-sufficient-to-FeduGO-the buildup Of m0ictur On th ad: Orb r and HE?' ' :t:r-}, These Hydrogen Control Systems are consistent with the recom-mendations of Regulatory Guide 1.7, " Control of Combustible Gas Concentrations in Containment Following a LOCA," March 1971. The Hydrogen Mixing Systems are provided to ensure adequate mixing of the containment atmosphere following a LOCA. This mixing action will prevent i localized accumulations of hydrogen from exceeding the flammable limit.
/2.S.C PENETRAT@N-ROGM-EXHAUST A!!! CL"A"UP SYSTOi [CPTIONAL]
n. ---- -, ,-,
ine v r ;.,na i ut i e vi wuc
,c ._____=.un r cuc u a .__,___x_.
avvm mono .. n:_ o.,
.....r. e,...,__ ..m.. _.... ,s I
that radieaetiva materials-lesking-from-the -containment atmosphece-through ::n-l tai ::nt p:nctraticn: f0!!Owing : LCCA cre filtered and ad: rb d pri;r t r: ching one :nvircrment. Operatier of th: cyster with the h:: tars :p;r; ting for at l . ! V0GTLE - UNIT 1 8 3/4 6-5 1
h
\
ON CONTAINMENT-SYSTEMS s
\ -
N BASES
/ \
PENETRATION ROOM EXHAUST AIR CLEANUP SYSTEM [0PTIONAL] (Continued) least\10 continous hours in a 31-day period is sufficient to reduce e buildup of mois'ture on the adsorbers and HEPA filters. The operation of th system and the resultant effect on offsite dosage calculations was assu in the LOCA analyses. ANSI N510-1975 and Generic Letter 83-13, "Clari ation of SurveillancixRequirements for HEPA Filters and Charcoal Adsorbpr Units in Standard Technical Specifications on ESF Cleanup Systems" ar /used as procedural guides for surveillance testing. 3/4.6.7 VACUUM RELIEF VALVES [0PTIONAL] N The OPERABILITY o.f the primary containment to atmosphere vacuum relief valves ensures that the containment internal pressu e does not become more i negative than psig.xNThis condition is necess to prevent exceeding the containment design limit for internal vacuum of . psig.
'N \ \ \
l \/
\ \s \ / \ \ /
! / \
/ / \'s / \
/
/ 'N s
i / \
/ \ ,/ \
j / \ l / \ ! ! \
,/ \ ,/ \, / N - /
O / V0GTLE - UNIT 1 8 3/4 6-ts . e
-hw. i 1
m
) 3/4.7 PLANT SYSTEMS BASES 3/4.7.1 TURBINE CYCLE 3/4.7.1.1 SAFETY VALVES '3*4 ags TheOPERABILITYofthejmainsteamlineCodesafetyvalves/ensuresthat the Secondary System pressure will be limited to within 110% B-lOe psigfof its design pressure of {1000] psig during the most severe anticipated (ystem operational transient. The maximum relieving capacity is associated with a Turbine trip from 100% RATED THERMAL POWER coincident with an assumed loss of condenser heat sink (i.e. , no steam bypass to the condenser).
17Co7,Lto The specified, valve lift settings and relieving capacities are in accordarcewitht/ierequirementsofSectionIIIoftheASMEBoilerandPressure Code, 1971 Edition. The total relieving capacity for all valves on all of the steam lines is / lbs/h which is y % of the total secondary steam flow of , lbs/h at 100% RATED THERMAL POWER. A minimum of two OPERABLE safety valves per steam generator ensures that sufficient relieving capacity is avai,lable for the allowable THERMAL POWER restriction in Table 3.7-2. eco
/.f '14ci'^"'"" " " """ "'"^ " "" "' "" "' ""*' '" "*"'
within the limitations of the ACTION requirements on the basis of the reduction , Cj' in Secondary Coolant System steam flow and THERMAL POWER required by the
reduced Reactor trip settings of the Power Range Neutron Flux channels. The Reactor Trip Setpoint reductions are derived on the following bases, ba s is *.
For 4N-loop operation 50 = (X)-(Y)(V)xIl09V X rer N-1 leep-eperat4en
/V% _ /\/\/ff\
t
's'
_L A Where: SP = Reduced Reactor Trir Setpoint in percent of RATED THERMAL POWER, V = Maximum number of inoperable safety valves per steam line,
'J - Maxime.m number-ef inoperable safety va!ver per operating t:= nc, lO V0GTLE - UNIT 1 B 3/4 7-1 l -- - - . - -.m
.-l 1
l l l l I Q,m, PLANT SYSTEMS BASES SAFETY VALVES (Continued) 4 fl097 = Power Range Neutron Flux-High Trip Setpoint for-E&} loop operation, [7G) = tiaximum-percent-of-RATED-TRERMAL POWER per-is? 4hla hy P-3 S:tpoint for D!-1] 100p Operation, X = Total relieving capacity of all safety valves per steam line in lbs/ hour, and Y = Maximum relieving capacity of any one safety valve in lbs/ hour 3/4.7.1.2 AUXILIARY FEEDWATER SYSTEM The OPERABILITY of the Auxiliary Feedwater System ensures that the Reactor CoolantSystemcanbecooleddowntolessthanj350fFfromnormaloperating conditions in the event of a total loss-of-offsite power.
% Ot+e-driver auxiliary ' ecd":t:r pu p i: ::pel: cf d:l';;r-Och lectri I gh"# "'# [4~; a teta! feedeter *!ce ef-[350] gpr at : pre::ur; cf [11:2] psig to the
t:2 ge ersterr. he stes--d i'fea aux" ir y #ee9 2te pu p Y*#J/47%fcrt2cee't" 3 it cereb'c c' de'i'!er ~; 2 total fee &eter ce of [700] ;;p :t : pre::tr: ef i
' Ell::: pais to the entrance of the staa; generators. This capacity is suffi j :icnt t On ur that adequate feedwater ficw is availabic t: removc d;;;y heat iand redu the 902: tor C001:nt Sy t = t :peratur: to 1c:: than [250] I when ' th; h;idual ~~ic;t P,casval Sy;t; : y bc pl;;;d int Opcrati0n.
3/4.7.1.3 CONDENSATE STORAGE TANK The OPERABILITY of the condensate storage tank with the minimum water volume ensures that sufficient water is available to maintain the RCS at HOT l STANDBY conditions for 4 hours with steam discharge to the atmosphere l concurrent with total loss-of-offsite poweru The contained water volume limit incluces an allowance for water not usable because of tank discharge line location or other physical characteristics. , 1 b/!ewel by a coelsewn *c i 3/J.7.1.4 SPECIFIC ACTIVITY
'# H A /"i#i"###" Ned iYio" 5 l
The limitations on Secondary Coolant System specific activity ensure that the resultant offsite radiation dose will be limited to a small fraction of 10 CFR Part 100 dose guideline values in the event of a steam line rupture. This dose also includes the effects of a coincident 1 gpm reactor-to-secondary A tube leak in the steam generator of the affected steam line. These values are consistent with the assumptions used in the safety analyses. V V0GTLE - UNIT 1 8 3/4 7-2
(m Insert to Page B 3/4 7-2 The auxiliary feedwater system is capable of delivering a total feedwater flow of 510 gpm at a pressure of 1221 psig to the entrance of at least two steam generators while allowing for: 1) any possible spillage through the design worst case break of the main feedwater line: 2) the design worst case single failure and, 3) recirculation flow (applicable for turbine-dri'/en auxiliary feedwater pump only). This capacity is sufficient to ensure that adequate feedwater flow is available to remove decag heat and reduce the Reactor Coolant System temperature to less than 350 F, at which point the Residual Heat Removal System may be placed in operation. Because it is not desirable to inject cold auxiliary feedwater into the steam generators during power generation, the pumps must be tested on miniflow (recircula-tion). The surveillance acceptance criteria are based on the miniflow testing configuration and are specified to ensure the above limits are met during injection to the steam generators. O !O L-
PLANT SYSTEMS BASES 3/4.7.1.5 MAIN STEAM LINE ISOLATION VALVES The OPERABILITY of the main steam line isolation valves ensures that no more than one steam generator will blow down in the event of a steam line rupture. This restriction is required to: (1) minimize the positive reac-tivity effects of the Reactor Coolant System cooldown associated with the blowdown, and (2) limit the pressure rise within containment in the event the steam line rupture occurs within containment. The OPERABILITY of the main steam isolation valves within the closure times of the Surveillance Require-ments are consistent with the assumptions used in the safety analyses. 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION , The limitation on steam generator pressure and temperature ensures that the pressure-induced stresses in the steam generators do not exceed the maximum allowable fracture toughness stress limits. The limitations of f703*F and 12007psigarebasedonasteamgeneratorRTNDT of Q F and are sufficient to prevent brittle fracture. O 3/4.7.3 COMPONENT COOLING WATER SYSTEM The OPERABILITY of the Component Cooling Water System ensures that suf-ficient cooling capacity is available for continued operation of safety-related equipment during normal and accident conditions. The redundant cooling capacity of this system, assuming a single failure, is consistent with the assumptions used in the safety analyses. N:/C Lssiil SGMV/CC Cecu WG. 3/4.7.44 C '! ICE WATER SYSTEM p]uejeu se v tree Co a bg The OPERABILITY of the S--v kc Water System ensures that sufficient cooling capacity is available for continued operation of safety-related equip-ment during normal and accident conditions. The redundant cooling capacity of this system, assuming a single failure, is consistent with the assumptions used in the safety analyses. 3/4.7.5 ULTIMATE HEAT SINK [CPTIO"^.L] g g , , ,, c,,, m , , ,, /,.g
,do r,m e r e t offM&E r&n, The limitations on the ultimate heat sink / level and temperature,tensure that sufficient cooling capacity is available./either,: (1) provide normal cooldown of the facility or (2) mitigate the effects of accident conditions within acceptable limits.
O f V0GTLE - UNIT 1 8 3/4 7-3 _ ._ .a i . _ _ _ _ __ . - .-
,m PLANT SYSTEMS
(] BASES ULTIMATE HEAT SINK (Continued) 4 "2 "'P"' refedM
* *I' " & Cl'EMA84&%s The limit ions k onade&E-minimum water level,ee4 maximum temperaturelare based on providing a 30 dQ cooling water supply to safety-related equipment without exceeding its design basis temperature and is consistent with the recommend-ations of Regulatory Guide 1.27, " Ultimate Heat Sink for Nuclear Plants,"
March 1974. 3/4.7.C ILOOO ?ROTECTION [0PTIONAL] The Lini-tation-on-f-lood-protection :n'; crc: that-fac414ty pret :tive
icn: ui' 5 ::icr (and operation will b: t: 'nated) 'n th: : :nt ;f ;;xinun
" n St La.;l is t,ea d m. ..m fl;;d
ndition:. The ' 4-it of !: vat 4en
tien at *ich-faeility 'lced centrcl

:::ur:: pr:vid: pr:t::ti:r to r2'aty--e!:t2d equipment.
G F/L D2A m b/ 3/4.7.P CONTROL ROOM EMERGENCY AIR CLEANUP SYSTEM c' d OEoM' The a.IL[TY eid ofnt cmdiNonsthe Control Room Emergency Air Cleanup Syst R Q(/ that: (1) the ambient air temperature does not exceed the allowable temperature for co;1tinuous-duty rating for the equipment and instrumentation cooled by this systemy and (2) the control room will remain habitable for operations personnel during and following all, credible accident conditions. Operation of the
. w ?! e ' system with the heatergcper: ting for at least 10 continuous hours in a 31-day . periou is sufficient to reduce the buildup of moisture on the adsorbers and r/< u //
M"f'"^ HEPA filters. The OPERABILITY of this system in conjunction with control room design provisions is based on limiting the radiation exposure to personnel occupying the control room to 5 rems or less whole body., or its equivalent. This limitation is consistent with the requirements of General Design Criterion 19 of Appendix A,10 CFli Part 50. ANSI N510-F#5 will be used as a l procedural guide for surveillance testing.
/?to 7 M^6PEM M 4MNAM'AE W "#N'"L M' & 'SW M '
3 /4. 7. ECCS PUMP ROCP EXIlAUST AIR CLEANUP SYSTO' rip in a Fe n e +un.er, A en i tHra +i ,, aa,m;.m t nee, ' The OPERABILITY of the;EES-Pump R Or Exhau t "- Cleanup Systa= ensures that radioactive materials leaking from the.ECCS equipment within the S ,,re,/ - pump room following a LOCA are filtered prior to reaching the environment. l e -ty. c f Uperation of the system with the heatersfcper: ting- for at least 10 continuous hours in a 31-day period is sufficient to reduce the buildup of moisture on ener ?ue i'
' the adsorbers and HEPA filters. The operation of this system and the resultant i effect on offsite dosage calculations was assumed in the safety analyses. '
ANSI N510- will be used as a procedural guide for survel'Ilance testing. Ii 10 es,, A.inmc,,r m e e A.wic.s.' T
,t7 9.pp d. frA bi e re 700 M f st r, $ h V0GTLE - UNIT 1 8 3/4 7-4 l
PLANT SYSTEMS ASES 3/ .9 SNUBBERS snubbers are required OPERABLE to ensure that the structural integrity of the % actor Coolant System and all other safety-related systemsf is main-tained du 'ng and following a seismic or other event initiating dynamic loads. Snubber are classified and grouped by design and manufacturer but not by size. For exa. le, mechanical snubbers utilizing the same design features of the 2-kip,10-k1 and 100-kip capacity manufactured by Company "A" are of the same type. The s , design mechanical snubbers manufactured by Company "B" for the purposes ohthis Technical Specification would be of a different type, as would hydraulic snubbers from either manufacturer. f l . l
A list of individu snubbers with detailed infonnation of snubber location and size and of system a cted shall be availablg'at the plant in accordance with Section 50.71(c) of 1(CFR Part 50. The acdessibility of each snubber shall be determined and app ved by the [ Unit Review Group]. The determination shall be based upon the exist g radiation lev'els and the expected time to perform a visual inspection in ach snubberflocation as well as other factors associated with accessibility du *ng plant' operations (e.g., temperature, atmosphere, location, etc.), and t e reco'mmendations of Regulatory Guides 8.8 and 8.10. The addition or deletion f dny hydraulic or mechanical snubber shall be made in accordance with Sect n 50.59 of 10 CFR Part 50.
The visual inspection freque is sed upon maintaining a constant level of snubber protection to each safet grelated system during an earthquake or severe transient. Therefore{ the required inspection interval varies inversely with the observed siubber failureshn a given system and is determined by the number ~of inoperablejnubbers found during an inspection of each system. In order to establish the. inspection frequency kr each type of snubber on a safety-related system, it'was assumed that the frequency of snubber failures and initiating events is constant with time and tha( the failure of any snubber on that system could cause the system to be unprotectgd and to result in failure during an assumed initiating event. Inspections perfonned before that interval has elapsed may be.'used as a new reference point to det' ine the next inspection. However, the results of such early inspections performed efore the original ' required time interval has elapsed (nominal time less 25%) say not be used to lengthen the required inspection interval. Any inspection whose results require i a shorter inspection interval will override the previous sche . The acceptance criteria are to be used in the visual inspec on to determine OPERABILITY of the snubbers. For example, if a fluid port of a by aulic snubber 1s found to be uncovered, the snubber shall be declared inop rable and shall not be determined OPERABLE via functional testing.
functi'onal testing methods is used with the stated acceptance criteria

T 1
O I i V0GTLE - UNIT 1 B 3/4 7-5
\
</fyc/ the S~nu bbers are pro eided to ens ure +ha + the &ree-fura./ mrye is rea e ve r ccela e,f .s ys 4m and a // chwr Safefy- re64d sys+ erns ser%1.L cr c%r eved ir ridsa -lij ma in da ined clariy and h//cwiy a.
U d e va n 4 b a ,/.1. PL-ANT SYSTEMS Pe' r 6-eneei& LeWersVn3 BASES t 15 3/4.7.fSNUBBERS f "' nubber_ 2re required OPEMBLE te ensure that the structu"1 in+=grity ic - iataiaed j (cf the reacter ceelant syster reicric and er al' ether ether event::fety related initiating cyctem: dynamic leads. j dur ng and fe!!ew ng i i j-9 Snubber: excluded ' rem this nspectier pregr:r are there installed er nencafety-i re!ated systems and ther enly i' thei" failure ep failure of the syste= ea
! 9fcb they are nstalled, weeld have ne adver:0 effect er any : fety-re!:ted i
l \ system The visual inspection frequency is based upon maintaining a constant level of snubber protectior,to systems. Therefore, the required inspection interval varies inversely with the observed snubber failures and is determined by the number of inoperable snubbers found during an inspection. Inspections performed before that interval has elapsed may be used as a new reference point to determine the next inspection. However, the results of such early i inspections performed before the original required time interval has elapsed (ncminal time less 25%) may not be used to lengthen the required inspection interval. Any inspection whose results require a shorter inspection interval will override the previous schedule. When the cause of the rejection of a snubber is clearly established and remedied for that snubber and for any other snubbers that may be generically susceptible, and verified by inservice functional testing, that snubber may be exempted from being counted as inoperable. Generically susceptible snubbers are those which are of a specific make or model and have the same design features directly related to rejection of the snubber by visual inspection or are similarly located or exposed to the same environmental conditions, such as temperature, radiation, and vibration. When a snubber is found inoperable, an engineering evaluation is performed, in addition to the determination of the snubber mode of failure, in order to determine if any safety-related component or system has been adversely affected by the inoperability of the snubber. The engineering evaluation shall determine whether or not the snubber mode of failure has imparted a significant effect or degradation on the supported component or system. To provide assurance of snubber functional reliability, a representative samola of the installed snubbers will be functionally tested during plant ; shutcowns at 18-month intervals. Selectien of a representative cample according to tnc cuprezien 35 (1 C) providc: a ccafidence ic.cl cf apprcxt::tely 95% that 9 N te 1005 ef the snubber: ia the plant "4" be OPE ESLE withia acceptance li : t:. Observed failures of these sample snubbers shall require functional testing of additional units. Hydraulic snubbers and mechanical snubbers may each be treated as a different entity fcr the above surveillance programs. vcCa24~- CAWA W YT5 B 3/4 7-5 JTl 1 ; ;;;g
~~~
PLAN'T SYSTEMS d,a BASES wowww ew.coc1ssoe gvvowunuwwy f r _ . _ ._ a 1
1. Functionally test-10Lof-a-type of snubber with-an-additien:1 1C% tested for cach function:1 testing failure, er 2 c unctionally-test sample-s4ze-and-determine :mple acceptance er
.m .y_ 4 , n_ + 4. n. n , , e 4. n. ,, r i. m, ' "a
_ . __ -- a.7-7_, nr
2. Functionally te:t : representative ::mple size and deter #n :: pic acceptance er rejectier using the :tated equatier.
cigure 4 7-1 u:: developed c;ing "Wald'; Scquential Pr b:bility Rati;
?!cn" : dc:cribcd in " Quality Control and Industrial Statistics" by Ache:ca J. Dunc:a.
Permanent er cther exemption: ' rom-the curveillance pr; gram for individuel
bber; may be granted by the Cc--i;;ica if a Justifiable basis fec exemption is prc ented and, i' applicabic, snubber life destructive testing L;; performed to-qual'fy the snubber; for th: Opplicabla design conditions at either the cum pletion of their fabrication or at-a-subsequent datc. Snubber; ;; cx :pted

h:l' bc li ted in the list of individual snubbers indic; ting the mmtent
s of th; cx: pticn:.
The service life of a snubber is established via manufacturer input and information through consideration of the snubber service conditions and associated installation and maintenance records (newly installed snubbers, seal replaced, spring replaced, in high radiation area, in high temperature area, etc.). The requirement to monitor the snubber service life is included to ensure tnat the snubbers periodically undergo a performance evaluation in view of their age and operating conditions. These records will provide statistical bases for future consideration of snubber service life. 3/4.7.1/ SEALED SOURCE CONTAMINATION The limitations on removable contamination for sources requiring leak testing, including alpha emitters, is based on 10 CFR 70.39(a)(3) limits for plutonium. This limitation will ensure that leakage from Byproduct, Source, and Special Nuclear Material sources wi31 act exceed allowable intake values. Sealed sources are classified into tnree groups according to their use, with Surveillance Requirements commensurate with the probability of damage to a source in that group. Those sources which are frequently handled are required to be tested more often than those which are not. Sealed sources which are continuously enclosed within a shielded mechanism (i.e. , sealed sources within radiation monitoring or boron measuring devices) are considered to be stored and need not be tested unless they are removed from the shielded mechanism. n\ V0GTLE - UNIT 1 B 3/4 7-6
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to 3/4.7.14 AREA TEMPERATURE MONITORING < The area temperature limitations ensure that safety-related equipment will ; not be subjected to temperatures in excess of their environmental qualification l temoeratures. Exposure to exce:::ive temperatures may degrade equipment and can cause a loss of its OPERABILITY. The temperature limits include an allowance for
instrument error of *F. ,'
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t () Insert to B 3/4 7-7 i 3/4.7.11 ENGINEERED SAFETY FEATURES (ESF) RCCf! COOLER AND SAFETY-R2 LATED CHILLER SYSTEM ) The operation of the ESF Room Cooler and Safety-Related Chiller System er.sures that the ambient air temcerature does not exceed the allewable ter.perature fcr continuous duty rating for the equipment cooled by the system. t ) 3/4.7_12 REACTOR COOLANT PUMP THERMAL BARRIER Cv0LIkG WATER ISOLATIGN Tr.is isolation function is designed to prevent a spill of the reactor J coolant from a postult :ed breached thermal barrier should a break accur in i the tor. safety-related ACCW piping downstream of the isolation salve, i s 1
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! 3/4.8.1 3/4.8.2, and 3/ g V .C. SOURCES, D.C. SOURCES, and ONSITE POWER DISTRIBTION
The OPERA 8ILITY of tite A.L and D.C power sources and associated distribu- .
tion estems during eperation ensures that sitTicient power will be available to supply the safity-relate 6 equipment required for: (1) the safe shutdown of the f=.ciMty, and (2) the mitigation and control of accident conditions within the facility. The minimum qXcified independent and redundant A.C. and D.C. powee 'oers:e.,
, and distribution systems satisfy the re Design Criterion 17 of Appendix A to 10 CfR Port O. quirements of Cereral Apps.a,dte A th . ; The ACTION requirements fspecified Qr the levels of degr&@ tion af the power sources orovide restrid. ion upon continued facility operation commenwfate .
I with the 4 vel of degradtfoL i. The OPERABILITY of the pcser sources are con-sO tent with the initial con fition asxmptions of the safety analyses and are l based upon maintaining at least one redundant set of 'orsite A.C. ar.d 9.0. oower 1 sour p and asLoci&ted distribution systems OPERABLE during accident :onditicns ninchlent with an assumed loss-of-offsite pcser and single failure if the i otmW osite A.C. source. 7he A.C. and D.C. source allowable out-of-service i ti.f.es are based of Regulatopy Guide le 93, " Availability of 21ectrical Fower
Scurces," Decekeer 1974 and4eneric Letter 84-15, " Proposed Staff Poshion to j
Improve and Maintain Diesel Generator Reliability." When one dies.el pentrator
; is inoperable, there is an additional ACTION requiremett to verify that all ,
! required systems, subsystems, trains, components and devices, that d@end on i the remaining OPERABLE diesel generator 'as a source of emergen.cy power, are $ also OPERABLE, and that tne steam-driven auxiliary feedwater pump.is OFEBABLE. f i This requirement is intended to provide assurance that a loss-of-cifsite power l event will not retSt in a complete loss of safety function of-critical systems , during the period cce Of the diesel generators is inoperable. The ters, . verify, as used in this context means to administratively check by examining
logs or other information to determine if certain components are out-of-service
! for maintenance or other reasons. It does not mean to perform the Surveillance , j Requirements needed t!s demonstrate the OPERABILITT of the component. ? > l The OPERABILITY of the minimum specified A.C. and 0.C. power sources and i i associated distribution systems during shutdom and refueling ensures that: l (1) the facility ccn be maintained in the shutcown or refueling condition for : l extended time pteriodt, and (2) sufficient instrucientation and control capa-l bility is available for monitoring and maintaining the unit status. , l ba s sd. i
; phe Surveillance Requirement die el geneators are S = r t ,jt for "th demonstrating the recommendationsthe OPERABILITY of Regulatoryof the l Guides 1.9, "Selectfen of Diesel Generator Set Capacity for Sta6dby Pcwer ,
i Supplies," Msrch 10,1971; 1.10B, " Periodic Testing of Diesel boerator Units l Uged as Onsite Electric Power Systems at Nuclear Power Plants," Revision 1, l August 1977; and 1.137, " Fuel-Dil Systems fer Standby Diesel Generatcrs," Revision 1, October 1979, Generic Letter 84-15, and Generic Letter 83-25,
" Clarification of Surveillance Requirements for. Diesel Fuel Impurity Level Tests." .. y "
Appo,du A fe ; i V0GTLE - UNIT 1 B 3/4 8-1 [ t . f i .. . .. . i
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ELECT 81 CAL POWER SYSTEE alsEs. = _ l A.C. SOURCES, D.C. ,SchRCES, and OySITE POWER DISTRIBUTION (Continu'ed) The Surveillance Requifament for demonstrating the OPERABILITY of the station batteries are based on the recommendations of Regulatory Guide 1.129,
" Maintenance Testing and Replacement of large lead Storage Batteries /9e Nuclear Pcwer Platts," Februnty 1978, and IEEE Std 450-1980, "IEEE Recommended Practice for Maintenance, Testing, and Replacerrent of Large Lead Storage Batteries for Generating Stations and Substationsy%,g sf#.ff fg "gu,,,,,,nded Hon oHW_ nrue da.thrie., for m,,mn htcNce /cr ins Ad/Jun Lhran a,41n.sul/3 Verifying average electrdlyte temperature a battery was sized, total battery terminal voltags on float charge, connection ?*[,% .
resistance values, and the performance of battery service and discharge tests ensures the effectiveness of the charging system, the ability to handle high discharge rates, and compares the battery capacity at that time with the rated capacity. Table 4.0 2 specifies the normal limits for each designatea pilot cell , and each ccnacted ceM for electrolyte level, float voltage, and specific gravi ty. The limits for the designated pilot cells float voltage and specific - gravity, greater than 2.13 riolis and 0.015 below the manufacturer's full charge A specific gravity or a battery charger currest that had stabilized at a low value, is characteristic of a charged cell with adequate capacity. The notual i linits for each connected cell fcr float voltage and specific gravity, greater than 2.13 vous and not cre than 0.020 below the manufacturer's full charge specific ravity with an average specific gravity of all the connecte;f cells not mere than 0.010 celcw the manufacturer's fell charce specific gravity, ensures tne OEERABILITY and capability of the battery. $ - Cpsratfor. with a battery celP s paraneter cutsice the normal limit but witnin the albable value specified in Table 4.9-2 is permitted for op to 7 days, During this 7-day period: (1) the allowabla waites for electrolyte leal esures no pnysical damage ta the plates with an adequate electron ; transfer capability; (2) the . allowable value for tre average specific gravity 2 of ail the cells, tot more tman 0.020 below the manufacturer's recommended full charge specific gravity, ensures that the decreusa in rating will be less than i the safety margin provided ir. siring; (3) tM alhable vaice for .an individua7 1 cell's specific gravity, ensures that an indiskual cell's specific gravity - will not be more than 0.040 telcw tha manufact rer's fcil charge specific ~ grasity and that the overall capabilny of the battery will ha caintained within rn acceptable ' limit; aad (4) the allowable valce fcr an individual call's float voltage, greater than 2-07 voIts., ens eres the 54ttery's capability 4 2 sc ; I to . perform its design function. LO
YOGTLE - WIT L B 3/4 8-2
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ELECTRICAL POWER SYSTEMS J BASEE . __ _ _ _ 3/4.9.4 ELECTRlCAL. EQUJPgNT PROTjg3_0EVICES Contaifuhent slEcti'ical penetrations ehd par,etration conductors are pro-tected by either deenergizing circuite not requiPed dur{dg reactor operatiun or by deinonstrating the'0PERABIt.ITY of pripary and backup overi:urrent 'protec-tion circuit treakers durin0 periodic surveillance. The Survefilance Requirements applicable to lower voltage circuit breakers and-Jusu provide assurance of breaker a&ftise reliability by testing at lesst one representativa sample of each macufacturer's brand of circuit breaker, anc/c- fe::. Each manufacturer's molded case and metal case circuf t breakers und/u fLm are grouped into representative samples which are thien tested on a rotating basis to ensure that all breakers :nd/er fu m are tested. If a wide variety exists within any manufacturer's brand of circuit breakers,aeW+r Mes, it is necessary to divida that manufacturer's breakers ender- fu;;; into jroups and treat each group as a separate type of breaker r fu; : for surveillance purposes. f1ceptdan.ny &%p pdpr cdic khy The C O ACILITi .-] [bypassir o overload protection./c[ ontinuous1ylker]fthemotoroperatedvalvesthermal [during :: ident tenditi n:] [by imw o , v.y g . - ice.] ensures that the thermal overload protection [during O V a :id:nt ::nditi:0:} will not prevent safety related valves from performing their fuc-tfon. fTgeSurveillanceRequirementsfordemonstratir.gthe _j w mou - i ; &2 t bypassing 3[of the thermal overload protection fcontinuously7 [and] [:r] Eduring :::ident tenditi:n:] are in accordance with Regulatory Gdide 1.106, " Thermal Ovt'rinad Protection for Elettriq Hotors on Dr.or Operated Valves," Revidon 1, MafG 1377. }' l 1 i l O l O!iTLE - @IT 1 8 3/4 8-3 i t *
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7Ae s'o ck, obr c/cse.L o ' ?he repaired v,tlye.t da. $in/ rc d ionfa naf/ialcheopterfron,% r'< //ed. pre e AJrs the sossiai/r qs ef aneo.dro/kd. borea,. S pe < tissas e / Me. fer e-fer ee,/a - ? S AC S e l ra> } s ra .hd wa.fe r by c
;ng h>ys Reny&hs fera . 7Ais from wJio<,
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6 ASIS .- - __ _ _ 3/4.9, Q Q CONCENTRATI0t The limitaticns on reactivity conditions during REfuELIf;G ensure that: (1) the raattor will remain subcritieral during CORE ALTERATI'JNS, ant) (2) a I f uniton boron concentration is c:aintained for reactivity control in the water vcTume haling direct access to the reactor vessel. W :: ' m*4eM-res-see eem4 tea *-*4 tr. th; in
Galwedi tiins =;;3a.M for- tc;9erse d+ ken a ..4- d;;f:t{
7 [ f r *.N: ::'e W 4 m . The value of 0.G5 or less for Kgff includes a 1% Mk ecnservative allowence for uncertainties. Hailarly, the cocon concentration vaice af (2000] ppm or greater includes c conservative unc.ertainty allowance of 50 ppm baron. The lockin0 closed of the required valves during refu9 ling operations creclude the cossibility of uncontrolled boron dilution of the filled portion at the RCS. This action prevents flow to thg P.C5 of unborated water by closing flew paths trem sourcet of unborated water. I 3/4.9.2 INSTRUMENTATION The OPERABILITY of the Source Range Neutron Flux RonP. ors arsures that redundant monitoring capability is available to detect changes in the i reactivity condition of the core. 3/4.9.3 DECAY TIME The minimen requirement for reactor subcriticality prior to movement of
; f rradiated fuel assemblies in the reactor vessel ensures that sufficient time has elop' sed to alicw the radioactive decay of the short-lived fission products. This decay time is consistent with the assumptions used in the Safaty anal /Ses.
3/4.9 4 7CONTA?NMENT _ BUILDING _ PENETRATIONS The raquirements on containient buildfag penetration closure and OPERABILITY , ancure that a releace cf radicactive material within containment will be restricted from leakage to tPc environment. The OPERABILITY and cicsure restrictions are sufficient to restrict radioactive material release fror, a fuel element ,upture based uoan the lack of contairwnt pressurization putential while in the REFUELIfiG MGDE. i 3/4.9.5 CC'/MUNICATIONS The requirement for corm:unicaticns ca;:atility ensures that refueling 1 station carsennel can ts promptly informed of significant changes in the I f acility status or core reactivity conditions during CORE ALTERATIONS, !O j V0GTLE - UNIT 1 . B 3/4 3-1 1
h 1 REFUELING OPERATIONS BASES _ REFG f*L /M3- AM c n /NS de4 clashr Gntrol A ssea,h/resourA) 3l%.9l M ' . TheOPiRASI).ITYrequirement3fortheUrMile :IcVensurethat:
/r /afh,,9 g m:dpub"* MM will be used for TCvement of dr%e re& and fuel assem-he e. t1MSaeoceane- has sufficient load capacity to lift a dr4 rc6or fuel assembly, and (3) the core internals and reactor vessel are protec ted from i excessive lifting foreo in the event they are inadvertently engaged '*
dering lifting operations. Rccd 3/4._9.7 _ CRANE TRAVEL - SP(NT FIE STORA_GE AREAS The restriction on movetaent of loads in acess of the nomind weight of & fuel and control red 9ss6mbly and associated handling tcol Over 'other fuel assevblies in the storage pool enWres that in the nent this had is dropped: (1) che activity release will be liciited to tnat ccatained in a single fuel assemely, and (2) any possible distortion of fue'l in the storage racks will not result in a critical array. This assumption is consistent with the activity - release assuaed ic the safety analyses. i 7/4.9.8 RESIOUAL HEAT REMOVAL AND COOL ANT CIRCULATION en. , n O' Tne requirestant that at least one residual heat removal (%HR).l w be in operation ensures that: (1) sufficient cecling capacity is availabla to remove decay heat and maintain the water in the reactor vessel below 14'fF as required curing the REFUELING f40DS, ano (2) sufficien,t coolant circulatf or. is maintained through tne c. ore to minimize the effect of a bcron dilutic,n incHent and prevent bo on stratification. train twins
- Tne requirement tolhave two RHR 4eoes GFERABL,E vhen there is less than 23 feet of watgr acevte/the reactor vessel flange ensures tr,at a single failure of the operating RHR 44ep will not result 10 a torptete loss of residual daat a
remaval dapability. With the reactor vessel head removed and at least 23 feet of water above the reactor pressure vessel flarge, a large heat sick is avail-able for cpee cooling. Thus, in th'e event of a failure of the operating RhR 24q, adequate time is provided to initi&te eeerger.cy procedures to cool . the core.strz;n
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3/4.9.9 CONTAINMENT MWe-AND-E#%%uirT ISOLATION SYSTEM The CPERABILITY of this system ensures that the containment vent and purge p0netratioc will be automati'cally f sciated upen detection of high radiation levels within the contairment. The QFERABILITY of this system is l reouired to restrict the release of radioactive material from the containment atmosphere to the environment. V0GTLE - UNIT I B 3/4 9-2 : ' l 1 t --- -
^ REFUELING 09ESA_TIONS BASES 3 /4. 9.10 aed __3 /4. 9.11 WATE*4 LEVil'- P.EACTCR VESSEL and STORAGE POOL The restrictions on minirnum water level ensure that sufficiedt water depth is available to remove 99% of the assumed 10% iodine gap activity released from the ructure of an irradiated fuel assembly. The minimum water depth is consistent with the assuntions of the :afety analysis.
3/2.0.12 STORAGE-FOGL-VW4L-AEON SYNEM i The limitat4 ens-on-the-5tocoge-F-ool-Venti!: tion Sy:te : cacure th:t 2!' 1:dicactive-eatef4aMclea';cd.frx O&-irradistr:d fuel -istembiy -will be filts.i ed thmugh the WEFA--f41ters-and-charcea! ad:crber pri;r t;-discharge to the-i atrrphere,-Oper4tice c' the speter-vitN44e -beaters -cperating-f4r at least , 10-ccr. tint.as-tours in-e-al-dcy-peded i; :Wisent-to-reduce the buildup of actstwe-en-the-edscrber.; and4iE;Ve-H4tsr . The{sPERA&! LIT'l of-this systes ae64ht-resulting-iodine-reenoval-capacity,we-c4Atr4+ teat-with the ass 4ept4 cat of-.the-svety-enalyses. ANSH610-1975Hn441-be-used= s a pr'adur brpie fw-surve4-14 nce testtag, l l l 1 i 1 l i O ) B 3/4 9-3 l V0GTLE - UNIT I l l l
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4 3/4.10 sPECIAL TEST CXCEPTIONS BASES __ 3/4.10.1 SHUTOOWN MARGIN . This special test exception provides that a minimum amount of control red worth is imrediately available for reactivity control when tests are performed for control rad vorth measurement. This special test exception is require d to parmit the periodic verification of the actual versus predicted core reactivity condition occurring as a result of fuel burnup or fuel cycling operations. 3/4.10.2 GROUP HEIGHT, INSERTION, AND POWER DISTRIBUTION LIMITS This special test exception permits individual control rods to be positioned outside of their normal group heights and insertion limits during the performance of such PHYSICS TESTS as those required to: (1) measure control rod worth, and (2) determine the reactor stability index and damping factor under xenon
, oscillation ccnditio^s.
3/4.10.3 PHYSICS TESTS Tnis special test exception permits PHYSICS TESTS to be perfomed at less than or equal to 5% of RATED THERMAL POWER with the RCS T avg sH gh0y lower than normalif allowed so that the 70ndamental nuclear characteristics of the OA core and relatec in$trumentation can be verified, In Drder for various chart.c-
> teristics to be accurately measured, it 16 at times necessary to cperate outside the norm:1 restrictions of these Technical Specifications. For instar,ce, to measure one moderator temperature coeffitfent at BOL, it is necessary to position the various control rods at heiDhts which may not normally be allowed by Specification 3.1.3.6,eich in turn =y :uw the RCS T"V9 Ih fall siichtly below the minlmum temperature of Specification 3.1.1.4 'LAL '"")'
3/4.10.4 REACTOR C00_LANT LOOPS Dis special test exception permits reactor criticality uncer no flow conditions and is required to perfom certain STARTUP and PHYSICS TESTS while at icw THERMAL. POWER levels. 3/4.10.5 POSITION INDICATION SYSTEM - SHUT 00W This special test exception permits the Position Indication Systems to be inoperable during rod drop time measurements, The exception is required since ' the data necessary to determine the rod drop time Are derived from the induced voltage in the position indicator coils as the rod is dropped. This induced voitage is small compared to the normal voltage and, therefors, cannot be observed if the Position Indication Systems remain OPERABLE. .' .i i V0GTt.E - UNIT 1 8 3/4 10-1
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1 3/4.11 RADI0 ACTIVE EFFLUENTS BASES _ 3/4.11.1 LIQUID EFFLUENTS 3/4.11.1.1 CONCENTRATION This specification is provided to ensure that the concentration of radio-active materials released in liquid waste effluents to UNRESTRICTED AREAS will be less than the concentration levels specified in 10 CFR Part 20, Appendix 8, Table II, Column 2. This limitation provides additional assurance that the levels of radioactive materials in bodies of water in UNRESTRICTED AREAS will result in exposures within: (1) the 3ection II.A design objectives of Appendix I,10 CFR Part 50, to a MEMBER OF THE PUBLIC, and (2) the limits of 10 CFR Part 20.106(e) to the population. The concentration limit for dissolved or entrained noble gases is based upon the assumption that Xe-135 is the controlling radioisotope and its MPC in air (submersion) was converted to an equiv? lent concentration in water using the methods described in Inter-national Commission on Radiological Protection (ICRP) Publication 2. This specification applies to the release of radioactive materials in liquid effluents from all units at the site. The required detection capabilities for radioactive materials in liquid waste samples are tabulated in terms of the lower limits of detection (LL0s). Detailed discussion of the LLD, and other detection limits can be found in Currie, L.A., " Lower Limit of Detection: Definition and Elaboration of a Os Proposed Position fer Radiological Effluent and Environmental Measurements," NUREG/"-4007 (September 1984), and in t.he HASL Procedures Manual, HASL-300 (revised annually). 3/4.11.1.2 DOSE This specification is provided to implement the requirements of Sections II.A, III.A, and IV.A of Appendix I, 10 CFR Part 50. The Limiting Condition for Operation implements the guides set forth in Section II. A of Appendix I. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to as'sure that the releases of radioactive material in liquid effluents to UNRESTRICTED AREAS will be kept "as low as is reasonably achievable." Also, for fresh water sites with drinking water supplies that can be potentially a'fected by plant operations, there is reasonable assurance that the operation of the facility will not result in radionuclide concentrations in the finished drinking water that are in excess of the requirements of 40 CFR Part 141. The dose calcula-tion methodology and parameters in the ODCM implement the requirements in Sec-tion III. A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data, such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated. The equations specified in the 00CM for calculating the doses due to the actual release rates of radioactive materials in liquid effluents are consistent with the methodology provided in Regulatory Guide 1.109, " Calculation of Annual Doses to Man from Routine Releases of V0GTLE - UNIT 1 B 3/4 11-1
,.l 1
l l O b RADI0 ACTIVE EFFLUENTS BASES DOSE (Continued) Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," Revision 1, October 1977 and Regulatory Guide 1.113, " Estimating Aquatic Dispersion of Effluents from Accidental and Routine Reactor Releases for the Purpose of Implementing Appendix I," April 1977. This specification applies to the release of radioactive materials in liquid effluents from each unit at the site. For units with shared Radwaste Treatment Systems, the liquid effluents from the shared system are to be proportioned among the units sharing that system. 3/4.11.1.3 LIQUID RADWASTE TREATMENT SYSTEM The OPERABILITY of the Liquid Radwaste Treatment System ensures that this system will be available for use whenever liquid effluents require treatment prior to release to the environment. The requirement that the appropriate portions of this system be used when specified provides assurance that the releases of radioactive materials in liquid effluents will be kept "as low as is reasonably achievable." This specification implements the requirements of 10 CFR 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50 and the design objective given in Section II.0 of Appendix I to 10 CFR Part 50. The specified limits governing the use of appropriate portions of the Liquid Radwaste Treatment System were specified as a suitable fraction of the dose design objectives set forth in Section II.A of Appendix I, 10 CFR Part 50 for liquid effluents. This specification applies to the release of radioactive materials in liquid effluents from each unit at the site. For units with shared Radwaste Treatment Systems, the liquid eff.?% nts from the shared system are to be pro-portioned among the units sharing that system. 3/4.11.1.4 LICUID HOLDUP TANKS The tanks listed in this specification include all those outdoor radwaste tanks that are not surrounded by liners, dikes, or walls capable of holding the taak contents and that do not have tank overflows and surrounding area drains connected to the Liquid Radwaste Treatment System. Restricting the quantity of radioactive material contained in the specified tanks provides assurance that in the event of an uncontrolled release of the tank's contents, the resulting concentrations would be less than the limits of 10 CFR Part 20, Appendix B, Table II, Column 2, at the nearest potable water supply and the nearest surface water supply in an UNRESTRICTED AREA. l u ; V0GTLE - UNIT 1 8 3/4 11-2 l _. _ _ _ . _ . _ _ . . _ _ _ . __ _ ._._ . _ , ~ ~
4
' RADIOACTIVE EFFLUENTS BASES 3/4.11.2 GASEOUS EFFLUENTS .
i 3/4.11.2.1 DOSE RATE l This specification is provided to ensure that the dose at any time at and. I 4 beyond the SITE B0UNDARY from gaseous effluents from all units on the site will be within the annual dose limits of 10 CFR Part 20 to UNRESTRICTED AREAS. -; The annual. dose limits are the doses associated with the concentrations of 10 CFR Part 20, Appendix B, Table II, Column I. These limits provide reasonable assurance that radioactive material discharged in gaseous. effluents will not : result in the exposure of a MEMBER OF THE PUBLIC in an UNRESTRICTED AREA, either within or outside the SITE BOUNDARY, to annual average concentrations exceeding the limits specified in Appendix B, Table II of 10 CFR Part 20
(10 CFR Part 20.106(b)). For MEMBERS OF THE PUBLIC who may at times be within
+ the SITE BOUNDARY, the occupancy of that MEMBER OF THE PUBLIC will usually be 4 sufficiently low to compensate for any increase in the atmospheric diffusion factor above that for the SITE BOUNDARY. Examples'of calculations for such MEMBERS OF THE PUBLIC, with the appropriate occupancy factors, shall be given in the 00CM. The specified release rate limits restrict, at all times, the corresponding gamma and beta dose rates above background to a MEMBER OF THE j PUBLIC at~or beyond the SITE BOUNDARY to less than or equal to 500 mrems/ year to the whole body or to less than or equal to 3000 arems/ year to the skin. These release rate limits also restrict, at all times, the corresporiding
thyroid dose rate above background to a child via the inhalation pathway to l
'less than or equal to 1500 mrems/ year.
i ! This specification applies to the release of radioactive materials in gaseous effluents from all units at the site. The required detection capabilities for radioactive material in gaseous l- waste samples are tabulated in terms of the lower limits of detection (LLDs). ~ "-- I Detailed discussion of the LLD, and other detection limits can be found in ! Currie, L.A., " Lower Limit of Detection: Definition and Elaboration of a Pro-
posed Position for Radiological ' Effluent and Environmental Measurements,"

NUREG/CR-4007 (September 1984), and in the HASL Procedures Manual, HASL-300 (revised annually).
, 3/4.11.2.2 DOSE - N0BLE GASES This specification is provided to implement the. requirements of Sections
II.8, III. A and IV. A of Appendix 'I,10 CFR Part 50. The Limiting Condition i for Operation implements the guides set forth in Section I.B of Appendix I. -
J The ACTION statements provide the required operating flexibility and at the same time implement- the guides set forth in Section IV. A of Appendix I to assure that the releases of radioactive material in gaseous effluents to UNRESTRICTED AREAS will be kept "as low as is reasonably achievable." The Surveillance' Requirements implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calcula- ! tional procedures based on models and data such that the actual exposure of a i MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated. The dose calculation methodology and parameters established V0GTLE - UNIT 1 B 3/4 11-3
. . - . - - ... - = - - . - - - . . _ _ _ _ , . _ _ . , - _
1 i I RADI0 ACTIVE EFFLUENTS BASES DOSE-NOBLE GASES (Continued) in the 00CM for calculating the doses due to the actual release rates of radioactive-noble gases in gaseous effluents are consistent with the methodology
; provided in Regulatory Guide'1.109, " Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance witn 10 CFR Part 50, Appendix I, " Revision I, October 1977 and Regulatory Guide 1.111, " Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water Cooled Reactors,"
Revision 1, July 1977. The ODCM equations provided for determining the air < doses at and beyond the SITE BOUNDARY are based upon'the historical average atmospheric conditions. This specification applies to the release of radioactive materials in i gaseous effluents from each unit at the site. For units with shared radwaste i treatment systems, the gaseous effluents from the shared system are proportioned among the units sharing that system. l 3/4.11.2.3 DOSE - 10 DINE-131, IODINE-133, TRITIUM, AND RADIOACTIVE MATERIAL IN PARTICULATE FORM This specification is provided to implement the requirements of Sections II.C, III.A and IV.A of Appendix I, 10 CFR Part 50. The Limiting Conditions ! for Operation are the guides set forth in Section ~II.C of Appendix I. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive materials in gaseous effluents to UNRESTRICTED AREAS will be kept "as low as is reasonably achievable." -The 00CM calculational methods specified in the Surveillance Requirements implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I 4' be shown by calculational procedures based on models and data such that the
actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated. The ODCM calculational methodology and parameters for calculating the doses due to the actual release rates of the 4
subject materials are consistent with the methodology provided in Regulatory Guide 1.109, " Calculation of Annual Doses to Man from Routine Releases of j Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, i Appendix I," Revision 1, October 1977 and Regulatory Guide 1.111, " Methods for j Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors," Revision 1, July 1977. These equa-tions also provide for determining the actual doses based upon the historical average atmospheric conditions. The release rate specifications for Iodine-131 Iodine-133, tritium, and radionuclides -in particulate form with half-lives greater than 8 days are dependent upon the existing radionuclide pathways to man
' in the areas at and beyond the SITE BOUNDARY. The pathways that were examined in the development of the calculations were: (1). individual inhalation of air-borne radionuclides, (2) deposition of radionuclides onto green leafy vegetation i with subsequent consumption by man, (3) deposition onto grassy areas where milk
animals and meat producing animals . graze with consumption of the milk and meat I by man, and (4) deposition on the ground with subsequent exposure of man.
t O V0GTLE - UNIT 1 8 3/4 11-4 I
RADI0 ACTIVE EFFLUENTS BASES DOSE - IODINE-131, IODINE-133, TRITIUM, AND RADI0 ACTIVE MATERIAL IN PARTICULATE FORM (Continued) This specification applies to the release of radioactive materials in gaseous effluents from each unit at the site. For units with shared radwaste treatment systems, the gaseous effluents from the shared system are proportioned among the units sharing that system. 3/4.11.2.4 GASEOUS RADWASTE TREATMENT SYSTEM
&sssoas wasrs Pacc Gsmcr The OPERABILITY of the 'JAST CAS "0 LOU.' SYSTEM and the VENTILATION EXHAUST TREATMENT SYSTEM ensures that the systems will be available for use whenever gaseous effluents require treatment prior to release to the environment. The requirement that the appropriate portions of these systems be used, when specified, provides reasonable assurance that the releases of radioactive materials in gaseous effluents will be kept "as low as is reasonably achievable." This specification implements the requirements of 10 CFR 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50 and the design objectives given in Section II.D of Appendix I to 10 CFR Part 50. The specified limits governing the use of appropriate portions of the systems were specified as a suitable fraction of the dose design objectives set forth in Sections II.8 and II.C of Appendix I,10 CFR Part 50, for gaseous effluents.
This specification applies to the release of radioactive materials in O gaseous effluents from each unit at the site. For units with shared radwaste treatment systems, the gaseous effluents from the shared system are proportioned among the units sharing that system. 1 3/4.11.2.5 EXPLOSIVE GAS MIXTURE crisseous wnsre prLocsssove Thisspecificationisproviaedtoensurethatthe/concentrationofpoten-tially explosive gas mixtures contained in the 4'ASTE GAS HCLDUF SYSTEM is maintained below the flammability limits of hydrogen and oxygen. [ Automatic control features are included in the system to prevent the hydrogen and oxygen concentrations from reaching these flammability limits. These automatic control features include isolation of the source of hydrogen and/or oxygen, automatic diversion to recombiners, or injection of dilutants to reduce the concentration below the flammability limits.] Maintaining the concentration of hydrogen and oxygen below their flammability limits provides assurance that the releases of radioactive materials will be controlled in conformance with the requirements of General Design Criterion 60 of Appendix A to 10 CFR Part 50. DscM 3/4 11.2.6 GAS STORAGE TANKS The tanks included in this specification are those tanks for which the quantity of radioactivity contained is not limited directly or indirectly by another Technical Specification. Restricting the quantity of radioactivity contained in each gas :te :g tank provides assurance that in the event of an uncontrolled release of the tank's contents, the resulting whole body exposure to a MEMBER OF THE PUBLIC at the nearest SITE BOUNDARY will not exceed 0.5 rem. Iecay P V0GTLE - UNIT 1 B 3/4 11-5 I
RADIOACTIVE EFFLUENTS
, BASES This is consistent with Standard Review Plan 11.3, Branch Technical Position ETSB 11-5, " Postulated Radioactive Releases Due to a Waste Gas System Leak or i Failure," in NUREG-0800, July 1981.
3/4.11.3 SOLID RADI0 ACTIVE WASTES This specification implements the requirements of 10 CFR 50.36a and General Design Criterion 60 of Appendix A to 10 CFR Part 50. The process parameters included in establishing the PROCESS CONTROL PROGRAM may include, but are not limited to, waste type, waste pP, waste / liquid /SQLIDIFICATION agent / catalyst ratios, waste oil content, waste principal chemical constituents, 1 and mixing and curing times. 3/4.11.4 TOTAL DOSE f <thai b recMy .sapport f h e- prehe.fion of
/e c f, , c 2/ f 4aer 4r gabcc. u se, This specification is provided te meet the dose limitations of 10 CFR Part 190 that have been incorporated into 10 CFR Part 20 by 46 FR 18525. The specification requires the preparation and submittal of a Special Report when-ever the calculated doses ue to releases of radioactivity and to radiation from uranium fuel cycle sources exceed 25 mrems to the whole body or any organ, except j the thyroid, which shall be limited to less than or equal to 75 mrems. For sites containing up to four reactors, it is highly unlikely that the resultant dose to a MEMBER OF THE PUBLIC will exceed the dose limits of 40 CFR Part 190 if the individual reactors remain within twice the dose sign objectives of O Appendix I, and if direct radiation doses from the units neluding outside storage tanks, etc.) are kept small. The Special Report will describe a course of action that should result in the limitation of the annual dose to a MEMBER OF THE PUBLIC to within the 40 CFR Part 190 limits. For the purposes of the
! Special Report, it may be assumed that the dose commitment to the MEMBER of
the PUBLIC from other uranium fuel cycle sources is~ negligible, with the i exception that dose contriDutions from other nuclear fuel cycle facilities at the same site or within a radius of 8 km must be considered. If the dose to any MEMBER OF THE PUBLIC is estimated to exceed the requirements of 40 CFR Part 190, the Special Report with a request for a variance (provided the

release conditions resulting in violation of 40 CFR Part 190 have not already l been corrected), in accordance with the provisions of 40 CFR 190.11 and 10 CFR j 20.405c, is considered to be a timely request and fulfills the requirements of i 40 CFR Part 190 until NRC staff action is completed. The variance only relates
! to the limits of 40 CFR Part 190, and does not apply in any way to the other j requirements for dose limitation of 10 CFR Part 20, as addressed in Specifi-cations 3.11.1.1 and 3.11.2.1. An individual is not considered a MEMBER OF THE l PUBLIC during any period in which he/she is engaged in carrying out any operation that is part of the nuclear fuel cycle. i O V0GTLE - UNIT 1 8 3/4 11-6
A w.- d t I 3/4.12 RATI0 LOGICAL ENVIRONMENTAL MONITORING .wdb.e w 4 e-pneed o s s. BASES 3/4.12.1 MONITORING PROGRAM . myswnf , l The Radiological Environmental Monitoring f i specificationprovidesrepresentativemeasureme/rogramrequiredbythis nts of radiation and Of radio-l active materials in those exposure pathways an'd for those radionuclices that I' lead to the highest potential radiation expos 6re of MEMBERS OF THE PL6LIC resulting from the plant operation. This moniitoring program implements 4 Section IV.8.2 of Appendix I to 10 CFR Partj50 and thereby supplements the Radiological Effluent Monitoring Program byncrifying th:t the ererM1e 7
concentrations of radioactive materials and levels ofr radiatio 4 teae expected on the basis of the effluent measurements and the.modeling of the environmental exposure pathways. Guidance for this monitoring program is provided by the Radiological Assessment __ Branch Technical Position on Environ- "
mental Monitoring,g 1, November 1979) Th; inittelly sp;;ified seniter - wJ-progra e!" be e. . ... . v: Mr :t 1 ::t the fir:t 3 y::r: Of : - :rci:1 operation.-Fo13owing-tM: peried, pregr:: ch:ng:: ::y b; i;iti;ted be;;d sa 4 cp:r tional-experiense, 4 The required detection capabilities for environmental sample analyses are tabulated in terms of the lower limits of detection (LLDs). The LLDs required 1- by Table 4.12-1 are considered optimum for routine environmental measurements J in industrial laboratories. It should be recognized that the LLO is defined
.as an a priori (before the fact) limit representing the capability of a measure-ment system and not as an a posteriori (after the fact) limit for a particular measurement. _ _ _ _ ,
l Detailed discussion of the LLD, and other detection limits, can be found ) ' /in Currie, L. A. , " Lower Limit of Detection: Definition and Elaboration of a \ Proposed Position for Radiological Effluent and Environmental Measurements," \ NUREG/CR-4007 (September 1984), and in the HASL Procedures Manual, HASL-300 ' I (revised annually). - - N um %- , 3/4.12.2 LAND USE CEN = su R V W
,- m ay f This specificationiis provided to ensure that changes in the use of areas
) at and beyond the SITE BOUNDARY are identified and that modifications to the , l Radiological Environmental Monitoring Program are made if required by the results of this' census, iThe b::t infer : tion 're: the deer-te-deer survey, fre rial = survey er from-consult 4ng-with 10c:1 :gricultur ! zutherities sham-be-used. This sens4s ' satisfies the requirements of Section IV.B.3 of s""2 Appendix I to 10 CFR Part 50. Restricting the sendes to gardens of greater ' tnan 50 m2 provides assurance that significant exposure pathways via leafy vegetables will be identified and monitored since a garden of this size is the minimum required to produce the quantity (26 kg/ year) of leafy vegetables assumed in Regulatory Guide 1.109 for consumption by a child. To determine l this minimuu garden size, the following assumptions were made: (1) 20% of the - garden was used for growing broad leaf vegetation (i.e., similar to lettuce and cabbage), and (2) a vegetation yield of 2 kg/m2 , l O 1 V0GTLE - UNIT'1 B 3/4 12-4
. .g_
RADIOLOGICAL ENVIRONMENTAL MONITORING BASES 3/4.12.3 INTERLABORATORY COMPARISON PROGRAM The requirement for participation in an approved Interlaboratory Comparison Program is provided 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 in order to demonstrate that the results are valid for the purposes of Section IV.B.2 of Appendix I to 10 CFR Part 50. O lO V0GTLE - UNIT. 1 8 3/4 12-2 l
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4 Y i 4 1 l (
- JUSTIFICATIONS FOR DEVIATIONS FROM STS I SECTIONS 3.0 AND 4.0 BASES l f i General:
l f These bases have been revised to be consistent with the Technical . Specifications as marked in this draft. In addition, certain revisions have ; been made for the purposes of amplification and clarification. 4 i 1 1 1 i-4 i } i i l l l I i l \ I ! l r I 4 9 h 0092v l l . _. _ _ _ _ _ . _ . _ _ _ _ _ _ _ _ _ . . . . - . - - - - - .
l I i 4 1 i i l l
> l j SECTION 5.0 l DESIGN FEATURES i 1
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S.0 DESIGN FEATURES 5.1 SITE EXCLUSION AREA 5.1.1 The Exclusion Area shall be as shown in Figure 5.1-1[ LOW POPULATION ZONE 5.1. 2 The Low Population Zone shall be as shown in Figure 5.1-2[ MAP DEFINING UNRESTRICTED AREAS AND SITE BOUNDARY FOR RADI0 ACTIVE GASEOUS AND LIOUID EFFLUENTS 5.1. 3 Information regarding radioactive gaseous and liquid effluents, which will allow identification of structures and release coints as well as defini-tion of UNRESTRICTED AREAS within the SITE BOUNDARY that MEMBERS OF THE PUBLIC, shall be as shown in Figure f5.1- [are accessible to
.L The definition of UNRESTRICTED AREA used in implementing these Technical Speci-fications has been expanded over that in 10 CFR 20.3(a)(17). The UNRESTRICTED AREA boundary may coincide with the Exclusion (fenced) Area boundary, as defined in 10 CFR 100.3(a), but the UNRESTRICTED AREA does not include areas over water bodies. The concept of UNRESTRICTED AREAS, established at or beyond the SITE SOUNDARY, is utilized in the Limiting Conditions for Operation to keep levels of radioactive materials in liquid and gaseous effluents as low as is reason-O ably achievable, pursuant to 10 CFR 50.36a. ;rj e 5,ye 4,,,,y,7/,
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5.2.1 The containment building is a steel-lined, reinforced concrete building of cylindrical shape, with a dome roof and having the following design features:
a. Ncminal inside diameter = /4c feet.
.b. Nominal inside height = 2 24. feet,
c. Minimum thickness of concrete walls = $ feet,9 inc Ae s.

d. Minimum thickness of concrete roof = _y_ feet, e inchd5-b.t s eir,a ,

e. Minimum thickness of concrete fhcr pad--= k feet,f ta /es-
, ,w,,inw , rhiexness ei concrete i,rs r< w e n t e . ,, c a ;iy , ~e7,a r' 2 ,f. Nominal thickness of steel liner = A> inches. /,. g. Net free volume = m odcubic feet.
DESIGN PRESSURE AND TEMPERATURE 5.2.2 The containment building is designed and shall be maintained for a maximum internal pressure of f.g_ psig and a temperature of4ec*F. O V0GTLE - UNIT 1 5-1
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- This figure shall consist of a map of the j \ '- site area showing the Low Population Zone boundary. Features such as towns, roads, s' industrial areas and recreational areas shall ./
1 be indicated in sufficient detail to allow- /
! Nidentification of significant shifts in j population distribution within the LPZ. 'N i \ ! \\
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[ Qi[A / NORTH 0 % 1 2 5 MILE RADIUS l I I I I i O VOGTLE a n t- s men n Georgia Powerku ELECTRIC GENER ATING PLANT unit uuo unit 2 uw ap"a"u/nca ws FIGURE 2.1.1- 3 5. /-2
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\ / \. This figure shall consist of a map of the site areaf 'g showing the SITE BOUNDARY and locating points wittif'n , N the SITE BOUNDARY where radioactive gaseous and }iquid N effluents are released, as well as where radioactive ' liquid effluents leave the site. If onsite areas sub-ject to radioactive materials in gaseous or liquid eff.luents are utilized by the public for recreational or other purposes, these areas shall be outlined on the map and identified by occupancy factors.and the licen-see's method of occupancy control (if any). The figure shall be4sufficiently detailed to allow identification 4 of structures and release point locations and elevations, as well as ' definition of UNRESTRICTED AREAS within the SITE BOUNDARY-that are accessible'by MEMBERS OF THE PUBLIC. The m'ap scale shall be'on the order of 2-3"/ mile.
SeeNUREG-0133foradd.tionaVguidance. t
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1 ,' Ns FIGURE 5.1-3 1-7 N( I e UNRESTRICTED APEAS AND SITE BOUNDA M F'OR RADI0 ACTIVE GASEOUS \
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V0GTLE - UNIT 1 5-4
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DESIGN FEATURES , 1 5.3 REACTOR CORE FUEL' ASSEMBLIES 5.3.1 The core shall contain / /J fuel assemblie6 with each fuel' assembly containing ec,+ fuel rods clad with,TZircaloy-4K Each fuel rod shall have a nominal active fuel length of /4+ inches and contain a maximum total weight of e7n grams uranium. The initial core loading shall have a maximum enrichment of u eight w percent U-235. Reload fuel shall be similar in physical design to tne initial core loading and shall have a maximum enrichment of n weight percent U-235. CONTROL ROD ASSEMBLIES 5.3.2 The core shall contain r5 full-1;ngth control rod assemblies. The A// fuli--lengt.h control rod assemblies shall contain a nominal 142 inches of absorber material. The n;;';;l ;;?;;; :' :t::rt:r ::t:r':? ch:!' 5: Y s m_ v 4.2 4 ' - -" "-eedwen. All control rods shall be clad with stainless steel ttrbW. s h, e. c.e m p e.s die r, 3 h a // ],e_ 9S.S "4 najura./ ha //ro inm a ,,/ t .s "le nn.tura::ireonium. 5.4 REACTOR COOLANT SYSTEM DESIGN PRESSURE AND TEMPERATURE 5.4.1 The Reactor Coolant System is designed and shall be maintained:
a. InaccordancewiththeCoderequirementsspecifiedinSectionTh.2 of the FSAR, with allowance for normal degradation pursuant to the applicable Surveillance Requirements,

b. For a pressure of: n psig, and

c. For a temperature of oso 'F, except for the pressurizer which is c ic *F.
VOLUME 5.4.2 The total water and steam volume of the Reactor Coolant System is c.2+o
+ . .- r i .
cubic feet at a noninal T^*9 of E50id'F. sM l ! 5.5 METEOROLOGICAL TOWER LOCATION 5.5.1 The meteorological tower shall be located as shown on FigureJ5.1-1% O
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V0GTLE - UNIT 1 5-t
DESIGN FEATURES ! i 5.6 FUEL STORAGE CRITICALITY i 5. 6.1.1 The spent. fuel storage racks are designed and shall be maintained I with: j ' a. A k,ff equivalent to less than or equal to 0.95 when flooded with ] unborated water, which includes a conservative allowance of j [2.5]% ik/ for uncertainties as described in Sectionf4.37 of the FSAR, and
/0
ie i b. A nominal E4 6 inch center-to-center distance between fuel j .
assemblies placed in the storage racks. I 5.6.1.2 The k f for new fuel for the firsJ corejloading stored dry in the spentfuelstof$geracksshallnotexceedfD.987whenaqueousfoammoderation j is assumed. DRAINAGE ] ! 5.6.2 The spent fuel storage pool is designed and shall be maintained to l . prevent inadvertent draining of the pool below elevation /r+'-if ". C_APACITY i 5.6.3 The spent fuel' storage pool is designed and shall be maintained with a j storage capacity limited to no more than2ss fuel assemblies. i j 5.7 COMPONENT CYCLIC OR TRANSIENT LIMIT 5.7.1' The components identified in Table 5.7-1 are designed and shall be maintained within the cyclic or transient limits of Table 5.7-1. I 1 1 1 i l !O . 1 s l V0GTLE - UNIT 1 5-fr i i
O O O TABLE 5.7-1 COMPONENT CYCLIC OR TRANSIENT LIMITS , E DESIGN CYCLE h CYCLIC OR TRANSIENT LIMIT OR TRANSIENT
^
C0llPONENT 200 H Reactor Coolant System [25&] heatup cycles at < 100*F/h Heatup cycle - T"V9 from 5 200*F and [250] cooldown cycles at to > $50*F. CooTdown cycle - T from
~< 100 F/h.' .2co N > 550*F to S 200*F 2so
[250] pressurizer cooldown cycles Pressurizer cooldown cycle at < 200 F/h. temperatures from 2 650*F to
< 200 F.
10 E100] loss of load cycles, without > 15% of RATED THERMAL POWER to immediate Turbine or Reactor trip. 0% of RATED THERMAL POWER. 4c-y [50] cycles of loss-of-offsite Loss-of-offsite A.C. electrical - yo A.C. electrical power. ESF Electrical System. yo E100] cycles of loss of flow in one Loss of only one reactor reactor coolant loop. coolant pump. Ann [500] Reactor trip cycles. 100% to 0% of RATED THERMAL POWER.
/ ,103auxiliaryspray Spray water temperature differential actuation cycles. > 320*F.
2co ) Pressurized to > 12485})psig. [5&] leak tests. _ so 3107 [5-] hydrostatic pressure tests. Pressurized to > {3100] psig. Secondary Coolant System Ig/steamlinebreak. Break in a > 6-inch steam line. 10 /4 RI E6-] hydrostatic pressure tests. Pressurized to >'[13503 psig.
i. 1 JUSTIFICATIONS FOR DEVIATIONS FROM STS j SECTION 5.0 l i ' 5.1.3: This specification was revised to reflect the f act that the site bcundary ! j lines. plant property lines, and the exclusion area are all the same. See Figure 1.1-1 of the VEGP FSAR. { l 4 3 5.3.2: This section was revised to reflect plant-specific RCCA design. Referer.ce , I to full-length rods was deleted as redundant since VEGP does not utilize part-length rods. ; 1 i I
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I SECH ON 6.0 ; t i ADMINISTRATIVE CONTR01.5 i s j i 4 1 G , f A I i
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i ACM NISTRATh E CCN7ECLS - Qe - g.1 AESPONSIBILQ g,,r TheGpiut=Str,Nmfat+' em> <ahu nu e r - 4 aHe .We Asa eepe,a 6 ens (&MVMd - 6.1.1 edenti sns11 he responsibleccera- for overall m tien ard shall delega*.e in writing the succession to this responsibility durirg his absence. 6.1.2 The-Shidt-Superdsor-fow-de4NJ44s-ebsexe4rorr 4he <cn4rstvcce,a dasigeated indtAJuaL)-srel1-be-respens4Me-fee-the-eof.t-robM= ::= -d Met 406~-4-menegament-directive-to-thi; ef feet:--srignsd by the--Ehi; hat bwel Weorporate-venagement-bsMbee-refwuad--to-aWstaitfen-ps ._, .. .. a _ mnes' bat h. It _ rA e G,vt me%;// an,,ac di), reissue a. diece fat. Mr er.,ps,astus 6_. 2 0RGeNIZAT10N , N,e pr, ana y ,,way# me,., e re.,7,,,,aa b//dy s/ w e,, air;itspam.han, say srvisee Gr ,ta/,ny 4a af.se,ee rhm % s c,.rrIemr,f.Q M na 0FFSTTE , 9 j,, ;,,,,, !,j g, ,ugu ,,e ,;,, c,,,,,a ,,s ,s.,,a na,,,) /r* r u,re apera,s,, ird 44c >?t n * :gro der .s // er redif ft n 3 en l'!5 thi*Y a r*d ?As N CltArb 6.2.1 The offsite oMahrza'tioti'f6c WiCmahagement' ana ' technical support! estad 5 ' shali de as shown in Figure 6.2-1. A ,, e 7.7 r.,rna,4,,/. L_ a'wh es . n.: s ,-E UNIT ET'M i+t
6. 2. 2 Thee unit r.
organization shall be as shown in Figure 6.2-2 ar.c:
a. Each on-duty shif t shall be cor. posed of at least tu minir,un shif t ;
' crew composition shown in Table 6.2-1; , o e.HAer t Hh e re se. fo r
o. At least one licensed Operator shall be in the control roco when fuel is ir. +-reactor. In addition, while th =Wis in MJDE 1, 4
2, 3, or 4, at least one licansed Senior Operator shall be in the control room: ln w J , . h t/ % // /;* ,
. i A Ji;;1".h P4ys4n4echnic h avka,ement *s ll be on rad;# fden site wh[n fuel is in thereie<++,yvxedar,es reactor; i
d. T,11 CORE ALTERATIO!IS shall be observed and directly supervised by eitrer a licensed Senior Operator or licensed Senior Operator Limited to Fuel Handling who s has no other concurrent respensibilities duric.g this operatian; is in ne o n tain ne nt b ai/ dim' ei Me a // e *. feel re.t e ter a nd .
m.-- ' :!t: cin: Erin: cf :t h::t "e ::d:r:* :bli k r.'ntd nd c;; s soie et 21' t' ::. ?M cire Brigad: :hil n:t !--l d the 2' ft. S e m4sor-and tre [two4-othe- " - d + " w -"- ^ 4 % m necessary #cr safe 9ttt- ,:' 15: c ' :r d :n3 pe r:c ene', req d re d fer :tb:r :::tr.t 4 :' frcticm SrM; 2 '4r: : cr;ercy, W l inJinhas./ sw'a ried. ro impa'ernent radiaNon preferNon r*resda
*The Heekh-Ptyttet-TecHIMegf ar.1 C'*e+';2d e mpetit'e may beelase thaa SM-aMr ::pf:w+e s for r a period of time not to exceed 4 hours, in ordor to accom.nodste unexpected absence, provided immediate action / is t0 ken to fill the reghed positions ( j j
Akstrt? V'JGTLE - LflIT 1 6-1 l
- + - - - - . . - - ,. -.. , .,, _ _ _ _ __ ,_ _ _
- - .- . ~ . ~ - . - , - - - . . _ - * .sw
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i i i ACMt4ISTMTIVE CONTROLS (' . CNs t w in p er /erav e ce. s / I M T STAFF (Continued) p /a r t . i e- +. Acciaistrative precedures shdil be developed and impleeented to limit the working hours of.d ft staff .:h: pc.f:r: safety-related T g,/ geg,ty, gj;F3 'r.:.nctions (e.g. , if censed Wing Senit.r Operators, licensed Operators,
'th phydcists,$ gey no,, operators, and key maintendoce pa 1
fec k ,ci a s' .fices,sel LThe emtaf-ovectime-workepy-w4tataff --t erc ; rb=Sg l safety-rahted-fanctions shhbe-14nited-in--accordance-teRH*e-4FC ' Pol icy-Statement-en-working-WGeneMc-Eettec-h.- SM2 ' .] 1 or
!' p/u t ; Adequate shift coverage shall be maintained withat!t routi::e heaiy '
use of overtime. The objective shall be to have operating ::ersonnel work a arxl 0 ,50er ty, 40-hour week while the udtt is operating.* , t However, in the event that unforaseen problems require substantial t xounts of overtime to be used, or during extended periods of shut , . down for refueling, major maintenance, or major plant modification, oc. a terporary basis the following guidelines shall be followed: t n se we,c w. e 2 my ee,,.si.r* ef12 -soue sJuf+ sel.e</a ks3 e j a 1. An indiridual should not be permitted to work more than 16 hours straight, excluding shift t'arnover time. , i i 2. An individual shculd not be permitted to work Jeore than 16 hours in any 24-hour period, not more than 24 hours in any 48-hour t period, not more than 72 hours in any 7-day period, all excluding j shift turnover time. l
3. A break of at least 8 hours should te allowed between work ;
pericds, including shift tur:over time. , q ! 4. Except during extended shutdown periods, the use of overtisa !
- should be comidered on an individual basis tod not for tho' entire staff on a shift. ,
4 l u, , Any deviation from the above guidelines shall be autnorin.1 by the ! ;f,l,, , ( p.t , a:c ,r? L rQ1;ut %per%%ctent] cr-hh-dep.ty, or higher levels of manage-m,,, , e ,, - n onrl 8'ent, in accordance witn established procedures and with doc wenta- , i tion of the basis fer granting the deviation. Cutrois shall 9e arcess 3 included in the procecures such that individual +oortinr snali te ' ;
re.fewed conthly by the [?' r t S perkteed m Q or his d Wignee to '
i assure t9at excessive hourphere net her* Md;'H. ?^"tfa' 4 t.i;t.'= fix W ee m g i tlir,6; fe not sein.dheG.j ' jewraf sheper* b At Wists *ar Cpr va}sen: , s - i aere aJAeriud. arJ 1%f i i Mey to ed beco.nz roane.. , I I 1 I ' l ! V0GTLE - USIT 1 6-2 ; I
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l l n This ff gure shd1 show the organizational strucNre and lines of responsibility for tha offsite groups that provida technical and xansgeneat suppcet for the unit. The organizational arrangement for O performing a"d maitoring qualf ty assurance activities shall also be indicatec. ftGUAE 6.2-1 0FF3ITE CRGAi41.'A710.i lO I V0GTLE - UXIT 1 6-3 , l l
2 t I ( 1 i l 1 This figure shall show tbo organizational structure and lines of responsibility for the unit staff. Pcsitions to be staffed by - licensed personnel shall be indicated, The organizational arrangr, ment for parforming and monitoring quality assurance activitics shall O also be indicated. t 1 l
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f 4 W 4 FIGURE 6,2-2 i UNIT ORGANIZATIGN O V0GTL2 - LNIT 1 6-4
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_ _ . _ ~ . ._ . _ _ N\ /
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TABLE E 2-1 ( NINIMUM SWIFT CREW CDMPOSITION f
/
! SINGLE UNIT FACILITY
/
POSITION NUMBEROFINDIVIDU,ALSREQUIREDTOFIt.LPOSIJ10N
'N MODE 1, 2, 3, or 4 MODE 5or,f SS 1 1 /
SEO 1 None l'
, R0 \ 2 A0 N 2 /1
' 1* , Nor.e STA \, i g
.55 - Shift'Sapervisor with a Senior Operator license on Vnit 1 SRO - Individual with a Senior Operator license on Unit 1 < 0 - .'ndividual Vth .an Operator license /cn Unit 1 j AC - Auxiliary' Operator /
5TA - Shift 7echnical A@isor j/ The shift crew cocposition m$'y be cae less th'in the minimum requiresents of Table 6.2-1 for a parfod cf tirm not to exceed 2 l'ours in crder to ace.oumedate
' unexpected absence of on-duty shift crsv.menbers provided finnediate acticn is tate, to restore the ch)ft crew ccmosition to within the minimum *.equirerents of Table 6.?-1. This provision doessnot permit any stift crew position to be ~
l l t.reanned upon shif t change due to an . coming shift crewman being late or i aDSefit. i \ Ourina any absence of the Shif t [upervf sokfrom the control roon while tre i, unit is in MCDE 1, 2, 3, or 4, /an individuaK(sther tran the Shift Technical a l Advis ar) with,a walid Senior Operator licens.e\shall ce designated to assnie j the control room caeand function. Durinc xy absence of the f,hift Supervisor i from the control room while the unit.1s in HL.E 5 or 6, an individual with a valid Senior Operator license or Ope ator license} hall be designated to as seme the control rocra ccmnand furc+. ice. x
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*1'l'e STA position shell ce manred in MCDCS 1, 2, 3, and 4 unless tr.e SMft Supervisor or the i.1dividual witA a Senior 09ertter license meets the quiifiestions f:r +.ho STA as recuired by the NRC. 'y \
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voorLE - UNIT 1 S-9 \
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~ 5 e' <' sr'[t/ ? !c 15 ) l^?Aag (, ~< \ j MIMM." Sr!EL4RE! CC"%EI'!W M $ VIII',A CO W { c N I M _ 500$
7
@AIT-!? - - 903 ? S I I VID44l$-EEQOISED IO 'id IC$IUM d:T" 'a'I+fHN - - Bom-dNUS-kN-- -ONE-tm1T-IN-WM-1. 2, &c-*
40"E 2 , 2 , 3 , EDE & cr 5 - ^NO er ' : 4R SEFUEG).-----CNE UN!T " "00E- 5 x 0 a XT'.'CLC
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ST*
- /?^ - _ - - - _none 1N r.aw a rs u - %' Supervisor w nh a Senior Operator ifG:nte SRO - Individual with a Senior Operatur license RO - Individual with an Operator license m- nc - naSVaw) J+v;w ,w-1. ice,,,eJ. cy s..) ,i-STA - Shif t Technical Advisor 4
The shift crew composition riay be w.ts less thanhhe minimum reauirements of Table 5.2-1 for perig of time ont to exceed 4-hours in order to accommodate caexpected 1bsence of en-duty shif t crew members provided irimediate action is ' taken to restore the stif *t crew composition to within the toinimum requirefients of Tas M 6.2*1. This provis:en dces not permit any shift crew pos'. tion to be unmanned y,on shirt char.go due to ar oncoming shift crownan being lace or absent. a e , nu Ouring uy absence of the Mbr HmsSupervisor fecm the controi room while %e r% is ir, MG 1, 2, 3, or 4, an individual f.#tMc--thethe-SM ft-TechMesi
20:9 with a valid Senior Operstcr license shall be desi0ratad to assume the co1 trol com canmerd function. During any absence of the L%f4 Superv'sor from tre central roo8 while tic jnit is in MODE 5 or 6, an individual with a valid Senior Cperator if cense og Qgarator licerse shall be desig1ated to 6
assume the control recm command <fonction. cpun bos 9 gi h r#_4c.'er
At least oFe of the required individuals must be assigned to the designated position for eacn un(1.
" At least one licensed Soniar Oprator or licensed Senior Operatof Limited to Fuel Handling must be prosent during CORE AL'TERATIDNS on either cait, who has no other concurrent responsibilities. *" The STA position shall be manned in M00E51, 2, 3, and 4 unless the Shif t Supervisor or the individual with a senior Operator licenae meets the qualifications for tne STA as required by the NRC, O
V0GTLE - ljNti 1 6-sap i u- -
fns. er to fy w ~ l 6-# - TAB 1,E 6.21 MINjMUM SHlFT CR6W COMPOSITIOtt (qj TWO UNITS WITH A COMMON CONTROL ROOM UNIT 1 IN MODES 1,2,3, OR 4 UNIT 2 IN MODE; POCITION 1,2,'3,CR4 l 5OR5 __ DEPUELED 1 1 OS 1 1 1 SRO 1 3 3 2 RO 3 2 l NLo 3 1 STA* 1 1 UNIT 1 IN MODE 5 OR C UNIT 2 IN MODE-
~'
POSITION 1,2,3 OR 4 5OR8 OEFUELED ( OS 1 1 1 SRO 1 0 0 RO 3 2 1 NLO 3 3 1 STA* 1 0 0 UNIT 1 DEFUELED
, f UNIT 2 IN MODE:
POSITION { i,2,3, OR 4 l 5OR6 OEFUELED OS 1 1 0 SRO 1 0 0 RO 2 1 0 l 0 NLO 2 1 ST A* 0 0 !O
THE ST A is tjoT REQUiHED ON CHlf T IF THE QU ALIFICATIONS OF THE CPERATIONS SUPEAVISOR.ARd UPGRAOED TO 7y gg, fyl. FILL THE HEGU)RETAENTS OF THE STA POSITION.
mn 1.c.UNrr 1 i . . , . _ . . - . . - - . . . . . - .- , - - .
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N\ TABLE 6.2-1b MINIMUM SHIFT CREW COMPOSITION , g
\ TWO UNITS WITH TWO SEPARATE CONTROL ROOMS \ - \ WITH UNIT [2] IN MODE 5 OR 6 OR DEFUELED ' POSITION ' NUMBEROFINDIVIDUALSREQUIREDTOFILLP.k!S? TION N MODE 1, 2, 3, or 4 MODE 5,d'r 6 SS ',
la '1* SR0 N 1 None 2 1 R0 N [' A0 N 2 2** STA
\
g 1*** None N / WITH U$kT [2] IN MODE 1, 2/3, OR 4
\ /
POSITION NUMBE FINDIVID,UILSREQUIREDTOFILLPOSITION MODE 1, 2 4 3, ,d 4 MODE 5 or 6 O SS 1
\
1* None SRO R0 2 \ 1 A0 2 s 1 None STA 1* *** 'N The shift crew compositionday be one less thansthe minimum requirements of
- Table 6.2-1 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 restore the shift crew composition to within the minimum requirements of Table 6.2-1. This' provision does not permit any shift crew position to be unmanned upon shif,t change due to an oncoming shift crewman being late or absent.
During any absenc'e of the Shift Supervisor from the control room while the unit is in MODF1, 2, 3, or 4, an individual (other than the Shift Technical Advisor) with< a valid Senior 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 5 or 6, an individual with a ~ valid Senior Operator license or Operator license shall be designated to assume the control room command function. s
*/ Individual may fill the same position on Unit [2]. ** One of the two required individuals may fill the same position on Unit [2]. / "* The STA position shall be manned in MODES 1, 2, 3, and 4 unless the Shift , Supervisor the individual with a Senior Operator license meets the \
l / qualifications for the STA as required by the NRC. N f
\ /, V0GTLE - UNIT 1 6-5b N N
NN
ADMINISTRATIVE CONTROLS 6.2.3 INDEPENDENT SAFETY ENGINEERING GROUP (ISEG) FUNCTION g/.tn/ plant 6.2.3.1 The ISEG shall function to examine we4 operating charact. eristics, NRC i uances, industry advisories, Licen;;; Event o epe m , and other sources of design and operating experience information, including crit: Of :i-ilar dn ign, which may indicate areas for improving u4t safety. The ISEG shall y make detailed recommendations for revised procedures, equipment modifications',' maintenance activities, operations activities, or other means of improving FM"/--tre4 safety to [: 'igh level cerper:t cf#ici:1 4- : techaic:11y criented cc;ition he i; ,ct i- the sage crt ch:in fer power productionb rAcSede Wec. President- 44:Jea.e dpemNens n.roays ne Manaye,.geejea,. p'e,.6,, nance. COMPOSITION 4,, x 4,, / j, . 6.2.3.2 The ISEG shall be composed of at least five, dedicated, full-time engineers.lec:ted er cite. Each shall have a bachelor's degree in engineering or related science and at least 2 years professional level experience in his field, at least 1 year of which experience shall be in the nuclear field. RESPONSIBILITIES p/an/ operxWons end-iro n in / en a.nc e. 6.2.3.3 The ISEG shall be responsible for maintaining surveillance ofAwe4 activities to provide independent verification
that these activities are performed correctly and that human errors are reduced as much as practical.
RECORDS 6.2.3.4 Records of activities performed by the ISEG shall be prepared, main-O' tained, and forwarded each calendar month to [a high level corporate official in a technically oriented position who is not in the management chain for power production]. 6.2.4 SHIFT TECHNICAL ADVISOR 6.2.4.1 The Shift Technical Advisor shall provide advisory technical support to the Shift Supervisor in the areas of thermal hydraulics, reactor engineering, and plant analysi: with regard to the safe operation of the unit. The Shift Technical Advisor shall have a bachelor's degree or equivalent in a scientific or engineering discipline and shall have received specific training in the response and analysis of the unit for transients and accidents, and in unit design and layout, including the capabilities of instrumentation and controls in the control room. S.2 U" U STAFF CUALICICATIONS [Mi-imu qualification; for me:ber: Of-the unit Ot:f# ch:1' be speci#ied by use Of cr ever:1' qual 4#ic: tier statement referencing :r aug7 St ng 73 3ccept3373 t0 the "9C taff cr alternately by :p ifying individual p;sition qualifica-tion:. Centrally, th: #irst ::thod i: prefer:ble; h; wever, th: ::cend method ir edeptable to there unit staff: requf 9; special qu:1'#ic:t; n :t:terant+ becauce of urique Orgari:: tion:1 tructure.]
*Not responsible for sign-off function.
V0GTLE - UNIT 1 6-6
Rep /nc e wir% .Z~n.,r er f' A O.7"
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ADMINISTRATIVE CONTROLS
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6.4 TRAINING br*""./codeof d,C w h i. 6.4.1 A retraining and replacement trainjng program for the a-t staff shall be maintained under the direction of the][p c~+<--
;iti n titic]- _,e.._ nd ; hall mer ;;t :r av mna +wo re_r,m4 --m,+_ _ _ _s m. a_m_ r_ ~_-_~__,a__,+4.... m.<.
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a $eyl Ace. wiHe .Z~n serf 8 fo 6.5 REVIEW AND AUDIT /'Af e b7 - r,m. .m_ + h. . a_ _ s_,i , a 4. , .,.. 4 m. a_m.p ne.ng_r.g.t,. _ ,.m. _a _s m_ a_4_ m.
+. ,. m_ . 4. +. m. m, m. . _, + 4. _ . .4. ,. ?cc- p'ished ay take-ons-of.-several-. forms. The 'is n:~e :y eith^ :::f p ._a :_2____;__. ,___ +we.. <m .-....+:, +,.-.. . .,4.,.,,.+.4.,., ..g_.. .. . ._.4.+. . .,..,y._..._ , . , . .. ..,.p,,,,,.... . . . . . ..u. _. y i . .pJ .. . . + 414 . , ,+- Af_a f6+m.
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FUNCTION l'23 6.5.1.1 The EGG 3 shall function to advise the ["'G m yWCOnt ,,aperi ;t;ndent] on 4L1 matters related to nuclear safety. . l CCMPOSITION
/$ep/aee w/Me . & sert C. . . . . , . . . , v u. .- r p-c,, , ..x.,,.,u...,,,_~,.,-,_a,,,+wm_. . _ .- .. .
fe - 7. fa paye. J Ch2i-+an: [Pl:nt Superint;nd:nt] 7 i va* hee- ~0per!.ti n; 5 p;rvi;;r] --_' u _ _ u . _. . rs v. echn4o,. e,. . m. ._ - _ . -. 2 u__u,_._,- r u .,. 4. w
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_. r-_*_--_9 u__L_._.
. . . . P m N.1. .. U..._i...-_. . %..._a 9
Ma.c. k., . rvu. __ _ ,... L. . MmL,. .. _ !. . .. . ; ALTERNATES PR8 6.5.1.3 All alternate members shall be appointed in writing by the EE G3 Chairman to serve on a temporary basis; however, no more than two alternates Q shall participate z.s voting members in EEG3 activities at any one time. P2B V0GTLE-bHIT1 6-7
i 4 () Insert A to 6-7 Personnel will meet the minimum education and experience recommendations of ANSI /ANS N18.1-1971 before they are considered qualified to perform ! all duties independently. Prior to meeting the recommendations of , ANSI N18.1-1971, personnel may be trained to perform specific tasks and will be qualified to perform those tasks independently. Insert 3 to 6-7 Personnel will meet the minimum education and experience recommendations- ! of ANSI /ANS N18.1-1971 and, for licensed staff, Appendix A of 10CFR55 before they are considered qualified to perform all duties independently. 3 Prior to meeting the recommendations of ANSI /ANS N18.1-1971, personnel ^ may be trained to perform specific tasks and will be qualified to perform those tasks independently. Insert C to 6-7 () 6.5.1.2 The PRB shall be composed of, as a minimum, a supervisor or equally qualified individual from the departments listed below: Operations
Maintenance j Quality Control (QC)
Health Physics i Regulatory Compliance j Plant Engineering and Services The Chairman and his alternate and other cembers -
of the PRB shall be desianated by the CMVNO.
i t r k i d .i e O v- -
-,-.-+.r --w, -, _ , , , . - . , ,,-,.,-,,,,.,,ww , , . - , - - -..y.,,y.e,w..g.,.,,,_ ,y. ,ym. ,,yy ,.p7 g , -,,,,,g-3 ,.9., ,j% .a,p, ,-,pg.7-9. pp9., 9
a - ADMINISTRATIVE CONTROLS V MEETING FRE00ENCY M8 6.5.1.4 The EW e3 shall meet at least once per calendar month and as convened by the SE 3 Chairman or his designated alternate. PR8 OUORUM PR8 PR8 6.5.1.5 The quorum of the [-WM-] necessary for the performance of the (4RG) responsibility and authority provisions of these Technical Specifications shall consist of the Chairman or his designated alternate and four members including alternates. RESPONSIBILITIES 5 2 e A r er-f to p aj e s ' - s a ,d. 4-9 kr Hen,s a ar,/ b. Pts 6.5.1.6 The (@ne) shall be responsible for: a
a. 0eview of: (1) :1' propc;cd procedures requwed L,y Speuiiicai.ivo 6.5

nd ch:r,q:: therete, (2) :l' preposed progr::: required by Speci'ic:ticr 5.S :nd change; therete, and (3) :ny other preposed preceduces er changes theretc :: deter ~ined by the [Pl:nt Super ntendent] te i

ffect nucle?" sfety, c . -b.. Review of all proposed tests and experiments that affect nuclear (n
uJ
) safety; M e.
d. +. Review of all proposed changes to Appendir. "A" Technical Specifications; 6.-6. Review of all proposed changes or modifications to unit systems or

equi,pment that affect nuclear safety; 5+. Investigation of all violations of the Technical Specifications, including the preparation and forwarding of reports covering evalua-tion and recommendations to prevent recurrence, to the ['!i

0 'recident-Nuclear Cp^ :tions] and to.the [ Company "ucle:r evice 2nd Auditg 5 sa r'e Group 3 ;Vice freddentand[eneralMange,-thefea, af;,n, cpa,p Rened doa r Review of all REPORTABLE EVENTS;

f. -f .
e valu.a.+ ions of l A+ Review ofs e-i-t operationsO to a.ndetect t- potential hazards to nuclear safety; 4 4. Performance of special reviews, investig.tions, or analyses and reports thereon as requested by the [ Plant Su5Or#7tendent] or the Nucice.r Review :nd Audit Group]; frM VA/o [Caparc$Sa.fe ty Ae vies dea <d.
/+ Review of the security Plan and implementing procedures and submittal of recommended changes to the [Cc pany uuclear Revice and A.udit C?;GP1; (yMVA/0 and Hse Sa fefy Aeded 8en.ed.-
1
/
l V0GTLE - UNIT 1 6-8
ADMINISTRATIVE CONTROLS O RESPONSIBILITIES (Continued) R.f, Re/iew of the Emergency Plan and implementing procedures and submittal of recommended changes to the'C$ empany Nuclear Review and ,^udit Group]; GA VNo and he .&fery' Berica don.ed rr I.' 9:vica cf 2ny accidental, unplanned, er uncontrolled dicactive See /oserr- 6 release 4acludia; the preparatier of reporte ceveria; eve!ustien,
*ecc--eadat# cas, and disperitier of the corrective action to prescat /*4e ad f>s /o 8r recu-rence and t'e #c~ardiag cf thece report: te the [Vice orc ident- ' "uclear Operation ] and to the [ Company Nucicar Revic., and Audit 8'" "'"-
bd Creu;]; 2nd m.+. Review of changes to the PROCESS CONTROL PROGRAM, the OFFSITE DOSE CALCULATION MANUAL, and the Radwaste Treatment Systems.
- pad 6.5.1.7 The M shall:
a. Recommend in writing to the [N' n VNoant Superintendent] approval or dis-approval of items considered under Specification 6.5.1.6a. through I. g prior to their implementation;

b. Render determinations in writing with regard to whether or not each item considered under Specification 6.5.1.6a. through g. constitutes f
an unreviewed safety question; and Vice Pe.tMed ud GeneralMa er-Afvelear erdtaa,s
c. Provide written notification within 24 hours to the =aViced-esiQeat-f _
ana t+ nennn Nucicar Operatiens-] and the [Cc=peay Mur'ee" Deviewandthe["!mt of disagreement however, the [ Plant between the @sent] shall hav Superinten responsibility for resolution of such disagreeme spursuanttgSpecification6.1.1. RECORDS S # V'/# ### &# VN" SdeG Se&& b*"'d-PM8 ARB RtB 6. at 5.1. 8 The {@G-] a minimum, shall maintain document written of 21'[ minutes of each E-WM-3 meeting the results [URC] activities performed under the responsibility provisions of these Technical Specifications. Copies shall be provided to the [Vice President-Nuclear Operations] and the [ Comp:ny Muclear a,,
. u.. .,a. . n. . ,,m. . c. - . , a 6.5.2 [ C C"" " " N U C L E.^ o 9EV!"c! ^"0 ^.UDIT CROUP (C" *.C)] r 4rsry ps//sd SeAfzo Bes)
FUNCTION sad 6.5.2.1 The [CN9^C] shall function to provide independent review and audit of designated activities in the areas of:
a. Nuclear power plant operations,
! b. Nuclear engineering, V0GTLE - UNIT 1 6-9 l
.a.-. \
tj INSERT FOR Pages 6-8 and 6-9
a. Review of 1) procedures which establish plant-wide adminis-trative controls .to implement the Q. A. program or Technical Specification surveillance program, 2) procedures for changing plant operating modes, 3) emergency and abnormal operating procedures, 4) procedures for effluent releases of radiolo-gical consequence, 5) fuel handling procedures.

b. Review of 1) programs required by Specification 6.8.4 and changes thereto, 2) proposed procedures and changes to pro-cedures, which in'colve an unreviewed safety question as per 10 CFR 50.59.

1. Review of any accidental, unplanned, or uncontrolled radio-active release in excess of I C1, excluding dissolved and entrained gasses and tritium for liquid effluents, and in excess of 150 Ci for noble gasses or 0.02 Ci of radioiodines for gaseous effluents; and disposition of the corrective action to prevent recurrence and the forwarding of these reports to the Vice President and General Manager-Nuclear Operations and the Safety Review Board.

n. Review of the Fire Protection Program and implementing pro-cedures and submittal of recommended changes to the GMVN0.
r U)
ADMINISTRATIVE CONTROLS O d FUNCTION (Continued)
c. Chemistry and radiochemistry,

d. Metallurgy,

e. Instrumentation and control,

f. Radiological safety,

g. Mechanical and electrical engineering,

h. Quality assurance practices, and

. [0ther cpprcprict 'icid; c::ccieted with the urique chcrectcrictice of the nuclecr pcuer plent.]
sen or yg The [:NRA;] shall report to and advise the,[Vice President-Nuclear Operations] on those iireas of responsibility specified in Specifications 6.5.2.7 and 6.5.2.8. COMPOSITION 6.5.2.2 The [CNP,AC] shall be com;50 sed of -t-he: 4 fn/nt,nm ,/ # eve pe rs,nf n __ __ e-_ , . truaswivu i.wic;
,, ,_, wl>*i As ayowf*, preside the experfise vi4wwwva.
[?O5itiac,Titic] f* ke vie" '"A "" d/f th' oper* fi*" */
"622 .
O um__.
,~~w. " mb c r-r n _ ._ u. . :.. ._ r,_.u. . ,. n_ , a.
v n a e ja , ,o,,,,, p ,j , f, 72, c.ja;,m, , [Pc:itica Tit?c) ^^4 other mem4*re s ha i/ l'e o ppoin /e4 "embe r- [Pecition Titic] by fbe .Senioe Vr'ce fres; dew' Wucleat ALTERNATES O P" *'*"* *' '"'I' afh e r; -erson as he a As/l desa'9no.fe The caseposi+ ion >f fAe SR8 shall
- 6.5.2.3 All alternate memeen shall be appointed in writing by the [CMAC] fnce/ de ,0irc:ter to serve an a temporary basis; however, no more thanftwo ,lte nates p 7,,,,g 0] activities at hny one ime.
(shallparticipateasvotingmembersin[ ,
< na , ernrn a m;,,,,.;fl af S x o jf.7-/f?o, 9
CONSULTANTS 6.5.2.4 Consultants shall be utilized as determined by the (CuoAG) Directer fef 6 CAa./riner to provide expert advice to the CNRAC [Sil8]. i lMEETINGFRE00ENCY
,s az e 6.5.2.5 The [C"DAG] shall meet at,least once per calendar quarter during the l initial year of operation following fuel loading and at least once per 6 lmonthsthereafter. ;jg l
0 A z,g,.r h pye 4-io Aere - 1 V0GTLE - UNIT 1 6-10
~~ ' - ,er- g , _-
4 i 1 i a Insert to page 6-10 However, in extenuating circumstances, the senior Vice President- ; Nuclear Operations may designate the use of additional alternates ' with voting authority when regular members are not available within necessary time constraints. I r
! I i
i t i l 1 I ( . 1 i i j i 4 i 4 4 i 1 I
- - , - - - - - y..,,, , , , , -p, e-,,-9,,v-+,-w+ , - -rw--www---,,-,.v,m, ,,cwe,,,m-aw.w._ _e g -w ,, , e s_ .-,www.-,e
,- D.e tamis4 l
[ i ADMINISTRATIVE CONTROLS i i f b] \ QUORUM See LeB 6.5.2.6 The quorum of the [Ct!RAC] necessary for the performance of the [CN"C] l review and audit functions of these Technical Specifications shall consist of the Direct ^ or his designated alternate and at least [f;ur C"RAC] members including lternates. No more than a minority of the quorum shall have line responsibi ity for operation of the un-i4. , y. d /h p Chainna.n f'IA " I s.z8 6.5.2.7 The [C"RAC] shall be responsible for the review of:
a. The safety evaluations for: (1) changes to procedures, equipment, or systems; and (2) tests or experiments completed under the provision of 10 CFR 50.59, to verify that such actions did not constitute an unreviewed safety question;

b. Proposed changes to procedures, equipment, or systems which involve an unreviewed safety <;uestion as defined in 10 CFR 50.59;

c. Proposed tests or experiments which involve an unreviewed safety question as defined in 10 CFR 50.59;

d. Proposed changes to Technical Specifications or this Operating License; (w\ e. Violations of Codes, regulations, orders, Technical Specifications, license requirements, or of internal procedures or instructions having nuclear safety significance;

f. Significant operating abnormalities or deviations from normal and exp5cted performance of un.it equipment that affect nuclear safety; p lant

g. All REPORTABLE EVENTS;

h. All recognized indications of an unanticipated deficiency in some aspect of design or operation of structures, systems, or components that could affect nuclear safety; and PMa

i. Reports and meeting minutes of the [4M-3 AUDITS plan t 6.5.2.6 Audits of' west activities shall be performed under the cognizance of the [C!;;AC]. Th;s; cudit; :hcll cr.00mp:::: Se e insert for pape 4 -it.
f Esf
a. The conformance of f we+t operation to provisions contained within the Technical Specifications and applicable . license conditions at least once per 12 months; O l V0GTLE - UNIT 1 6-11
l I INSERT for Page 6-11 l l Each inspection or audit shall be performed within the specified time
interval with

l 1. A maximum allowable extension not to exceed 25% of the inspection j audit interval. j l 2. A total maximum combined interval time for any 3 consecutive inspection j ! or audit intervals not to exceed 3.25 times the specified inspection or i, audit interval. , These audits shall encompass: l l 3 i t t 3 i [ I l 1 $ f I i l9 i i I t 1 I i ! i l ! I i 4 1 ll f l !
4
} ! i l ! l O l
l . ADMINISTRATIVE CONTROLS AUDITS (Continued)
b. The performance, training, and qualifications of the entire i staff at.least once per 12 months;

c. The results of actions taken to correct deficiencies occurring in 3
p/anrun.i.t equipment, structures, systems, or method of operation that affect nuclear safety, at least once per 6 months;_ i
d. The performance of activities required by the Operational Quality l
Assurance Program to meet the criteria of Appendix B, 10 CFR Part 50,
at least once per 24 months; j e. The fire protection programmatic controls including the implementing procedures at least once per 24 months by qualified licensee QA ,
personnel; j f. The fire protection equipment and program implementation at least once per 12 months utilizing either a qualified offsite licensee
fire protection engineer or an outside independent fire protection
! consultant. An outside independent fire protection consultant shall
{ be used at least every third year;
g. The Radiological Environmental Monitoring Program and the results thereof at least once per 12 months; O h. The OFFSITE DOSE CALCULATION MANUAL and implementing procedures at i least once per 24 months;

i. The PROCESS CONTROL PROGRAM and implementing procedures for processing l
and packaging of radioactive wastes at least once per 24 months; 3 j. The performance of activities required by the Quality Assurance Program for effluent and environmental monitoring at least once per i 12 months; and
k. ^ny cther are: cf unit Op:rcti:n ::::ider:d .ppr:grict; by the
[:NPdC] er the [' lice Preside t .t. clear Operatier.;]. ;
'See insert for pay. e. 6 -12 \
RECORDS SM8 , 6.5.2.9 Records of [CM ^.C] activities shall be prepared, approved, ar.d dis- l tributed as indicated below: Sen to r SM
a. CNPJ.C] meeting shall be prepared, a Minutes forwardedof to h tlAh [jVice President-Nuclear Operations /pproved, an within 14 days-

following each meeting; r,ter

b. Reports of reviews encompassed by Specif cation 6.5.2.7 shall be prepared, approved, and forwarded to the Vice President-Nuclear-O- Operations)/ within 14 days following completion of the review; and -
V0GTLE-bHIT1 6-12 l
._e i
i Insert for page 6-12 i 1 4 1. The Emergency Plan and implementing procedures (at least once per 12 months); I and 'l
m. The Security Plan and implementing procedures (at least once per 12 months).
i i l i ! i 8
. t 5 I l,.
l 4 l l I - 3 4 4 l } } i 3 i .l I I i . l l i ! i l I 1O l l
ADMINISTRATIVE CONTROLS RECORDS (Continued)
#* * ~ #b .ser,ior
c. Audit re s encompassed by Specification 6.5.2.8 shall be forwarded to the icePresident-NuclearOperations7andtothemanagementposi-tions responsible for the areas audited within 30 days after comple-tien of the audit by the auditing organization.
! 6.6 REPORTABLE EVENT ACTION 6.6.1 The following actions shall be taken for REPORTABLE EVENTS: /er ; a. The Commission shall be notified and,a report submitted pursuant to the requirements of,Section 50.73 to 10 CFR Part 50, and se e.tton so. 72. and pga 'b. Each REPORTABLE EVENT shall be reviewed by the EeRe-3, and the results of this review shall be submitted to the [CNPAC] ##8 and the
[Vice President-Nu:!: r Oper tion;3 Vice. henden t an.s Genera./ Nayr - Alseleu Cpera.+/ ens 6.7 SAFETY LIMIT VIOLATION i 6.7.1 The following actions shall be taken in the event a Safety Limit is violated: i N a. In accordance with 10 CFR 50.72, the NRC Operations Center shall be s notifi'ed by telephone as soon as practical and in all cases within one hour after the violation has been determined. The 9:n:;; r,
<Nuci:Or Production and th: NFSC shall be notified within 24 hours, ne. Vice President ud. Genern/ Ma. nag er-Aln/ea rOpern+ ions, theSM, PR$ ad H,e
b. A Licensee Event Report snall be prepared in accordance with CrNNO 10 CFR 50.73.
MAsu Gy VNo
c. TheLicenseeEventReportshallbesubmittydtotheCdmmissionin
; accordance with 10 CFR 50.73, and to the PGRG, the N4C, and the M:::;;;r, L lcar Production within 30 days after discovery of the event. gN ,, ,p,,,;g,,y ,,, g,,,,,_, 99,,,],, _ Apvelear Cper'* fi n s
d. Critical operation of the unit shall not be resumed until authorized by the Nuclear Regulatory Commission.

6. 8 PROCEDURES AND PROGRAMS
; 6.8.1 Written procedures shall be established, implemented, and maintained covering the activities referenced below:
a. The applicable procedures recommended in Appendix A of Regulatory Guide 1.33, Revision 2, February 1978; i b. The emergency operating procedures required to implement the require-i ments of NUREG-0737 and Supplement 1 to NUREG-0737 as stated in Generic Letter No. 82-33; V0GTLE - UNIT 1 6-13
7 i () ADMINISTRATIVE CONTROLS i i PROCEDURES AND PROGRAMS (Continued)
c. Security Plan implementation;

d. Emergency Plan implementation;
; e. PROCESS CONTROL PROGRAM implementation; i
5
; f. OFFSITE DOSE CALCULATION MANUAL implementation; and +
S~.> e in s e r r
! /,,- o ta a g. Quality Assurance for effluent and environmental monitoring. & - t i. h. ff*** IWl*""!'"Yi'"'
1 '67372-+Eschfire p-ecede-e ProofftcHen Specificatica 6.9.1, ead cheages the-ete, she!' be i reviewed by the [URG}-and-shall be :ppreved by the [Pl:nt Superintendent] prior to imp!: mentation and revir::d periodically as set forth 4a adm'-f:tr:- tive precedures. 6.8.3 Temporary changes to procedures of Specification 6.8.1 may be made pro- -! vided:
a. The intent of the original procedure is not altered; i

b. The change is approved by two members of the plant management staff, i at least one of whom holds a Senior Operator license.:n the unit sotambnA. snA i in aceanlance esi/A 61 L

c. The change is documented, reviewed by th: ["Pf], and approved 4by the-
[P!:nt Superintendent] within 14 days of implementation. ] 6.8.4 The following programs shall be established, implemented, and maintained: i i a. Primary Coolant Sources Outside Containment i a p p,/tc a /e pe ette n s e + c.o nwn me nr AprogramtoreducelTakagefromthoseportionsofsystemsoutside containment that coutd contain highly radioactive fluids during a serious transient orlaccident to as low as practical levels. The systems include [thejrecirculation spray, Safety Injection, chemical i and volume control, g : : tripper, and, hydrogen receatiners]. The program shall include the following: Mi residul Nxt remeva.1 susfeats i 1) Preventive maintenance and periodic visual inspection requirements, and j 2) Intege:ted- leak test requirements for each system at refueling
cycle intervals or less.

b. In-Plant Radiation Monitoring i i A program which will ensure the capability to accurately determine i the airborne iodine concentration in vital areas under accident conditions. This program shall include the following:
1 i V0GTLE - UNIT 1 6-14 l
l i l Insert for Page 6-14 !' ( } 6.8.2 Each procedure of 6.8.1 above, and changes thereto, shall be approved by either the GMVN0 or the department head of the responsible department prior to implementation with the exception of the following which shall be approved by the GMVNO:
1) procedures which establish plant-wide administrative controls (which implement the quality assurance program and the Technical Specifications surveillance program),

2) unit operating procedures (UOPs)

3) emergency operating procedures (E0Ps)

4) abnormal operating procedures (A0Ps)

5) procedures for implementing tha escurity plan, emergency plan, and the fire protection program, and

6) fuel handling procedures.
PRB responsibilities for review of procedures are delineated in 6.5.1.6. Additionally, procedures will be reviewed periodically as set forth in-administrative procedures.
)
e d O
n ADMINISTRATIVE CONTROLS (m )I PROCEDURES AND PROGRAMS (Continued)
b. In-Plant Radiation Monitoring (Continued)

1) Training of personnel,

2) Procedures for monitoring, and

3) Provisions for maintenance of sampling and analysis equipment.

c. Secondary Water Chemistry A program for monitoring of secondary water chemistry to inhibit steam generator tube degradation. This 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 points).?ich th !' ' c'ude scrit;cing the di::hcrg: Of the cend::::te pu p: #c ev'de'~e a f 20nden r # -lech:g^,,

4) Procedures for the recording and management of data, q 5) Procedures defining corrective actions for all off-control Q 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.

d. Bac keo-Hethc d ' r D e te r ' ' ; Eb c co ! ' n; " ;#7 U A: wit.n ; :ing?: nnnnei Of : ritering M::rrent:t cn3 #
A-progr e "ich "i encure th: : pability t: Occur:tely ::riter the ka-ter-Coolant SysteSsubsoe44eg margin Thi: ;;r:gra : hall inide 4ae-fol-lo4:
1) Training-cf per:ennel, 2nd
) Procedure: for 0riter#ng.
d +. Post-Accident Samoling A program which will ensure the capability to obtain and analyze reactor coolant, radicactive iodines and particulates in plant gaseous effluents, and containment atmosphere samples under accident conditions. The program shall include the folicwing:
1) Training of personnel,

2) Procedures for samoling and analysis, and

3) o rovisions for maintenance of sampling and analysis equipment.
V0GTLE - UNIT 1 6-15 ywr - p-
ADMINISTRATIVE CONTROLS
-O V ~
6. 9 REPORTING REQUIREMENTS ROUTINE REPORTS 6.9.1 In addition to the applicable reporting requirements of Title 10, Code
! of Federal Regulations', the following reports shall be submitted to the Regional , Administrator of the Regional Office of the NRC unless otherwise noted. STARTUP REPORT 6.9.1.1 A summary report of plant startup and power escalation testing shall be submitted following: (1) receipt of an Operating License, (2) amendment to the license involving a planned increase in power level, (3) installation of fuel that has a different design or has been manufactured by a different fuel i supplier, and (4) modifications that may have significantly altered the nuclear,
, thermal, or hydraulic performance of the unit.
The initial Startup Report shall address each of the startup tests iden-1 tified in the Final Safety Analysis Report and shall include a description of the measured values of the operating conditions or characteristics obtained during the test program and a comparison of these values with design ,,redictions and specifications. Any corrective actions that were required to obtain satis-factory 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/or 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 events have been completed. ANNUAL REPORTS
i plar,t 6.9.1.2 Annual Reports covering the activities of the un44 as described below for the previous calendar year shall be submitted prior to March 1 of each year. The initial report shall be submitted prior to March 1 of the year following initial criticality.
Reports required on an annual basis shall include

olan t
a. A tabulation on an annual basis of the number of RM4+e, utility, !
and other personnel (including contractors) receiving exposures lan+ ,
*A single submittal may be made for a multiple unit _p"t5. The submittal l should combine those sections that are common to all A at the g .
( ho+4 reacfors i i V0GTLE - UNIT 1 6-16 i l l , 1 A ._,,-,..---,,_,m. = , . , . . . - . , . _ , .. -, - --_ -.
4 4+J &mi-- ,- a 4.J a- --t g 4 aAJ . ,a..., A. ..a.a - 24. ADMINISTRATIVE CONTROLS ANNUAL REPORTS (Continued) greater than 100 mrem /yr and their associated man-rem exposure
, according to work and job functions * (e.g., reactor operations and surveillance, inservice inspection, routine maintenance, special maintenance [ describe maintenancc], waste processing, and refueling).
The dose assignments to various duty functions may be estimated i 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 4
sources should be assigned to specific major work functions;
b. The results of specific activity analyses in which the primary coolant exceeded the limits of Specification 3.4.8. The following information j shall be included: (1) Reactor power history starting 48 hours prior-l to the first sample in which the limit was exceeded (in graphic and tabular format); (2) Results of the last isotopic analysis for radio-iodine performed prior to exceeding the limit, results of analysis while limit was exceeded and results of one analysis after the radio-iodine 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 hours prior to the first sample in which the limit was exceeded; (4) Graph of the I-131 concentration (pCi/gm) and one other radioidine isotope concentration (pCi/gm) as
,O i
a function of time for the duration of the specific activity above the steady-state level; and (5) The time d.uration when the specific
~
activity of the primary coolant exceeded the radiciodine limit. i See kres-r' 19epage. 6 -/7.
<c. [.^ny ofher unit unique--reports-required en :n :nnu:? b::ic.]
l su,zymycg acHviHes of +he ANNUAL RADIOLOGICAL ENVIRONMENTAL OPEi&T4NG REPORT ** M /o N Envi e ntedal 6.9.1.3 famit/d*"'" ") #U" RoutineAnnualRadiologicalEnvironmental9pece%,gReportscovering1 l
the Oper: tier of the unit during the previous calendar year shall be submitted prior to May 1 of each year. The initial report shall be submitted prior to 4
May 1 of the year following initial criticality and shall include copies of reports of ~the preoperational Radiological Environmental Monitoring Program of I the unit for at least two years prior to initial criticality. sumillance. The Annual Radiological Environmental 0;;;r: tin;; Reports shall include i summaries, interpretations, and an analysis of trends of the results of the radiological environmental surveillance activities for the report period, i
*This tabulation supplements the requirements of $20.407 of 10 CFR Part 20. **A single submittal may be made for a multiple unit ~ Me44en.
l p u.n r O. ! V0GTLE - UNIT 1 6-17
INSERT for Page 6-17 - ( )~
c. A report shall be prepared and submitted to the commission on an annual basis if sealed source or fission detector leakage tests reveal the presence of greater than or equal to 0.005 microcuries of removable ,
contamination.
d. An annual data report on diesel generator reliability will be sub-mitted and, in addition, the following information will be included:

1. A summary of all tests (valid and invalid) that occurred within the time period over which the last 20/100 valid tests were perfo rmed.

2. Analysis of failures and determination of root ca . of failures.

3. Identification of all actions taken or to be taken to 1) correct the root causes of failures defined in b) above and 2) achieve a general improvement of diesel generator reliability.
4 An assessment of the existing reliability of electric power to engineered-safety-feature equipment, j n V H I t l
hMu6 l The rad.to/syica/ leve/ of radionaelides whicI, aer. ADMINISTRATIVE CONTROLS
,swe AnuANCE }
ANNUAL RADIOLOGICAL ENVIRONMENTAL CPERATI M REPORT (Continued) as a-ppro ytate. including /a comp /rison with preoperational studies, with operational controls,
;ppr:priath and with previous environmental surveillance reports, and an assessment of t-he observed impacts of t-he plant operationson the environment. ,
go7 ',The reports shall also include the results of the Land Use Geews required by Specification 3.12.2. 5*"ven saveiItwee The Annual Radiological Environmental Opcr; ting Reports shall include the. results of analysis of all radiological environmental samples and of all ' environmental radiation measurements taken during the period pursuant to the locations specified in the table and figures in the Offsite Dose Calculation Manual, as well as summarized and tabulated results of these analyses and ' measurements in the format of the table in the Radiological Assessment Branch Technical Position, Revision 1, November 1979. A In the event that some indivi-dual results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results. The , missing data shall be submitted as soon as per:fb?: in a supplementary report.
.2 poin r modwe he+weer, +A e A.io P7" NC'ble-Tne re' ports shall also include the following: a summary description of tne , adiolcgical Environmental Monitoring Program; at 1 ::t tue legible maps coveqing all sampling locations keyed to a table giving distances and directions from tne cent:" " Of One reactor; the results of licensee participation in the Interlaboratory Comparison Program and the corrective action taken if the C') specified program is not being performed as required by Specification 3.12.3; / reasons for not conducting the Radiological Environmental Monitoring Program as recuired by specification 3.12.1, and discussion of all deviations from the sampling schedule of Table 3.12-1; discussion of environrental sample measure-ments that exceed the repo-ting levels of Table 3.12-2 but are not the result of plant effluents, purscant tc ACTION b. of Specification 3.12.1; and discussion af all analyses ir. which the LLD -equired by Table 4.12-1 was not aeMevaMe.
SEMIANN'JAL R;CICACT:VE EFFJ.;ENT K EASE REPORT *F d e Aie ved-6.9.1.2 Ret tine h miarnual Radicactive Effluent Release Reports covering the coeratf or of nc unit during the previou -6 nanths of operation shall be ; subMtted witrin 60 days af ter January 1 W July 1 of each year. The period of tie first report shall becin sitn the dete of initial criticality. The SeMannual Radioactive Effluent Release ieports shall include a ' sunary o* the quantities of radioactive liquid anc gaseous effluents and ' s:l d waste released from the unit as outlined in Regulatory Guide 1.21,
'Peasuring, Evaluating, and Reporting Radioactivity in Solic Wastes and Eeleases of Radioactive Materials in Liquid and Gaseous Eff uents from J gnt-Water-Ccoled nuclear Power Plants," Revision 1, June 19M, ,wi th data +w>-shi1 coe-stat 4oM-nea~the-SM&aGWGAW;+secenm i1 ; 9chde-twrerMMMM-s-tethm . Oad A sincie sutmittal may ne mace for a multiple unit it "'?'. The abmitte' snoul: zmoine these se tions that are cccccr, to all units at the station.;
J rc-ever, for uni n sith 4arate racwaste systems, the suamittal shall scaci ? =e re: eases of r::ioact ive material from each unit. V0GTLE - UNIl 1 6-1F
ADMINISTRAT3/E CONTROLS V SEHIANN"JAL Rf.DICACTIVE EFFLUENT RELEASE REPORT (Continued) scraarized on a cuartarly ba. sis following the format of Appendix B thereof. For solid wastes; the format for Table 3 in Appendix B shall be supplemented with three additional categori es: class of solid wastes (as defined by 10 CFR Part 61), type of container (e.g., LSA, Type A, Type C, Large Quantity) and SOLIDIFICATION agent cr absorbent (e.g. , cement, urea formaldehyde). Te Senianr, cal Radfuactive Effluent Release Report to bc submitted within 60 days af ter January 1 of each year shall include an annual summary of boutly oetearologica1 data collected cver the previous year. This annual sumary may me either in tre fom cf an hour-by-bcur listing on magnetic tape of wicd speed, wind direction, atmoseneric stability, and precipitation (if measttred), or in the for:n of joint frequency distributiens of wind speed, wind directicn3 and atnespbe -ic stability." This same report shall include an assessrr, ant of tne. radiation doses due to the radicactive liquid and gaseous efficenti released from the unit or station during the previous calendar year. This same raport shall also irclude an assessment of the radiation dcses from radioactive liquid and garecus effluents to MEMEERS OF THE PUSLIC due to their activities g3j ,inside the SITE SCUNCARY (Figure hE-t-43) during the report period. All assumptions used in making cese assessments, i.e., specific activity, exposure time, and location, shall be inf.uded in these reports. h meteorological conditions concurrent with the time of release of radicacti ve materials in gaseous effluer.ts, as determined by sampling frequency and' neasurement, shall be used for determirticg the gaseous pathway coses. The estessment of radiation f dosas shaB be perfomed in accordance with the methodology and parameters in k]/ the OFFSITE DOSE CALCULATION fWiUAL (ODCM). .wrver/c / uha/uuage me%<,/eg cas,d.w,s or-The Semiannual Radioactive Effluent Release Report to be submitted within 60 days af ter Canuary 1 cf esca year shall also include an assessment of radiation doses tc the If kaly most exposed MEMBER OF THE PUBLIC from reactor releases anc cther aE+ coy uranium fuel cycle sources, including doses from f primary effluent cathways and direct radiation, for/the previous calendar year to shew conformarce with 40 CFR Part 190, "Envirent/ ental Radiation Protection Standards for Nuciear Pover Operation." Acceptabl,e methods for calculating - the case contribution from liquid and gasecus eff/uents are given in Regulatory Guide 1.109, Rev.1, October 1977.
wHhin S km The Semiannual Radicactive Effluent Relesse Reports shall include a list and desc:iption of espianried releases frem the site to UNRESTRICTED AREAS of radioactive materials in gasecus and liquid effluents made during the reporting period.
The Semiannual Radioactive Effluent Release Reports shall include any changes made during the reporting period to the PROCESS CONTROL PROGRAM (PCP) and to the OFF5ITE DOSE CALCULATION WNUAL (00CH), pursuant to Specifica-tiens 6.13 ans 6.14, respectively, at well as any major change to Liquid, i
"In lieu cf sucaission with the Semiannual Radioactive Effluent Release Report, the licensee has the optica of retaining this summary of required
(_3 nateorological cata on site in a file that shall be prcvided to the NRC l t/ cpon request. VCGTLE - UNIT 1 6-19 l
t' ADMINISTRATIVE CONTROLS _ _ l
SEMIANNUAL PADI0 ACTIVE EFFLUENT R$ LEASE RE_ PORT (Continued)
Gaseous, Or Solid Radwaste Treatment .5ystems pursuant to Specification 6.15. - It shall also include a listing of new locations for dose calculations and/or , environ.tantal monitoring identified by the Land Usefc= pursuant to Speci- , fIcation 2.12.2. ,y ,.,,e y ,- l The Semiannual Radioactive Effluent Qelease Reports shall also include the following: an explanatiore as to why the inoperability of liquid or gaseous effluent ronitoring instrumentation was not corrected within the time specified in Specification 3.3.3.10 or 3.3.3.11, respectively; and description of the events leading to liquid holdup tanks or gas storage tanks exceeding the : iiraits of Specification 3.11.1.4 or 3.11.2.6, respectively. MONTHLY OPERATING REFORTS
; 6.9.1.5 Routine reports of operating statistics and shutdown experience, ,
ir;cluding docur.entation of all challenges to the PORVs or safety valves, shall ha submitted on a monthly basis to the Director, Office of Resource Management, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555, with a copy to the Regional At.ainistrator of the Regional Office of the NRC, no later than the 15th of each month following the calendar month covered by the report.
RADIAL PEAKING FACTOR LIMIT REPORT
6. 9.1. 6 Tne F xy limits for RATED THERMAL POWERxy (FRTP) shall be provided to $
the [lRC Regiona? Administrator with a copy to Director of Nuclear Reactor ; Reguiatica, Attention: Chief, Core Perfoemance Branch, U.S. Nuclear Regulatory - Cc mission, Washirgton, D. C. 20555, for all core planes contaioing Bank "D" cor.trcl rods and all unrodded core planes and the plot of predicted (F Pg ,)) - vs Axial Core Height with the limit envelope at least 60 days prior to each j cych initial critica?ity unless otherwise approved by the Connission by letter. l In additf on, in the et 91t that the limit should change requiring a new substar.- , tial or an amended submittal to tne Radial PeaMcg Factor Limit Report, it will be submitted 60 days prior to the date thy limit would become effeckive unless otherwise approved by the Commission by letter. Any information needed to suoport F,N will be by request froc the NRC and need not be 'ncluded in this report. s
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each mit Sed 'te M pr6caration 2nd 59 5-tte! . e dt.:!g@,e4 4-t'.: Ted-ic:1 . Sp::i'icatic,m.] 4
5. 9.2 Scecial reports sha}3 6e subcoitted to the Regional Actitistrator of the .
! Regional Office of the tiRC within the tiree pericd specified f.w e.ach report..
6.10 RECORD RETENTION 6.10.1 In acaition to the appHeable record retsotion recut ements of Title 10, Code of FederM Regulations, the followtrg records sosti de retained for at , Teast tPe mirian period indicated. , 5.10.2 T5e fc1'.owirs racords sta9 ce retairad for at least 5 years: ce
a. Records and logs of M~ + operation covering ti .e interval .at each <
power level;
b. Reccrds and legs of prircipa) . maintenance activtties, inscections.
{ s repair, and replaren.ent of principal itans of equip. tent related to nuclear safety;
c. All REPCRTABLE EVE.YT5;

d. Recores of surveilla.nca activities, inspections, and calibratiens repaired by these Tacnnica! 5pacifications;

e. Records of chroges inade to the procedures required by Specification 6.6.1;

f. Records of radioactive shipre'its;

g. Records of sealed source and fiufon detector leak tests and res.ults;
-,644
h. Records of annual physical inventory of all sealed sour:e caterf ai of record;udc ;
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4 ADP.INISTRATIYE CONTROLS
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RECORDRETENTION(Contincedl pid 6.10.3 The folicwing records shall be retained for the duration of the w4t Cperating License: p%#
a. Records and diawing changes reflecting Wt- design modifications

made to systems and equipment described in the Final Safety Analysis Repcrt; b, Records of new and irradiated fuel inventory, fuel transfers, and assemb;y burnup histories;

c. Records of radiation exposure for all individuals entering radiation cartrol areas;

d. Secords of gaseous and liquid radioactive caterial released to the ;
ensirons;
e. Pecorcs of transient or cperational cycles for these ett cociponents
; identi fied in Table 5.7-1; ,
f. Records of reactor tests and experfeents;

g. Records of training and qtralification for current owbers of the ggstaff;

b. ' Records of inservice insoections performed pursuant to these Technical Specificaticns; GS A L Records of qua14ty assurante activities required by the CprM.imai O'
i.
-Quefh7 2:u-:co hci;
j. Records of reviews performed for changes made ta procedures or equipeent or reviews of tests and e.xperiments pursuant to 10 CFR 50.59; L Records of maetings of the -h and the ((dC3; t 1 Re ords of the service livu of all hydraulic and mechanical 4
scubbers raquired by Specification 3.7.dfincluding the date at which the service 'ife commences and associated installation and inaistenance records; m R ecord s- of_ seccthwata r-- sampli ng- and-wate rgwMty+-a nd ra .+. Records of analyses rerpired by th.t Radio 1cgical Environmental Idenitoring Prgram that would permit. evaluation of the accuracy of the analysis at a later date. This snould include procedures effective at specified times and QA recards showing that these procedures were followad. 6.11 RADIATIG4 PROTECTION PROGRAN 6.11.1 Procedures for persennel radiation protection shall be prepared censistent with tha requirements of 10 CFR Pstt 20 and shall be approved, maintained, and achered to for all operations involving personnei radiation exposure. l l O VoGTLE - MI7 1 6-22 1 l 1
73 ACMINI$TRATIVE CONTROLS v- - .c , {w,
5. '.2_ Pld R_ADIATIOK AREA [0PT10Ml.J 6.12.1 Pursuant to paragraph 20.203(c)(5) of 10 CFR Part 20, in lieu of the
" control oevice" or " alarm signai," receired by paragraph 20.203(c),eact} high , radiation area, as definen in 10 CFR Part 20, in Wnich toe intensity of radia-a r e r- jtien 1sE2! to-cr less thar. '000 aR/h at 45 cm (18 ir.) from the radisticr. %m .
ce mv AsoEr_ce cr from any surface which the radiation penetrates shall ne barricaded W ,'and conspicJously pgsted as a high radiaticn area and entrance thefeto Shall be jcentroHed by requiring issuance of a Rsdiation Work Permit (RWP). Individuals
< qualified in riciation protection procedures (e.g., Hepith PhYd es Techniciard gh ff . or personncl-cent 4neoush-escorted by such individeals -est be execut frem the RWP Qssuance reguires,e_n_tj uring the performance of their assignud duties in high radla-tion neas with ex;osuro rateshther 163s than 1000 AR/h, provided they are otherwise following plant radiation protection procedures for entry into such high radiation areas. Any indivioual or gr6up of indiciduals permitted to enter such arar3 shall be provided with or accompanied by one or more of the following:
a, A faciation monitoring device which continuously indicates the radiatien dose rate in the area; or
b. A radiation monf torieg device which continuously integrates the radiation dose rata ir, the area and alarms when a preset integrated dose is received. Entry f ato such areas with this monitoring device may ba made after the dose rate levels in the area have been 1 established ar.d personnel have been made knowledgeable of them; or O) t V

c. An individual qualified in radiation protection procedures with a
, radistion dose rate conitoring device, who is responsible for providing &c d'M N h bW periodic radiation SurvGillance at the frequency spy:ified by the
{;positivecontrolovertheactivitieswithintheareaand;hallp [ Radiation Protection Nanager) in tha elWP. 5,12.2 D @4cn t' t"a r pirement: cf-Specincataea4-12.1, cre : ::: g ibi; 4+erscanet with - rad 4 tion !evels-greatee-thare4000 ;ti!!/Pr--at 'l : ; (10 in. ) - M - t % radiat4ca source cr 'rwxr::c o.: Men the r:diatisre s.eactrates M be powland-with-Mcked-doors-to-prevent-unauthori2ed-ntry, n.d A is st. ARM r.,+?nt& iced-under-the adMaistratire ^atre? ef the eh4 ft cereen ^^
-ity ud/or4ealth-physics-supervision,-Doors-shaWr:@ 1 :ked =:cpt irN r,c.iads-of--access-by-personne0-under-an-aporsved4Wp--which-sha!' pecify the -
das& rate leveis in the immediate wcrk areas-and the W em 20 0:12!: tey ; ti e 'ce fr4ivt ^ 'a + 5at-are a ImMeu of the etay t4- ma*'-"4" af I the-%prairect or- remote--(such as closed circuit -TV- camerd coat - i smeiUanca-aay-be-mada-by-personne-quam 4ed in radiathn protectick p rcee *o 7-revM positi'ze exposure - centre! - 0"- the- :: tit-itic; being p+ciarmed withi n the- was.
Fee 4ndividua'i high ra44at4on-areas-accestMe te Pr50=e1 '4th-rad 4at'^^
' W:-obgreater--than--1000-4/h-that-are4ocated-vithin-icrge rea:,-:uch :: ,
L2-soota43entri-+.here-no-enclosure exists-fospurpose+-ef leeMg, ?nd h: : x ,chrc- c:n te ra+ enemy-con +tructed--arsund th; t,dividual arca, e,6. V %:' cr: . :h:1' 5: M rricad M on:01:ece:1y pcete .', &b'heOii
. M : h ' ' bc ac t i v ' ete d - 0 ; c m*4eg -dey4+e . l m )
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k_m) INSERT FOR Page 6-23 Tre reouirements of 6.12.1, above, shall also apply to thcs: creas accessible to personnel with radiaticn levels such C'n at a major corticn of the body could receive a dose greater than 1000 meem/hr. In addition, lockec dcors shal! be provided to prevent unhothorized entry into such a rea s . Those areas fer wnich no enclosure exists for purposes of locking '- tand no enclosure can be reasonaoly constructed arour.d those area 5) shall te ropec-off, :anscicuously costed, and a flashing light shall be acti. vates as a warning cevice. Dose measbrenents 6re made at IB" from the source of radioactivity in these arecs. The keys shall be maintained under the administrative control of tne Operations Supervisor on duty ant /cr
ne Plant Health Physicist.
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[-] ADMINIS7RATIW CONTROLS L13 PRGCEss CONTROL PROGRef g _ 6-13.1 The PC.1:h " beepoeWed-by--the-4:=i; ise p-f or tz-%mmut4w, 1 6.13..C. Licensee-Initiated changes to the l'CP:
a. Shall be submitten to the Co!ttaission in the faisnnual Radicactive Effluent Release Report for the period in which the charCe(s) was made. This sutmittal shall contaic:

1) S;Jfficiehtly cetailed inform 2 tion to totalIy support the rationale for ths chance without ben'efit of additional or supplemental infor, nation;

2) A determination inat the change did not reduct the overall -
conformance of the solidified waste product to uisting criteria for solid wastos; and
3) Boct:santation of the fact that the change has been reviewed and found acceptable by the g .
/ C'i'NCA N !' (* 5-l $
D. Shall bEcoms affective upon review idb2ccefcCtra?ept,aA6;wlE5 --P{tRG,y. by ti.e-a , s. a,rirmt on accordace ins S epe.1, e, +, ar, s. . r.2 6.14 0FFSITE C055 CALCULATION MAMJAL (ODCM1 5.I'.1 The CDC" M' be pp eved by Mc certeder pde te "p.' m at:ti e. 6.14 Licensee-initiated changes to the ODCM:
a. Shall be seb:nitted to the Coonission in the Seniennual Radioactive Effluent RC ease Rep 3rt for the period in which the change (s) was r' ace effective. This subnittal shall contain:

1) Sufficiently oetailed informatica tc totally sLppcrt the rdtionale for th9 Change without benefit of additional or sLpplemental inforrtation. Infornation subnitted should consi.st of a package of those pcges of the OD".X to be changed with each pags numbered, dated and containin,q the revision nut.ber, togetner with acpropriate anslyses or evaluations justifying the change (s);

2) A determination that the charge will not reduce the accuracy or reliability of dose calculations or Satpoint determinations; and

3) Occumentation of the fact that the char.ge has been reviewed and found acceptable by the g.

b. Shall .beccme effective epon review W acc4tre by t.;; ['J."G].
in a cco rJ.1,rce u. irh sp.e i.4 cxum s .g. i. c of xcyrevd in ~L ee.. rix , c e. with Spe cifica Sen 4 f ,e . V0GTLE'- UNIT 1 6-24 i
u .. ADMINISTRATIVE CONTROLS 6.15 MAJOR CHANGES TO LIQUIE, GASEOUS, AND SOLID RADWASTE TREATMENT SYSTEMS
6.15.1 Licensee-initiated major changes to the Radwaste Treatment Systems (liquid, gaseous, and solid):
a. Shall be reported to the Commission in the Semiannual Radioactive Effluent Release Report for the period in which the evaluation was reviewed by the ftfRet. The discussion of each change shall contain:
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1) A summary of the evaluction that led to the determination that the change could be made in accordance with 10 CFR 50.59;

2) Sufficient detailed information to totally support the' reason for the change without benefit of additional or supplemental information;

3) A detailed description of the equipment, components, and processes involved and the interfaces with other plant systems;

4) An evaluation of the change, which shows the predicted releases of radioactive materials in liquid and gaseous effluents and/or
~
quantity of solid waste that differ from those previously predicted in the License application and amendments thereto;
5) An evaluation of the change, which shows the expected maximum exposures to a MEMBER OF THE PUBLIC in the UNRESTRICTED AREA and to the general population that differ from those previously estimated in the License application and amendments thereto;

6) A comparison of the predicted releases cf radioactive materials, in liquid and gaseous effluents and in solid waste, to the actual releases for the period prior to when the change is to be made;
'/) An estimate of the exposure to plant operating personnel as a result of the change; and
8) Documentation of the fact that the change was reviewed and found acceptable by the-[BRGh h accordarece <444 Spec /Ac, Won u.5.!

b. Shall become effective upon review and acceptance by tt.c [URC]r s'a Mc e r*.D rr c 4 w,*/is $p t c ificen fie e' 4. C . / . ,
*Lic'ensees may choose to submit the information called for in this Specification as part of the annual FSAR update.
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V0GTLE - UNIT 1 6-25
[%'Jl JUSTIFICATIONS FOR DEVIATIONS FROM STS SECTION 6.0 General:
" Plant Superintendent" was changed to "GMVNOD" to reflect plant-specific nomenclature.
The word " unit" was replaced by " plant" to reflect plant-specific nomenclature. 6.1.2: The STS wording was replaced with the wording found in paragraph 13.5.1.1.E of the VEGP FSAR. 6.2.2: All changes to 6.2.2 were made to reflect plant-specific nomenclature except the following:
1) " Health Physics Technician" was revised to " Individual qualified to
-['T
(/ implement Radiation Protection Procedures" on the hasis that the intent of this specification can be achieved by the qua'.f fication of other individuals outside the Health Physics organization (i.e. , any member
of the operations organization-05, .;, STA, R0, NLO) as long as the total number of onshift personnel is sufficient for the safe shutdown of the reactor (s) and the fire brigade.
2) The words "is in the containment building of the affected reactor and" were inserted to provide further definition of movement limitations necessary to meet the intent of " observed and directly supervised."

3) The deletion of 6.2.2.e was made on the basis that " personnel required for other essential functions during a fire emergency" is not well defined and the three members of the minimum shift crew will be able to achieve safe shutdown.

4) The time interval of 2 hours in footnote
was revised to 4 hours on the basis of plant location and personnel availability.
6.2.2.f: The words " normal 8-heur day" were deleted and reference to a 12-hour shift schedule was added to be consistent with plant policy. The intent of j Generic Letter 82-12 is maintained.
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[ " Manager, Plant Operations" was revised to " Operations Superintendent" to be consistent with the plant-specific organization. The worcs "or his deputy" were deleted to be consistent with plant policy. The words "who perform" were replaced with "in performance of" in order to ensure that hours worked in performance of nonsafety-related functions are not included in the scope of this specification. The term " health physicists" was replaced with ' Key Health Physics Technicians" and " auxiliary operators" was replaced with " key non-licensed operators" in order to better define the word " Staff." In addition, certain health physics technicians and nonlicensed operators may perform work that is not readily subject to degraded quality due to extended hours. The words "[ Plant Superintendent] or his deputy" were replaced with
" applicable department superintendent" to be consistent with paragraph 13.5.1.1.G of the VEGP FSAR. In addition, this paragraph was revised to clarify what the General Manager-Vogtle Nuclear Operations reviews with regard to excess overtime.
Although 12--hour shift schedules are in conflict with a " normal 8-hour day," these schedules meet the intent of the total number of hours worked during the 6 weeks rotating shift period. Fatigue should not be a factor as the 12-nour shift schedules allow for several consecutive days off during the 6-weeks rotating shift period. O v Tabic 6.2-1: > The time interval for unexpected absences was revised from 2 hours to 4 hours on the basis of plant location with reference tc perscenel
, availability.
The format of Table 6.2-1 was revised for the sake of readability. The content is unchanged. 6.2.3: The words " Licensee Event Report" were deleted from paragraph 6.2.3.1 te in consistent with item I.b.1.2 of NUREG-0737. Also, both the PRB and the 3B review Reportable Events. The words " including units of similar design" were deleted on the basis that "NRC issuances, industry advisories, and other sources of plant design and operating experience information" will include any information regarding units of similar design. Also, the deleted phrase is not included in NUREG-0737, item I.B.1.2. D (a
[~'T The words " located on site" were deleted since the five individuals will be
\~ / a combination of corporate and onsite personnel. This is consistent with NUREG-0737, item I.B.1.2. (NUREG-0737, item I.B.I.2 provides for a smaller onsita group when a utility has multiple sites.)
6.3: This statement is censistent with VEGP FSAR paragraph 13.1.3.1. 6.4.1: This specification has been revised to be consistent with subsection 13.2.2 of the VEGP FSAR. 6.5.1.2: This specification has been revised to be consistent with subsection 13.4.1 of the VEGP FSAR with the exception that we no longer have the title Vice-Chairman of the PRB. The chairman has an alternate who shall be cesignated by the GMVNO. 6.5.1.6: s-/ a) & b) The VEGP PRB reviews the administrative and controls required to develop plant procedures. This should be adequate to ensure that
~
acceptable procedures are generated. The VEGP PRB responsibilities are listed in subsection 13.4.1 of the VEGP FSAR. The revisions propoeed here are consistent with paragraph 13.4.1.B of the FSAR. d) " Appendix A" was deleted on the basis that the distinction between Appendix A and B is no longer warranted due to the incorporation of NUREG-0472 into NUREG-0452 and the redesignation of Appendix B as the
" Environmental Protection Plan."
h) This sp2cification was revised to be consistent with subsection 13.4.1 of the VEGP FSAR.
1) This specification was revised to be consistent with 13.4.1.k of the VEGP FSAR.
6.5.1.7: a) The change "(a) throug5 (d)" to "(a) through (e)" was made because of the change made to item 6.5.1.6.a. b) The change "(a) through (e)" to "(a) through (f)" was made because of tne change made to item 6.3.1.6.a. fh
\_ >
l
l l [) The " Nuclear Review and Audit Group" will be known as the " Safety Review Board." These changes are specific to the organization of GPC and the VEGP. 6.5.2: The Company Nuclear Review and Audit Group (CNRAG) will be known as the Safety Review Board (SRB). This revision has been made throughout the draft of Section 6.0. 6.5.2.2: The wording supplied in this section provides a better description of the composition of the SRB than does the STS and is consistent with subsection 13.4.2 of the VEGP FSAR. 6.5.2.3 and 6.5.2.4: The wording supplied in this section is consistent with subsection 13.4.2 of' the VEGP FSAR. 6.5.2.6: The wording supplied in this section is consistent with subsection 13.4.2 of the VEGP FSAR. 6.5.2.8: . The insert for page 6-11 was added to provide some flexibility for inspections and audits. 6.6.1: Reference to Section 50.72 was added because it requires immediata notification and the word "or" was added because a written followup report may not be required. 6.7.1: , Addition of PRB and SRB is for clarifica*. ion in that these bodies review Technical Soecification violations. (l u
[~N. 6.3.2:
\~ >
This section was revised to reflect the plant-specific method of review and approval of VEGP procedures and changes thereto. 6.8.3: b) VEGP Units 1 and 2 will be essentially identical and the operations personnel licensed on Unit I will be licensed on Unit 2 also. c) See the discussior for 6.8.2. 6.8.4: a) The system designations were revised to be consistent with plant-specific terminology. The word " integrated" was deleted to eliminate any confusion of this test with the containment ILRT requirements. c) (3) At VEGP, monitoring can be performed prior to discharge. This
<< is an accurate method for determining evidence of condenser in-leakage.
4 ~g Old item d) This item was deleted on the basis that VEGP is equipped ((.'/ with more than one channel of monitoring instrumentation. 6J: See justifications for .evisions to Specs. 3.4.8, 4.7.9.3, and 4.8.1.1.3. 6.9.1.3: This annual report which documents radiological environmental monitoring activit;es for the calendar year might more appropriately be entitled the Annual Radiological Environmental Surveillance Report; this title is used at our other nuclear plant. The addition to the first paragraph was made to emphasize the fact that the activities of the radiological environmental monitoring program are being reported rather than operation of the plant per se. In the second paragraph the words "as appropriate" should apply to all of the items that might be specifically reported. Also, it is more appropriate to say "any observed impacts" as opposed to "the observed impacts" since impacts of plant operation are often miniscule, and thus, difficult to detect and confirm. Finally, the tenn " land use survey" is believed to be more appropriate than " land use census." ((3_/ 3
h In the third paragraph, the inserted sentence is intended to prevent cluttering of the report with irrelevant data so that significant results will not be masked. The word "possible" was replaced with " practicable" on the basis that the submittal of missing data does not warrant the highest priority, especially when it may not be possible to substantiate that the priority was achieved. In the last paragraph, the necessary maps will be provided in the reports and their number need not be specified in the Technical Specifications. Also, all area maps have been constructed with the midway point between the two reactors as the origin. The centers of the two reactors are a few hundred feet apart. 6.9.1.?: The addition of the words " Historical annual average meteorological conditions or" is based on the fact that Section 3.3 of NUREG 0133 allows the use of historical annual averaga meteorological conditions for routine gaseous dose calculations. The word " nearby" as it modifies " uranium fuel cycle sources" was deleted and the phrase "within 8 km" was inserted to provide clarification. This is consistent with the bases for Specification 3/4.11.4. l 6.10.2, item "i": This item was removed from Specification 6.10.3 and placed here because we feel that in a 5 year timeframe a sufficient i. mount of data will be available for us to identify any negative trends in secondary water quality. 6.10.3, item "i":
"0A Manual" was replaced with "FSAR" because Chapter 17 of the VEGP FSAR governs VEGP's Quality Assurance activities.
6.10.3 item "m": See the justification for 6.10.2, item "1" for deletion of old item "m." 6.12.1: Inis section was revised to clarify the requirements of the STS. These changes are consistent with the requirements of paragraph 12.3.1.2 of the VEGP FSAR and section 12.3.1 of the SER. n
.- . .. _ . _ _ . . - - _ .- ~.
6.12.2: The words ". .. equal to or less than 1000 mR/h. . ." were changed to
"... greater than 100 mR/h but less thet 1000 mR/h..." to be consistent with the definition of high radiation area provided in Part 20.202 of 10 CFR.~
6.13.1 (STS): This specification was deleted because the NRC has informed VEGP that we must have an approved PCP prior to receipt of a license. This direction was. provided per a telephone conference call at 2:00 p.m. on 3/19/85. Therefore, 6.13.1 will be obsolete at the time of Technical Specification approval by the NRC, i.e., receipt of the VEGP Unit 1 License. 6.13.1.b: the words " acceptance by the PRB" were deleted on the basis that the PRB ] makes recommendations to the GMVNO for approval of changes to .the PCP. 1 6.14.1 (STS): This specification was deleted because the NRC provided written direction in their " Summary of September 11, 1984, Technical Specification meeting for O . Vogtle, Units 1 and 2" that ". . . the offsite dose calculation manual must-be approved by the staff prior to license issuance . . ..". Therefore, 6.14.1 will be obsolete at the time of Technical Specification approval by-1 NRC, i.e., receipt of the VEGP Unit 1 License. 6.14.1.b: i The words " acceptance by the PRB" were deleted on the basis that the PRB makes recommendations to the GMVNO for approval of changes to the ODCM. i . 6.15.1.a.8: The words "by the PRB" were deleted on the basis that the PRB makes reccmmendations to the GMVNO for approval / disapproval. 6.16: The justification for adding a specification to specify the date for which ' Technical Specifications changes are ta be implemented is that these type of i changes' affect procedures and programs and time will be required to
implement the changes (i.e. , incorporate them into procedures and programs
, and train personnel). Also, it will take time to get all of the copies of the Technical Specifications posted to ensure that personnel are aware of the changes. 0054v l r I . l [
x
.NUREG-xxx !
t i 1 i i1 nr i i I' i ' i TECHNICAL SPECIFICATIONS l' 'V0GTLE ELECTRIC GENERATING PLANT i : 1 4 UNIT 1 i 1 L DOCKET NOS. 50-424 !- APPENDIX "A" to i ! LICENSE NO. NPF-1 I ,
i issued by the U.S. Nuclear Regulatory Commission !
Office of Nuclear Reactor Regulation t i ' t 4 i 8
.c l
1 l ' i i i e I- f
INDEX O
\s_,l DEFINITIONS SECTION PAGE.
1.0 DEFINITIONS 1.1 ACTI0N........................................................ 1-1 1.2 ACTUATION LOGIC TEST.......................................... 1-1 1.3 ANALOG CHANNE L OP ERATIONAL TEST. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-1 1.4 AX I A L F LUX D I FF EREN C E . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-1 1.5 CHANN E L CA LI B RATI ON . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-1 1.6 CHANNEL CHECX................................................. 1-1 1.7 CONTAINMENT INTEGRITY......................................... 1-2
,-.g,-- ,_, .- -mm.,,. .-,
munnnum............................................ .m
..o uva.avuuus 4 71r9 CO R E A LT E RAT I O N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-2 /.?lr10 DOSE EQUIVALENT I-131........................................ 1-2
i. <c 3=41 E-AVERAGE DISINTEGRATION ENERGY.............................. 1-2

i. <f 1r12 ENGINEERED SAFETY FEATURES RESPONSE TIME.....................
1-3 1.f; 1r13 FREQUENCY N0TATION........................................... 1-3
,I 1.14 IDENTIFIED LEAKAGE........................................... 1-3
(~'S (-/ '
1. 15 MASTER RELAY TEST............................................ 1-3 1.16 MEMBER (S) 0F THE PUBLIC...................................... 1-3 1.17 0FFSITE DOSE CALCULATION MANUAL.............................. 1-3 1.18 OPERABLE - OPERABILITY....................................... 1-4 1.19 OPERATIONAL MODE - M0DE...................................... 1-4 1.20 PHYSICS TESTS................................................ 1-4 1.21 PRESSURE BOUNDARY LEAKAGE.................................... 1-4 1.22 PROCESS CONTROL PR0 GRAM...................................... 1-4 1.23 PURGE - PURGING.............................................. 1-4 1.24 QUADRANT POWER TILT RATI0.................................... 1-5 1.25 RAT ED TH E RMA L P0WE R . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-5 4
1.25 REACTUR TRIP SYSTEM RESPONSE TIME............................ 1-5 1.27 REPORTABLE EVENT............................................. 1-5 W_ -54r. e ,.....
, m , , n t. u~r .m ,ue.e .m F. ,."..................................... A _-
a
~
e.afir29- S HUTDOWN MARG I N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-5 e.4?4244 SITE B0UNDARY................................................ , 1-5
/.2cir31 S LAV E R E LAY T EST . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-6
1. i3 j dSEcus .sas7E Piloc.Es soxr sys7 st 1.3 V0GTLE - UNIT 1 I
INDEX (' DEFINITIONS PAGE SECTION 1-6
! /-3/ h-32 SOLIDIFICATION:..............................................
l 1-6 i 3.t M3 SOURCE CHECK................................................. 1-6 I
/ 53 M4 STAGGERED TEST BASIS.........................................
1-6 f . 3+ MS THERMAL P0WER................................................ 1-6 i 35MG TRIP ACTUATING DEVICE OPERATIONAL TEST....................... 1-6
/.n i-37 UNIDENTIFIED LEAKAGE......................................... .
1-7 j.s y MS UNRESTRICTED AREA............................................ i 1-7 i 511-39 VENTILATION EXHAUST TREATMENT SYSTEM......................... 1-7 I
/ df 1 VENTING...-.................................................. ..,2 L.os.m . r-- C,. e+, u, ~n--.
i n i i.n .e.v. .e.r. . .eu.
! 1-8 TABLE 1.1 FREQUENCY N0TATION......................................
1-9 TABLE 1.2 OPERATIONAL M0 DES....................................... I i a i f I J w V0GTLE - UNIT 1 II
-- --,,-% y _ .-, ,- +me . ~ . m ,,_, , - . - - + , -,,,.,sw w.%wy--,.. , _ , . , - . . , .y+ pr-,.__., -.r-. .- m., , , . , . - - - , . _ . . - - -
l INDEX l SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS PAGE SECTION r 2.1 SAFETY LIMITS 2-1 2.1.1 REACTOR C0RE................................................ 2-1 2.1.2 REACTOR COOLANT SYSTEM TRESSURE............................. t
!." OP:n T!O." 2-2 4
FIGURE 2.1-1 REACTOR CORE SAFETY LIMIIS- TOL': LOOPS THREE LOOPS II; CFERATIO". 20 FIGURE 2.1-2 REACTOR CCRC CATCTY LIMIT 2.2 LIMITING SAFETY SYSTEM SETTINGS 2-4 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETP0lNTS. . . . . . . . . . . . . . . 1 2-5 j TABLE 2.2-1 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS.... i 1 i BASES l PAGE
< = SECTION i
2.1 SAFETY LIMITS a B 2-1 i 2.1.1 REACTOR C0RE................................................ 8 2-2 2.1.2 REACTOR COOLANT SYSTEM PRESSURE............................. ' i f
2. 2 LIMITING SAFETY SYSTEM SETTINGS
, ..,. B 2-3 ! 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETP01NTS. . . . . . . . . . . . . . . i' ! \ ! III ! V0GTLE - UNIT 1 i i __ - . . _ _ _ .~. . . _ .__ _ _ . _ . _ . . . . _ _ _ . . _ . _ . . _ . _ , _ . _ _ . , - - . _ . ~ . , _ . . . _.._,_.
l INDEX n b' LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS PAGE SECTION 3/4.0 APPLICABILITY............................................... 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS i 3/4.1.1 BORATION CONTROL t Shutdown Margin - T avg Greater Than 200*F................ 3/4 1-1 Shutdc i Margin - T Less Than or Equal to 200 F....... 3/4 1-3 avg Mode ntor Temperature Coefficient........................ 3/4 1-4 . Minimum Temperature for Criticality...................... 3/4 1-6 3/4.1.2 BORATION SYSTEMS Flow Path - Shutdown..................................... 3/4 1-7 Flow Paths - Operating................................... 3/4 1-8 Charging Pump - Sn.stdown................................. 3/4 1-9 i Charging Pumps - Operating............................... 3/4 1-10 , Borated Water Source - Shutdown.......................... 3/4 1-11 i \ Borated Water Sources - Operating........................ 3/4 1-12 3/4.1.3 MOVABLE CONTROL ASSEMBLIES i Group Height............................................. 3/4 1-14 TABLE 3.1-1 ACCIDENT ANALYSES REQUIRING REEVALUATION IN THE 3/4 1-16 EVENT OF AN INOPERABLE TULL L"%cMntos. en SHurpeuMMCT" R00.................I. Posi tion Indi cation Systems - Operating. . . . . . . . . . . . . . . . . . 3/4 1-17 Position Indication System - Shutdown.................... 3/4 1-18 Rod Drop Time............................................ 3/4 1-19 Shutdown Rod Insertion Limit............................. 3/4 1-20 Control Rod Insertion Limits............................. 3/4 1-21 FIGURE 3.1-1 RCD BANK INSERTION LIMITS VERSUS THERMAL POWER i:UR LOOP 0P:P", TION...................................... 3/4 1-22 FMURE-3rl-240-BAMK-INSERHON-MMITS-VERSUS-THERMAL POWER THREE-t00P 0PERATION..................................... 3/4 1-23 4 1 l l w VCGTi.E - UNIT 1 IV 4 1 -
INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX DIFFERENCE.................................... 3/4 2-1 FIGURE 3.2-1 AXIAL FLUX DIFFERENCE LIMITS AS A FUNCTION OF RATED THERMAL P0WER...................................... 3/4 2-3 3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR.".k(4). . . . . . . . . . . . . . . . . . 3/4 . . . 2-4 FIGURE 3.2-2 K(Z) - NORMALIZED F (Z) AS A FUNCTION OF. CORE HEIGHT. 3/4 2-5 q 3/4.2.3 -20', FLC':! o4TE ".O NUELEAR ENTHALPY Rf5t HOT CHANNEL FACT 0 R . .p y. p . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 2-8 FICURE 3.2-3---RCS-TOTAt-FEOW RATE VERSUS-R - FOUR LOCPS IM 1PERATION............................................. 3/4 2-0 3/4.2.4 QUADRANT POWER TILT RATI0................................ 3/42-1-1/o 3/4.2.5 DNB PARAMETERS........................................... 3/4 2-14/J 1 TABLE 3.2-1 DNB PARAMETERS........................................ 3/4 2-M /t 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION...................... 3/4 3-1 TABLE 3.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION................... 3/4 3-2 TABLE 3.3-2 REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES.... 3/4 3-9 TABLE 4.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS............................................. 3/4 3-11 3/4.3.2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION........................................ 3/4 3-16 TABLE 3.3-3 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION.......................................... 3/4 3-18 TABLE 3.3-4 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETP0INTS........................... 3/4 3-M.29 TABLE 3.3-5 ENGINEERED SAFETY FEATURES RESPONSE TIMES............. 3/4 3-M & TABLE 4.3-2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILL\NCE REQUIREMENTS................ 3/4 3 +I 3/4.3.3 MONITORING INSTRUMENTATION Radiation Monitoring For Plant Operations................ 3/4 3-+7 4 i O V0GTLE - UNIT 1 V
1 INDEX i LIMITING CONDITIONS FOR'0PERATION AND' SURVEILLANCE REQUIREMENTS SECTION PAGE TABLE 3-3-6 RADIATION MONITORING INSTRUMENTATION FOR PLANT OPERATIONS..................................... 3/4 3 4-7 TABLE 4.3-3 RADIATION MONITORING INSTRUMENTATION FOR PLANT OPERATIONS SURVEILLANCE REQUIREMENTS..................... 3/4 3-50 41 : Movable Incore Detectors................................. 3/4 3-51 s o Seismic Instrumentation.................................. 3/4 3-52 Si TABLE 3.3-7 SEISMIC MONITORING INSTRUMENTATION.................... 3/4 3-53 5A TABLE 4.3-4 SEISMIC MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS............................................. 3/4 3-54 SJ Meteorological Instrumentation........................... 3/4 3-55 5+ l TABLE 3.3-8 METEOROLOGICAL MONITORING INSTRUMENTATION............. 3/4 3-56 65 TABLE 4.3-5 METEOROLOGICAL MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS............................................. 3/4 3-5-756
^ Remote Shutdown !a;t 1 tti: .f/ ATE.W. . . . . . . . . . . . . . . . 3/43-5867 TABLE 3.3-9 REMOTE SHUTDOWN MONITORING INSTRUMENTATION............ 3/4 3-59-62 4
TABLE 4.3-6 REMOTE SHUTDOWN MONITORING INSTRUMENTATION i SURVEILLANCE REQUIREMENTS................................ 3/4 3-60-6 f Accident Monitoring Instrumentation...................... 3/4 3-616 0 TABLE 3.3-10 ACCIDENT MONITORING INSTRUMENTATION.................. 3/4 3-62(of TABLE 4.3-7 ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS............................................. 3/4 3-44-(,3 Chlorine Detection Systems............................... 3/43-6664 Fhe-setect i c a I ns tru. ;; ntati on. . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 3-07 l T*1LE 3.3-11 FIRE DETICTICM IMSTR',' MENT",T:0M. . . . . . . . . . . . . . . . . . . . . . 2/i 3-50 Loose-Part Detection System.............................. 3/43-4966 Radioactive Liquid Effluent Monitoring Instrumentation... 3/4 3-70 66 i si ! TABLE 3.3- W RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION 3/4 3- M 47 > TABLE 4.3-8 RADIOACTIVE LIQUID EFFLUENT MONITORING 4 INSTRUMENTATION SURVEILLANCE REQUIREMENTS................ 3/4 3-74-70 Radioactive Gaseous Effluent Monitoring Instrumentation.. 3/4 3-N75 rL TABLE 3.3- M RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION.......................................... 3/4 3-76 7f TABLE 4.3-9 RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS................ 3/4 3-6690 l < 4 3/4.3.4 TURBINE OVERSPEED PROTECTION.............................. 3/4 3-4187 (~ Mi L*n e BrW. Z.fola.h'en Senso r.r . . . . . . . 3/4 3-65 Mynsu 3.$h - aM Enere58 orney uar 6xrn msrmurAnod.. 3/+ s-eto V0GTLE - UNIT 1 VI i
INDEX b
~# LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS PAGE SECTION 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION Startup and Powe r Ope rati on. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 4-1 Hot Standby.............................................. 3/4 4-2 Hot Shutdown............................................. 3/4 4-3 Col d Shutdown - Loops F i11 ed. . . . . . . . . . . . . . . ,l . . . . . . . . . . . . 3/4 4-5 Cold Shutdown - Loops Not Filled........ ................ 3/4 4-6 - 4sohted-Leop............................................ 3/' '-7 4 sele ted-Loop-Sta rtup . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/' '-8 3/4.4.2 SAFETY VALVES Shutdown............................................... 3/4 4-9 7 0perating............................................. 3/4 4-107 ~
3/4.4.3 PRESSURIZER.............................................. 3/4 4-13 9 3/4.4.4 RELIEF VALVES............................................ 3/4 4-1410 3/4.4.5 STEAM GENERATORS......................................... 3/4 4-1+ t L TABLE 4.4-1 MINIMUM NUMBER OF STEAM GENERATORS TO BE INSPECTED DURING INSERVICE INSPECTION............................. 3/4 4-19 I7 TABLE 4.4-2 STEAM GENERATOR TUBE INSPECTION....................... 3/4 4-20 lf 3/4.4 6 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems................................ 3/4 4-3117 Operational Leakage...................................... 3/4 4-22 00 TABLE 3.4-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES...... 3/4 4-444.A 3/4.4.7 CH EM I ST RY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 4-26 43 TABLE 3.4-2 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS............... 3/4 4-96.24 TABLE 4.4-3 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS SURVEILLANCE REQUIREMENTS............................................. 3/4 4-27.16 3/4.4.8 SPECIFIC ACTIVITY........................................ 3/4 4-tt.2 0 FIGURE 3.4-1 DOSE EQUIVALENT I-131 REACTOR COOLANT SPECIFIC l ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL POWER WITH THE REACTOR COOLANT SPECIFIC ACTIVITY > 1pCi/ gram DOSE EQUIVALENT I-131.................................... 3/4 4-30 17 l TABLE 4.4-4 REACTOR COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS - l PR0 GRAM.................................................. 3/4 4-3-112 l i O V0GTLE - UNIT 1 VII
INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS PAGE SECTION 3/4.4.9 PRESSURE / TEMPERATURE LIMITS Reactor Coolant System................................... 3/44-3-35o pnessutr-renewruaE DMiT5 w'aws i.e*fm FIGURE 3.4-2 REACTOR COOLANT SYSTEM "EAiur L:"IT/.T!O"5 - A wo eo. W e e n e w _ EFPV. 4#.GA9? ) passw f.YR4MM."c ??.8T.W"'~ 3/4 4-3431 c AFFLICABLE UF TG
'F"E u m 4#
FIGURE 3.4-3 REACTOR COOLANT SYSTEM feGE90WM LIMI STIOM " MPPtifABLE UP TO Ei r'//- m s T.S 't?.*. W . f n ! W .. " f7 f 5 3/4 4-35r3 t TABLE 4.4-5 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM - WITHDRAWAL SCHEDULE...................................... 3/4 4-3633 Pressurizer.............................................. 3/4 4-97 3+ Coed.0ver s t e, r.s . . , . . . . . . . . . . . . . . . . . . 3/4 4-38 3 6 ur vst mane f c sisuru 3/4.4.10
3. .s-+au4n.au.u pressure Aucwas ProtectionSTRUCTURAL _ _ _ - .
INTEGRITY 4 9W.3 ?.v d s.4-s44 f/ 3/4.4.11 REACTOR COOLANT SYSTEM VENTS............................. 3/4 4-41-39 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.S.1 ACCUMULATORS............................................. 3/4 5-1 3/4.5.2 ECCS SUBSYSTEMS - T,yg GREATER THAN OR EQUAL TO 350 F.... 3/4 5-5.5 3/4.5.3 ECCS SUBSYSTEMS - T . ..... 3/4 : 0
*?9 LESS - . - THAN - . - -- 350*F_. -. . - . - - - - - - 3/45-7 fcc.5 Msp% - - - 3/+ S
9 Sde 6 Z,iee-hen Amy.s . - - - - - - - - - - - - -
/t C.4--BOROW-!!tJECTI0n 5Y.2 4 cn Beren -I nje c tion-Tan k . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 5-11 Heat-Tracing . . . . . . . . . . . . . . . . . . . . . . . . . .................. 3/4-5-12 s
REFUELING WATER STORAGE TANX............................. 3/4 5-13 /C 3/4.5.) V0GTLE - UNIT 1 VIII . t%
1 t i. INDEX a O A LfMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE o _
,., . .im,. .,, .m.i. .- o n , v.uor ..,, -u.,.- .,,~ .,. .mm u, ,
3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIYARY CONTAINHENT Containment Integrity.................................... 3/4 6-1 Containment Leakage........................, ............ 3/4 6-2 Conta i nmen t Ai r Loc ks. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 6-5 Gente fement-Isolat ion -Val ve-and-Channe Hhrhi Pres suri-z ati :n Sy;t e. .; . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2/' S-7
o rerm a e., swe . . . .
_da u-1 3/4 6-4f Air Temperature.......................................... l Containment Structural Int gri ty. . . . . . . . . . . . . . . . . . . . . . . . . 3/4 6-19 9 Containment Ventilation System......................'..... 3/4 6-p /2. 3/4.6.2 OEPRESSURIZATION AND COOLING SYSTEMS j Containmsnt Spray System................................. 3/4 SE /+ Spray Additive System.................................... 3/4 6-42-/5 0% Containment Cooling System............................... 3/4 6-23 16 A 2 S a f.
.,,f,M . . . . , ,9.% l' t F,A,.k.tn.i ?9 f,VSTEM. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ,. R, .T .t.t r. m m .e .n.e, 3/4. 6.,i' CONTAINMENT ISOLATION VALVES............................. 3/4 6-B-/ 7 i TABLE 3.6-1 CONTAINMENT ISOLATION VALVES.......................... 3/4 6-19 /7 4
3/4.6.,5 COMBUSTIBLE GAS C0riTROL Hydrogen Monitors........................................ 3/4 6-3022. Electric Hydrogen Recombiners............................ 3/4 6-n 2.3 Hydrogen Purge-Cleanup-System..........rn............... 3/4 6-32 Hyd rogen-Mi x i ng- Sy s tes., .1rn . . . . cn mm. . . . . . . . . . . . . . . 3/4 6-34 3/4r64--P ENETRATION-R00M-EXHAUST-AIR -CLEANUP -SYSTE". . . . . . . . . . . . . . 3/4 5-35 j -3/4dr7-VACUUM-R EH EF-VA LVES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 5-37 k l l. n v l VCGTLE - UNIT 1 IX _. . . , .. . . . . . . - - . ~
INDEX (n) v LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE Safety Va1ves................................... ........ 3/4 7-1 TABLE 3.7-1 "AXIMU" ALLOWAOLE POWER PJNGC NCUTRON FLUX "!C" SETPGINT-WETH-INOPERABLE-STEA" LINE SAFE"I VALVES DURI'G FCUR LOOP OPERATION. .F.W.dM.?'.M .#f.58~YXMF.5. Afd.Mo# 3/4 7-2 TABLE 3.7-2 MAXIMUM ALLOWABLE POWER RANGE NEUTRON FLUX HIGH SETPOINT WITH INOPERABLE STEAM LINF,tVALVES DURING THREE+ LOOP OPERATION...................5.A.TffY................. 3/47-2-3 TABLE 3.7-3 STEA". LINE SAFETv VALVES PER LOOP.. . 3/A 7-3 Auxiliary Feedwater System............................... 3/4 7-4 Condensate Storage Tank.................................. 3/4 7-6 Specific Activity........................................ 3/4 7-7 TABLE 4.7-1 SECONDARY COOLANT SYSTEM SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PR0 GRAM..................................... 3/4 7-8 Main Steam Line Isolation Va1ves.......................:. 3/4 7-9 3/4.7.7. STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION.......... 3/4 7-10 3/4.7.3 COMPONENT COOLING WATER SYSTEM........................... 3/4 7-11 3/4.7.4 xue SERVit n x sesuNM.e. .e. . caous
< Ee WATER-SY5 . . . . . . .................. .ya reiz.(ais ew) -sT.srgM ..... 3/4 7-12 3/4.7.5 ULTIMATE HEAT SINK....................................... 3/4 7-13 W4-7. S 3/4 7 14 - - FLOCO PR0TECTICN..................d.
n ru ..................... 3/4. 7.3., r CCNTROL ROOM EMERGENCY m A!yvugA"UP CL. SYSTEM. . . . . . . . . . . . . . . . 3/4 7-15/+
-s 3/' "' . S ECG-S-PUMP-ROGM-EXHAUST-AIR-GEEANUP SYSTE". . . . . . . . . . . . . . . . 3/4 7-10 er 3/4.7.9 SNUBBERS................................................. 3/4 7-2917 i FIGURE 4.7-1 SAMPLE PLAN 2) FOR SNUBBER FUNCTIONAL TEST........... 3/4 7-26.Lt 3/4.7.lb SEALED SOURCE CONTAMINATION.......................... ... 3/4 7-2 M 5 L 3/a - 1 ftPiAh5 PE,v EiTLA rsc.V A 454 PH- 772 /*ric,t/ Ana al+ 1-17 5;rHAv.sr sys7EsA nv V0GTLE - UNIT 1 X
INDEX O V LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS PAGE SECTION
.,n....., , , r , , ,- ,, . , -..nues - - - , - wn , - .o .~ .....
s . mo
" r: uppre::i c . L20t r Sy:te.r... ..... ........... 3/* 7-20 Spray and/or Sprinkler Systa=0...................... .. . 3/" 7-31 ~. ,.
ww: a,m,sw___.............................................. ai , , sa
*E
9. ,/,. 9, a .,
u, ~* i. .n a.
. e.,s,,4..-...,..
s..
.e... .u.,.,.. .. .. .e. . . s. ,.,..... , 1, ,, .,e ,..., s,. ,, ,. -,mr ,mer ,,..,.,m,, .,.n .un,................................ ... , n, ,_,, .n.. ..a . ,v e m Yard c're "ydr:nt: :nd Hydr:nt-He:: "cu::s............... 3/4 7--38 fl v s o.n. e ,..s..e uv. n.. o. n.u,.e .u ie .e.n. e. vn i,.re, ,mer e a m u.,.- uws. o "m "v a ~ .
mu.m nnn. si , < o-- I
m. .. ,. e ., . , _ e. . . ... . . ..r. m . m. o
.,n,...-,4-10 .e v .. ..en. , r, naaw.m merru er.r. .................................... ai ,- 1 m u-3/4.7.r3 AREA TEMPERATURE MONITORING.............................. 3/4 7-4t 27 "
TABLE 3.7-6 AREA TEMPERATURE MONITORING........................... 3/4 7-4-32.7 3/4.1.ll 5 ^!&tNE3/LGD s;4peyy pgarurzg,$ (egp);tes,u cxign y/4 7- 2.9 e.) AND S4FP7Y-(2 tArwD cMiLLErz. SysrsA1 1 3/4.8 ELECTRICAL POWER SYSTEMS G b 3/4.8.1 A.C. SOURCES 0perating................................................ 3/4 8-1 TABLE 4.8-1 DIESEL GENERATOR TEST SCHEDULE........................ 3/4 8-8 Shutdown................................................. 3/4 8-9 3/4.8.2 D.C. SOURCES 0perating................................................ 3/4 8-10 TABLE 4.8-2 BATTERY SURVEILLANCE REQUIREMENTS..................... 3/4 8-12 Shutdown................................................. 3/4 8-13 3/4.8.3 ONSITE POWER DISTRIBUTION 0perating................................................ 3/4 8-14 Shutdown................................................. 3/4 8-16
~ ., 3 s , 7. n. .? :*R 7n vt ce cc,v,7 l'?;g merz. stas. gysp.i giz cm.,,;y 31.s. 7 3o w rDL /GovWi >.',/
VCGTLE - UNIT 1 XI
- ,, ~-. ._
J' $
(. . __. _ f INDEX g LIMITING CONDITICNS y OPERATION AND SURVEILLANCE RE0UIREMENTS PAGE SECTION 3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES A.C. Circuit: Inside o- N ry Ceatai-ment. . . . 9-17 Containment Penetration Conductor Overcurrent /7 Protective Devices.4t'.4.fr.'. Mr. .&f.Hers. .fe.Drhr'~ 3/4 8-10 j r
, n < ms now.e. e n 1 to V ch s s i; a pyme,,7 7,.,o-,,...rm,. .t.. -r..,,..-.. ---m,,.- - - - - - . . - - - - ,~_.m, .-,r , . . - n mma , < ma m . am 4 v n wunuww un vvenuunncni ,/9 (") _ 89 0 D U. . N. T.
T..U.O.f CU. T. f.*
.. f. .C . . .. .. .
Bypa si Dw'ce.1
% & .- .&we.LMotor-Operated Valves Thermal Overload Protection.j . . . . . . 3/4 8-M/9 ,or ~ .. - -~m - - - - - - , . . - ,~,,, ,.m_,m ,un ,. , ,m - , r .,un, m eu , u. mr e .unonu , , r . e, , ,.nm v u nuuna rnviuw.iun .,n o_.,-. . e, m n o .- , ..,.- v.- . ,.7 e, - n.eu...ee. - .. .... . .... . .. .... . .
3/4.9 REFUELING OPERATIONS 1 3/4.9.1 BORON CONCENTRATION. .................................... 3/4 9-1 3/4.9.2 INSTRUMENTATION.......................................... 3/4 9-2 3/4.9.3 DECAY T!ME. ... ......................................... 3/4 9-3 j i' 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS........................ 3/4 9-4 3/4.9.5 COMMUNICATIONS........................................... 3/4 9-5 gyunius/c, ::!,<.m e" 3/4.9.6 Matt! Ptf TAT O R-C RArtt . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 9-6 M CAS . 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGEf"CL "UIL4MG. . . . . . . . . . 3/4 9-7 3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION High Water Level.................... .................... 3/4 9-8 Low Water Level.................-........................ 3/4 9-9 O VCGTLE - UNIT 1 XII
)
INDEX _) LJMITING CONDITIONS FOR OPEAATION AND St3RVIILLANCE SEQUIREMEfiTS PAGE
.S_ECT~ 0N y'c.a/rr JA TMu 3/4 9-10 3/4.9.5 CONTAINMENT +tEI 50 EFE^?S-T ISQLATION SYSTEM. .. .. . . .. . .
2;': ; 11 3/4.9.20 VATER LEVEL - SEACTOR VESSEL-
. . . . . . . . . 3p pit
ser ifrumbke .s. .. . . . . . . . .
. . . . . . . . . . . . /+ 9-f b C c e rrv l A'e d S . .e y/4 3 9 -ML / 5 3/403.21 VAT ER LEVE L - STDuTE PO OL . . . . . . . . , . . . . . . . . . . . . . . . . . . . . .
M4a2 -FUEL- 5%4E-700L-AIR-CLEANOR-S%5TO' .. .e - ......rvr 7/4 Pl?- _I/4.13 SPECIAL TE.C EXCEPTIONS 3/3.10.1 SHOTC04N MAPGIN.......................................... 3/4 10-1 3/'.10.2 GRO.'1? HEIGHT.1%5ERTION, WO POWER DISTRIBUTION LIMITS... 3/4 10-2 3/4.10.3 PHYSICS 7ESTS............................................ 3/4 10-3 3/4 10.4 REACTOR COOLANT 1,0C?$...................... .... ........ 3/4 10-4 2/4.10.5 FOSITIO N INDICATION $ (5 TEM - SHbTC0W. . . . . . . . . . . . . . . . . . . . 3/4 10-5 2c: 11 M MCT.!YE g-- EFFUJENTS 3/ :1L 1 LI^e1C EFFLUEN!S Ccreentation............................................ 3/4 11-1 TMLE 4.11-1 SA010 ACTIVE LIQ'JIO VASTE SAMLING rND ANALYSIT e FRC:WAM.................................................. 3/4 11-2
a. ................................................ 3/4 11-5
. Liq. : aucwasta Treateent Syst.am... ..................... 3/4 11-6 L i c m d 16 wp T o .1k s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 11-7 A
r J. VCGILE - UNIT . XIII I l
INDEX
^-
Q D
-LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS =
PAGE SECTION { 3/4.11.2 GASEOUS EFFLVENTS Dose Rate. ............................................ . 3/4 11-8 4 I TABLE 4.11-2 RADI0ACTJVE GASEGUS WASTE SAMPLING AND ANALYS153/4 11~9 PR0GQM...........-...................................... 3/4 11-12 Oose - Nobi. Gases...........,..............<............ . ' Dose - Iodine-131, Iodine-133, Tritium, and Radioactive 3/4 11-13 Material in Particul ate form. . . . . . . . . . . . . . . . . . . . . . . . . .. . . Gaseous Radwaste Trestrent 5ystem........................ 3/4 11-14 Explosive Gas Aixture.................................... 3/4 11-15 Da e.a 3/4 11-16 Gas M arage Tanks........................................ 3/4 11-17 3/4.11.3 SOLID RADI0 ACTIVE WASTE 5.................................
........... 3/4 11-18 3/4.11.4 TOTAL 00SE...................................
i 3/4.12 RADIOLOGICAL ENVINONMENTAL MONITORING 3/4 12-1 3/4.12.1 MON I TO RING P R0GT.AM . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 12-3 TABLE 3.12-1 RADICLOGICAL ENVIRONMENTAL MCNITORING PROGRAM........ l TABLE 3.12-2 REPORTING LEVELS FOR CAOI0ACTIVIT) CGNCEN7%TIONS 3/4 12-9 IN ENVIRChMENTAJ. SAMPLES.................................- i TABLE 4.12-1 0~TECTION CAPABILITh"S FOR ENVIRONF.5NTAL SAMPLE ANALY SIS ,' M4M /,'et e.'.96.PfEWW (4 AQl . . . . . . . . 3/4 12-10
SJk.wu !.'/4 12-13 3/4.12.2 LAND USE GEM 505..........................................
3/4 12-15 i 3/4.12.3 INTERLABO~d.TdF:Y COMP ARISON PROGRAM. . .. . . . . . . . . . . . . . . . . . . . 1 i t V0GTLE - UNIT 1 XIV
o I
- INDEX EASES ..
SECTICN PAGE i . 3/4.0 Ar e LI c Ae n 171. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . s 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS ' 3 3/4.1.1 BORATION CONTR0L.............................;............ B 3/4 1-1 ; 3/4.1.2 BORATION SYSTEMS.......................................... B 3/41-2 4 f 3/4.1.3 MOVAB LE CONTROL ASSEMBLIES. . . . . . , . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 1-4 1- \ , 3/4.2 PadER CISTRT30 TION LIMITS................................... E 3/4 2-1 , i
! 3/4.2.1 AXIAL FLUX 01FFERENCE..................................... B 3/4 2-1 i ,.- 3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL Fr4 TOR rd ECS FLO" f
(4 B3/42-ff 1 l 4WrE- AND NUCLEAR ENTHALPY-M-SE- HOT CHANNEL FACTCR. . . . . . . ,
-i FIGURE B~3/4.2-1 TYPICAL INDICATED AXIAL FLUX DIFFERENCE VERSUS TifERMAL PCWER............................................. B 3/4 2-3
us l

'3/4.2.4 7JAOFXtT POWER TILT RATI0................................. S 3/4 7-5 i
i 3/4.2.5 DNB PARAMETERS............................................ B 3/4 2 6 1 3 /4.3 INSTRUFENTATION 3/4.3.1 ar.c 3/4.3.2 RE.;CTOR TRIP SYSTEM tnd ENGINEERED SAFETt FEATU AES ACTUATICN SYSTEM INSTRCHENTATION. . . . . . . . . .. . . . . B 3/4 3-1 J , 3/4.3.3 NCNITO RING INSTP.L'NENT ATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 3-3 i 3/4.3.4 TUR3IMI CVERSPEED P90TECTION.............................. B 3/4 3-7 l I i
)
i l V0GTLE - UNIT 1 XV j i
INDEX p - 9 1 BASES PAGE
SECTION 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION.............
B 3/4 4-1 3/4.4.2 SAFETY VALVES............................................. B 3/4 4-2 3/4.4.3 PRESSURIZER.................................... .......... B 3/4 4-2 3/4.4.4 RELIEF VALVES............................................. B 3/4 4-3 3/4.4.5 STEAM GENERATORS.........'................................. B 3/4 4-3 3/4.4.5 REACTOR COO LANT SYSTEM LEAKAGE. . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 4-4 3/4.4.7 CHEMISTRY................................................. B 3/4 4-5 3/4.4.8 SPECIFIC ACTIVITY......................................... B 3/4 4-5 3/4.4.9 PRESSURE / TEMPERATURE LIMITS............................... B 3/4 4-7
,wa rcre tA t. PAoPGrzit ES TABLE B 3/4.4-1 REACTOR VESSEL TOU0HNCSS. . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 4-9 FIGURE B 3/4.4-1 FAST NEUTRON FLUENCE (E>1MeV) AS A FUNCTION OF i
l FULL POWER SERVICE LIFE.................................. B 3/4 4-10 FIGURE B 3/4.4-2 EFFECT OF FLUENCE AND COPPER CONTENT ON SHIFT OF RT FOR REACTOR VESSELSh -- EXPOSED TO 550*F. . . . . . . . . . . . B 3/4 4-11 NOT sreEL $ :x.stataw A7 B 3/4 4-16 3/4.4.10 STRUCTURAL INTEGRITY..................................... 3/4.4.11 REACTOR CCOLANT SYSTEM VENTS............................. B 3/4 4-16 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS.............................................. B 3/4 5-1 3/4.5.2 and 3/4.5.3 ECCS SUBSYSTEMS............................... B 3/4 5-1 3/4.5.4 00RCN INJECTICN SYSTEM.................................. . S 3/' 5 2 3/4.5.5 REFUELING WATER STORAGE TANK.............................. B 3/4 5-2 O s-V0GTLE - UNIT 1 XVI . l \ l l._. _ __ _ _ _ , _ _ _ _ ____ . . . . _ _ _ _ _ ,, _ _ . _ , _ .-__ _ ___ _.-_-_..___i
1,
! I INDEX BASES SECTION PAGE m..,,.... ,.,r.,. . . . . . . . . . _ . . _ ~s... mum wun,nuiriwi. g 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT....................................... 8 3/4 6-1 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS...................... B 3/4 6-3 S-/4. C. 3 IODI.'iE CLEANUfLSYSTEM............... .... _ 5/4 6-;-
3/4.6.,i CONTAINMENT ISOLATION VALVES.............................. B 3/4 6-5
.s.
3/4.6.,5 COMSUSTIBLE GAS CONTR0L................................... B 3/4 6-5 3/4. 6. 6-PENETRAMetHt00M-EXHAUST-AIR-ELEANUP SYSTEM. . . . . . . . . . . . . . . 0 3/4 5-5 gy .. g,, y .. . -
-N ,...--.....,~.mg I I I O 1. ecd .J............................... O M[ 7 W W U
f O V0GTLE - UNIT 1 XVII
- -n - -- --y.n - , ,, 4 -c. .-- ,w- ,._.y , . , _ , _ - .-myv - y,- swy,y y , , ,--w.,--.w-,-ae iv--. mr cm.---,
9 INDEX
/%
k BASES SECTION . PAGE 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE............................................. B 3/4 7-1 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION........... B 3/4 7-3 3/4.7.3 COMPONENT COOLING WATER SYSTEM............................ B 3/4 7-3 tiv e w a rt scruicE cocuece 3/4.7.4.I SERVICE-WATER SYSTEM...................................... B 3/4 7-3 3/4.7.5 ULTIMATE HEAT SINK........................................ B 3/4 7-3 Sf 4. 7. 5 FLOOD-PROTECTION.......................................... B 3/4 7-4 0 cit 77tn rio.J 3/4.7.7 CONTROL ROOM EMERGENCY AM-GLEANUP- SYSTEM. . . . . . . . . . . . . . . . . B 3/4 7-4 E'#.!%verstarms/ Ans=A nernes r/os/ wo EMAvsrSySTEM 3/4.7.,51 ECCS PUMP ROCM EXHAUST AIR CLEANUP SYSTEM. . . . . . . . . . . . . . . . . B 3/4 7-4 3/4.7.hSNUBBERS.................................................. B 3/4 7-5 o sJ 9 3/4.7.itrSEALED SOURCE CONTAMINATION............................... B 3/4 7-6
-3/ ' . 7.11 FIRE-SUPPR CS SION SYSTEMS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 7-7 ..,,,, .,. ,.e ev.nr . n n,umuwe neswo n , r em ru., , , r s...............................
r
..... o of*--/t 10 3/4. 7.-H AREA TEMPERATURE MONITORING. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 7-7 & J / 3/4.8 ELECTRICAL POWER SYSTEMS Q
3/4.8.1', 3/4.8.2, and 3/4.8. 3 A.C. SOURCES, D.C. SOURCES, and ONSITE POWER DISTRIBUTION............................... B 3/4 8-1 3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES................... B 3/4 8-3 p.
.. El4 1 Il E,vCr /^/6[/ LED SdfETy FERrur26.$ (Esf) p.sw cootat- B 3l+ 1 '1 <
ls ,t ,v o sM- G ry - K e L)TfD c siiLLGK s yS7 tan O 2 IL 12EHcri.irL c=w.W Pup r' THerwa. ageiEte- a 3)q 1-1 Co0LINC- ;.fT6fL /Suprn .J , 1 l V0GTLE - UNIT 1 XVIII l
-_ - - _ _ _ _ - - - - - _ _ _ - . - - - _ _ - _ - _ _ - _ _ _ _ . - . _ _ _ - . _ _ _ - - - _ - _ _ . . - - - . . - __-.__..._..__..-.__-a- _ . _ . - - . _ _ . _ _ - - . - _ _ - . - _ _ - - _ _ . _ _ - - - - - - . - - - - - - . _ _ _ _ - _ _ - - - _ - _ - - . _ _ _ _ - - -
l l 1 INDEX BASES PAGE SECTION 3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION....................................... 8 3/4 9-1 3/4.9.2 INSTRUMENTATION........................................... 8 3/4 9-1 3/4.9.3 DECAY TIME...................................'............. B 3/4 9-1 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS......................... B 3/4 9-1 3/4.9.5 COMMUNICATIONS............................................ B 3/4 9-1 st'df~ufu AG- AttkH/d[ B 3/4 9-2 3/4.9.6 ttAMI P" LATO R C Rf ."E . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . AREM 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE B&RBMG. . . . . . . . . . . . . . . . B 3/4 9-2 3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION. . . . . . . . . . . . . B 3/4 9-2
! ~
i lie M77LA 770Al 3/4.9.9 CONTAINMENT FURGE t,NO EMiA' J ST ISOLATION SYSTEM. . . . . . . . . . . . B 3/4 9-2 3/4.9.10 and 3/4.9.11 WATER LEVEL - REACTOR VESSEL and STORAGE P00L............................................ 8 3/4 9-3 l
.tf*-9-12-STORAGE-P00t-VENTItATIOM-SYSTEM..................... ..... " 3/4 0-3 3/4.10 SPECIAL TEST EXCEPTIONS 4
3/4.10.1 SHUTDOWN MARGIN........................................... B 3/4 10-1 3/4.10.2 GROUP HEIGHT, INSERTION, AND POWER DISTRIBUTION LIMITS.... B 3/4 10-1 3/4.10.3 PHYSICS TESTS............................................. B 3/4 10-1 3/4.10.4 REACTOR COOLANT L00PS..................................... B 3/4 10-1 3/4.10.5 POSITION INDICATION SYSTEM - SHUTD0WN..................... B 3/4 10-1 i l l O
V0GTLE - UNIT 1 XIX f
s 4
i INDEX
)/
BASES 3/4.11 RADIOACTIVE EFFLUENTS ^ 3/4.11.1 LIQUID EFFLUENTS........................................ B 3/4 11-1 3/4.11.2 GASEOUS EFFLUENTS....................................... B 3/4 11-3 3/4.11.3 SO LID RADI0 ACTIVE WASTES. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 11-6 3/4.11.4 TOTAL 00SE.............................................. B 3/4 11-6 3/4.12 RADIOLOGICAL; ENVIRONMENTAL MONITORING + 3/4.12.1 MONITORING PR0 GRAM...................................... B 3/4 12-1 i s vez vs/ 3/4.12.2 LAN D U S E GENS US . : . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 12-1 3/4.12.3 INTERLABORATORY COMPARISON PR0 GRAM. . . . . . . . . . . . . . . . . . . . . . B 3/4 12-2 i i , i k YOGTLE-UNIT 1 XX
INDEX O, A-
/1 DESIGN FEATURES SECTION PAGE 5.1 SITE 5.1.1 EXCLUSION AREA.............................................. 5-1 5.1.2 LOW PO P U LAT I O N Z0 N E . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-1 5.1.3 MAPS DEFINING UNRESTRICTED AREAS AND SITE BOUNDARY FOR RADI0 ACTIVE GASEOUS AND LIQUID EFFLUENTS.................... 5-1 FIGURE 5.1-1 EXCLUSION AREA....................................... 5-2 FIGURE 5.1-2 LOW POPULATION Z0NE.................................. 5-3 FICURE 5.1-3 UNRESTRIGTE0 AREA AND SITE SOUNCARY FOR RA&IBAETIVE-GASEGUS-EFFLUENTS. . . . . . . . . . . . . . . . . . . . . . . . 9-+
FIGURE 5.-1-4---UNRESTRICTED-AREA-AND-SITE-BOUNDARP10R
;5 RA&IOAGTIVE-LIQUID-EFFLUEMT-5. . . . . . . . . . . . . . . . . . . . . . . . .
5.2 CONTAINMENT 5.2.1 CONFIGURATION............................................... 5-1 5.2.2 DESIGN PRESSURE AND TEMPERATURE............................. 5-1 5.3 REACTOR CORE 5.3.1 FUEL ASSEMBLIES............................................. 5,64 5.3.2 CONTROL ROD ASSEMBLIES...................................... 5-94 1 5.4 REACTOR COOLANT SYSTEM 5.4.1 DESIGN PRESSURE AND TEMPERATURE............................. 5-64
5.4.2 V0LUME...................................................... 5 f 5.5 METEOROLOGICAL TOWER LOCATION................................. 5-6 f l
5.6 FUEL STORAGE 5.6.1 CRITICALITY................................................. 5-7:6 5.6.2 DRAINAGE.................................................... 5-715 5.6.3 CAPACITY.................................................... 5-79 5.7 COMPONENT CYCLIC OR TRANSIENT LIMIT........................... 5-749 ! TABLE 5.7-1 CCMPONENT CYCLIC OR TRANSIENT LIMITS.................. 5-9 6 i, I L V0GTLE - UNIT 1 XXI 1 i
i. .
INDEX V ADMINISTRATIVE CONTROLS PAGE SECTION 6.1 RESPONSIBILITY.............................................. 6-1 6.2 ORGANIZATION................................................ 6-1 6-1 6.2.1 0FFSITE................................................... c.vs ire = 6.2.2 IMIf-5 fA FF. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-1 6-3
. FIGURE 6.2-1 0FFSITE ORGANIZATION...............................
FIGURE 6.2-2 UNIT ORGANIZATION.................................. 6-4 TABLE 6.2-1 MINIMUM SHIFT CREW COMPOSITION...................... 6-5 6.2.3 INDEPENDENT SAFETY ENGINEERING GROUP Function.................................................. 6-6 Composition............................................... 6-6 Responsibilities.......................................... 6-6 , Records................................................... 6-6 6.2.4 SHIFT TECHNICAL ADVIS0R................................... 6-6 o.a .... un. M AF.r ......flen...~..ituna................................... vunti ou
t. 3 FLAV 7 SntFF QUM/ ftCA MNS t. f 6.4 TRAINING.................................................... 6-7 neu,-,
~- A00,, ............................................ o ,-
6.-. ~_ p.' M Y,,R,,.a,=vi 60 &eMt o m, . - , , . 6.5.1 un o ne , m _onna m Function.................................................. 6-7 Composition........ ...................................... 6-7 Alternates................................................ 6-7 Meeting Frequency......................................... 6-8 6-8 Quorum.................................................... Responsibilities.......................................... 6-8 Records................................................... 6-9 V0GTLE - UNIT 1 XXII
i l l INDEX p y ADMINISTRATIVE CONTROLS SECTION
, .-C4 . . . ,.F. . . mEYY... REV/SW 60A/2 D -- -
1 6.5.2 . worna, nu m nn n - - nnu nuun unuur Function.................................................. 6-9 Composition............................................... 6-10 l Alternates................................................ 6-10 Consultants............................................... 6-10 Meeting Frequency............................. ;.......... 6-10 Quorum.................................................... 6-11 Review.................................................... 6-11 Audits.................................................... 6-11 Records................................................... 6-12 6.6 REPORTABLE EVENT ACTI0N..................................... 6-13
6. 7 SAFF.TY LIMIT VIOLATION...................................... 6-13 6.8 PROCEDURES AND PR0 GRAMS..................................... 6-13 6.9 REPORTING REQUIREMENTS 5.9.1 ROUTINE REP 0RTS........................................... 6-16 Startup Report............................................ 6-16
, Annual Reports............................k............... / n e.
6 16 Annual Radiological Environmental ~mm,ing 0g ro i. Report. . . . . . . . . 6-17 Semiannual Radioactive Effluent Release Report............ 6-18 Monthly Operating Report?................................. 6-20 i Radial Peaking Factor Limit Report. . . . . . . . . . . . . . . . . . . . . . . . 6-20 6.9.2 SPECIAL REP 0RTS........................................... 6-21 6.10 RECORD RETENTION........................................... 6-21 i f
O XXIII V0GTLE - UNIT 1
,,g im=*e . 8*
__ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ____..______m. _ _ _________-.________.__._____.__.____.__.-____________________-____.__._.__.._._______m_ _ _ _ _ _ _ _ . _ _
j 1-J INDEX i * ! ADMINISTRATIVE CONTROLS 1 1, 4 SECTION f l 6.11 RADIATION PROTECTION PR0 GRAM............................... 6-22 1 6.12 HIGH RADIATION AREA........................................ 6-22-A3 1 I 6.13 PROCESS CONTROL PROGRAM (PCP).............................. 6-21 2.4 1 6.14 0FFSITE OOSE CALCULATION MANUAL (0DCM)..................... 6-24 i i 6.15 MAJOR CHANGES TO LIQUID, GASEOUS, AND SOLID RA0 WASTE TREATMENT SYSTEMS................................. 6-25 l L is A ntra o m E N 1 s -rn TfcHNic A L _fpf?r cir J r'inMe , . . VZb i } I i I 3 i i l V0GTLE - UNIT 1 XXIV i l I
6 - - ,-,, A Le aL 4-. o A,wan-~<a' sam- 4.a n--- - .-MiY,w-Ae --. + -sm-&4a-,.,,as m,.4 -.4- , - g. I I !9 l 1 ! I I 4 1 e l i l SECTION 1.0 DEFINITIONS 1 1 i 9 . i b, , t i .I A ) .i j J l l 1 I i l b, i 4 f O f 1 l I
1.0 DEFINITIONS (v The defined terms of this section appear in capitalized type and are applicable throughout these Technical Specifications. ACTION 1.1 ACTION shall be that part of a Technical Specification which prescribes remedial measures required under designated conditions. ACTUATION LOGIC TEST 1.2 An ACTUATION LOGIC TEST shall be the application of various simulated input combinations in conjunction with each possible l'nterlock logic state and verification of the required logic output. The ACTUATION LOGIC TEST shall include a continuity check, as a minimum, of output devices. ANALOG CHANNEL OPERATIONAL TEST 1.3 An ANALOG CHANNEL OPERATIONAL TEST shall be the injection of a simulated signal into the channel as close to the sensor as practicable to verify OPERABILITY of alarm, interlock and/or trip functions. The ANALOG CHANNEL OPERATIONAL TEST shall include adjustments, as necessary, of the alarm, inter-lock and/or Trip Setpoints such that the setpoints are within the required range and accuracy. g g. g 4 A_XIAL FLUX DIFFERENCE # /A* c*'"o.r "'"
.Ls "#+y") #
pueen 1.4 AXIAL FLUX DIFFERENCE shall be the difference in no = 11:cd flux signals a between the top and bottom halves of a two section excore neutron detector. CHANNEL CALIBRATION 1.5 A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel such that it responds within the required range and accuracy to known values of input. The CHANNEL CALIBRATION shall encompass the entire channel including the sensors and alarm, interlock and/or trip functions and may be performed by any series of sequential, overlapping, or total channel steps such that the entire channel is calibrated. CHANNEL CHECK 1.6 A CHANNEL CHECK shall be the qualitative assessment of channel behavior during operation by observation. This determination shall include, where possible, comparison of the channel indication and/or status with other indications and/or status derived from independent instrument channels measuring the same parameter. O V V0GTLE - UNIT 1 1-1
DEFINITIONS b CONTAINMENT INTEGRITY 1.7 CONTAINMENT INTEGRITY shall exist when:
a. All penetrations required to be closed during accident conditions are either:

1) Capable of being c.losed by an OPERABLE containment automatic isolation valve system, or

2) Closed by manual valves, blind flanges, or deactivated automatic valvessecuredintheirclosedpositionsfeueptasprovidedin Tatrie [3.P1] cf-Spec 4f4catrion-9,64]J

b. All equipment hatches are closed and sealed,

c. Each air lock is in compliance with the requirements of Specification

43. 6.1. 3)(

d. The containment leakage rates are within the limits of Specification f(3. 6.1. 2;}',' and

e. The sealing mechanism associated with each penetration (e.,g., welds, bellows, or 0-rings) is OPERABLE.
p V CONTROLLEO LEA GGI 1.0 00NTR0tifD-tEAKAGE-shall be that-seel-water-flow :uppli:d to the re:cter coolant p=p ::al:. CORE ALTERATIONS I d CORE ALTERATIONS shall be the movement or manipulation of any component within the reactor pressure vessel with the vessel head removed and fuel in the vessel. Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe conservative position. DOSE EOUIVALENT I-131
../
h-10 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcurie / gram) which alone would oroduce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134, and I-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in [T2 Lie III ef T10-14944, "Gelettlet4en-of-Mstance-Faetcr: for P =:r :nd T::t Reseter Sites" er Table E-7 of NRC Regulatory Guide 1.109, Revision 1, October 1977,}:' I - AVERAGE DISINTEGRATION ENERGY
../C M1 I shall be tre average (weighted in praportion to the concentration of each radionuclide in the sample) of the sum of the average beta and gamma energies per disintegn tion (MeV/d) for the radionuclides in-the-sample, , . j ,., ,, y
r. r.. r L Cr . '<- / , . c . t w //,a.n i+ ,,,;,,u r,,, ,,,,, ,;,,,4 4 Q .i ? .<v? ,ic 7. ; r. ,<
n: . , :. r, . i .4 .; 6,n ,' ,,o,,, a j;o
,,,c s,od;,,e. a e s'ivi/y ire H e sa nsp /t. .
V0GTLE - UNIT 1 1-2
1 i DEFINITIONS ENGINEERED SAFETY FEATURES RESPONSE TIME
/.//
1-H The ENGINEERED SAFETY FEATURES (ESF) RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ESF Actuation Setpoint. at the channel sensor until the ESF equipment is capable of performing its safety function (i.e. , the valves travel to their required positions, pump ! discharge pressures reach their required values, etc.). Times shall include ' j diesel generator starting and sequence loading delays where applicable. FREQUENCY NOTATION
/. /2 1-H The FREQUENCY NOTATION specified for the performance of Surveillance l
f,,,g Requirements shall correspond to the intervals defined in Table 1.1.
^
4 IDENTIFIED LEAKAGE 1.14 IDENTIFIED LEAKAGE shall be: i A%e+ar /?o / tad Pu mp .%' hako&*
a. Leakage (exceptJECNT"0LL " LEAC OC) into closed systems, such as pump seal or valve packing leaks that are captured and conducted to a sump j or collecting tank, or i

b. I.eakage into the containment atmosphere from sources that are both '
l specifically located and known either not to interfere with the operation of Leakage Detection Systems or not to be PRESSURE BOUNDARY LEAKAGE, or 1
c. Reactor Coolant System leakage through a steam generator to the i
Secondary Coolant System. MASTER RELAY TEST
)
1.1'5 A MASTER RELAY TEST shall be the energization of each master relay and 'l verification of OPERABILITY of each relay. The MASTER RELAY TEST shall include a continuity check of each associated slave relay. MEMBER (S) 0F THE PUBLIC i l 1.16 MEMBER (S) 0F THE PUBLIC shall include all persons who are not occupa-tionally associated with the plant. This category does not include employees .I of the licensee, its contractors, or vendors. Also excluded from this category ' i are persons who enter the site to service equipment or to make deliveries. This category does include persons who use portions of the site for recre-l i ational, occupational, or other purposes not associated with the plant. ,
; /
l OFFSITE 00SE CALCULATION MANUAL 1.17 The 0FFSITE DOSE CALCULATION MANUAL (00CM) shall contain the methodology and parameters used in the calculation of offsite doses due to radioactive t gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring Alarm / Trip Setpoints, and in the conduct of the Environ-mental Radiological Monitoring Program. i ! 'V0GTLE - UNIT 1 1-3 1
INSERT 1.13 GASEOUS WASTE PROCESSING SYSTEM 1.13 A GASE0US WASTE PROCESSING SYSTEM shall be any system designed and installed to reduce radioactive gaseous effluents by collecting Reactor Coolant System offgases from the Reactor Coolant System and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment. O O .
DEFINITIONS (3 V
)
OPERABLE - OPERABILITY 1.18 A system, subsystem, train, component or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified function (s), and when all necessary attendant instrumentation, controls, electrical power, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component, or device to perform its function (s) are also capable of performing their related support function (s). OPERATIONAL MODE - MODE 1.19 An OPERATIONAL MODE (i.e., MODE) shall correspond to any one inclusive combination of core reactivity condition, power level, and average reactor coolant temperature specified in Table 1.2. PHYSICS TESTS 1.20 PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics o bthe ,rp ctor core and related instrumentation: (1) described in Chapter,[14.0f of the FSAR, (2) authorized under the provisions of 10 CFR 50.59, or (3) otherwise approved by the Commission. PRESSURE BOUNDARY LEAKAGE 1.21 PRESSURE BOUNDARY LEAKAGE shall be leakage (except steam generator tube leakage) through a nonisolable fault in a Reactor Coolant System component O() body, pipe wall, or vessel wall. PROCESS CONTROL PROGRAM 1.22 The PROCESS CONTROL PROGRAM (PCP) shall contain the current formulas, sampling, analyses, tests, and determinations to be made to ensure that prccessing and packaging of solid radioactive wastes based on demonstrated processing of actual or simulated wet solid wastes will be accomplished in such a way as to assure compliance with 10 CFR Parts 20, 61, and 71 and Federal and State regulations, burial ground requirements, and other require-ments governing the disposal of radioactive waste. PURGE - PURGING 1.23 PURGE or PURGING shall be any controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is required to purify the confinement. V0GTLE - UNIT 1 1-4
DEFINITIONS e 4 V QUADRANT POWER TILT RATIO 1.24 QUADRANT POWER TILT RATIO shall be the ratio of the maximum upper excore detector. calibrated output to the average of the upper excore detector cali-brated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, whichever is greater. With one excore detector inoperable, the remaining three detectors shall be used for computicg the average. RATED THERMAL POWER 1.25 RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 3d//MWt. REACTOR TRIP SYSTEM RESPONSE TIME 1.26 The REACTOR TRIP SYSTEM RESPONSE TIME shall be the time interval from when the monitored parameter exceeds its Trip Setpoint at the channel sensor until loss of stationary gripper coil voltage. REPORTABLE EVENT 1.27 A REPORTABLE EVENT shall be any of those conditions specified in Section 50.73 of 10 CFR Part 50. 5"IELO CUILDING-IttTEGRMY ( 1.20 0HIELO OUILOINC INTECRITY shell exist when; a---Esch-door in ::ch ::. s: Opening is cle::d :xcept when th: caets sp:ning i: bei ng-used-40e-norm:I tr:ns4t-entry :nd exit, ther-at least one door shall be cle::d,
b. The Shield Sai4 ding-H4trat4cn-Gystem is-OPERASLE, and

c. The sealing mechanism seseefsted-w Rh-eech-peneteati n (e.g., .::!d:,
balisws, or 0 rings) is OPERAatE. SHUTCOWN MARGIN
/. 2 7 1799 SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present conditio^
assuming all fwil leng;r, rod cluster assemblies (shutdown and control) are fully inserted except for the single rod cluster assembly of highest reactivity worth which is assumed to be fully withdrawn. SITE BOUNDARY t.19 1730 The SITE BOUNDARY shall be that lin b;y:nd ^ich th: 1:nd 4: n0ither ewn d, cer leased, n r otherw+se-contro44ed4y-the 'ic?n&**, pmperp //ne 5AsWn in f m.rc S M c / /h e s .t 7e~ ch nic.a./ Spec / Nc.t. ho n s. 9
-)
V0GTLE - UNIT 1 1-5
q OEFINITIONS O - SLAVE RELAY TEST
, . Jo 1-91 A SLAVE RELAY TEST shall be the energization of each slave relay and verification of OPERABILITY of each relay. The SLAVE RELAY TEST shall include a continuity check, as a minimum, of associated testable actuation devices.
SOLIDIFICATION
/.3/
1-92 SOLIDIFICATION shall be the conversion of wet wastes into a form that meets shipping and burial ground requirements. SOURCE CHECK .
$ A SOURCE CHECK shall be the qualitative assessment of channel response when the Channel sensor is exposed to a source of increased radioactivity.
51AGGEREDTESTBASIS
/. )]
h34 A STAGGERED TEST BASIS shall consist of:
3. A test schedule for n systems, subsystems, trains, or other designated components obtained by dividing the specified test interval into n equal subintervals, and

b. The testing of one system, subsystem, train, or other designated component at the beginning of each subinterval.
THErtMAL POWER
/J+
1-95 THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant. TRIP ACTUATING DEVICE OPERATIONAL TEST s.35 1--36 A TRIP ACTUATING DEVICE OPERATIONAL TEST shall consist of operating the Trip Actuating Device and verifying OPERABILITY of alarm, interlock and/or trip functions. The TRIP ACTUATING DEVICE OPERATIONAL TEST shall include aajustment, as necessary, of the Trip Actuating Device such that it actuates at the required Setpoint within the required accuracy. UNIDENTIFIED LEAKAGE
/ 34 1-37 UNIDENTIFIED LEAKAGE shall be all leakage which is not IDENTIFIED LEAKAGE.
r CO T"0LL:0 LEAKAO . O V0GTLE - UNIT 1 1-6
DEFINITIONS
^
UNRESTRICTED AREA t37 -! 1r36 An UNRESTRICTED AREA shall be any area at or beyond the SITE BOUNDARY access to which is not controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials, or any area within the SITE BOUNDARY used for residential quarters or for industrial, commercial, institutional, and/or recreational purposes. VENTILATION EXHAUST TREATMENT SYSTEM til 1r39 A VENTILATION EXHAUST TREATMENT SYSTEM shall be any system designed and installed to reduce gaseous radioiodine or radioactive material in particulate form in effluents by passing ver.tilation or vent exhaust gases through charcoal adsorbers and/or HEPA filters for the purpose of removing iodines or particu-lates from the gaseous exhaust stream prior to the release to the environment. Such a system is not considered to have any effect on noble gas effluents. i Engineered Safety Features Atmospheric Cleanup Systems are not considered to be VENTILATION EXHAUST TREATMENT SYSTEM components. VENTING
\ !. 39 .
140 VENTING shall be the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration, or other operating condition, in such a manner that replacement air or gas is not pro-vided or required during VENTING. Vent, used in system names, does not imply a VENTING process. WMirTC OZ N0LDU? SYSTEM i 1.41 A 5'ASTE GAS HGLOGA-SYSTEM-sha14-be :ny y:ter 4 :ign:d :nd 'n t:11:d t: redue: c:df rective-gaseous-eff4uentsr-by-sel4eetring ":::t:r 0::hnt Oy:tr j offg:::: fron-the-Reactoc-Coolant Sy:tre : M pr:eiding f:r &l:y er Mi t; f r th: purpose-e&-reducing the tet:1 redirectivity pri:r te rel ::: te the i ....e.. - .. f s . I ,t 4 O VCGTLE - UNIT 1 1-7 i l
- . . - . . - . . . .. . ._- - . . . - . . - - . . . . - . . .- .- .~
i I i 1. TABLE 1.1 !O FREQUENCY NOTATION j f NOTATION FREQUENCY ! S At least once per 12 hours, i i i D At least once per 24 hours. . 1 l W At least once per 7 days. . I
M At least once per 31 days. i j
At least once per 92 days. Q l j SA At least once per 184 days. ! i i R At least once per 18 months. ; i S/U Prior to each reactor startup. l i. N.A. Not applicable. j f P Completed prior to each release, f i 1 O -
)
i i i i i : 4 i 1 { 4 i i l j -
. i
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i
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1 4
l !
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! f I f i I; i i V0GTLE - UNIT 1 1-8 ! l t j > l 2 l 1 ,
l 1. ! TABLE 1.2 1 1 OPERATIONAL MODES i . j
! REACTIVITY % RATED AVERAGE COOLANT MCDE THERMAL POWER
TEMPERATURE CONDITION. K_ff 3
j 1. POWER OPERATICN > 0.99 > 5% > 350*F I
2. STARTUP > 0.99 1 5% y,350*F
') j 3. HOT STANDBY < 0.99 0 > 350*F i i i 1 4 HOT SHUTDOWN < 0.99 0 350*F > T
> 200*F avg l '
I 1
5. COLD SHUTCOWN '< 0.99 0 1 200*F !

6. REFUELING ** $ 0.95 0 $ 140*F ,
1 i i i I j
Excluding decay heat. 'l'
}
** Fuel in the reactor vessel with the vessel head closure bolts less than fully j .
tensioned or with the head removed. 1 . 1 J i 1, : t l 1 i t j i i i ! i
i i ;
! I i ! I I L i V0GTLE - UNIT 1 1-9 l 1 i , j i
- . . - ~ - . - - - - _ - _ . - = _ - _ . . _ - . - . - _
4 l 4
JUSTIFICATIONS FOR DEVIATIONS FROM STS SECTION 1.0 l f
, Definitions: ! 1.4: ,
i This definition was revised so as to give definition to the tern
"ncrmalized" and relieve a potentially ambiguous situation. Without [
l qualification, the term normalized could imply that the flux signals ce i
normalized to their sum, their full power sum, or some other value. ,
i i 3 1.7.
)
See the justification provided for Specification 4.6.1.1.a. 1 i 1.8: t I
; The definition of CONTROLLED LEAKAGE was deleted as a result of changes made i t in specifications dealing with limiting ficwrates through the RCP seal water i i injection line. ;
I i 4 The purpose of the STS limit on controlled leakage is to ensure that the i flow through the seal water injection line is less than that assumed in the ' j accident analysis. This ensures that sufficient centrifugal charging pump l i injection flow is directed to the baron injection tank and, ultimately, to ; the RCS. ! j The generic STS surveillance (4.4.6.2.1.c) ensures the above by measuring l i CONTROLLED LEAKAGE at a specified RCS pressure with the modulating flow i
; control valve (121) fully open. This method is sensitive to conditions in (
the chemical and volume control system, particularly the position of the - charging flow valve (182). The proposed technical specification changes , j would provide a more quantitative measurement of the seal water line flow by [ i limiting the flowrate across the RCP seal water injection throttle valves at t j a specified pressure drop across the valve. This will result in making the ; l measurement independent of charging flow valve (182) position, modulating f 4 flow control valve (121) position, or CVCS system conditions. l The proposed surveillance requirement H.5.2 9 3) would ensure proper RCP ( 4 seal water injection throttle valve position and, hence, flow. The STS f i definition of CONTROLLEO LEAKAGE was deleted since it is no longer i ! necessary. The definitions of IDENTIFIED LEAKAGE and UNIDENTIFIED LEAKAGE l j were revised to account for the deletion of CONTROLLED LEAKAGE. Limits and i
surveillances on CONTROLLED LEAKAGE were deleted and a new surveillanca !
! requirement (4.5.2.g.3) implemented to control flow through the RCS seal j
< water injection lines. ,
' l O ! i r
O This change was made to incorporate the correct reference for the dose conversion factors utilized in the VEGP safety analysis. See paragraph 1.9.109.2 and subsection 11.3.3 of the VEGP FSAR. 1.10: The bases 'ar Specification 3/4.4.8 state that the radionuclides in the typical reactor coolant have half-lives of less than 4 minutes or greater tnan 14 minutes. The 10-minute half-life cutof f seems to be arbitrary espec' ally in light cf the fact that, depending on which reference you use, 11-13 is reported to have a half-1tfe cf 9.97 minutes to 10.5 minutes. Establ#scing the cutoff at 14 minutes relieves this amb1 Cutty while remaining consistent with the intent of the specification. L.U: Tne chinge in terminology (from waste gas holdup system to gaseous waste pecesstr.g system) was nade to reflect plant-specific nomenclature. See section 11.3 cf the VEGC FSAR.
' 12:
The de'irttion of ICENTIFIED LEAKASE vas revised to account for the deletion of CONTR0' LED LEMNIE. (See the justification for the deletion of the term CONTCi. LED l. ear. AGE,)
.V1.(STD:
T4 cefinitten of Sh!5LD EUILDING DifEGRITY was deleted because it is not ap.M icable to VECP. The design of VEOP does not include a shield building. 2.21: 7% .ords "f ull-ler,Sth" were delete on the basis that all RCCAs at VEGP are fuli-!ergth. See swyrerb 4.2.2.3 of the VEGP F5AR. r 1.29- : TMis enaige was mace to spect/tcally identify only the land associated with ! the VEGP site s.1d eclude ot.ner lacd owned ty GPC ddjacent to the site l bandary. j l t l l O 1
. l . J
4
. i j 1.36:
{ The term " CONTROLLED LEAKAGE" was deleted from the definition of
~
UNIDENTIFIED LEAKAGE on the basis provided for the deletion of the i definition of " CONTROLLED LEAKAGC." i ) 1.41 (STS): i. l This definition now appears as 1.13 with the new title " Gaseous Waste Processing System," See the justification provided for definition 1.13. r
.j a
I 1 i i ,! O l i i J i l I h k i i l 1 $ ) i 004b/ i i i l
4 i l 1, l l l i i i i I SECTION 2.0 l . i SAFETY LIMITS i AND f !, LIMITING SAFETY SYSTEM SETTINGS 4 l I l t i 4 I 1 1 ! i i i 1 i l i )
. . _ , _ _ . . . . . . - - . _ ~ . _ , _ .
i 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS i ( PZ-c us A, a,
C, PE-c +-S4 A, 8, w C, PI- o +S7 ), 3, wC, PI-e+SF A, 6, *C i
i 2.1 SAFETY LIMITS l
% c 4 /, N'- ce 4 2, Al- 00 +3, U- e c 4 &)*= '
REACTOR CORE
1 2.1.1 The combination of THERMAL POWER pressurizer pressure, and the highest !
! operating loop coolant temperature (T,yg shall not exceed the limits shown in l Figure [2.1-1.; .d 2.1-2 fer a end a-1 1::,\:;;r;ti;n, r;;;;;tiv:1y. Zoop 2 rr - c 4 < 2 ;
, APPLICABILITY: MODES 1 and 2. L eop rz-o+22 .
I '# ! U### # ' ACTION: Leop 1 TZ' *r + + 2 Whenever the point defined by the combination of the highest operating loop i average temperature and THERMAL POWER has exceeded the appropriate pressurizer l j pressure line, be in HOT STANDBY within 1 hour, and comply'with the require- < q ments of Specification 6.7.1. ! < t i I REACTOR COOLANT SYSTEM PRESSURE (Fr oect, Pr-o#/s, pr-es2 t, pr-o #3 7 ) I 2.1.2 The Reactor Coolant System pressure $shall not exceed 2735 psig. t'
APPLICABILITY: MODES 1, 2, 3, 4, and 5.
i ACTION: MODES 1 and 2: Whenever the Reactor Coolant System pressure has exceeded 2735 psig, be j in HOT STANDBY with the Reactor Coolant System pressure within its limit 1 within I hour, and comply with the requirements of Specification 6.7.1. i i l MODES 3, 4 and 5: J Whenever the Reactor Coolant System pressure has exceeded 2735 psig, J j reduce the Reactor Coolant System pressure to within its limit within i 5 minutes, and comply with the requirements of Specification 6.7.1. i l 1 i i lO V0GTLE - UNIT 1 2-1 l .
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J / \ i / \, 2 N I / 1 / } ,',' FICURE 2.1-1 REACTOR CORE SAFETY LIMIT - FOUR LOOPS IN OPERATION
/ /
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O/ - V0GTLE - UNIT 1 2-2
/
64/41/10733
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= -
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_ . ..g
. _ 3 A C C E PTA B LE . .__ a__ .._.: .:. . .-..__.=...=._.....____.._._-..__m.__.__ . .. . =. . rr. . : . :. .. . .r :. . . .. ._1.__.._.
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t: :.4 ;:.--).n :". :: . . 3.. i. : . .~ -. . . = . . - - . .:_ _~. }. :.; ; ; ._: . . ___
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_ - = - - -=r* :_~ ..---+. tu: : _ :_ 560 .. . . ' . . . . . n!.._ =: : ......__. .._- .. ..!. .__.._. __ . . - : ......=... - : = __._._ 2__=n..e..r_=__.=_-i....._.._.....__...a
= _u.[_.: . = = t : . - . -- t : . . . . _ ,
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y ,4 c~ ~i - ! 4 :iH ] i l -i i + i i. i. l : l i { .i C 0.2 0.4 0.6 0.8 1.0 1.2 FR ACTION OF RATED THERMAL POWER Figure 2.1 1. Reac:or Core Safety Limits (O, a
' r -r # , z. TL t- - ll p.!* - .7 ~ 2.
j O 4 N% , i N\ s N i N
/
[
\\ /
l \
/ 'x N / ! N /
J \N 1 N N / i \s / I \
\
i !O s i j s
\ / s j
j
/ ,
t
/ N I / \ / / \,
i '
/ FIGURE 2.1-2 N ,
i N
\
REACTOR CORE SAFETY LIMIT - THREE LOOPS IN OPERATION l N\
/ \
O/ / V0GTLE - UNIT 1 2-3 l
i SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.2 LIMITING SAFETY SYSTEM SETTINGS REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS l 2.2.1 The Reactor Trip System Instrumentation and Interlock Setpoints shall be set consistent with the Trip Setpoint values shown in Table 2.2-1. APPLICABILITY: As shown for each channel in Table 3.3-1.
ACTION:
; a. With a Reactor Trip System Instrumentation or Interlock Setpoint less conservative than the value shown in the Trip Setpoint column but more conservative than the value shown in the Allowable Value column of Table 2.2-1, adjust the Setpoint consistent with the Trip Setpoint value.
b. With the Reactor Trip System Instrumentation or Interlock Setpoint less conservative than the value shown in the Allowable Values column of Table 2.2-1, either:

1. Adjust the Setpoint consistent with the Trfp Setpoint value of Table 2.2-1 and determine within 12 hours that Equation 2.2-1 -
was satisfied for the affected channel, .or
2. Declare the channel inoperable and apply.the applicable ACTION statement requirement of Specification 3.3.1 until the channel is restored to OPERABLE status with its Setpoint adjusted consistent with the Trip Setpoint-value.
Equation 2.2-1 Z + R + S 1 TA Where: Z = The value from Column Z of Table 2.2-1 for the affected channel, R = The "as measured" value (in percent span) of rack error for the affected channel, S = Either the'"as measured" value (in percent span) of the sensor error, or the value from Column S (Sensor Error) of Table 2.2-1 for the affected channel, and TA = The value from Cohnn TA (Total Allowance) of Table 2.2-1 for the affected channel'. , O V0GTLE - UNIT 1 2-4 i
. - - . . . , ~, - - , , - . . , . , , ,,, .--.....-,,-.,.......,..,n- , , . ~ . , - . . , . - , . _ , ,
O O O TABLE 2.2-1 6 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETP0lNTS S . l~, . / \/- on-e t. N-co s 2. , N-co .u, n evn ) SENSOR
. TOTAL ERROR (S) TRIP SETPOINT ALLOWABLE VALUE i FUNCTIONAL UNIT ALLOWANCE (TA) Z c-Manual Reactor Trip N.A. N.A. N.A. N.A. N.A.
! h- 1. e* i 2. Power Range, Neutron Flux r/ /' / / V' y ff, . f
~~-~"'a. High Setpoint [7.5] .t4.56] O ${109]% of RTPa* 1E414r2-]% of RTP** "' v' V v' z -1.1 l b. Low Setpoint [8.3] 4.56 0 1{25]% of RTP** 1E2hE]% of RTP**
i t. & t l _v/ d Y' &.3
3. Power Range, Neutron Flux, [2.03 {0.5J 0 ${5]% of RTP** with IE6:8]% of RTP** with High Positive Rate a ti a M me constant i
(N- o o y /, d-o o + 2., N- co v 5, N-oo4 + ) 1(2]yconstant seconds 1[2] seconds ' J J
t. & $/J L.3

4. Power Range, Neutron Flux, [t0] [0.5] 0 $[5]% of RTP** with $[6r83% of RTP** with a timp constant 1
High Negative Rate a A. ige constant (n- c o /, .v'-co n z , ,V- oo 43, Al-co 4 4 ) >
^{2[ seconds N(seconds i '? gl ) yl v y' 30.9 '
5. Intermediate Range, {17. 0]y' 18.41_1 0 ${25]% of RTP** 1[S1-]% of RTP**
Neutron Flux I A/-cMf Al-co36 ) yl v J (Lafer) CL &er) 6.
i Source Range, Neutron Flux 117.0_y/ 3 {10.01T0 $E4GL] cps $[L
x M 5] cps l
i (A -co sf Al-co s 2 ) l 4. S 3.4 I J
7. Overtehperature AT [6 7] E2-793 f0.87 See Note 1 See Note 2 (rf-o+s/ 4 + B, 7'E*-o 42 / A #8, 76 -D43),H 8,1C-04y/ A rB ) l

8. Overpower AT E4r3]s:o / *3 Eir1] f0.2J}SeeNote3 See Note 4 19&o 19

9. f Pressurizer rE -oest &B, if-oez / Ar8, rE-093/
Pressure-Low E6.83Av6, 5./ TE ows,A 18)/ /10.71] 11.5)j 1E190a3 1E4 psig psig (PZ- o fic ),8 t C, RZ'-o *S4 A,8,K, Pf-o #ST A,8,rt, RI-o FS 2 A r 8 vc ) y J d 1)
10. Pressurizer Pressure-High 43.1]v .(0. 71PyE1.8V 1[2385T psig 1[239frpsig pr- o 4st A, & +c 1

11. (Pf-o 95G A,3, Pressurizer vC, Pf-Water o 954 d, d,f Level-High ,
+c., 5. 0]<PI-o957 A,8 d2.+C,18Pfl.5J 1[;92]%)of instrument9 1[,3.8%j of instrument Sp n span
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ML20137X316

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