Text

Proposed Tech Specs,Deleting Section 1.39 Re Solidification, 1.46 Re Ventilation Exhaust Treatment sys,3.4,3,7 Re Radioactive Liquid Effluent Monitoring & 3/4.11.2 Re Gaseous Effluents
	ML20086G130
Person / Time
Site:	Fermi
Issue date:	11/26/1991
From:	; DETROIT EDISON CO.
To:	;
Shared Package
ML20086G127	List: ML20086G130; ML20086G135; NRC-91-0131, Application for Amend to License NPF-43,changing Tech Specs to Relocate Radiological Effluent Tech Specs to ODCM or Process Control Program,Per Generic Ltr 89-01;
References
	NUDOCS 9112040212
	Download: ML20086G130 (154)
	v • d • e

O ENCLOSURF. 3 PROPOSED TECl!NICAL SPECIFICATIONS MARK-UP O

O propi!MQp c

O

=

Di rit11110 tis SECT 10t1 DEF1 tilt 10 tis (Conti.ued)

IE1 1.26 OPE RAT 10!iAL C0fiDIT 10ti - C0tiDIT 10fi..........................

1-4 1.27 PHYSIC 3 TEST 5..............................................

14 1.28 PRES 50RE BoutiDARY LEAKAGE..................................

1-4 1.29 PRIMARY C0!1T AltiME t41 If1T EGRI T Y..............................

14 1.30 PROCESS C0f(TROL PR0 GRAM....................................

1-5 1.31 PURGE.PURG111G..............................................

1-5 1.32 RAT E D T H E RM AL P0W E R........................................

1-5 1.33 REACTOR PROTECTloti SYSTEM RESP 0fiSE TIME....................

1-6

)

1.34 RE PORT AB L E E V Ell 1................................. t.........

1-6 O

i.as ROD Otfisiiv................................................

1-6 1.36 SEC0fiDARY C0 tit Al!4ME!!T liiT E GRlT Y............................

16 1.37 SHUTDOWil MARG 1ti............................................

1-6 1.3B SITE B0VfiDARY...................................

16 m -. m... e.

.. ~

1.t0 SOURCE CHECK...............................................

17 1,41 ST AGGE RED T E ST B A515.......................................

1-7 1.42 1HERMAL P0KER..............................................

1-7 1.43 TURB1 tie BYPASS SYST EM RESP 0ftSE TIME........................

17 1.44 Util DE til l f l E D L E AK A G E.......................................

17 1.45 UtiRESTRICTED AREA..........................................

1-7 1-S----

1. 4 VEIM4MT@4-Ewr "ErME"' SYSTEM

.,s..

)

1.47 VEtaltiG....................................................

18 O

FERM1 Utili 2 ii

JNDEX 11M1 TING CONDITIONS FOR OPERATION AND SURVEltt ANCE RE0VIRFMENTS h

f/d{

t,a SECTION 3/4,3 INSTRUMENTATIOL REACTOR PROTECTION SYSTEM 1HSTRUMENTAT10N........... 3/4 31 3/4.3.1 ISOLATION ACTUATION INSTRUMENTATION.................

3/4 3 9 3/4.3.2 EMERGENCY CORE COOLING SYSTEM ACTUATION 3/4.3.3 I NST RUMENT AT I ON..................................... 3/4 3 ATWS RECIRCULATION PUMP TRIP SYSTEM ACTUATION 3/4.3.4 INSTRUMENTATION.....................................

3/4 3 32 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION 3/4.3.5 INSTRUMENTATION.....................................

3/4 3 36 3/4.3.6 CONT ROL ROD BLOCK I NSTRUMENT AT I ON................... 3/4 3 MONITORING INSTRUMENTATION 3/4.3.7 Radiation Monitoring Instrumentation................

3/4 3 47 Seismic Monitoring Instrumentation..................

3/4 3 51 Meteorological Monitoring Instrumentation........... 3/4 3 54 Remote Shutdown System Instrumentation and Controls. 3/4 3 57 Accident Monitoring Instrumentatton.................

3/4 3 60 Source Rang e Monitors............................... 3/4 3 64 Traversing In Core Probe System.....................

3/4 3-65 Chlorine Detection System........................... 3/4 3 66 Del e t e d............................................. 3/4 3 6 7 loose Part Detection System......................... 3/4 3 70 o

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....'.. M e........................................ ;/ " 0-7;_

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t Monitoring I nstrument ati on..................................... 3/4 3 7 6 l

FEEDWATER/ MAIN TURBINE TRIP SYSTEM ACTUATION l

3/4.3.9 I NST RUMENT AT I ON........................................ 3 l

3/4.3.10 RESERVED.,...............................................

3/4.3.11 APPENDIX R 1.LTERNATIVE SHUTDOWN INS 1RUMENTATION f.

FERMI - UNIT 2 v

Amendment No.- ;;, ;;, ;;, 7;-

INDEX 1

i

[]M] TING C0f4D1110NS FOR OPERA 110N AND SURVElllANCf Rf001RfMENTS

- SECTION fM{

3/4.11 RADIDACTIVE EFFtVENTS J

3/4.11.1 LICUID EffLUEN15

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3/4 11-7 3/4.11.2 GASEOUS EFFLUENTS

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3/4 11-16 Main Condenser...........................................

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INDEX l

l Basis fAE l

SECTION INSTRLMENTATION (Continued)

M3NITORING INSTRUMENTATION (Continued)

Meteorological Monitoring Instrumentation....... B 3/4 3-4 Remote Shutdown System Instrumentation and Controls.........'...............................B3/43-4 AccidentMonitoringInstrumentation.............B3/43-4 Source Range Monitors........................... B 3/4 3 4 Traversing In-Core Probe System................. B 3/4 3 4 Chlorine Detection System....................... B 3/4 3 5 Del e t e d......................................... B 3/4 3 5 Loose-Part Detection System..................... B 3/4 3 5 n. 21.......

4-m 2 e r st.... u.. a.. a.,,.

......_.....'.'...........'.......'..'............,,3-6-==

[$E[akE.

_m e#A NEEE T.;. ',n;t..r Ca;;;;; iff'.nt Monitoring newsive.

(

Instrumentation.................................

B 3/4 3 6 3/4.3.9 FEE 0 WATER /MAlH TURBINE TRIP SYSTEMS ACTUATION INSTRUMENTATION.................................B3/43-6 3/4.3.10 RESERVED 3/43.11 APPENDIX R ALTERNATIVE SHUTDOWN I NST RUMENT AT I ON................................. B 3/4 3 6 3/4.4 REACTOR C00t Atti SYSTD4 3/4.4.1 REclRCULATION SYSTEM............................ B 3/4 4 1 3/4.4.2 S AF ETY/REll EF VALVES............................ B 3/4 4 l a 3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems....................... B 3/4 4 2 Ope rational Le akag e............................. B 3/4 4 2 3/4.4.4 CHEMISTRY.......................................

B 3/4 4-2 3/4.4.5 SPECIF E ACT1VITY............................... B 3/4 4 3 3/4.4.6 PRES $URE/ TEMPERATURELlHITS..................... B 3/4 4 4 3/4.4.7 MAI N ST E AM LINE I SOLAT ION V ALVES................ B 3/4 4 5 3/4.4.8 STRUCTURAL INT EGRITY............................ B 3/4 4 5 (f

3/4.4.9 RES100 AL BE AT REH0 VAL........................... B 3/4 4 5 FERMI - UNIT 2 xiii Amendment No.-;", ;; ;; ;;. S

INDEX

(

BASES PAGE SECTION 3/4.11 RADI0 ACTIVE EFFLUENTS B 3/4 11-1 3/4.11.1 LIQUID EFFLUENTS.........................................

B 3/4 11-3 3/4.11.2 GASEOUS EFFLUENTS........................................

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S i1_f b A h t & A P T t t f f" (J A f, f*

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3/4.12

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xvii FERMI - UNIT 2

-.... - - --. -. -.. - - -.... ~ - - - --

INDEX h

ADMINISTRATIVf CONTROLS SECT 104 PAGE 6.1 RESP 0NSIBillTY...........................................

6-1 6.2 ORGAN 17AT10N.............................................

61 OFFsorc. twn ots rc onnm za rton 5 6.2.1 0: ^M L..........................................

61 6.2.2 UNIl staff........................................

61 6.2.3 INDEPENDENT SAFETY ENGINEERING GR0VP..............

66 fUNCT10N..........................................

66 COMPOSIT10N.......................................

6-6 RESP 0NSIBILITIES..................................

66 AUTHORITY.........................................

66 6.2.4 SHlfT lECHNICAL ADVIS0R...........................

6-6 (2) UNIT staff OVAllflCAT10NS................................

6-7 6.4 T R A I N I NS.................................................

67 W

6.5 REVlfW AND AUDIT.........................................

67 6.5.1 ONSITE REVIEW ORGANIZATION (OSRO)

FUNCT10N..........................................

6-7 COMPOSIT10N.......................................

67 ALTERNATES........................................

68 MEETING FREQUENCY.................................

6-8 GU0 RUM............................................

68 RESPONSIBILITIES..................................

6-8 REC 0RDS...........................................

69 l

6.5.2 NUCLEAR SAFETY REVIEW GROUP (NSRG)................

69 I

IUNCT10N..........................................

6-9 COMPOSIT10N.......................................

6-10 ALTERNATES.......................................

6-10 i

CONSULTANTS.......................................

6-10 MEETING FREQUENCY.................................

6-10 QU0 RUM............................................

6-10 1

1 1

r mg FERMI - UNIT 2 xix

_ _ _ _ ~

INDfX

()

ADMINISTRATIVE CONTR0ls SECTION PJ.E REVIEW..............................................

6 10 AUDITS..............................................

6 11 REC 0RDS.............................................

6-12 6.5.3 TECHNICAL REVIEW AND CONTR0L........................

6-12 ACTIVITIES..........................................

6-12 REVIEW..............................................

6 13 SAFETY EVALUAT10NS..................................

6 13 QUAllF* CAT 10NS......................................

6-13 REC 0RDS.............................................

6 13 6.6 REPORTABLE EVENT ACT10N....................................

6 13 37 SAFETY llMIT V10lAT10N.....................................

6 14 6.8 PROCEDURES AND PR0 GRAMS....................................

6 14 6.9 REPORTING RE0VIREMENTS.....................................

6 16 6.9.1 ROUTINE REP 0RTS.....................................

6 16 STARTUP REP 0RT......................................

6-16 ANNUAL REP 0RTS......................................

6-17

)

MONTHLY OPERATING REP 0RTS...........................

6-18

. O ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT..

6 18 SEM1 ANNUAL RADI0 ACTIVE EFFLUENT RELEASE REPORT......

6 19 6.9. 2 -

SPECIAL REP 0RTS.....................................

6 21 6.9.3 CORE OPERAT ING L IM115 REP 0RT........................

6 21 6.10 RECORD RETENT104..........................................

6 21

{2 11 _ RADIATION PROTECTION PR0 GRAM..............................

6 22a 6.12 HIGH RADlATION AREA.......................................

6 22a 6.13 PROCESS CONTROL PR0 GRAM...................................

6 23 6.14 0FFSITE DOSE CALCULATION MANUAL...........................

6 24

' ' ' ~ '

~

..,,,r'.,.

3v -

I O

UNIT 2

~

xx Amendment No.-;;,

FERH]

INDEX

( )

LIST OF TABLES (Continued)

PAGE TABLE 3.3.7.9-1 0ELETED.........................................

3/4 3 68

2 '2: :

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cwam y c nrs 3.3.7.12-1

'A'u:^A^;:V; ^A^;^U^ ;T^ie NT-MONITORING M S..............................

3/4 3-77 INSTRUMENTATION asnr><.s ve 4.3,7.12-1

' ' ~ ~ ^ ^ ^ ^ ^ ^ ^ ^ ^ ~ ^ "' "' ' "'"* MONI TO R I NG INSTRUMENTATION SURVEILLANCE REQUIREMENTS......

3/4 3-81 3.3.9-1 FEE 0 WATER / MAIN TURBINE TRIP SYSTEM ACTUATION I NST RUMENT AT ION................................

3/4 3-87 3.3.9-2 FEEDWATER/ MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENT ATION SETPOINTS......................

3/4 3-88 4.3.9.1-1 FEE 0 WATER / MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS......

3/4 3-89 3.3.11-1 APPENDIX R ALTERNATIVE SHUTDOWN INSTRUMENTATION.. 3/4 3-91 4.3.11.1-1 APPENDIX R ALTERNATIVE SHUTDOWN INSTRUMENTATION SURVEILLANCE REQUIREMENTS.......

3/4 3-92

)

/~s U-REACTOR COOLANT SYSTEM PRESSURE ISOLATION 3.4.3.2-1 VALVES.........................................

3/4 4-12 3.4.3.2-2 REACTOR COOLANT SYSTEM INTERFACE VALVES LEAKAGE PRESSURE MONITORS..............................

3/4 4-12 3.4.4-1 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS........

3/4 4-15 4.4.5-1 PRIMARY COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM...............................

3/4 4-18 4.4.6.1.3-1 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM--

WIT H0RAWAL SCHEDULE............................

3/4 4-22 i

4.6.1.1-1 PRIMARY CONTAINHENT ISOLATION VALVES / FLANGES L

LOCA1ED IN LOCKED HIGH RADIATION AREAS.........

3/4 6-Ib l

3.6.3-1 PRIHARY CONTAINMENT ISOLATION VALVES...........

3/4 6-22 3.6.5.2-1 SECONDARY CONTAINMENT VENTILATION SYSTEM AUTOMATI C ISOLATION D AMPERS....................

3/4 6-53 3.7.3-1 SURVEY POINTS FOR SHORE BARRIER................

3/4 7-12 3.7.7.5-1 0ELETED.........................................

3/4 7-32 3.7.7.6-1 DELETE0.........................................

3/4 7-37 4.8.1.1.2-1 DIESEL GENERATOR TEST SCHEDULE.................

3/4 8-8 FERMI - UNIT 2 xxiv Amendment No. 59, l,<'-

~.

i I

JNDEX 10 us1 Or = Es mene o TABLE EAGE 4.8.2,1-1 BATTERY SURVEILLANCE REQUIREMENTS.............

3/4 8 12 3.8.4.2 1 PRIMARY CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES................

3/4 8 19 3.8.4.3-1 MOTOR OPERATED VALVES THERMAL OVERLOAD PROTECTION....................................

3/4 8-21 3.8.4.5 1 STAtlDBY LIQUID CONTROL SYSTEM ASSOCIATED ISOLATION DEVICES 480 V MOTOR CONTROL CENTERS...

3/4 8 27 m.m.,.,....,

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. w-i B3/4.4.6 1 RE ACTOR VESSEL TOUGHNESS.......................

B 3/4 4 6 5.7.1 1 COMPONENT CYCLIC OR TRANSIENT LIMllS...........

57 6.2.2 1 MINIMUM SHIFT CREW COMPOSITION.................

65 l:

l t.

LO FERM1 - UNil 2 xxv Amendment No. -:'.

i-lE

_.,_.,__..g.,,,___

3 p

4

. ions r

e#

I k..

i

' inoM CRl11 CAL POWER RATIO D

9 The h(NIMUM CR111 CAL POWER RATIO (MCPh) shall be the smallest CPR which

.ists.in the core.

1.

gad (REATM6' (vsTEM

[

3 An OFF-GAS UT SYSTEM is any system designed and installed to reduce i j.[

radivact.vo y, r.a ef fluents by collecting reactor coolant system offgases A

frue the r(actor coolant and providing for delay or hoidup for the purpose e

g' y}

ef reducing the total radioactivity prior to release to the environment.

l SITE DG;E CALCULATIO _N KtNUAL

~

L 24 M " N 0G. CALCULATION MANUAL (00CM) shall contain t Btruct

-'

used %n the cal o site doses WITH /#50RT 5.

methodology ano pm u uents, in the calculation J.Z'j (SEE due to radioactive gaseous uent monitoring a

~

oints, and NEXT PME).

of gaseousonduct of the r diological environmental monitorin'

~.

m OPERAglE-OPERABILITY 1.25 A system, subsystem, train, component or device shall be OPERABLE or have OPERABIllTY when it s 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 requi,ed for the system, subsystem, train, component or device to perform its function (s) are also capable of performing their related support function (s).

j OPERATIONAL CONDITION - CONDITION 1.26 An OPERATIONAL CONDITION, i.e., CO*' TION, shall be any one inclusive combination of mode switch positi, nd average react 6r coolant temperaturc as specified in Table 1.2.

PHYSICS TESTS 1.27 PHYSICS TESTS si* be those tests performed to measure the fundamental h

stics of the reactor core and related instrumentation nuclear chat 4 h

and (1) des c,! in Chaoter 14 of the FSAR, (2) authorized under t e provisions oi '0 CFR 50.59, or (3) otherwise approved by the Commissio PRESSURE BOUNDARY LEAKAGE 1.28 PRESSURE BCUNDARY LEAKAGE shall be leakage through a nonisolable fault in a reactor coolant system component body, pipe wall, or vessel wall.

PRIMARY CONTAINMENT INTEGRITY, 1.29 PRIMARY CONTAINMENT INTEGRITY shall exist when:

Al'. primary containment penetrations required to be closed during a.

accident conditions are either:

l.

Capable of being closed by an OPERABLE piimary containment automatic isolation system, or 9

1-4 FERMI - UNIT 2

~

INSERT 1.24 O

XJ OFFSITE DOSE CALCULATION MANUAL 1.24 The OFFSITE DOSE CALCULATION MANUAL (ODCH) shall contain the methodology and parameters used in the calculation of offsite doses resulting from radiorctive gaseous and 11guld effluents, in the calculation of gaseous and liquid effluent monitoring Alarm / Trip Setpoints, and in the conduct of the Radiological Environmental Monitoring Program. The ODCM shall also contain (1) the Radioactive Effluent Controls arid Radiological Environmental Monitoring Programs required by Section 6.8.5 and (2) descriptions of the information that should be included in the Annual Radiological Environmental Operating and Semiannur ~ ladioactive Effluent Release Reports required by Specification. i.9.1 7 and 6.9 1.8.

INSERT 1 30 1 30 - The PROCESS CONTROL PROGRAM (PCP) shall contain the current formulas, sampling, analyses, test, and determinations to be made to ensure that processing 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, Stat.e regulations, burial ground requirements, and other requirements governing the disposal of solid radioactive uaste.

l r

DEFINITIONS (O

2.

Ciosed by at ieast one maeual vaive, biank fiange, or deactivated automatic valve secured in its closed position, except as provided in Table 3.6.3-1 of Specification 3.6..

b.

All primary containment equipment hatches are closed and sealed.

c.

Ead prirary conhinment rir lock is in compliance with the requirements of Specification 3.6.1.3.

d.

The primary containment leakage rates are within the lim'ts of Specification 3.6.1.2.

e.

The suppression chamber is in compliance with the requirement of Specification 3.6.2.1.

f.

The sealing meciianism associated with each primary containment penetration, e.g., welds,. bellows, or 0-rings, is OPERABLE.

g.

The suppression chamber to reactor building vacuum breakers are in compilance with Specification 3.6.4.2.

WM i M.

E

[

-_P N M W THE PROCESS CONTROL PROGRAM prmytous pvt:)

1.30 7i ROCESS CONTROL PROGRAM (PCP) shall contain the provisions to ass that OLIDIFICATION of wet radioactive wastes results in a w fctm that meet the requirements of 10 CFR Part 6

. of with prope e

h low-level radio ve waste disposal sites.

The PCP s identify

W process parameters 1

.ncing SOLIDIFICATION, su s pH, oil content, H,,0 content, solids conte ratio of solidi-nion agent to waste ahd/or necessary additives for h ty anticipated waste, and the l

acceptable boundsry conditions for rocess parameters shall be identified for each waste ty ased on oratory scale and full scale testing or experience.

CP shall'also in e an identification of l

conditions that zu satisfied, based on full s testing, to assure that dewateri bead resins, powered resins, and fi sludges will result i umes of free water, at the time of disposal, wi.

the lim f 10 CFR Part 61 and of low-level radioactive waste disp es.

PURGE - PURGING 1.31 PURGE or PURGING is 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 required to purify the t.onfinement.

I RATED THERMAL POWER L

1.32 RATED THERMAL POWER shall be o tr'_..c r tor core heat transfer rate to eeactor coolant ot ' 3293 WT.

l-

'O V

l.

FERMI - UNIT 2 1-5 Amendment No.-+0-

O oErinttions 1:1.:,::::n :=

.. Z 3DM ON shall be the immobilization of wet radioactive w

.M as evaporator t resins, sludges

.,mosis concen-trates as a result of a proceo v mixing the waste type with a solidificati orm a free stan in th chemical and

^

aracteristics specified in the PROCESS CONTROL P u

s.

SOURCE CHECK 1.40 A SOURCE CHECK shall be the qualitative assessment of channel responst when the r.hannel sensor is exposed to a radioactive source.

f7AGGERED TEST BASIS 1.41 A STAGGERED TEST BASIC shall censist of:

A test schedule for n systems, subsystems, trains or other designated a.

components obtained by dividing the specified test interval into n equal subintervals, b.

The testing of one system, subsystem, train or other designated component at the oeginning of each subinterval.

THERMAL POWER 1.42 THERMAL POWER shall be the total reactor core heat transfer rate to the O

reector cooiant.

TURBINE BYPASS SYSTEM RESPONSE TIME 1.43 The TURBINE BWASS SYSTEM RESP 0MSE TIME shall be that time interval fror when the turbine bypass control unit generates a turbine bypass valve flow signal until the turbine bypass valves travel to their required positions.

The response time-may be measured by any series of sequential, overlapping or total steps such that the entire response time is measured.

7

\\

UNIDENTIFIED LEAKAGE 1.44 UNIDENTIFIED LEAKAGE shall be all learage which is not IDENTIFIED LEAKAGE.

i l

l UNRESTRICTED AREA l

1.45 An UNRESTRICTED AREA shall be any area at or beyond the SITE B0UNDARY p

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 l-recreational purposes.

I 3-i O

t FERMI - UNIT 2 1-7

i l

DEF1NITIONS

e7 TE4H4 L Ai!04 EX"A'J0T TRC AT'il"i SVGTEM--

VENTILATION EXHAUST TREATMENT SYSTEM shall be any system uesigned and '

ins o reduce gaseous radiciodine or radioactive ma aPTi) parti-nts by passing ventilation or en haust gases culate-form n through charcoal.adsor e d or HEPA fi e

or the purpose of re:noving iodines or particulates from the exhaust stream prior to the release to the environme.

ch a system sidered to have any effect on noble uents. Engineered Safety SF) etmospheric cleanu s are not considered to be VENTILATION EXHAUST

~.SV components.

VENTING 1,47 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 replac ment air or gas is not provided or required during VENTING. Vent, used in system names, does not imply a VENTING process.-

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3.7.11 The radioactive liquid effluent monitoring instrumentation channe sh in Table 3.3.7.11-1 shall be OPERABLE with their alarm / trip setroi set ensure that the limits of Specification 3.11.1.1 are not exceed The alarm /th setpoints of these channels shall be determined and adjus d in accordance ith the methodology and parameters in the OFFSITE DOSE ALCULATION MANUAL (ODC APPLICABILITY:

A 11 times.

ACTION:

a.

With a radioac iquid effluent monit ing instrumentation channel alarm / trip setpoint ss conservative an requircd by the above specification, immedia y suspend release of radioactive liquid effluents monitored by th affect channel, tr declare the channel i

inoperable, or change the s t so it is acceptably conservative.

l no b.

With less than the minimum be f radioactive liquid effluent monitoring instrumentati channels PERABLE, take the ACTION shown in Table 3.3.7.11-1.

store the ino able instrumentation to OPERABLE status wit

30 days and, if u uccessful, explain why this inoperabilit as not corrected in a ely manner in the next Semiannual Radi ctive Effluent Release Repor c.

The provisi s of Specifications 3.0.3 and 3.0.4 a not applicable.

SURVEILLANCE R IREMENTS s

4.3.7.

Each radioactive liquid effluent monitoring instrumentation channe shal

e. demonstrated OPERABLE by perfotrmance of the CHANNEL CHECK, SOURCE CHE l

C NEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations at the equencies shown 'n Table 4.3.7.11-1.

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GE MINIMUM E

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INST OP LE ACTION

=

1.

GROSS RADI0 ACTIVITY MONITOR-VIDING ALARM AND AlITOMATIC TERMINATION OF RELEA.

a.

Liquid Radwaste Effluent Line Dil-N 1

110 2.

GROSS RADI0 ACTIVITY MONITORS PROVIDI RM BUT NOT PROVIDING AUTOMATIC TERMINAT F RELEASE

/

a.

Circulating Water R voir Decant Line D11-N402 1

111 3.

FLOW RATE MEASU T DEVICES w

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With the number of channels OPEP,ABLE less than required by th Minimum Channels OPERABLE requirement, effluent releases fr a this pathway may continue provided that prior to initiati a

release:

At least two independent samples are analyzed accordance with Specification 4.11.1.1.1, and b.

least two tectmically qualified indi duals independently ver the release rate calculations

,d discharge line vaivin -

Otherwise, suspe release of radio ive effluents via this pathway.

ACTION 111 -

With the number of chan s OP LE less than required by the Minimum Channels OPERABLE uirement, effluent releeses via this pathway may continue r ded that grab samples are collected and analyzed least ce per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for gross radioactivity (betao gamma) at ower limit of detection of at least 10- micro rie/m1, for Cs-Otherwise, suspend release of radio ive effluents via t pathway.

ACTION 112 -

With the num.

of channels OPERABLE less th required by the O

ainim#m ch eis DeERABLE requiremeat, effiuea eleases via this pat ay may continue provided the flow rate estimated at lea once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months during actual releases, perfom-ance utves generated in place may be used to tstimate ow.

Ot rwise, suspend release of radioactive effluents via s

thway.

O FERMI - UNIT 2 3/4 3-73 Amendment No. 24

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A "f":^ ^^7:% L:;S'" CiLZ"' T!?'"!T !" T""N"*"':^" CV"" :LLT."^t ^:"C:" Z :'" -

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L GROSS RADIDA077/ITY RS PROVIDINr; ALARM AND AUTOMATIC TE.

10N OF RELEASE

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GROSS BETA OR GAMA RADI0 ACTIVITY MONITORS PROVIDING ALARM BUT NOT PROVIDING AUT TERMINATION OF RELEASE a.

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) The CHANNEL FUNCTIONAL TEST shall also demonstrate that automatic isola on of this pathway occurs if any of the following conditions exists:

1.

Instrument indicates measured levels above the alarm / trip se oint.

2.

' cuit failure.

(2) The CHANN FUNCTIONAL TEST shall also demonstrate that c trol room alam annunc' tion occurs if any of the following condi ons exists:

1.

Instrument dicates measured levels above the arm setpoint.

Circuit failure.

2.

3.

Instrument indicate downscale failu 4.

Instrument controls not et in ope te mode.

(3) The initial CHANNEL CALIBRATION-be performed using National Bureau of Standards traceable sources, e standards shall permit calibrating the system over the range of en<

gy measurement expected during normal operation and anticipated ope

.ional o urrences.

For subsequent CHANNEL p

CALIBRATION, sources that h e been relat to the initial calibration or v

are National Bureau of St dards traceable all be used.

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(4) CHANNEL CHECK shall c sist of verifying-indica ' n of flow during periods of release.

CHANNE HECK shall be made at least ce per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months on dhys on which con uous, periodic, or batch releas are made.

(5) The CHANNEL F TIONAL TEST shall also demonstrate that ntrol room alarm annun tion occurs if any of the following conditi exists:

l 1.

In ument indicates measured levels above the alarm setp nt.

2.

freuit failure.

1 Instrument indicates a downscale failure.

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[1MITINGCONDITIONFOROPERATION tC V P12 The - 3xr p =ut as

h=t monitoring. instrumentation channel +-

.I im 3.3.7.

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t = =: :

shown in Table 3.3.7.12-1 shall be OPERABLE with alaraAe4p-setpointe. set to ensure.that th.e limits of Koecification 0."..~. '1Fre,not e,xceede,d,. -%e-

. s.

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33. 2. L -

APPLICABILITY: As shown in Table 3.3.7.12-1 ACTION:

w Emeuvn a.

With ; ndh::th: g2s:=: : h rt monitoring instrumentation channel alars/4e+p. setpoint less conservative than required by the above Specification, 4 ed'ete'y tripe.d the - '-"..c' redf r:tir ;n==:

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declare tha channel inoperabig or change the setpoint so it is acceptably conservative.

tIW(eswE. GR5 b.

With less than the minis.um number of red'retf= ; trr: ef'hr-t monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 3.3.7.12-1.

Restore the inoperable Instrumentstion to

( )

OPERABLE status within 30 days and, if unsuccessful explain why this inoperability was not corrected in a timely na er; t tr.: cri On.Or.c..;1 hdhn',h; E'f h=t hin;; ";prt.

c.

The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS

-Exnouve 4.3.7.12 Each red h rti r gesr r: cr-* monitoring instrumentation channel shall be demonstrated OPER.EE by performance of the CHANNEL CHECK,40 WAGE--

0:: 0X, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations at the frequencies shown in Table 4.3.7.12-1.

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I MSLE 3.3.7.i2-1_lContinued)

B BLE NOTATICNS at mm.,-

- During main condenser offgas treatment system operation.

- ^^^0erintgereika of 'he-main-condenser-eir eject +N-88#.Sriq : tratier of the standby p: tr::t= ;t :y:ter.

he Schd d & T:bh ?.3.7.5 2, It:: 13.:.

ACTION STATEMENTS ACTION ::1 5 th the number of channels UPERABLE less than required by the Mini Ch is OPERABLE requirement, effluent releases via this pathw

.ay contin rovided grab samples are taken at least once per ours and these s.

es are analyzed for gross activity with hours.

Otherwise, susp release of radioactive effluent a this pathway.

- ACTIO" l20

.With the number of chann OPERABLE one than required by the Minimum Channels OPERABLE re ement riuent releases via this pathway may continue provided tha hin 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months samples are continuously collected with a iary ling equipment as required in Table 4.11.2.1.2-1.

a.CTIOV 222 With the numberi annels OPERABLE less than req by the Minimum Channels OPE

{ requirement, effluent releases via th athway may continue vided the flow rate is estimated at least once p 4

hour Otherwise, suspend release of radioactive effluents via

)

n way.

Uj ACTION 124 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, operation of main condenser offgas tres,tment system raay continue provided grab samples are collected at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and analyzed within the following 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

Otherwise, suspend release of radioactive effluents via this pathway.

-ACTIO" :2C - Wit he number of channels OPERABLE less than required by t /

Minimui annels OPERABLE requirement, effluent releases this pathway ma ontinue provided grab samples are taken east once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and t samples are analyzed for gross vity within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Otherwise, uspend release of radioact effluents via this pathway.

ACTIO" 12C - With the number of channels RABL ess than required by the Stinimum Channels OPERABLE requirement, ases via this pathway to the environment may continue for to ays provided that:

a.

The offgas syst s not bypassed, an

~

b.

The reac butiding exhaust plenun noble ga ffluent (dow

'eam) ennitor is OPERABLE; 0'%A4ey-te h-:t-least-HOT-STAN98V-wMhk 10 l-r.

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r~ 4) FERMI - UNIT 2 3/4 3-83 i TABLE 4.3.7.12-1 (Continued) TABLE NOTATIONS fJ al ', t ;;eem During main condenser uitgas treatment system operation. '" v vi n i3 vesiosivo vi ,stv, sus mm o s v u m m., s, o,, 1 visi u t. vi vI shs lui Ivy s w l.. s 11 ;j m t ;- - I m y., s, NNELFUNCTIONALTESTshallalsodemonstratethatcontrolroom 1 s 'y annuncia ecurs if any of the following conditions exist - e i alarm setpoint. ured levels ab 1. Instrument indicate 2. Circuit failure. 3. Instrume cates a downscale failure. u pc 7 J. b ..,..-os.., ous aus in vesi u sc mvuc s a i oi m vi ( tial CHANNEL CAllBRATION shall be performed using National A' of Stan a le sources. These standards sh a 1brating expected during normal the system over the ra n i es. For subsequent CHANNEL operation and anticipate CAllBRATI. iat have been rela *.ed to ) .alibration or O enai aureeu of Steaderds treceehie snaii de used. i (3) The CHANNEL CAllBRAl;0N shall include the use of standard gas samples containir.g a nominal: 1. One volume percent hydrogen, balance nitrogen, and 2. Four volume percent hydregen, balance nitrogen. ii % ft%QANNEL FUNCTIONAL TEST shall also demonstrate that automatic iso occurs ion level and that control room alarm annunciation if any of the follo onditions exists: 1. instrument indicates. red levels - the alarm setpoints. 2. Circuit failure. 3. Instrum icates a downscale failure. i instrument controls not set in the operate mode (alarm or ~ i FERMI - UNIT 2 3/4 3-84 e a&a ,,a % /A 1 1 P. A h t A A f* T T 1 f f P P P f 6 f f L!? f s IA 1 1 1 i f Af f T h P P P t lihlfT P ...-. m. _. _ _ _... --;;;a;mi U:0"- L :""C C C"""' 0" T CC C PU !. 0,. 1.1.1 The concentration of radioactive material released in liquid effluen to STRICTED AREAS (see figure 5.1.3-1) shall be limited to the concentr tions. .ified in 10 CFR Part 20 Appendix B, Table II, Column ? for r nuclides o r than dissolved or entrained noble gases, for disso1/ or entrained nob ases, the concentration shall be limited to 2 x - microcuries/ml to activity. APPLICABILITY: At all s. ACTION: With the concentration of radioactive al released in liquid effluents its, immediately restore the to UNRESTRICTED AREA 5' exceeding the ab concentration to within the above ts. .SdRVEILLANCEREQUIREMENTS m fw 4.11.1.1.1 Radi ive liquid wastes shall be sampled and analy according ( )g to the sampli and analysis program of-Table 4.11.1.1.1-1. 4.11.1 The results of the radioactivity enalyses_shall be used in at ance - wi e methodology and parameters in the ODCM to assure that the concentra s the point of release are maintained within the limits of Specification 3.11,1. 3/4 11-1 FERMI - UNIT 2 t j - m,, s m~ m.,, .s s, r m _. ~._,..me.- ~. m / \\ Lower Lim Minimum of Detec on Liqu Release Sainpling Analysis Type of Activity (L.' z Ty Frequency frequency Analysis (t 1/mi) A, Batch ste P P 7 Release Each Batch Each Batch PrincipajGamma x10 Emitters Sample Tanks (3) -6 1-131 lx10 -5 M Dissolve [nd 1x10 One tch/M Entrain, Gases (Gamm mitters) -5 P M N d Each Bhtch Composite -7 [rossAlpha lx10 -8 p Q Sr-89, Sr-90 5x10 Each Batch Comp eit ~ 1 Fe lx10 nV -7 M Principa}Garr= 5 x10 B. Continuoys d Releases omposite mitters General Service NA -6 Water System I-1\\ 1x10 (G5W) (if -5 Dissol k and 1x10 Contaminated) w M Grab ample Entrainet set (Gamma Emi rs) ) -5 \\ 1x10 H-3 M -7 d NA Composite Gross Alpha x10 9 Q Sr-09, Sr-90 5x10 d NA Composite -6 Fe-55 1x10 i O' FERMI -~ UNIT 2 3/4 11-2 l ,,,,,7_.r_ g m,, ,,m., 6. L % e LLO is defined, for purposes of these specifications, as the smalle< ct' centration of radioactive material in a sample that will yield a ne , above system background, that will be detected with 95% proba' lity cou lon with nly 5% prcbability of falsely concluding that a blank observ represt ts a "real" signal, for a part cular measurement system, which may include radiocF aical separation: 4' b LLO = Y exp (-Aat) E V 2,22 x 100 Where: LLD is the "a priori lower limit of detec on as defined above, as microcuries per uni mass or volume, s is the standard deviat n of the be ground counting rate or of b appropriate, as counts per tne counting rate of a blan. sample a

minute, E is the counting efficiency, a ounts per disintegration, V is the sample size in unite of ma or volume, 2.22 x 10s is the number disintegrat'ons per minute per microcurie, Y is the fractional r ochemical yield, w n applicable, A is the radioacti decay constant for the p ticular radionuclide, and at for plant e luents is the elapsed time between he midpoint of sample col'a ion and time of counting.

Typical v ues of E V, Y, and At should be used in the alculation. It should b recognized that the LLD is defined as an a priori fore a nc' the fact) imit representing the capability of a measurement syste not as. a_ posteriori (after the fact) limit for a particular meas. ement, b i release is the discharge of liquid wastes of a discrete volume. A ba Pri to sampling for analyses, each batch shall be isolated, and then roughly mixed by a method described in the 00CM to assure representativ tt ampling. O FEPlil - UNIT 2 ".'1 31-3 y w y = ur s TAhi P 1 !" T # T 1 P.11 ~ principal gamma emitters for which the LLD specification applies i exc ely are: Mn-54, F e-59, Co-58, 00-60, In-65, Mo 99, Cs-134, Cs-Ce-141, Ce-144. This does not mean that only these nuclides a' o be considered. er peaks that are identifiable, together with ae of the above nuclides,= 1 also be analyzed and reported in th miannual Radioactive Effluent ease Report pursuant to Spec otion 6.9.1.8. dA composite sample is one in wi the qu y of liquid samples is proportional to the quantity of liq ste discharged and in which the method of sampling employed resu n ecimen that is representative of the liquids released. Th ay be accom hed through composites of grab samples obtaine or to discharge aft he tanks have been recirculated. 'A contin - release is the discharge of liquid wastes of a no 'screte vol , e.g., from a volume of a system that has an input flow dur the inuous release, O-O roast - unil 2 3/4 11-4 -. m. m.,,, m c err,,,-m,, _, -V. _,o,,,m7 7un,,,go rgn yre. 11.1.2 The dose or dose commitment to a MEMBER OF THE PUBLIC f rom rad - ve materials in liquid effluents released, from each reactor unit, o a t. UNR RICTED AREAS (see Figure 5,1.3-1) shall be limited: During any calendar qtarter to less than or equal to 1 mrems to the a. tal body and to.less than or equal to 5 mrems to a organ, and b. Duri any calendar year to less than or equal t mrems to the total dy.and to less than or equal to 10 mrer to any organ. APPLICABILIT : At a times. J ACTION: - a. With the calculate ose from the r ease of radioactive materials in liquid effluents eeding an-f.the above limits, prepare and submit to the Commissic withi 0 days, pursuant to Specifica-tion 6.9.2, a Special Rep t at identifies the cause(s) for exceeding the limit (s) and fines the corrective actions that ' lave been taken - to reduce he eleases and the proposed corrective actions to be taken to sure t t subsequent releases will be in h compliance with the ve limits. This Special Report shall also include (1) the res ts of r'adiolog al analyses of the drinking water source and ) the radiological pact on finished drinking water supplies th regard to the requi ents of 40 CFR Part 141, ' Safe Drinkin ater Act.* b. The provi ons of Specifications 3.0.3 and 3. are not applicable. SURVEILLANCE IREMENTS s 4 4.11.1 Cumulative dose contributions from liquid effluents for the urrent cale ar quarter nd the current calendar year shall be determined in rdance with the methodology and parameters in the 00CM at least once r ac days.

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S I T.

Li F.A h P A"7 AT T

k 11.1.3 The liquid radwaste treatment system shall be OPERABLE and appropriapd po ' tons of the system shall be used to reduce the radioactive materials in liqu wastes prior to their discharge when the projected doses due to the liquic 'ffluent, from earn reactor unit, to UNRESTRICTED AREAS (see Figurc 5.'.3-1) would exceed 0.06 mrem to tae total body or 0.2 m-om t any organ 'n a 31-day period.

APPLICABILITY: t all times. ACTIC!{: a. With radioact "e liquid waste being dischargt and in excess of the above limits an. any portion of the liquid dwaste treatment syste9 not in operation, repare and t itxnit to t Commission within 30 days pursuant to Specifi tion 6.9.2 a Speci Report that includes the following information. 1. Explanation of why 1 uid r aste was being discharged without i complete treatment, id ti cation of any inoperable equipment or subsystems, and the rea for the inoperability, () 2. Action (s) taken to . store t inoperable equipment to OPERABLE status, and 3. Summary desar' tion of action (s) ta a'i to prevent a recurrence, b. The provisions f Specifications 3.0.3 and 9.4 are not applicable. SURVEILLANCE REQUlpfMENTS r r 4.11.1.3.1 oses due to liquid releases from each reactor unit to ' RESTRICTED AREAS sh be projected at least once per 31 days in accordance with e method ogy and parameters in the 00CM. . ed + 4. . 1. 3. 2 The installed liquid radwaste treatment sy; tem shall be demonst <ERABLE by meeting Spacifications 3.11.1.1 and 3.11.1.2. t /\\ FERMI - L'dIT 2 3/4 11-6 u iG '; - K ;-.I..: a _ ;,.., 7 r. c g, g r r, 7 -;;;-:- T '!! , ": ' :. :-C ',S' ' C' " :S O M " ' ' : :- The dose ra'e due to radioactive materials released in gaseor .2.1 mts from the sit' '> areas at and beyond the SITE 70VNDARI (se eff4 Figure, 1.3-1) shall be limited to the following: a. Fc noble gases: Less than or equal to 500 m, ems /yr so the total body d less than or equal to 3000 mrems/yr to t. skin, and b. For iodin 31, iodine-133, tritium, and for i radionuclides in particulate nrm with half-lives greater t i S days: Less than or equal to 1500 ems /yr to any organ. APPLICABILITY: At all times. AC110N: With the dose rate (s) exceeding the ap '. limits, immediately restore the releaseratetowithintheabove1[i.t(s) O SURVEILLANCE REQUIREMENTS x 4.11.2.1.1 The dose ate due to neble gases in gaseous e luents shall be determined to be w' nin the above limits in accordance with Se methodology and garameters the ODCM. ~ all radio-4.11.2.1.2 1e dose rate due to iortine-131, iodine-133, tritium, a nt. lides ' particulate form with half-lives greater than 8 days in g< . with the ous , shall be determined to be within the above limits in accordan efflue.21ogy and parameters in the ODCM by obtaining representative samples md meth '+ied arming analyses in accordance with the sampling and analysis program spe pe ' Table 4.11.2.1.2-1. FERMI - UNIT 2 3/4 11-8 f O .O O a ' _pr > :: nn7c--o ,,.cvc c,un,er ..m ,e,..,,,e n,m,-. 3, 7 -,c g,,r N Lower Limit r Detection. tD)* Minimum E Sampling Analysis Type of Z Gaseu.- Release Type Frequency Frequency Activity Analysis (pCi ' i ) m I I P, S P, S b 4 i A. Containme

URGE Each PURGE Each PURGE Principal Gamma Emitters 1x10-6 1x10 (Pre treatm

) Grab Sample P H-3 M 'd'* M Principal Gamma Emitter 1x10[6 D C C f 8. Reactor Buildin-1x10 Exhaust Plenum Grab Sample M H-3 Standby Gas Tgeat-ment System 1x10_f ~ D C. Radsaste Building M M Principal .ea Emitters 1x10 Turbine Building Grab Sample M H-3 R Service Building On-site Storage l y Facility -12 f 1x10 D. All Release Types Continuous W9 I '.31 -10 1x10 as listed in B Adsorbent I-133 and C above. Sample Ix10'll D I Continuous Prin 'nal Gamma Emitters P>rticulate (I-131, thers) Sample 1x10'll Gross Alpha \\ Continue > M Composite Particulate Sample 1x10'II Continuous # Q Sr-89, Sr-90 Composite Particulate Sample 1x10 Continuous # Noble Gas Noble Gases Monitor Gross Beta o-Gamma ~ ./ ... _. ~ - - - - ,.m ,,,,,,,, m y- ,-.--j

wa 5e LLD is defined, for purposes of these specifications, as the smalier 1

cc centration of radioactive material in a samole that will yield a ne , above syst em background, that will be' detected with 95% proba' lity coui with 'ly 5% probability of falsely concluding that a blank observ ton represe s a "real" signal. For a parti lar measurement system, which may include radiot mical separation: \\ 4.66 s %g b LLO = 6 Y

exp (-Aat)

E*V- .22 x 10 Where: LLD is the "a priori" ower limit of det tion as defined above, as microcuries per unit ass or volume s is the standard devie.tio of the ackgtnund counting rate or of btne counting rate of a blank imp as appropriate, as counts per t

minute, E is the counting efficiency as c nts per disintegration,

volume, V is the sample size in its of mass e

O S is the nur er of disintegratio per minute per microcurie. 2.22 x IO Y it the fraction radiochemical yield, when pplicable, l f A=is the radi ctive decay constant for the parti ilar radionuclide, and i at for p nt effluents is the elapsed time between the idpoint of -i j sample ollection and time of counting, i l Ty cal values of E V, Y, ano at should be used in the calc ation. It 5 Ould be recognized that the LLD is defined as an a priori (befo the fa ) limit representing the capability of a measurement system and not as ., a posteriori (after the fact) limit for a particular measurement. l l FERMI - UNil 2 3/4 11-10 ~ _ e r. _ n .a _-.,,,,,4 .x. ..~ ,b-w m., 'Tc principal gamma emitters for which the LLD wecification applies exclus'i y ai - the following radionuclides: Kr-87, Kr-88, Xe-133, Xe-133m, Xe-135 - ar Xe-18 in noble gas releases and Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99 l Cs-134, Cs-137. Ce-141 and Ce-144 in iodine anc' particulate reler.es. This 1 -t does not mean that only these nuclides are to be considered Other ga. a peaks that are identifiable, together with those of the ,sove nuclides, hall also be analyzed and reported in the Semiannual Ra' oactive Effluent Re 'ase Report pursuant to Specification 6.9.1.8. Sampling and at lysis shall also be performed following shut n, startup, c or a THERMAL POW change exceeding 15% of RATED THERMAL P0_ R within a 1-hour period. Th s requirement does not apply if (1) an ysis shows that the DOSE EQUIV, ENT I-131 concentration in the pri ry coolant has not increased more th. a factor of 3; and (2) the no > gas monitor shows that effluent at 'vity has not incrt.ased more ian a factor of 3. Tritium grab samples shall . taken at least on per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when either d the reactor well or the drye separator stora pool is flooded. ' Tritium grab samples shall be t en at lear. once per 7 days from the ventilation exhaust from the spen fuel ol area, whenever spent fuel is in the spent fuel pool. I The ratio of the sample flow rate t t sampled stream flow rate shall -be known for the time period cover d by ach dose or dose rate calculation 1 O meee 40 eccoreeace ith sPecifi ioas 3. .2.1. 3.11.2.z. emo 3.12.2.3. 95amples shall be changed st _ast once oer 7 Tys and analyset shall be completed within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months er chang ".g, or a er removal from sampler. Sampling shall also be pe.ormed at least once p 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for at least 7 days following each s tdown, startup or THERMA POWER change exceeding 15% of RATED THERMAL WER in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and analyses s 11 be completed within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months of anging. When samples collected or 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ate analyzed, the corr ponding LLDs may be increased by-a actor of 10. This requirement oes not apply if (1) analysis shows th the DOSE EQUIVALENT I concentration in the primary coolant has at increased more than a f ctor* of 3; and (2) the noble gas monitor show that l effluent ac* vity has not increased more than a factor of 3. hRequire when the SGTS is in operation. I ontainment shall be sampled and analyzed within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months prior t the The s rt of any VENTING or PURGING and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during NTING or PURGING of the drywell through other than the SGTS. I FERMI - UNIT 2 3/4 11-11 l O ~ ..L ~ ' ~ ~ ::= i.,C[w. '^^ i C' [ l i l i. 1 es. cir :ar 11.2.2 The air dose due to noble gases released in gaseous effluents, fre reactor unit, to areas at and beyond the SITE BOUNDARY (see figure 5 .3-1) eat shal Se limited to the following: a. ring any calendar quarter: Less than or equal to 5 mr 5 for gamma ra 'ation and less than or equal to 10 mrads for beta adiation and, b. During v calendar year: Less than or equal to mrads for gamma radiation d less than or equal to 20 mrads f, beta radiation. APPLICABILITY: At all t as. ACTION a. With the calculated air se fr a radioactive noble gases in gaseous effluents exceeding any of F. above limits, prepare and submit to the Commission within 30 's ursuant to Specification 6.9.2, a Special Report that ide fies - cause(s) for exceeding the limit (s) and defines the corr ive action + hat have been taken to reduce the releases and t proposed correc 've actions to be taken to i assure that subs utnt releases will be n compliance with the () above limits, b. The prov. ons of Specifications 3.0.3 and 3. are not applicable. SURVEILLANC EQUIREMENTS x 1 4.11. 2 Cumulative dose contributions for the current calendar quare r and cu ent calendar year for noble gases shall be determined in acccrdance th ne methodology and parameters in the ODCM at least once per 31 days. e k O FERMI - UNIT ? 3/4 11-12 1 i -O -

^ 2

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10;a L.J.; .m , ;n' : :m 7 ::, m., m' ' ' : " ^ ^ :.'. : : C ' T^r ^.'^:^ w 11.2.3 The dose to a MEMBER OF THE PUBLIC from iodine-131, iodine-1 ater ium, and all radionuclides in particulate form with half-lives g tr days in gaseous efflucata released, from each reactor unit, o areas than ed to the eyond the SITE BOUNDARY (see figure 5.1.3-1) shall be li at an. followin a. Du. g any calendar quarter: tess than or equal.

7. 5 rei; to any orga nd,.

b. During an ' calendar year: Less than or eq to 15 mrems to any organ. APPLICABILITY: At all 'mes. ACTION: a. With the calculated d from release of iodine-131, iodine-133, i particulate form with half lives greater tritium, and radionucli ' Jents exceeding any of the above limits, than 8 days, in gaseous e c. itsion within 30 days, pursuant te prepare and submit to th Specification 6.9.2, a ecial' eport that icentifies the cause(s) for exceeding the lin ' and defi s the corrective actions that have () been taken to redu the releases 1 the proposed corrective actions to be taken to a re that subsequen eleases will be in compliance with the above

mits, b.

The provisi s of Specifications 3.0.3 and 0.4. ave not applicable. SURVEILLANCE RE IREMENTS \\ Cumulative dose contributions-for the current calendar q "ter and .4.11.2.3 lides' calendar year for iodine-131, iodine-133, tritium, and radiot. curren ed ticulate form with half-lives greater than 8 days shall be deter, in p-in ccordance with the methodology and parameters in the 00CM at least on j _ r 31 days. [- l. 1 FERMI - UNIT 2 3/4 11-13 l l l ..-o,c c rr, o -. m. m,, t. c g -crr ===c: :::::- g.,,,y 7,7,,3.,,3 7 ~,, 11.2.4 The OFF-GAS TREATMENT SYSTEM shall be OPERABLE and shall be op ation. APPLIC. ITY: Whenever the main condenser steam jet air eject, are in operation. ACTION: With the -GAS TREATMENT SYSTEM inoperab for more than 7 days, prepare and Smit to the Commission wi n 30 days, pursuant to Specification b. .2, a Special Report at includes the following information: 1. Identification o Se inor able equipment or subsystems and the reason for the 1 -

ability, 2.

Action (s) taken to estor the inoperable equipment to OPERABLE status, ar.d 3. Summary des ption of action (s) ken to prevent a recurrence, i-b. The provisi of Specifications 3.0.3 and 0.4 are not applicable. c. The pro sions of Specification 4.0.4 are not a, icable, SURVEILLAN REQUIREMENTS r x 4. .2.4 The OFf-GAS TREATMENT SYSTEM shall be demonstrated OPERABLE by m ting ' ecifications 3.11.2.1, 3.11.2.2, and 3.11.2.3. 4 i na9.1f=_a eT TtiF F f* f= g t, f= g, y r F LtY T f ATT 11 F V [] A f I f T T FnTk PieT F if P T f" L A

":':':C CCt'T:C: FC? Cr:r":::N -

11.2.5 The VENTILATION EXHAUST TREATMENT SYSTEM as described in the ODr sh 1 be OPERABLE and appropriate portions of the system shall be used redus radioactive materials in gaseous waste prior to their dischar. when the projec d doses due to gaseous effluent releases from the site to U' STRICTED AREAS (se Figure 5.1.3-1) would exceed 0.3 mrem to any organ in y 31-day period. , APPLICABILITY: t all times. ACTION: a. With radioact 'e gaseous waste being disc rged in excess of the above limits and any rtion of the VENTILATI EXHAUST TREATMENT SYSTEM not in operation, pre re and submit to t Commission within 30 days pursuant to Specifi tion 6.9.2 a 5 cial Report that includes the following information. 1. Identification of an 'n erable equipment or subsystems, and the reason for the ino

ability, 2.

Action (s)-taken to estore e inoperable equipment to OPERABLE status, and 3. Summary desc ption of action (s) ken to prevent a recurrence, b. The provision of Specifications 3.0.3 and .0.4 are not applicable. SURVEILLANCE REQUI NTS s 4.11.2.5.1-oses due to gaseous releases from the site shall be ojected at least onc er 31 days in accordance with the methodology and param ers in the ODCM, w .1 any portion of the VENTILATION EXHAUST TREATMENT SYSTEM is in us 4. . 2. 5. 2 The VENTILATION EXHA'JST TREATMENT SYSTEM shall be demonstrated ERABLE by meeting Specifications 3.11.2.1, 3.11.2.2, and 3.11,2.3. I l FERMI - UNIT 2 3/4 11-15 l l l 3/N // RADIOACTIVE EFFLUENTS 314.#. 2. G ASEcu 5 GF Pt.V F. N T b ( }' EXPLOSIVE GAS MIXTURE LIMITING CONDITION FOR OPERATION 3.11.'.6 The concentration of hydrogen in the main condenser of fgas treatment system shall be limited to less than or equal to 4% by volume. APPLICABILITY: Whenever the main condenser steam jet air ejectors are in operation. ACTION: a. With the concentration of hydrogen in the main condenser offgas treatment system exceeding the limit, restore the concentration to within the limit within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. SURVEILLANCE RE0VIREMENTS 4.11.2.0 The concentration of hydrogen in the main condenser of fgas treatment system shall be determined to be within the above limits by continuously monitoring the waste gases in the main condenser offgas treatment system with the hydrogen monitors required OPERABLE by Table 3.3.7.12-1 of Specification O 3.3.7.12: FERMI - UNIT 2 3/4 11-16 {} l RADI0 ACTIVE EFFLUENTS MAIN CONDENSER LIMITING CONDITION FOR OPERATION 3.11.2.7 The gross radioactivity rate of noble gases measured at the discharge of the 2.2 minute delay piping shall be limited to less than or equal to 340 millicuries /sec after 30 minute decay. APPLICABILITY: OPERATIONAL CONDITIONS 1, 2*, and 3*. ACTION: With the gross radioactivity rate of noble gases at the discharge of the 2.2 minute delay piping exceeding 340 millicuries /sec after 30 minute decay, restore the gross radioactivity rate to within its limit within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least STARIUP, with all main steam lines isolated, within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . SURVEILLANCE REOUIREMENTS 4.11.2.7.1 The radioactivity rate of noble gases at the discharge of the f-2.2 m.inute delay piping shall be continuously monitored in accordance with (. O The gross radioactivity rate of noble gases from the main condenser 4.11.2.7.2 steam jet air ejector shall be determined to be within the limits of Specifica-tien 3.11.2.7 at the following frequencies by performing an isotopic analysis of a representative sample of gases taken at the discharge of the 2.2 minute delay piping: a. At least once per 31 days. b. Within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months fol % ing an increase, as indicated by the Offgas Radiation Monitor of greater than 50'., after factoring out increases due to thanges in THERMAL POWER level, in the nominal steady-state fission gas release from the primary coolant. The provisions of Specification 4.0.4 are not applicable. c. (x0 "When tne main condenser air ejector is in operation. FERMJ - UNIT 2 3/4 11-17 Amendment No.-49 D10 ACTIVE EFFLUENTS 'VE ING OR PURGING LIMIilk CONDITION FOR OPERATION 3.11.2.8 -VL TING or PURGING of the Mark I containment shall be throug he standby gas t 'atment system or the reactor building ventilation sys APPLICABILITY: 5enever the co 'ainment is vented or purged. ACTION: a. With the re irements of the above specificati not satisfied, suspend all VENTING a PURGING of the drywell, b. The provisions o Specifications 3.0.3 J 3.0.4 are not applicable. SURVEILLANCE REQUIREMENTS s s 4.11.2.8.1 The containment shall se ermined to be aligned for VENTING or PURGING through the standby gas tr aent system or the reactor building ventilation system within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months r to start of and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during VENTING or PURG v of he containment. 4.11.2.8.2 Prior to use of e purge sy *em through the standby gas treatment system assure that: a. Both standb gas treatment system ains-are OPERABLE whenever the purge sys m is in use, and i b. Whene r the purge system is in use du 'ng OPERATIONAL CONDITION 1 or " or 3, only one of the standby gas t eatment system trains may used. 4.11.2 .3 The containment drywell shall be sampled a analyzed per Tabl .11.2.1.2-1 of Specification 3.11.2.1 within 8 h rs prior to the st-of and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during VENTING ar PURGING of the 5 ywell through other than the standby gas treatment syste I FERMI - UNIT 2 3/4 11-18 l -RA..-s I5 [ o,,, cm ,n , e -,, m. 1,-

' : *c T ^ " : "

c s

1.3 Radioactive wastes shall be SOLIDIFIED or dewatered in accordance wi the 70 CESS CONTROL PROGRAM to meet shipping and transportation requireme during vansit, and disposal site requirements when received at the di ssal site. APPLICABillT). At all times. ACTION: a. With 50L101 'ATION or dewatering not meetin isposal site and shipping and ti.'sportation requirements, spend shipment of the inadequately proct. ed wastes and corr the PROCESS CONTROL PROGRAM, the procedures and/o he solid was system as necessary to prevent recurrence, b. With SOLIDIFICATION or dewa ng not performed in accordance with the PROCESS CONTROL PROGR ,( demonstrate by test or analysis that the improperly pro ssed was in each container meets the requirsments f or tr. ortation to t disposal site and for receipt at the disposal s and (2) take appro ate administrative action to prevent recurr e. c. The provis is of Specifications 3.0.3 and 3.0, are not applicable. SURVEILLANCE R IREMENTS O s s 4.11.3 PROCESS CONTROL PROGRAM shall be used to verify that the operties of th gackaged waste meet the minimum stability requirements of 10 CFk ' rt 61 and s her requirements for transportation to the disposal site and receipt .e disposal site. O FERMI - UNIT 2 3/4 11-19 ""':C/ :T : /: C ir wi ^- O." '. '. " T O T % C : 4 ":": ::Z:T: ,T CimTL - PUBLICduetoreleasesofradioactivityandtoradiationfromurani-[ fue The annual (calendar year) dose or dose corra.itment to any MEMBER 11.4 T cyc ' sources shall be limited to less than or equal to 25 mrems to th total body any organ, except the thyroid, which shall be limited to les han or equal 75 mrems. APPLICABIL TY: At all times. ACTION: a. With e calculated doses from the release of ra oactive materials in liqu or gaseous effluents exceeding twice ie limits of Specifica-tion 3.1 .2a., 3.11.1.2b., 3.11.2.2a., 3.11 .2b., 3.11.2.3a., or 3.11.2.3b., calculations should be made inc1 ing direct radiation contribution from the reactor units and f m outside storage tanks to determine ther the above limits of ecification 3.11.4 have been exceeded. f such is the case, pr 2are and submit to the Commission within 10 days, pursuant t pecification 6.9.2, a Special Report tha defines the corr tive action to be taken to reduce subsequent re ases to prev t recurrence of exceeding the above limits and incl .s the sc ule for achieving conformance with the above limits. his Sp tal Report, as defined in 10 CFR 20.405c, shall include an nal is that estimates the radiation O expo ure (eose) to e "tsae the eus'ic <ro urea 4u <uei cvcie sources, including all eff' t pathways and direct radiation, for the calendar year that in ude the release (s) covered by this report. It shall also scribe evels of radiation and concentra-tions of radioactive er erial inv ved, and the cause of the exposure levels or c centrations. f the estimated dose (s) exceeds the above 1. its, and if the elease condition resulting in violation of 40 C Part 190 has not a eady been corrected, the Special Report all include a request r a variance in accordance with the provi ons of 40 CFR Part 190. bmittal of the report is considered a imely request, and a variance is granted until staff action on . request is complete. b. The prov tons of Specifications 3.0.3 ana 4.0. are not applicable. ' SURVEILLANCE R IREMENTS s 4.11.4.1 oulative dose contributions from liquid and gaseous e luents shall be determ' ed in accordance with Specifications 4.11.1.2, 4.11.2.2, and 4.11.2.3, and in cordance with the methodology and parameters in the ODCM. 4.11 .2 Cumulative dose contributions from direct radiation from the actor uni ~ arid from outside storage tanks shall be determined in accordance vi t-methodology and parameters in the ODCM. This requirement is applicabit nly under conditions set forth in Specification 3.11.4, ACTION a. g O nRMI - um 2 u4 u-20 \\ 2," r N. 0" N D T : ':"; "^ :'0 7 ~ 7_.7, ., 7q.,, ms s g -.~ 3 - 3.'2.1 The radiological environmental moaitoring program shall be conduc J as ; ecified in Table 3.12.1-1. APPLICx '1 LITY: At all times. ACTION: a. With he radiological environmental monitoring progr< not being condut. d as specified in Table 3.12.1-1, prepare a submit to the Commiss 7, in the Annual Radiological Environmen i Operating Report required > Specification 6.9.1.7, a descriptio f the reasons for not conduct g the program as required and the, alans for preventing a recurrence, b. With the level o radioactivity as the re It of plant effluents in an environmental s.pling medium at a s. cified location exceeding the reporting levels of Table 3.12.1-2 hen averaged over any calendar quarter, prepare and mit to the C alssion within 30 days, pursuant to Specification 6,9.2, Special port that identifies the causc(s) for ex.ceeding the limit (s and do nes the corrective actions to be O taken to reduce radioactive ffl ents so that the potential annual dose

to A MEMBER OF THE PUB is less than the calendar year limits of Specifications 3.11.1.2,

.2.2, and 3.11.2.3. When more than one of the radionuclices i abl 3.12.1-2 are detected in the sampling medium, this rep,t shal he submitted if: concentration (1 concent, tion (2) reporting level ll, reporting vel (2) - 1.0 + ...) When radionuclide other than those in Tab 3.12-2 are detected and are the result plant effluents, this repo shall be submitted if the potential anual dose

to A MEMBER OF THE

'BLIC from all radio-nuclides is ual to or greater than the calenda ear limits of Specificati ns 3.11.1.2, 3.11.2.2, and 3.11.2.3. is report is not required the measured level of radioactivity was at the result of luents; however, in such an event, the condi 'on shall be plant e report J and described in the Annual Radiological Envir mental Operr ing Report pursuant to Specification 6.9.1.1. W' h milk or fresh leafy vegetable samples unavailable from 'e or c. , ore of the sample locations required by Table 3.12.1-1, ident. to the specific locations for obtaining replacement samples and add the, radiological environmental monitoring program within 30 days. The specific methodology used to estimate the potential annual dose to a MEMBER OF THE

T'JBLIC shall be indicated in this report.

O FERMI - UNIT 2 2/4 12-1 1 f U; ;; m s^ ?m [N. : RO, ': NT L " : ' O P ' s. O ,a,.e e v-,~,ev re-e en.,,m. r e. ._, : ~._ n " TION: (Continued) locations frcm which samples were unavailable may then be dele' - the monitc ring program. Pursuant to Specification 6 4.8, the cause of the unavailability of samples ar dentify iden u the new ^ation(s) for obtaining replacement sam ' in the next Semiannual

n. ioactive Effluent Release Repor

. rsuant to Specifi-cation 6.9.1.8 ' also include in the rer a revised figure (s') and table for the 1 reflecting th s location (s). d. The provisions of Specifica 3.0.3 and 3.0.4 are not applicable. SURVEILLANCE REQUIREMENTS 4.12.1 The iological environmental monitoring samples sha. e collected Table 3.12.1-1 from the specific locations given in table and pursuan v s ) in the ODCM, and shall be analyzed pursuant to the requiren. "s of fi " s . ale 3.12.1-1 and the detection capabilities required by Table 4.12.1-i 1 O FERMI - UNIT 2 3/4 12-2 f] O O. o ~ m m_,.., 7 m.. - m _ m. m. ,-.,c.,,, ...-.,s.. Number of g Representative sure Pathway Samples and Sampling and Type and Frea acy a a or Sample Sample' Locatirns Collection Frequency of An_alys', b 1. DIR RADIATION 37 routine monitoring stations, Quarterly Gamma quarterly. with two or more dosimeters placed as follows:

1) an inner ring of stations in the general area of the SITE UNDARY and additional rings roximately 2, 5, and 10 at i

miles, ith a station in at other meteorological least ev sector for h ring with the w2 exception of t e sectors over Lake Erie. The b nce of.the i

7 stations, 8, should laced in special interest are such as population centers, nea. residences, schools, and i c or 3 areas to serve as - rol stations. t 2. AIRBORNE Radiciodine and Samples from locations. Continuous npler Radioiodine Cannister: Particulates operation wit. ample 1-131 analysis weekly.

a. 3 s., ales from close to collection weekly, or

$.. 3 SITE BOUNDARY loca-more frequently if ions, ir, different sectors, required by dust '1rticulate Sampler: of the highest calculated loading. oss beta radioactivit* ana f s folloging i annual average ground-level D/Q. filter ange;

b. I sample from the vicinity of Gamma isotop. analysis' a community having the highest of composite (by calculated annual average ground-location) quarterly.

level D/Q. O o O "f E 1 ? ' ? ! ' t : ^ - ~ -9 ? ^^^:CLC": J L :^ : "C " " T " ' "i ' "" ' ' " : "^ ' C "; ^.-

o3 Number of Representative.

i c < nosure Pathway Samples and Sampling and Type and frequer, 5 a.. r Sample Sample, locations" Collection Frequency of Analysi w c. I sample from a control ~ location, as for example 15-30 km distant and in direction.grevalentwind the least 3. WATERBORNE a. Surface

a. I sample up am.

Composi sampfeover Gamma isotopic analysis' f

b. I sample dowas m.

1-mo period monthly. Composite for tritium analysis c;uarterly. b. Ground Samples from l'or 2 sources Quarterly Gamma isotopic' and tritium i only if ikely to be analysis quarterly if ground water flow reversal is noted. affected y c. Drinking

a. I sample of each of to Comp

'te sample I-131 analy!is on each 9 3 of the nearest ter over 2 w k period composite when the dose supplies that uld be when 1-131 lysis calculated for the consump- +hly tion of the water is greater affected b its discharge. is performed, .m composite otherwis thar. 1 mrem per year. Com-

b. Is e from a control esIte for gross beta and isotopic analyses

1 ation.

go. montha Composite for tritium a. vsis quarterly, d. Sedimen I sample from downstream area Semiannually Gamma isotopic an.' sis' fro with existing or potential semiannually. ..oreline recreational value. g O' O o n. _n m -.m. m ~. _ _-.-. _..~. 3 Number of Representative a g Hosure Pathway Samples.and Sampling and Type and Frequern 3 a..J'or Sample Sample Locations Collection Frequency of Analysi" r m 4. INGL ON a. Milk 'a. Samples from milking animals Semimonthly when G isotopic

and I-131 in 3 locations within 5 km animals are on analysis semimonthly when distance having the highest pasture, montnly at animals are on pasture; dose potential.

If there other times monthly at other times. e none, then, 1 sample fr milking animals in each of 3 - s between 5 to 8 km distant e doses are calculated eg ater-than 1 mrem per w 1

b. I sample from milkin animals at a contro o -

7 tion 15-30 km d' . ant and in the 1 at prevalent wind dire on. b. Fish and

a. I s<

e of each commercially Samp in season, or Gamma isotopic analysis' if they on edible portions. Inverte- - recreationally important semiann - brates species in vicinity of plant are not seaa 'ai discharge area. '

b. I sample of.ame species in areas not influenced by plant discharge.

Food a.1 sample of each principal At time of harvest) Gamma topic analyses' Products class of food products from on edible tion. i any area that is irrigated j by water in which liquid plant wastes have been discharged. i \\ (3, O v-O,~ ~. .g ..nir fr._.:__n- -^B

9*T"It I k!? f T 00 ALMP LIT F T LAO h! T T O O T kl_M 0 0 0. O f tA 11 i

F T y n, x Number of 25 Representative sure Pathway Samples and Sampling and Type and Freo" .j - 25 anoi o 'n le Sample Locations Collection Frequency of An '_,s a c: omma isotopic" and I-131. c. Food

b. Samples of 3 different. kinds Monthly when

[l Products road leaf vegetation grown available analysis. of two different (cont' d) neares offsite loca f highest e predicted annual ave. + level D/Q if milk sam 3 is performed. c. - se of each of the similar Monthly when Gamma isotopic'.and I-131 road leaf vegetation grown available a.. ,-is. 15-30 km distant in the least prevalent wind direction if milk sampling is not performed.

a t

v e e+ r ' : : : '"- :

C_

,~7 7. y_ _ ' aecific parameters of distance and direction sector from the centerline reactor, and additional description where pertinent, shall be provi d on. for sch and every sample location in Table 3.12.1-1 in a table and fi re(s) in the ODCM. Refer to NUREG-0133, " Preparation of Radiological Effl nt Tech-nical S cifications for Nuclear Power Plants," October 1978, and Radio-logical essment Branch Technical Position, Revision 1, Novembe 1979. Deviations e permitted from the required sampling schedule if specimens are unobtainable ue to hazardous conditions, seasonal unavailabi ty, malfunction of automatic s-ling equipment and other legitimate reason If specimens are unobtainable due o sampling equipment malfunction, every fort shall be made to complete corre ve action prior to the end of the ne sampling period. All deviations from e sampling schedule shall be doc ented in the Annual Radiological Environm tal Operating Report pursuant a Specification 6.9.1.7. It is recognized that, times, it may not be poss' le or practicable to continue to obtain sample of the media of choice t the most desired location or time. In these instance suitable specific ernative media and locatinns may be chosen for the parti lar pathway in stion and appropriate substi-tutions made within 30 days in he radiologi 1 environmental monitoring program. Pursuant to Specification 6.9.1. identif the cause of the unavailability of samples for that pathway and i ntify te new location (s) for obtainir.g nn 1 Radioactive Effluent Release replacement samples in the next Sem Report and also include in the report revised figure (s) and table for the ODCM reflecting the new locatio s). O 0ne or more 4nstruments, such a, ,, esse, .d ion chamber, fo, measuring eed 9 recording dose rate continuous may be used 'n place of, or in addition to, integrating dosimeters. For ,e purposes of t s table, a thermoluminescent dosimeter (TLD) is consider tc be one phospho two or more phosphors in a packet are considered as o or more dosimeters. ilm badges shall not be used as dosimeters for asuring direct radiation, e frequency of analysis or readout for TLD sys 'ms will depend upon the chara eristics of the specific system used and shou be selected to obtain optimum do information with minimal fading. The purpose of is sample is to obtain background informati If it is not C practical to tablish control locations in accordance with th. distance and wind directi n criteria, other sites that have valid background ta may be substitute. i d particulate sample filters shall be analyzed for gross beta Airbor radio etivity 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or more af ter sampling to allow for radon and th "on dau ter ' decay. 'If gross beta activity in air particulate samples is gro ter t n 10 times the yearly mean of control samples, gamma isotopic analysis all be performed on the individual samples. O FERMI - UNIT 2 3/4 12 7 ,,,,,7. m eu e M ,.,.7,3 mma isotopic analysis means the identification and quantification s from' ogam.. -emitting radionuclides that may be attributable to the efflu-the fau'lity. The "upstre sample" shall be teken at a distance beyond gnificant influence of t. discharge. The " downstream" sample sh, be taken in an the mixing zone. " Upstream" sam es in an estuary must area beyond but n 'r be taken far enough ' tream to be beyond the pla influence. 9A composite sample is one ', which the quant (aliquot) of liquid sampled is i proportional to the quantity f ficwing 11 id and in which the method of sampling employed results in a..cimen iat is representative of the liquid. i aliquots shall be collected at time

flow, in this program composite i

intervals that are very short (e. urly) relative to the compositing period (e.g., monthly) in order o assu obtaining a representative sample. Groundwater samples shall taken when this 'urce is tapped for drinking or irrigation purposes in eas where the hydrauli 'radient or recharge proper-ties are suitable fo ontamination. The dose shall calculated for the maximum organ and a - group, using the methodology parameters in the ODCM. lf harv occurs more than once a year, sampling shall be perfo d during I each screte harvest. If harvest uccurs continuously, sampling s u 1 be f] m ehly. Attention shall be paid to including samples of tuberous an 'oot V ood products. FERMI - UNIT 2 2/4 12-8 l l ~ - O O }o 7,^.; L: 2.::..I ^ _ rra"-*"- :""r*: "^^ ^^2:^^r: : '" ^ ^ ^ : ^ "" ^ ^ ' I ^^ :2 : E7 I CEI" ^ o___ 9 7...,._... Water Airborne Particulate Fish Milk Food Pr - ts ro Analys (pC1/2) or Gases (pC1/m ) (pCi/kg, wet) (pCf/1) (pC ' g, wet) 2 H-3 20, Mn-54 1,000 30,000 t Fe-59 400 10,000 Co. 8 .1,000 30, i M 10,000 Co-60 300 x-C 000 Zn-65 300 <O Zr-Nb-95 400 3 100 I-131 2 0.9 Cs-134 30 10 1,000 1,000 Cs-137 20 2,000 70 2,000 300 Ba-La 200 \\ drinking water samples. This it. 40 CFR Part 141 value. O O_ o m' Q a ervervenu can ovirvere enn roue nn yc ur a, e -n. r tea.uc c ,msb,c , n,.c n,,-,, mc 5 r r, c _,,,._, s Water Airborne Particulate Fish Milk Food Products Seo. - s Analysi (pCi/2) or Gas (pCi/m ) (pCi/kg, wet) (pCi/1) (pCi/kg, wet) / /kg. dry) 3 gross beta 0.01 H-3 2000 Mn-54 15 130 Fe-59 30 260 { Co-58,60 15 y Zn-65 30 260 s Z r-Nb-95 15 d 1 I-131 I O. Cs-134 15 0.05 130 15 60 150 Cs-137 18 0.06 150 18 80 80 15 Ba- -140 15 r + -O + = = = = = 7..,. - m y e-is list does not mean that only these nuclides are to be considere, -O er peaks that are identifiable, together with those of the abov-ides, shall also be analyzed and reported in the Annual Radio gical nuEnvi,nmental Operating Report pursuant to Specification 6,9.1.7 Require ^etection capabilities for thermoluminescent dosime rs used b 1 for envirc ental measurements are given in Regulatory Guid 4.13, The LLD is de 'ned, for purposes of these specification as the smallest c concentration o radioactive material in a sample that ill yielo a net count, above sys background, that will be detecte ith 95% probability -with only 5% proba lity of falsely concluding that blank observation represents a "real"

ignal, For a particular measu ent system, which may clude radiochemical separation:

4' S b LLD = - ex -Aat) 2.22 V E Where: LLD is the "a priori" lower 1 ,1 of detection as defined above, as picocuries per unit mass -vol e, s is the standard deviat n of the b kground counting rate or of sttie. counting rate of a b nk sample as propriate, as counts per

minute, E is the counting ef ciency, as counts per isintegration, V is the sample-s e in units of mass or volumt 2.22 is the.nu. er of disintegrations per minute r pidocurie,

~ Y is the fr ional radiochemical yield, when applic le, l A is the dioactive decay constant for the particular r dionuclide, f and At fo environmental samples is the elapsed time-between sam le coll ction, or end of the sample collection period, and time o co ting pical valves of E, V, Y, and at should be used in the calculatio i- / O FERMI - UNIT 2 3/4 12-11 l-1 l i ,,3, rg_g_g. ,,nir s i.g,, g i7 _ 1, n, r <hould be-recognized that the LLO is defined as an a pricri (be - tM imit representing the capability of a measurement em and not as an a riot'i (after the fact) limit for a pa .o ar measurement. Analyses shall be med in such a manner tha stated LL0s will be achieved under routine c ons. Occasie-y background fluctuatiot.s. unavoidable small sample sizes, nce of interfering nuclides, or other uncontrollable circumsta. ey er these LLDs unachievable. In such cases, the contri ig factors sha identified and described in the Annual Radio al Environmental Operating rt pursuant to specification .7. d ~ drinking water samples. L i O nam - um 2 m u-u n o n, yf_- m t go, n y.m e m e..,.~ ; 7. m r? ' ":0 'Jr Ut:JC ,o ,+q rnunp y rgn gnrni gu . 12.2 A lano use census shall be conducted and F il identify within dL ance of 8 km (5 miles) the location in each of the 16 meteorologi I sec rs of th? nearest milk animal, the nearest residence and the ne est garde of greater than 50 m (500 ft ) producing broad leaf veget< ion. 2 2 APPLICAB TY: At all times. ACTION: a. With land use census identifying a location (s that yields a calcu-lated e or dose commitment greater than the alues currtntly being calculate in Specification 4.11.2.3, identi the new location (s) in the next St. iannual Radioactive Effluent Re ase Report, pursuant to-Specificatio 6.9.1.8, b.~ With a land use ensus identifying a 1 ation(s) that yields a calculated dose o dose commitment (v' the same exposure pathway) 20% greater than a location from nich samples are currently being obtained in ac rdance with ecification 3.12.1, add the new location (s) to the adiologi 1 environmental monitoring program within 30 days, he sa ling location (s), excluding the control station location, v g the lowest calculated dose or dose commitment (s), via the same posure pathway, may be deleted from this monitoring program af r ctober 31 of the year in which this land use census was condu ed. ursuant to Specification 6.9.1.8, identify the new locati (s) in se next S.miannual Radioactive Effluent Release Repor and also i lude in the report a revised figure (s) and table f r the O'CM re _cting the new location (s), c. The provisions of pecifications 3.0.3 nd 3.0.4 are not applicable. SURVEILLANCE REQUIREMEN 4.12.2 The land u census shall be conducted during the g wing season at least once per 12 onths using that information that will pr-ide the best i results, such a y a door-to-door survey, visual survey, aeri survey, or. by consulting cal agriculture authorities. The results of the and use census shall e included in the Annual Radiological Environmental nerating Report purs nt to Specification 6.9.1.7.

Broad af vegetation sampling of at least three different kinds of veg ation may t. performed at the SITE BOUNDARY in each of two dif ferent direction ctors wit the highest predicted D/Qs in lieu of the garden census.

Specificati s f broad leaf vegetation sampling in Table 3.12.1-1, Part 4.c., shall be llowed,-including analysis of control samples. FERMI - UNIT 2 2/1 12-13 r l l t ,. g e g. n a v r-a t_ _ u e vig n ig rane,naeu-uv. =('s. ...as.n,enu ,4. ,creo, nen..nnt n,-.u_ .m.v mg -ey,, ,,r,- . 2,3 Analyses shall be performed on radioactive materials supplied as pa of Interlaboratory Comparison Program that has been approved by the H Commis>n. I APPLICABILIT). At all times. ACTION: a, With analyses no ming performed as r aired above, report the corrective actions n to preven recurrence to the Commission in the Annual Radiologi Envi .imental Operating Report pursuant to Specification 6,9.1.7. b, The provisions of Spe ications .3 and 3.0.4 are not applicable. SURVEILLANCE REQUIREMr s s f 4.12.3 T nterlaboratory Comparison Program shall be describe the 00CM. A summ. of-the results obtained as part of the above required Inte labr tory Comparison Program shall be included in the Annual Radfulogi O .vironmental Operating Report pursuant to Specification 6.9.1.7. FERMI - UNIT 2 3/4 12-14 (]) INSTRUK!NTATION BASES _ MONITORING INSTRUMENTAT_IR (Continued) 3/4.3.7.8 CHLORINE DETECTION SYSTEMS 1.ie OPERABli.!TY of the chlorine detection sy; tem ensures that en accidental chlorine release will be detected promptly and the necessary protective actions will be automatically initiated to provide protection for cont-c1 room personnel. Upon detection of a high concentrati.n of chlorine, the contr( <a emergency ven-tilation system will automatically be placed in the chlori' snde of opera-tion to provide the required protection. The detection system r, quired by this specification is ansistent with the recommendations of Regulatory Guide 1.95 " Protection of Nuclear Power Plant Control Room Operators against an Accidental Chlorine Release". Revision 1 January, 1977. 3/4.3.7.9 DELE 1ED 3/4.3.7.10 LOOSE-pART DETECTION SYSTEM The CPERABILITY of the loose part detection system ensures that sufficient capability is available to detect loose met:llic perts in the primary system and avoid or mitigate damage to primary system components. The allowable out-of-service tires 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. W

/..'.;; i'2: 1'fT:": ^" 0 T r i":!" "^ M " ^ ^ T ! !$ ? """E "* I ^"

dioactive liquid effluent monitoring instrumentation is provi. s in monitor ano as applicable, the releases of radioacti liquid effluents dor 1 or potential releas d effluents. The e calculated and adjusted in alarm / trip setpoints for these n the ODCH to ensure that the accordance with the methodo1 ra alarm / trip will o to exceeding the lin CFR Part 20. The OPERAB use of this instrumentation is consistent w requirements ,.etG neral Design Criteria 60, 63, and 64 of Appendix A to 10 CFR a l { O EtRsi. us11 2 Amenemeet ne. e o 3,4 3-s w y --,,-.,--e s JNSTRLMENTATION O l" tiQS110 RING INSTRLMENTATION (Continued) L eu r.ie E ffE^JJ"YI,. GAS =', ITTn= My4110 RING INSTRLMENTATION 3/4.3.7.12 yn r The-nf.r:qp=. gas== ;fn;,a; monitoring instrumentation is provided to monitor r f n t=1, n ;;;;i nti;, tt; 7:1;;;;; ef 7;di;ntin sta;;;;- nr eff!r t i t =txt : pt=ti:1 nt:== :f=== cff!=:t:. in :mr=/t '; et;&t: f;r tin: intrr=t: &?? 5: eld!&d 29 Th

d'=td in Me ter dthth: sti;f;;;;y =d-

= :t;r; fr. Ot; ^^t" -i;titi:ing th : :t= d=';n. ?!= =t= = :;=ifI=d ir ^'#" ! n dr: --Lie ? M=

n==: ti= :-thd i: erf Srrn &.tr' ! 6: ;/te!; &

nt: trer r at tr'r. Th': d?? erer & 6-o.nrr -se 1: r r d* ; t'- 't e' !^ "e a -t ?^. '58: ' e t---t et. a..d r '-d s

rr'&M '-- ---'te-f e; rd nitr"'t; the concentrations of potentially explosive gas mixtures in the main andanser offgas treatment system. The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Ortin;; 60gy4+ of Appendix A to 10 CFR 1

Part 50. c.micetma 3/4.3.9 FEE 0 VATER /KAIN TURBINE TRIP SYSTEM AETUATION INSTRtHENTATION The feedwater/ main turbine trip system actuation instrumentation is j provided to initiate action of the feedwater systern/s,ain turbine trip system O in the event of a high reactor vessel water level due to failure of the feedwater controller under maximum demand. 3/4.3.11 APPENDIX R ALTERNdIVE SHtfTD0VN INSTRLHENTATIQN The OPERABILITY of the alternative shutdown system ensures that a fire will not preclude achieving safe shutdown. The alternative shutdown system instrumentation is independent of areas where a fire could damage systems normally used to shutdown the reactor. Thus, ti.e system capability is consistent with General Design Criterion 3 and-#ppendix R to 10 CFR 50. i l' g O FERMI - UNIT 2 .B 3/4 3 6 M endment No.-)ft-M- 3/4.11 RA010ACilVE fif ttlINi$ BA$f 5 3/4.11.1 t! QUID EfflufNTS ai,. u.a., . L : r ^~::^ This specification is provided to ensure that the concentration of radio 've materials released in liquid waste effluents to UNRESTRICT REAS will be than the concentration levels specified in 10 CfR Par . Appgn-dix B. Table Column 2. This limitation provides additiona surance that the levels of ra tive materials in bodies of water in U. TRICTED AREAS will result in expos within (1) the Section !!.A de ' n objectives of Appen-dix 1, 10 CFR Part 50, t MEMBER Of THE PUBLIC an ) the limits of 10 CfR Part 20.106(e) to the popula The concentr n limit for dissolved or entrained noble gases is based u the assi ion that Xe-135 is the control-ling radioisotope and its MPC in ali rsion) was converted to an equivalent concentration in water using the met scribed in International Commission on Radiological Protection (ICR alicat The required detecti capabilities for radion ve materials in Itauld waste samples are tab ed in terms of the lower limi f detection (LLDs). Detailed discussi f the LLD, and other detection limits be found in HASL Procedure *.anual. HASL-300 (revised annually). Currie, L. . " Limits for Quallt e Detection and Quantitative Determination - Applica ' n to Radioc stry," Anal. Chem. 40, 586-93 (19';8), and Hartwell, J. K., ection O' for Radioanalytical Counting Techniques," Atlantic Richt telet Hanfo Lim ipany Report ARH-SA-215 (June 1975). s,,,u...< is specification is provided to implement the requi mm nts of T imiting Secti '.A. Ill.A and IV.A of Appendix 1, 10 CFR Part 50. Condition neration implements the guides set forth in on 11.A of Appendix 1. The ON statements provide the required rating flexibility and at the same time ' ment the guides set fortl Section IV.A of Appen-dix ! to assure that the r es of radioacti interial in liquid effluents to UNRESTRICTED AREAS will be k "as lo is reasonably achievable." Also, fe.r fresh water sites with drinking - supplies that can be potentially affected by plant operations, th is re. ble assurance that the operation of the facility will not re, in radionucli centrations in the finished drinking water that are excess of se requiremen 40 CfR Part 141. The dose calculation m.. dology and parameters in the ODCM ment the require-des of ments in Secti

11. A of Apper. dix ! that conformance with th shown by calculationel procedures based on models an ta, such Appendix. actual exposurt of a MEMBER OF THE PUBLIC through appropriati:

s that niikely to be substantially underestimated. The equations specified in O FERMI - UNIT 2 3 3/4 11-1 .. _ _ _ _ _. _ _. ~. _ RADl0ACilVE ffflufN15 O eAsfs = "T^^; i u ' calculating the doses due to the actual release rates of r d materials ni .a provided in ffluents are consistent with the m Regulatory Guide 1.109, ion of Ann a Man from Routine Releases of Reactor Effluents f -- of Evaluating Compliance with 10 CFR Part 50, Appen evision 1, Octo nd Regulatory Guide 1.113 "Estimati-ispersion of Effluents from Acciden tine Reactor > for the Purpose of Implementing Appendix 1," Aprl) 1977. - " :: :: ::U w an N he OPERABIU

of the liquid radwaste treatment system ensures that avt P able for use whenever liquid effluents requir ment system prior to their

- to the environment. The requiremen the appropriate portions of this syste. d, when specified, r assurance that the releases of radioactive materi liquid nts will be kept "as low as is reasonably achievable." This spe n implements the requirements of General Design Criterion 60 .g ndix A to part 50 and the design objective given in 5 1.0 of Appendix ! to 10 50. The specified limits govern ,, use of appropriate portions of the liqu 1ste.reatment system specified as a suitable fraction of the dose design ob e .

Set i in Section !!.A of Appendix 1, 10 CFR Part $0, for liquid effluents.

I 3/4.11.1.4 t! QUID,HOLOUP TANKS The tanks listed in this specification include all those outdoor radwaste-tanks that are not surrounded by Ifners, dikes, or walls capable of holding the tank contents and that do not have tank overflows and surrounding area drains connected to the liquid radwaste treatment syt. tem. Restricting the quantity of radioactive material contained in the specified tanks provides assurance that in the event of an uncontrolled release of the tanks' contents, the resulting concentrations would be less than the limits of 10 CFR Part 20, Appendix 0. Table II, Column 2, at the nearest potable water supply and the nearest surface water supply in an UNRESTRICTED AREA. l O rtR"i - u"iT 2 8 2/4 11-2 h RADIDACTIVE EFFLt!EN15 BASES 3/4.11.2 GASLOUS ffflufNf5 7_ This specification is provided to ensure that the dose at any time at a beye the SITE BOUNDARY from gaseous effluents from all units on the site ill be wit the annual dose limits of 10 CrR Part 20 to UNRESTRICTED AREAr The annual do limits are the doses associated with the concentrations - 0 CFR Part 20, Ap dix B. Table !!, Column 1. These limits provide re iable assurance that dioactive material discharged in gaseous efflu s will not result in the em. ure of a MEMBER Of THE PUBLIC in an UNRES

TED AREA, either within or ou ide the SITE BOUNDARY, to annual aver e concentrations exceeding the limits cified in Appendix B Table !!

10 CfR Part 20 (10 CFR Part 20.106(b)). or MEMBERS OF THE PUBLIC - may at times be within the SITE BOUNDARY -the occ ncy of that MEMBER OF ' fE PUOLIC will usually be sufficiently low to compensat. for any increase the atmospheric diffusion factor above that for the $11E NDARY. Ex es of calculations for such i MEMBERS OF THE PUBLIC, with the ap priat-ccupancy factors, shall be given in the ODCH. The specified release imits restrict, at all times, the co respending gamma and beta dose ra ove background to a HEMBER Of THE Ptvl!C at or beyond the SITE COUND to 's than or equal to 500 mrems/ year to the total body or to less th or equal t '4000 mremshear to the skin. These release rate limits als restrict, at al imes, the corresponding 1 O thyroid dose rate above ba round to a child via e inhalation pathway to less than or equal to 1 mrems/ year. The required ection capabilities for radioactive terials in gaseous tabulated in terms of tM lower limits o tection (LLDs). waste samples ar Detailed dise sion of the LLO, and other detection limits can 4e found in NASL Procedures nual, HA$t-300 (revised annually), Currie, l. A., its for Qualita e Detection and Quantitative Determination - Application Radio-chemi y," Anal. Chem. 40, 586-93 (1968), and Hartwell, J. K., "Dete. for. d Lir s for Radioanalytical Counting Techniques," Atlantic Richfield Han mpany Report '@H-SA-215 (June 1975). ,m .www w m p w--, ---w- --g-is specification is provided to implement the requirements Sections 11.A and IV.A of Appendix 1, 10 CFR Part e timiting Condition for Opr implements the guides set f Section II.B of utred operating flexibility Appendix !.- The ACTION s s provide and at the same time implement the set forth in Section IV.A of Appendix ! to assure that the releases oactive 1 in gasenus effluents to UNRESTRICTED AREAS w ept "as low as is reasom j hievable." The Sur-veillance Re its implement the requirements in Section of Appendix I tance with the guides of Appendix ! be shown by calculatto ec-that s based on models and data such that the actual exposure of a HEMBER Of FERMI - UNIT 2 03/411-3 -. = i r ? ,s R_A0!0 ACTIVE EFFLUENTS g BASES

ME - 9' E C^!E? SM' _ A "AllC through appropriate pathways is unlikely to be substantially u.d-o estimat e dose calculation methodology and parameters establ n the ODCM for Calcu he doses due to the actual release r radioactive noble gases in gaseous

' nts are consistent wi methodology provided in Regulatory Guide 1,109, "Ca ion of oses to Man from Routine Releases of Reactor Effluents for th - e of Evaluating Compliance with 10 CFR Part 50, Appendix on 1, Octoy 77 and Regulatory Guide 1.111 " Methods for Estimati uuspheric Transport and 0 - n of Gaseous Effluents in Routine Re1 rom Light-Water Cooled Reactors," Rav s July 1977. The OD tons provided for determining the air doses at and bey

e uuNDARY'are based upon the historical average atmospheric conditions,

~ ~ ~ " This specification is provided to implenent the requirements of Secti. !! C.111. A and IV. A of Appendix 1,10 C' R Part 50. The Limi. ig Conditio for Operation are the guides set fortt in Section ll.C o 4ppendix 1. The ACTION tements provide the required opera,,ing flexibility d at the same time imp ent the guides set forth in Section IV.A of Ap dix ! to assure that the eases of radioactive materials in gaseous fluents to ) UNRESTRICTED AREAS 11 be kept "as low as is reasonably levable," The ODCM calculational me ds specified in the Surveillanc equirements implement the requirements in Sect !!!,A of Appendix ! that nformance with the guides of Appendix ! be shown by c .ulational procedures ased on models anv data, such that the actual-exposure a MEMBER OF TH -UBLIC through appropriate pathways is unlikely to be subst ially unde stimated. The 00CM calculational methodology and parameters for calc ting e doses due to the actual release rates of the subject materials are cor ent with the methodology provided in Regulatory Guide 1,109, " Calculation ual Doses to Man from Routine Releases of Reactor Effluents for the Purpos of Eva ting Compliance with-10 CFR Pu t 50 -Appendix 1," Revision 1, October 77 and Reg tory Guide 1,111, " Methods for Estimating Atmospheric Transp and Dispersion Gaseous Effluents in Routine Releases f rom Light-Water-C. ed Reactors," Revisio , July 1977. These equations also provide f determining the actual dos based upon the histgrical average atmospheric ce tions. The release rate specif tions for iodine-131, iodine-133, tritium nd radionuclides in particulate form .h half lives greater than 8 da-are dependent upon the existing radionuci pathways to man, in the ar at and beyond the SITE BOUNDARY. The pathways at were examined in e development of these calculations were: (1) indivi 51 inhala- -tion of a orne radionuclides,-(2) deposition of radionuclides onto en leafy -vegetat. with subsequent consumption by man, (3) deposition onto grass, mreas wher i lk animals and meat producing animals graze with consumption of the mi and meat by man, and (4) deposition on the ground with subsequent exposui

man, O

FERMI - UNIT 2 6 3/4 11-4 y rvwr ,w,.m --r,m--y-~. e-,--- .-m%,,.m-e-.-,,.

w-

,,w,m

-,v..

+,m

-..,o,-c-..--

-e.-.<w-

.----,-...<..~,---wn,-----,

m

RAD 10ACTlVE EFFLUEN15 BASES

, ^

.;.I.^

W...

m, -

OPERABILITY of the Off-GAS TREATMENT SYSTEM ensures that the s will be 4.

Sie for use whenever gaseous effluents require tre prior to release to ti

- ronment.

The requirement that the

. viriate portions of t'ese systems be use.

specified, providee

.onable assurance t aht the releases of radioactive ma in effluents will be kept "as ification imniements the require-low as is reasonably achievable."

0 of Appen o 10 CFR Part 50, and the ments of General Design Cri - -

design objectives bection 11.0 of Appendix n CfR Part 50.

The specified li vverning the use of appropriate portions o systems were speci

> a suitable fraction of the dose design objectives set in ons !!.B and 11.C of Appendix 1, 10 CFR Part 50, for gaseous effluen

.i..

~

^^:-

~^ ^ n;n 1; m

drement that the appropriate portions of this system be us -

specified, prov able assurance that the relea ive is reasonably achievable."

materials-in gaseous effluen

. ortions of the systems The specified limits gover o <

set forth e fraction of the dose des gn e

were specifi-a 1.0 and II.C of Appendix 1, 10 CFR Part 50, for gaseous e is O

s'4.12.2.6 cxa'osivo cas sixtuat This specification is provided to ensure that the concentration of poten-tially explosive. gas mixtures contained in the main condenser offgas system is Maintaining maintained below the flammability limits of hydrogen and oxygen.

the concentration of hydrogen below the flammability limit 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, 3/4.11.2.7 MAIN CONDENSER Restricting the gross radioactivity rate of noble gases from the main condenser provides reasonable assurance that the total body exposure to an individual at the exclusion area boundary will not exceed a small fraction of thelimitsof10CFRPart107intheeventthiseffluentisinadvertently This specification discharged directly to the e.aironment without treatment.

implements-the requirements of Genert) Design Criteria 60 and 64 of Appendix A to 10 CFR Part 50.

i g,, ;

3 g m y 7n r.yg

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FERMI - UNIT 2 B 3/4 11-5

_-,.,_-,-e-g..,v m,-e-,-e-e------..,__,,._-..-...e.,

r.m.-m-...

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v.,,.sr-e - r--

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RADIDACTIVE trrLUENTS O

BASES m,.;;.;

4 R

"'"""'l i

cification impicments the requirements of General Desi n en of Appendix

' 40.

The process par W n establishing

^

~

the PROCESS CONTROL PROGRAM ma not limited to waste type, waste pH, waste /11 ui ' ^'

on agent / catalyst ra nil content, waste iemical constituents, and mixing and curing times.

7,,7,.

This specification is provided to meet the dose limitations of 40 C Part 0 that have been incorporated into 10 CFR Part 20 by 46 FR 185'.

The spe 1 cation requires the preparation and submittal of a Speed Report I

whenever t calculated doses from plant generated radioact ve e uents and direct ra tion exceed 25 mrems to the total body or any gan, except the thyroid, wh shall be limited to less than or equal 5 mrems.

For sites containi, up to 4 reactors, it is highly unli. y that the resultant dose to a HEMBER OF PUBLIC will exceed the dose 1 ts of 40 CFR Part 190 if the individual reactu remain within twice the sedesignubjectivesof Appendix I, and if direct lation doses from t. reactor units and outside storage tanks are kept small, he $pecial Re t will describe a course of action that should result in the mitatio f the annual dose to a MEMBER OF THE PUBLIC to within the 40 CFR Pa 90 mits. For the purposes of the i

Special-Report, it may be assumed tha he dose commitment to the HEMBER OF O_

THE PUBLIC from other uranium fuel cle nurces is negligible, with the exception that dose contribution rom othe nuclear fuel cycle facilities at the same site or within a rad of 8 km must considered.

If the dose to any MEMBER OF THE PUBLIC i stimated to exceed requirements of 40 CFR Part 190, the Special Re - t with a request for a v 'ance (provided the release conditions re' ting in violation of 40 CFR P.

190 have not already been corrected), i ccordance with the provisions of 40 R Part 190.11 and 10 CFR Part 20.4 c is considered to be a timely request a fulfills the requirements 40 CFR Part 190 until NRC staff action is comp ted. The variance o relates to the limits of 40 CFR Part 190, and does t apply in any way the other requirements for dose limitation of 10 CFR Par

'0, as addre d in Specifications 3.11.1.1 and 3.11.2.1.

An individual is ne co dered a MEMBF3 0F THE PUBLIC during any period in which he/she is en qed carrying out any operation that is part of the nuclear fuel cycle.

i FERMI - UNIT 2 B 3/4 11-6

. - - -. - - _ - - - _.,,,-. _ _ _.- _., _ - -,-_,_ _ _.~. _ - _ ___.. _ -,__. ~.. -.. -,- _---.-. -.

~-

O=n

=

.. ~ _

a,.u The radiological environmental monitoring program required by this ication provides representative measurements of radiation and of ra ' -

l spec active terials in those exposure pathways and for those radionuclide ist lead to t highest potential radiation exposures of MEMBERS Of THE LIC resulting fr the station; operation. This monitoring program im ments Section l'! li.2 Appendix ! to 10 CFR Part 50 and thereby sup ments the radiolog M 1 iff) t monitoring program by verifying that measurable concentrations of ra ' active materials and levels of ra ion are not higher than expected on the ba s of-the effluent measurement nd the modeling of the environmental exposur athways. Guidance for s monitoring program is provided by-the Radiological sessment Branch Te nical Position on Environ-mental Monitoring. The initia specified no oring program will be effective for at least the first 3 years of mmercia

peration, following this period, program changes may be initiated bas o

perational experience.

i The required detection capabil es environmental sample analyses

]

are tabulated in terms of the lo. limits o etection (LL0s). The LL0s 1

required by Table 4.12.1-1 ar onsidered optim for routine environmental measurements in industrial oratories, it shou be recognized that the LLO is defined as an a ori (before the fact) limi epresenting the capa-bility of a measureme system and not as an a posterio - (after the fact) limit for a partic r ' me a suremer.t.

Detailed scussion of the LLO, and other detection limits, an be found in HASL Pr dures Manual, HASL-300 (revised annually), Currie, L.

" Limits for Qua ' ative Detection and Quantitative betermination - Applicati to Radi - emistry," Anal. Cheen 40, 586-93 (1908), and Hartwell, J. K.,

ction L '.

s for.Radioanalytical Counting Techniques," Atlantic Richfield Hanfor mpany Report ARH-SA-215 (June 1975).

This specification is provided to ensure that changes in the use of ar -

at-an

> nd the $1TE BOUNDARY are identified and that modification ie radiologica

ronmental monitoring program are made if requi y the results of this c The best information f rom the do

-door survey, from aerial-survey, fro oal survey or from cons g with local agricul-tural authorities shall be us.

This census

-.sfies the requirements of Section IV 8.3 of-Appendix ! to 10 so.

Restricting the census to gar-

' dens of greater than 50 m provider. ur hat significant exposure pathways 8

via leafy vegetables will be ntified and mon H since a garden of this site is the minimum re d to produce the quantity

/ year) of leafy vege-tables assumed i

. ulatory Guide 1.109 for consumption y,

- ild. To deter-tre ma e

1) 207,of mine this am garden size, the following assumptions the c n was used iJr growing broad leaf vegetation (i.e., $(milar to oce e

cabbage), and (2) a vegetation yield of 2 kg/m,

FERMI - UNIT 2 8 3/4 12-1

_ -,, - - ~_

l l

0 r ' s m :: -, '

_7

,r r

.1.

is s t i.t r m t a n. a; an-r omarirmu es r. n _- n i n _

W reouirement for participation in an approved Interlaboratory Com h

nsure that independent checks on tha u

and accu-Program s,

.v ronmental sample racy of the measurements o

' ma r

nee program for anviron-matrices are performed as

r me qua for the r to demonstrato that, the resu mental monito Section IV.B.2 of Appendix 1 to 10 CFR Part 50.

O I

O FERMI - UNIT 2 0 3/4 12-2

- ^^

7 ADMINISTRATIVE CONTR01S 6.1 RGPON11DJLl))

6.1.1 The Plant Manager shall be responsibic for overall unit safe operation and shall delegate in writing the succession to this responsibility during his absence.

The Plant Manager shall have control over those onsite activities necessary for safe operation and maintenance of the plant.

6.1.2 The Nuclear Shift Supervisor or, during his 6bsence from the control room, a designated individual shall be responsible for the control room command function.

A management directive to this effect, sighed by the Assistant Vice President and Hanager - Nuclear Production, shall be reissued to all station personnel on an annual basis.

6.2 ORGANIZATIOf{

6.2.1 OffSITE AND ONSITE ORGANIZATIONS Onsite and offsite organizations shall be established for unit operation and corporate management, respectively.

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

a.

Lines of authority, responsibility, and communication shall be established and defined for the highest management levels through in'.ermediate icvels to and including all operating organization positions. These relationships shall be documented and updated, as appropriate, in the form of organization charts, functional descri)tions of departmental responsibilities and relationships, and jo) descriptions for key personnel positions, or in equivalent forms of documentation.

These requirements shall be documented in the Updated final Safety Analysis Report.

b.

The Senior Vice President shall have corporate responsibi'ity for overall plant nuclear safety and shall take any measures eded to ensure acceptable performance of the staff in operating, maintaining, and providing technical support of the plant to ensure nuclear safety.

c.

The individuals who train the operating staff and those who carry out health physics and quality assurance functions may report to the appropriate onsite manager; however, they shall have 1

sufficient organizational freedom to ensure their independence from operating pressures.

6.2.2 UNil staff a.

Each on duty shift shall be composed of at least the minimum shif t crew composition shown in Table 6.2.2-1; b.

At least one licensed Operator shall be in the control room when fuel is in the reactor, in addition, while the unit is in

(-

OPERATIONAL CONDITION 1, 2 or 3, at least one licensed Senior Operator shall be in the control room; FERMI - UNIT 2 6-1 Amendment No.-J1, y,

m

ADMINISTRAllVE CONTROLS O

PROCEDURES AND PROGRAMS (Continued)

)

1.

Training of personnel, 2.

Procedures for monitoring, and 3.

Provisions for maintenance of sampling and analysis equipment.

c.

Post accident Sampling A program which will ensure the capability to obtain and analyze reactor coolant, radioactive lodines and particulates in plant gaseous effluents, and containment atmosphere samples under accident conditions.

The program shall include the following.

1.

Training of personnel, 2.

Procedures for sampilng and analysis, and 3.

Provisions for maintenance of sampling and analysis equipment.

d.

High Density Spent Fuel Racks ADD lt/RRT A program which will assure that any unanticipated degradation of the 605e (SEF.

high density spent fuel racks will be detected and will not compro-M PN

~ise the integrity of the racks.

m

~

r-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 Administ,rator of the Regional Of fice 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 l

supplier, and (4) modifications that may have significantly altered the nuclear, thermal, or hydraulic performance of the unit.

6.9.1.2 The startup report shall address each of the tests identified in Subsection 14.1.4.8 t f the Final Safety Analysis Report and shall include a description of the measured values of the operating conditions or characteristics obtained during the test program and a comparison of these values with design predictions and specifications. Any corrective actions tnat were required to I

lO l

FERMI - UNIT 2 6-16 Amendment No.:', !! 30 -

1 i

INSEllT 6.11.5.o c.

Radioactive Effluent Controla Program A program shall be provided conforming with 10 CFH 50 36a for the control of radioactivo effluents and for maintaining the doses to HEMBERS OF Tile PUBLIC from radioactivo effluents as low as reasonably achievable. The program (1) shall be contained in the ODCH, (2) shall be implemented by operating procedures, and (3) shall include remedial actions to be taken whencycr the program limits are exceeded. The program shall include the following elements:

1)

Limitations on the operability of radioactivo 11guld and gaseous monitoring instrumentation including surveillance tests and setpoint determination in accordance with the methodology in the ODCH, 2)

Limitations on the concentrations of radioactive material released in liquid effluents to UNHESTHICTED AREAS conforming to 10 CFit Part 20, Appendix B, Table 11, Column 2, 3)

Honitoring, sampling, and analysis or radioactive liquid and gaseous effluents in accordance with 10 CFR O

20.106 and with the methodology and parameters in the ODCH.

4)

Limitations on the annual and quarterly doses or dose commitment to a HEMBER OF Tile PUBLIC from radioactive materials in liquid effluents released from each unit to UNBESTRICTED AREAS conforming to Appendix ! to 10 CFR Part 50, 5)

Determination of cumulative and projected dose contributions from radioactive effluents for the current calendar quarter and current calendar year in accordance with the methodology and parameters in the ODCH at least every 31 days, 6)

Limitations on the operability and use of the liquid and gaseous effluent treatment systems to ensure that

- the appropriate portions of these nystems are used to i

reduce releases of radioactivity when the projected i

O

6.8.5.e Cont.

O doses in a 31-day period would exceed 2 percent of the guidelines for the annual dose or dose commitment conforming to Appendix 1 to 10 CFR Part 50, 7)

Limitations on the dose rate resulting from radioactive material released in gaseous effluents to areas beyond the SITE BOUNDARY conforming to the doses associated with 10 CFR Part 20, Appendix B Table !!, Column 1, 8)

Limitations on the annual and quarterly air doses resulting from noble gases released in gaseous effluents from each unit to areas beyond the SITE BOUNDARY conforming to Appendix 1 to 10 CFR Part 50, 9)

Limitations on the annual and quarterly doses to a HEMBER 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 beyond the SITE BOUNDARY conforming to Appendix I to 10 CFR Part 50,

10) Limitations on venting and purging of the Mark I containment through the Standby Gas Treatment System or the Reactor Building Ventilation System to maintain releases as low as reasonably achievable, O

11) Limitations on the annual dose or dose commitment to any MEMBER OF THE PUBLIC due to releases of radioactivity and to radiation from uranium fuel cyclo sources conforming to 40 CFR Part 190.

f.

Radiological Environmental Monitoring Program A program shall be provided to monitor the radiation and radionuclides in +.he environs of the plant. The program shall provide (1) representative measurements of radioactivity in thi highest potential exposure pathways, and (2) verification of the accuracy of the effluent monitoring program and modeling of environmental exposure pathways. The program shall (1) be contained in the ODCM, (2) conform to the guidance of Appendix I to 10 CFR Part 50, and (3) include the following:

O 1

6.8.S.f. Cont.

1)

Honitoring, sampling, analysis, and rcporting of radiation and radionuclidos in the environment in accordance with the methodology and parameters in the

ODCH, 2)

A 1.and Oso Census to ensuro that changes in the use of areas at and beyond the SITE BOUNDARY are identiflod and that modifications to the monitoring program are made if roouired by the results of this consus, and, 3)

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

O O

A ADMINISTRATIVE CONTR0tS A

ANNUAt REPORTS (Continued)

V (1) reactor power history starting 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months prior tJ the first sample in which the limit was exceeded; (2) results of the last isotopic analysis for radiolodine performed prior to exceeding.he limit, results of analysis while limit was exceeded and results of one analysis af ter the radiciodine activity was reduced to less than limit (each result should include date and time of sampling and the radiolodineconcentrations);

(3) clean-up system flow history starting 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months prior to the first sample in which the limit was exceeded; (4) graph of the I-131 and one other radiotodine isotope concentrations in microcuries per gram as a function of time for the duration of the specific activity above the steady-stato level; ar.d 1

(5) the time duration when the specific activity of the primary coolant exceeded the radiciodine limit.

MONTHLY OPERATING REFORTS 6.9.1.6 Routine reports of operating statistics and shutdowrr experience shall be submitted on a monthly basis to the Director, Of fice of Resource Management, I

U.S. Nuclear Regulatory Commission, Washington, D.C. 20555, with a copy to the

]

Regional Administrator of the Regional Office of the NRC, no later than the 15th of each month following the calendar month covered by the report.

REPWE WITH D I$d*d ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT (SEE HEtT PAG &

6.9.1.7 Routine Annual Radiological Environmental Operating Reports covert peration of the unit during the previous calendar year shall be su ted prio May 1 of each year. The-initial report shall be submitte or te, May 1 of ear following initial criticality.

The Annual Radiolo 1 Environmental Operating Repor all include summaries, interpretations, and a lysis of trends of the uits of the radiological environmental surveillance vities for the ort period, including a compar-ison as appropriate, with preop ional es, with operational controls, and with previous environmental.surv nee reports, and an assessment of the observed impacts of the plant ope on he environment. The reports shall also include the results of 1 use censuse uired by Specification 3.12.2.

The Annual Radiological ronmental Operating ts shall include the results of analysis of all r ogical environmental samples of all environmental radiation measur

.ts taken during the period pursuant t e locations spect-fled in the e and Figures in the ODCM, as well as summariz no tabulated i

l results hese analyses and measurements in the format of the ta in the Radi gical Assessment Branch Technical Position, Revision 1, November 9.

he event that some individual results are not available for inclusion L

FERMI - UNIT 2 6-1B Mendment No..H.

. - _ ~.

pSElrr 6.Q.1.*(

O ANNUAL RADIOLOGICAL ENV!HONMENTAL OPERATING REP 0fft 6 9 1.7 The Annual Radiological Environmental Operating Report covering the operation of the unit during the previous calender year shall be submitted before May 1 of each year.

The report shall include summaries, interpretations, and analysis of trends of the results of the Radiological Environmental Monitoring Program for the reporting period. The material provided shall be consistent with the objectives outlined in (1) the ODCH and (2) Sections IV.B.2, IV.B.3, and IV.C of Appendix ! to 10 CFR Part 50.

SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 6.9.1.8 The Semiannual Radioactive Effluent Release Report covering the operation of the unit during the previous 6 months of operation shall be submitted within 60 days after January 1 and July 1 of each year. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit. The material provided shall be (1) consistent with the objectives outlined in the ODCH and PCP and (2) in conformance with 10CFR 50 36a and Section IV.B.1 of Appendir I to 10 CFR Part 50.

O l

l l

O 1

ADMINISTRAl!VE CONTROLS ANNUAL RADIOLOGICAL ERVlRONMENTAL OPERAT]NG REP 0_RT (Continued) report, the report shall be subeltted noting and explaining the reas a #

for ssing results.

If possible, the missing dste shall be s ed as soon as le in a. supplementary report.

The reports shall also the following:

ary description of the radiological environmental mon)

at least two legible maps

cover-ing all sampling locations keye Iving distances and directions from the centerline of o r; the resu licensee participation in the Interlaborato rison Program, required ification 3.12.3; discussion deviatJons from the sampling schedule o le 3.12.1-1; disc of all analyses in which the LLD required by Table

-1 was achievable.

SEM1AhWUAL RADIDACTIVE EFFLUENT RELEASE REPORT **

6.9.1.

outine Sealannual Radioactive Effluent Release Reports covering t operation the unit during the previous 6 months of operation shall be submitted wit ' 60 days af ter January I and July 1 of each ysar.

T eriod of the first rep shall begin with the <! ate of initial critical The Semiannual Radioac Effhnnt Release Reports shall ude a summary of the quantities of radioac li p s and gaseous eff1 s and solid waste released from the unit as out d in Regulatory G 1.21,* Measuring, Evaluating, and Reporting Radios ity in Soli astes and Releases of Radio-active Material 8 in Liquid and Gase Effl s from Light-Water-cooled Nuclear Fower Plants, Fevision 1 June 1974,

.ata summarized on a quarterly basis Q

following the forsat of Aopendix B t o.

The Semiannual Radicactive Eff nt Release Repo be subeltted within 60 days after January 1 of h year shall include nnual summary of bo9tly meteorologit.a1 data col ted over the previous year, a annual summary may be either in the fo f an hour-by-hour listing on magnet tape of wind speed, wind direction, spheric stab 11My, and precipitation (if

ured),orir

,the form of t frequency distributions of wind speed, wind di' lon, and atmospher tability.*** This same report shall include an asses

' of th radiat doses due to the radioactive liquid and gaseous effluents re ed fra he unit or station during the previous calendar year.

This same rep p

also include an assessment of the radiation doses from rVioactive liqu map shall cover stations near the SITE 900NDARY; a se:end shall i the tant stations.

- A single submit be made for a multiple unit s The subnittal should combine those see that are common units at the station; however, for units with separa; stems, the subelttal shall specify the releases of radioactive ma unit.

- - In lieu of submissi the f t:st half year M Radioactive Effluent Release Repo e licensee has the nption of retaining ary of rt:Quir eorological data on site in a file that shall be prov C upon request.

FERMI - UNIl 2 6-19 Amendment No. Il

i ADMINISTRATIVE CONTROLS SEM1 ANNUAL DA010 ACTIVE EFFLUENT RELEAS[ REPORT (Continued) nd gaseous ef fluents to MEKBERS Of THE PUBLIC due to their activities insi n'

t. ' S!TE BOUNDARY (Figure 5.1.3-1) during the report period.

All assui tir s use in making these assessmerts, i.e., specific activity, exposure time d

loca 'on, shall be included in these reports.

The assessment of radiat' n doses. hall be perfomed in accordance with the zethodology and parav ers in the FSITE DOSE CALCULATION MANUAL (ODCH).

The Semiann il Padioactive Etfluent Release Report to be submitte 60 days af ter January 1 of each year shall also include an assessment o radiation doses to the 1.ely most exposed HIMBER OF THE PUBLIC from rea or releases and other nearb sranium fuel cycle sources, including doses rom primary effluent pathways nd direct radiatinn, for the previous c' 'endar year to show conformance with 40 FR Part 190, Environmental Radiation rotection Standards for Nuclear Power Opt 9 tion.

The assessment of radiati doses shall be perfomed in accordant with the methodology and param ers in the ODCH.

Tt.e Semiar.ncal Radioactiv Effluent Release Reports hall include the following infomation for e -h class of solio was (as defined by 10 CFR Part 61) shipped offsite duri the report perio.

a.

Cont 41ner volume, b.

Total curie quantity (sp ify wb ber deterinined by meascrement or estimate),

c.

Principal radionuclides (spe fy whether detemined by measurement O

or esti=>te).

d.

Source of waste and proce sing e loyed (e.g., dewatered spent resin, co7 acted dry waste, ev, orator bo tom!),

e.

Type of container (e., LSA, Type A. Type B, Large Quantity), and f.

Solidification age oc absorbent (e.g.

cement, urea formaldehyde).

4 The Semiannual Radioactiv Effluent Release Reports all include a list and description of unplanned rcleases from the site to UN 'STRICTE0 AREAS of radio-active materials in ga ous and liquid effluents made d -ing the reporting period.

The Semiannuai 4a,oactive Effluent Release Reports shall in lude any thanges made durir.g the eporting period to the PROCESS CONTROL PROGR (PCP) and to the OFFSITE DO cat.CULATION KMMAL (ODCH), as well as a listin of new loca-Lions for do calculations and/or envirorsental monitoring ident'fied by the land use ce us pursuant to Specification 3.12.2.

The Semi nual Radioactive Effluent Release Reports shall also includ the follow g:

an explanation as to why the inoperability of liquid or gas us eff1 t monitoring instrumentation was not corrected within the time spt i-fie in Specification 3.3.7.11 or 3.3.7.12, respectively; and description t

events leading to liquid holdup tanks exceeding the limits of Specifica-en 3.1L I.4.

O FERMI - UNIT 2 6-20 Ame nd.me nt N o. 11, ~24

60M14151RATIvfCONTR0ts O

e.io arcono 8ttt"tio" - <<o"t4evem 6.10.3 The following records shall be retained for the duration of the un't Operating License:

a.

Records and drawing changes reflecting unit design modifications made to systems and equipment described in the final Safety Analysis Report, b.

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

c.

Records of radiation exposure for all individuals entering radiation control areas.

d.

Records of gaseous and liquid radioactive material released to the

environs, c.

Record; of transient ur operational cycles for those unit components identified in Table 5.7.1 1.

f.

Records of reactor tests and experiments.

g.

Records of training and qualification for current members of the unit

staff, h.

Records of inservice inspections performed pursuant to these Technical e

Specifications.

1.

Records of quality assurance activities required by the Operational Quality Assurance Manual, j.

Records of reviews cerformed for changes made to procedures or equipment or review: of tests and experiments pursuant to 10 J 50.59.

k.

Records of meetings of the OSRO and the NSRG.

1.

Records of the service lives of all hydraulic and mechanical snubbers required by Specification 3.7.5 including the date at which the service life commences and associated installation and maintenance records, m.

Records of analyses required by the radiological environmental monitoring program that would permit evaluation of the accuracy of the analysis at a later date.

This should include procedures effective at specified times and 0A records showing that these procedures were

followed, RE c rto s o F REVIEWS PEr4Fcr2Hr:D th c///Me&65 +o thE o

OFF5/TE Dose CALCUL.ATidM MAMUAL = the PRcmSS CONTROL PROGrd Am.

O FERM1 UN1' 2 6 22 Amendment Noc JJ, 5"

l ADMINISTRATIVE CONTROLS HIGH RADIATION ARIA (Continued)

A radiation monitoring device which continuously indicates !.he a.

radiation dose rate in the area.

b.

A radiation monitoring device which continuously integrates the f

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

c.

A health physics qualified individual (i.e., qualified in radiation protection procedures) with a radiation dose rate monitoring device who is responsible for providing positive control over the activi*

ties within the area and shall perform periodic radiation surveillance at the frequency specified by the Health Physicist in the RWP.

6.12.2 In addition to the requirements of Specification 6.12.1, areas acces-sible to personnel with radiation levels such that a major portion of the body could rective in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months a dose greater than 1000 Erems shall be provided with locked doors to prevent unauthorized entry, and the keys shall be maintained under the administrative control of the Nuclear Shift Supervisor on duty and/

or the health physic supervision. Doors shall remain locked except during periods of access by personnel under an approved RWP which shall specify the dose rate levels in the immediate work area and the maximum allowable stay time for individuals in that area. For individual areas accessible to person-nelwithradiationlevelssuchthatamajorportionofthebodycouldreceive O

in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months a dose in excess of 1000 mreas that are located within large areas, such as the containment, where no enclosure exists for urposes of locking, and no enclosure can be reasonably constructed around t e individual areas, then that area shall be roped off.. conspicuously posted, and a flashin light shall-be activated as a warning device. In lieu of the stay time spec fica-tion of the RWP, continuous surveillance, direct or remote (such as use of closed circuit TV cameras) may be made by personnel qualified in radiation protection procedures to provide positive exposure control over the activities within the area.

RER.NE w T H 6.13 PROCESS CONTROL PROGRAM (PCP)

%en, t 4, g 6.13.1 The PCP shall be approved by the Connission prior to implementation I

E 12.0 ti:::::: '-!ti;t:d :h;n"':: i: t': ?!'

hall be submitted to the Commission in the Semiannual ve ease Report for the period in which nge(s) was made.- This s hall contain:

ally support the 1.

Sufficiently de

- n orsa ratioria e change without benefit o nal or ments) information;

'Heasurement made at 18 inches from source of radioactivity.

FERMI - UNIT 2 6-23 Amendment No. -n

i r

O INSERT 6.13 I

6.13.2 FROCESS 00llfkOL PROGRAM (PCP) l Changes to t he TCP a.

Shall be docunicoted and s ecords of reviews pe fottoed shall be tetained as required by Specification 6.10.3n.

This doeuraentation shall contain:

1)

Suf ficient inf orination t o support the change together with the appropslate analyses or evaluations juotifying the change (s) and

2) A determination that the change will maintain the overall conf ormance of the solidified waste product to existing requir ements of Federal, State, or other applicable regulations.

b.

Shall becomo ef fective sites review and acceptance by the OSRO and the approval of the Plant Manager.

O Y

I L

lL L.

I O

i w.

,,ge,w,--,--n..,,,-.,,g.,..nd,,,rrwn,,,,-.ar,,-,.

wm--..,,n

,.-nrnn,v,..,

vnv.,.-,

,,,.ns,-,,,,,,.-.-n.--.-~w,,

,--,~n-.

--n,

l

, ADMIN 151RAllVE CONTROLS PROCCS$ CON 1ROL PROGRAM (PCP) (Continued)

A determination that the change did not te e overall c

nce of the solidified wast ct to existing criteria lid wastes; n

3.

Documentatio e fac he change has been reviewed and foun able by the 05RO.

iall become effective upon review and acceptance by 6.14 OrF511E DOSE CALCULATION KANUAL (ODCM) 6.14.1 The ODCM shall be approved by the Commission prior to implementation.

SW Lk:n::: 8-itht:d :h:n;;; t: th: ODC"-

g - Fh1outA.

a.

5 be submitted to the Commission in the Semiannual R active uTo lustic Efflue elease Report for the period in which the ige (s) was

(.. / y made effec This submittal shall contain:

ggg gg p 1.

Sufficiently led information to 11y support the rationale for the c e without fit of additional or supple-mental inforsation.

In ti ~ submitted should consist of a packkge of those pages of CH to be changed with each page numbered and provided v en ap 1 and date box, together with appropriate an es or evaluat untifying the change (s);

}

O 2.

A deter inati that the chaege wiii not redoc e accuracy or reliabili f dose calculations or setpoint dete tions; and 3

cumentation of the f act that the change has been reviewed and found acceptable by the OSRO.

Shall become effective upon review and acceptance by the OSRO.

SE C""C C E --7 0 " ",0 : 0 ' # i P/ 0 03 :0. C ' 'J 0 'J ;. '"?

0 0 0 Uf ' " ""'":"'

%(liqui,Licensee-initiated major changes to the radioactive waste s eous,andsolid):

a.

Shall be ted to the Commission Semiannual Radioactive Effluent Releas ort for t lod in which the evaluation was reviewed by the 0$RO..

scussion of each change shall contain:

1.

As of the evaluation ed to the determination that

.e change could be made in accorda.

ith 10 CFR 50.59.

LE:n:::: ::3 :h:::: t: ;drit the H=ethe ::1hd Sr 5 thh EF:: '4

-t':: ;; p.-t ;f U., _,,.d i'"I; y b,

s FERMI - UNIT 2 6-24 Amendment No. 11 I

l O

INSERT 6.14 6.14.2 0FFSITE DOSE CALCUld TION MANU AL ((11CM)

Changes to t he (DCM:

a.

Shall be document ed and records of r eviews per f ortned shall be t et ained no required by Specification 6.10.3n.

This document ation chall cent ain:

1)

Suf ficient inf ormation to support the change ton, ether with the appropriate analyses or evaluations justifying the change (s) and

2) A detetmination that the change will maintain the level of radioactive ef fluent control required by 10 CFR 20.106, 40 CFR Part 190,10 CFR 50.36a, and Appendix I t o 10 CFR Part 50 and not adversely itnpact the accuracy or reliability of ef fluent, done, or cetpoint calculations.

b.

Shell become ef f ective af ter review and accept ance by the OSR0 and the approval of the Plant Manager.

O c.

Sha11 he r.ehmitt ed te the C-mi iee 1e the f erm ei a complete legible copy of the entire 00CM ac part of or concurrent with the Semiannual Radioactive Ef fluent 3elease Report for the period of the report in which any change to the CDCM was made.

Each change shall be identified by markings in the margin of the af fected pages, clearly indicating the area of the page that was changed, and shall indicate the date (e.g., month / year) the change was implement ed.

O

4 e

o 10 aus== con =s

.;^ "..:,;;m;;. R:::., -,; :a ;;,Z: a m.mt a i s i.m n 2

' ' ' ^

^^

^

W ',i 2 :C; Suf ficient detailed information to totally support the reas l

for the change without benefit of additionel or suppleme 4

orsation; 3.

A det led description of the equipment, compone

, and involved and the interfaces with oth plant systems; process 4.

An evaluatio f the change, which shows e predicted releases of radioactive terials in liqufd and seous effluents and/or quantity of soli ste that differ om those previously predicted in the 11 se applicat and amendments thereto; 5.

An evaluation of the cht.,

ich shows the expected maximum exposures to a MEMBER OF i UBLIC in the UNRESTRICTED AREA and to the general populati, tha iffer from those previously estimated in the lic e applic on and amendments thereto; 6.

A comparison of e predicted releas of radioactive materials, in liquid an aseous effluents and in lid waste, to the actual releases f the period prior to when the anges are to be made; 7.

An es ate of the exposure to plant operating rsonnel as a re t of the change; and 8.

ocumentation of the fact that the change was reviewe nd found acceptable by the OSRO.

Shall become effective upon review and acceptance by the OSRO.

a i

l

'O M'ad**DL N

U FERMI - UNIT 2 6-25

, - - - - - ~ - -

1:nclosure 4 tc NRC-91-0131 Page 1 i

D l

Common TS Pages This table lists currently submitted, proposed TS amendments that include TS pages that are common to this amendment.

SUBJECT SURH1'IT Al.

AFFECTED 1)ATP.

PAGES Proposed Technical Specification (License 5/24/88 3/4 3 71 Amendment) Change - Appilcability (3/4.0) 3-76 Amendment), NitC.88-0062.

(see comment 1) 11-5 11 6

~~~ANI) 11 12 thru 11-15 O

Additional Information Concerning Proposed 2/27/91 11 18 thru Change to Technical Specification 3 0.4, 11-20 NRC.90-0185 (see comment 1) 12_2 12 13 12-14 Proposed License Amendment - Uprated 9/23/91 1-5 Power Operation, NRC.91 0102.

(see comrent 2)

Comment,s 1.

With the exception of page 3/4 3-76, all of the TS on these pages are deleted by the relocation of the radiologleal effluent TS as proposed in this submittal.

2.

The changes in the Uprated Power Operation submittal are independent of and not affected by the relocation of the radiological effluent TS as proposed in this submittal. Therefore. the proposed changes can be approved in any order.

O

O ENQ.05URE 5 TEQ'NICA1. f.PEGiFICATION

~~~~

alAtr,6f, PAGES O

O

1ROH DEFINITIONS sft1T0ff O-1.0 DEfil!1110HS EAR [

1.1 ACT10N.....................................................

1-1 1.2 AVERAGE PLANAR EXP0SURE....................................

1-1 1.3 AVERAGE PLANAR LINEAR HEAT GENERATION RATE.................

1-1 1.4 CHANNEL CALIBRAT10N........................................

1-1 l.5 CHANNEL CHECK..............................................

1-1 1.6 CHANNEL FUNCTIONAL TEST....................................

1-1 1.7 C O R E A L T C 21.T ! DN............................................

1-2 1.7a COR E OP E RAT I NG L l H I T S R E P0RT...............................

1 - 2 l

1.8 C RI T I C AL POW E R RA1 10.......................................

142 1.9 DOSE EQUIVALENT l-131......................................

1-2 1.10 E-AVERAGE DISINTEGRATION ENERGY............................

1-2 1.11 EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIM 2.........

1-2a j

(])

1.12 FRACTION Of LlHITING POWER 0ENSITY.........................

1-2a j

1.13 FRACTION Of RATED THERMAL P0WER............................

1-2a j

1.14 FREQUENCY NOTAT10N.........................................

1-3 1.15 IDENT!f!ED LEAKAGE,........................................

1-3 1.16 ISOLATION SYSTEM RESPONSE TIME.............................

1-3 1.17 LIMITING CONTROL R0D PATTERN...............................

1-3 1.18 LINEAR HEAT GENERATION PATE................................

13 1.19 LOGIC SYSTEM FUNCTIONAL TEST...............................

1-3 1.20 MAXIMUM FRACTION Of LIMITING POWER DENSITY.................

1-3 1.21 MEMBER (S) Of THE PUBLIC....................................

1-3 1.22 MINIMUM CRITICAL POWER RAT 10...............................

14 1.23 Off GAS TREATMENT SYSTEM...................................

1-4 1.24 Of f SIT E DOSE CALCULAT ION MANUAL............................

1-4

-(.-

1.25 OPERABLE - OPERABILITY.....................................

1-4 FERMI - UNIT 2 i

Amendment No. 64

INDEX DEFINITIONS SECTION DEFINIT [pNS (Continued)

PAGE 1.26 OPERATIONAL C6dDIT10N -

CONLIT10N....................

14 1.27 PHYSICS-TESTS..............................................

1-4 1,28 PRESSURE BOUNDARY LEAKAGE..................................

1-4 1,29 PRIMARY CONTAINMENT INTEGRITY..............................

1-4 1.30 PROCESS CONTRCL PR0 CRAM....................................

1 1.31 PURGE-PURGING.............................................

1-5 e

1.32 RATED THERMAL P0WER........................................

1-S 1.33 REACTOR PROTECTION SYSTEM RESPONSE TIME....................

1-6 1.34 REPORTABLE EVENT...........................................

1-6 1.35 ROC DENSITY................................................

1-6 l'.36 SECONDARY CONTAINMENT INTEGRITY............................

1-6

({}

1. 3 7 : SHUTDOWN MA " 'I N............................................

1-6 1,38 SITE BOUNDAR 1-6 i

1.

1.40 SOURCE CHECK...............................................

1-7 1.41 STAGGERED TEST BASIS.......................................

1-7 1.42 THERMAL P0WER..............................................

1-7 1.43 TURBINE BYPASS SYSTEM RESPONSE TIME........................

1-7 L

l' l.44 UNIDENTIFIED LEAKAGE.......................................

1-7 1.45 UNRESTRICTED AREA...........................

1-7 l

I i

1.47' VENTING..........................................

1-8 1'

l O

l F-ERMI - UNIT 2 ii Amendment No.

lilDE

=Q LiliLTJNG CONDITIONS FOR OPERATION AND SURVIillANCLREQUIRt]RTS SECTION l%Cf 3/4.3 INSTR)fENTATION 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTAi MN...........

3/4 3-1 3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION.................

3/4'3-9 3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION...................................... :/4 3-23 3/4.3.4 ATWS RECIRCULATION PUMP TRIP SYSTEM ACTUAT10t' INSTRUMENTATION................................,....

3/4 3-32 3/4.3.5 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION.....................................

3/4 3 45 3/4.3.6 CONTROL R00 BLOCK INST RUMENTATION...................

3/4 3-41 3/4.3.7 MONITORING INSTRUMENTATION Radiation Monitoring Instrumentation................

3/4 3-47 Seismic Monitoring Instrumentation..................

3/4 3-51 Meteorological Monitoring Instrumentation...........

3/4 3-54 Reme u Shutdown System Instrumentation and Controls 3/4 3-57 Accident Moritoring Instrumentation.................

3/4 3-60 Source Range - Moni tors...............................

3/4 3-64

_O Traversing In-Core Probe System....................

3/4 3-65 Chlorine Detection System.........................

3/4 3-66 D e l e ted............................................. 3/ 4 3 - 6 7

' Loose-Part Detection System.........................

3/4 3-70 1

I i

Explosive Gas Monitoring Instrument t,o 3/4 3-76 l

3/4.3.9 FEEDWATER/ MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION........................................

3/4 3-86 3/.3.10 RESERVED............................................

4 3/4.3.11 APPENDIX R ALTERNATIVE SHUTDOWN INSTRUMENTATION......

3/4 3-90 O

FERMI - UNIT 2 v

Amendment No. EJ,-E9, E/,7J,

EDil

(]

LIMITING CONDITIONS FOR OPERATION AND SURVEll1ANCE REQU_1 REM W IS SECTION P3LG1 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 RECIRCULATION SYSTEM Recirculation-Loops..................................

3/4 4-1 Jet Pumps............................................

3/4 4-4 Recirculation Pumps..................................

3/4 4-5 Idle Recirculation loop Startup......................

3/4 4-6 3/4.4.2 SAFETY / RELIEF VALVES Safety / Relief Va1ves.................................

3/4 4-7 Safety / Relief Valves low-Low Set function............

3/4 4-8 3/4 4.3 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems............................

3/4 4-9 Operat i on al Le a kag e..................................

3/4 4-10 3/4.4.4 CHEMISTRY............................................

3/4 4-13 3/4-.4.5 SPECIFIC ACTIVITY....................................

3/4 4-16

--0 3/4.4.6 eaESSuRE/ TEMPERATURE LIMITS Reactor Cool ant System...............................

3/4 4-19 Reactor Steam Dome...................................

3/4 4-23 3/4.4.7 MAIN STEAM LINE ISOLATION VALVES.....................

3/4 4-24 3/4.4.8 STRUCTURAL INTEGRITY.................................

3/4 4-25 3/4.4.9 RESIDUAL HEAT REMOVAL Hot Shutdown.........................................

3/4 4-26 Cold Shutdown........................................

3/4 4-28 i

3/4.4.10 CORE THERMAL HYDRAULIC STABILITY.....................

3/4 4-30 l

3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ECCS - 0PERATING.....................................

3/4 5-1 3/4.5.2 ECCS -

SHUTD0WN......................................

3/4 5-6 3/4.5.3 SUPPRESSION CHAMBER..........................

3/4 5-8 O

l FERMI - UNIT 2 vi Amendment No. 53

itiM2 Q

t.1MlllNGC0JiDJTj0NSFOROPERATIONANDSURVEILLANCERJ1URIMENTS j

SECTION EAGE 3/4.11 RADIDACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS 1

l I

I Liquid Holdup Tanks........................................

3/4 11-7 3/4,1) 2 GASEOUS EFFLUENTS i

l i

I I

1 O

I Explosive Gas Mixture......................................

3/4 11-16 Main Condenser.............................................

3/4 11-17 i

l I

i 3/4.12 DELETED l

l l

i O

FLRMI - UNIT 2 xi Amendment No.

JEf2

]

BASES SECT I.0J -

PAft[

3/4.0 A P P L I C A B l_ LUX............................................

B 3/4 0-1 3L4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 SHUTDOWN MARGIN...............................

B 3/4 1-1 3/4.1.2 REACTIVITY AN0MAllES..........................

B 3/4 1-1 3/4.1.3 CONTROL R0DS..................................

B 3/4

'-2 3/4.1.4 CONTROL R0D PROGRAM CONTR0LS..................

B 3/4 1-3 3/4.1.5 STANDBY LIQUID CONTROL SYSTEM................

B 3/4 1-4 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE..........................................

B 3/4 2-1 l

3/4.2.3 MINIMUM CRITICAL POWER RATIO..................

B 3/4 2-3 3/4.2.4 LINEAR HEAT GENERATION RATE...................

B 3/4 2-4

' O 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION.....

B 3/4 3-1 3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION...........

B 3/4 3-2 3/4.3.3 EMERGENCY-CORE-COOLING SYSTEM ACTUATION INSTRUMENTATION...............................

B 3/4 3 3/4.3.4 ATWS RECIRCULATION PUMP TRIP SYSTEM ACTUATION INSTRUMENTATION...............................

B 3/4 3-3 3/4.3.5 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION.....................

B 3/4 3-3 3/4.3.6 CONTR01. R0D BLOCK INSTRUMENTATION...............................

B 3/4 3 3 3/4.3.7

-MONITORING: INSTRUMENTATION Radiation Monitoring instrumentation..........

B 3/4 3-3 Seismic Monitoring Instrumentation............

B 3/4 3-4 O

FERMI - UNIT 2 xii Amendment No. EA, 69 s

a

-. ~ -

. _ ~

INDEX

@ASES SECTION-E!LG1 INSTRUMENTATION (Continued)

MONITORING INSTRUMENTATION (Continued) q Meteorological Monitoring Instrumentation....... B 3/4 3-4

)

Remote Shutdown System Instrumentation and C o n t rol s........................................ B 3/ 4 3 4 Accident Monitoring Instrumentation............. B 3/4 3-4 Source Range Monitors........................... B 3/4 3-4 Traversing In-Core Probe System................. B 3/4 3-4 Chlorine Detection System....................... B 3/4 3-5 Del e ted......................................... B 3/4 3 Loose-Part Detection System..................... B 3/4 3-5 l

Explosive Gas Monitoring Instrumentation........ B 3/4 3-6 l

3/4.3.9 FEEDWATER/ MAIN TURBINE TRIP SiSTEMS ACTUATION INSTRUMENTATION.................................

B 3/4 3-6 3/4.3.10 RESERVED 3/4 3.11 APPENDIX-R ALTERNATIVE SHUTDOWN INSTRUMENTATION.................................

B 3/4 3-6 3/4.4 REACTOR C00LAt1T SYSTEM 3/4.4.1

_ RECIRCULATION SYSTEM............................ B 3/4 4-1

'3/4.4.2 SAFETY / RELIEF VALVES............................ B 3/4 4-la 3/4.'4.3' REACTOR COOLANT SYSTEM LEAKAGE Leakag: Detection Systems....................... B 3/4 4-2 Operational Leakage............................. B 3/4 4-2 3/4.4.4 CHEMISTRY....................................... B 3/4 4-2 3/4.4.5 SPECIFIC ACTIVITY.............................. B 3/4 4 3/4.4.6 PRESSURE / TEMPERATURE LIMITS..................... B 3/4 4-4 3/4.4.7 MAIN STEAM LINE ISOLATION VALVES..............

B 3/4 4-5 3/4.4.8 STRUCTURAL INTEGRITY.................

... B 3/4 4-5 Q

3/4.4.9 RESIDUAL HEAT REMOVAL..............

......... B 3/4 4-5 FERMI - UNIT 2 xiii Amendment No.,49, EJ, E9, E7, /J,

INDG i

)

]

ILASES SEC110N PAGE 3/4.4 REACTOR CQ0LANT SYSTEM (ContinuedJ 3/4.4.10 CORE THERMAL HYDRAULIC STABILITY................ B 3/4 4 8 l

3/4.5-EMERGENCY CORE COOLING SYSTEMS 3/4.5.1/2 ECCS - OPERATING and SHUTD0WN...................

B 3/4 5-1 3/4.5.3 SVPPRESSION CHAMBER............................. B 3/4 5 2 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT-Primary Containment Integrity................... B 3/4 6-1 Primary Containment Leakage..................... B 3/4 6-1 Primary Containment Air Locks................... B 3/4 6-la MSIV Leakage Control System..................... B 3/4 6-2

' nary Containment Structural Integrity........ B 3/4 6-2 O

Drywell and Suppression Chamber Internal Pressure.................................... B 3/4 6-2 Drywell Average Air Temperature................. B 3/4 6-2 Drywell and. Suppression Chamber Purge System.... B 3/4 6-2 3/4.6.2 DEPRESSURIZATION SYSTEMS........................ B 3/4 6-3 3/4.6.3 PRIMARY CONTAINMENT ISOLATION VALVES............ B 3/4 6-6 3/4.6.4 VACUUM RELIEF.................................. B 3/4 6-6 3/4.6.5 SECONDARY CONTAINMENT........................... B 3/4 6-6 3/4.6.6 PRIMARY CONTAINMENT ATMOSPHERE CONTROL.......... B 3/4 6-7 O

FERMI - UNIT 2 xiv Amendment No. #, 59

i

'[

l!NDll BASES SECTION

.PAGE

.3/4.11 RA010 ACTIVE EffLUEN15 3/4.11.1 LIQUID EFFLUENTS...................................... B 3/4 11,1 3/4.11.2

. GASE0US E f flu ENT S..................................... B 3/4 11 -3 3/4.12 DELETED O

t

}

l.

1:

l-

.O FERMI - UNIT 2 xvii f.mendment No.

_.____m._.-._

180f1 DESIGN FEATURES SI.G_.Loli IWlE 51 SITE Exclusion Area......................................

5-1 Low Popul a ti on 20ne.......................................

5 - 1 Site Boundaries............................................

5-1 5,2 CONTAINMENT C o n fi g u ra t i on..............................................

5 - 1 i

Design Temperature and Pressure............................

5 1 Secondary Containment......................................

5-1 5.3 REACTOR CORE F u el A s s e mbl i e s............................................

5 - 5 T,o n trol Rod As s embl i e s....................................

5 - 5 5.4 REACTOR C00LANT SYSTEM Design Pressure and Temperature............................

5-5 Volume.....................................................5-5

'5.5-METEOROLOGICAL TOWER LOCAT10N..............................

5-6 5.6 FUEL STORAGE

. Criticality................................................

5-6 Drainage...................................................

5-6 Capacity...................................................

5-6 l

l 5.7 COMPONENT CYCLIC OR TRANSifNT LIMIT........................

5-6 y

O FERMI - UNIT 2 xviii

11M1

(])

ADMINISTRATIVE CONTR0lS ELC1LQH MGL 61 R E SEQB1Ln1L 11 Y...........................................

6-1 1

622 ORGANIZAll03.............................................

6-1 6.2.1 0FFSITE AND ONSITE ORGANIZATIONS..................

6-1 j

6.2.2 UNIT STAFF........................................

6-1 6.2.3 INDEPENDENT SAFETY ENGINEERING GROUP..............

6-6 FUNCT10N..........................................

6-6 COMPOSIT10N.......................................

6-6 RESP 0NSIBILITIES..................................

6-6 AUTH0RITY.........................................

6-6 6.2.4 SHIFT TECHNICAL ADVIS0R...........................

6-6 6.3 UNIT STAFF 0VAllFICAT10NS................................

6-7 6.4 TRAINING.................................................

6-7 6.5 REVIEW AND AUDIT.........................................

6-7 6.5.1 ONSITE REVIEW ORGANIZATION (0SRO)

([)

FUNCT10N..........................................

6-7 COMPOSIT10N.......................................

6-7 ALTERNATES........................................

6-8 MEETING FREQUENCY.................................

6-8 QU0 RUM...............................

6-8 RESP 0NSIBILITIES..................................

6-8 REC 0RDS..................................

6-9 6.5.2 NUCLEAR SAFETY REVIEW GROUP (NSRG)................

6-9 FUNCT10N..........................................

6-9 COMPOSIT10N.......................................

6-10 ALTERNATES........................................

6-10 CONSULTANTS.................

6-10 MEETING FREQUENCY.......................

6-10 QU0 RUM.....................

6-10 O

FERMI - UNIT 2 xix Amendment No.

g ADMINISTRAJ,lVECONTR0ls

.SEC110N PM[

REVIEW..............................................

6-10 AUDITS..............................................

6-11 REC 0RDS.............................................

6-12 6.5.3 TECHNICAL REVIEW AND CON 1ROL........................

6-12 ACTIVITIES..........................................

6-12 REVIEW..............................................

6-13 SAFO Y EVALUAT10NS..................................

6-13 i

QUALIFICATIONS..............................,...... 6-13 REC 0RDS.............................................

6-13 65 REPORTABLE EVENT ACTI0y....................................

6-13 6.7-SAFETY LIMIT VIOLAT10N.....................................

6-14 6.8 PROCEDURES AND PR0 GRAMS.,..................................

6-14 6.9 REPORTING RE0VIREMENTS.....................................

6-16b 6.9.1 ROUTINE REP 0RTS.....................................

6-16b STARTUP REP 0RT......................................

6-16b ANNUAL REP 0RTS......................................

-6 17 A U R

0 A

bhMkkikLbPkkkiib'kkPbki'.'

8

(]

SEMIANNUAL RADI0 ACTIVE EFFLUENT RELEASE REPORT......

6-18 j

6. 9. 2 -

SPECIAL REP 0RTS.....................................

6-21 6.9.3 CORE OPERATING LIMITS REP 0RT.........................

6-21 6.10 RECORD RETENTION..........................................

6-21 6.11 RADIATION PROTECTION PR0 GRAM..............................

6-22a 6.12 HIGH RADIATION AREA.......................................

6-22a 6.13 PROCESS CONTROL PR0 GRAM...................................

6-23 6.14 0FFSITE DOSE CALCULATION MANUAL...........................

6-24 i

l l O FERMI - UNIT 2 xx Amendment No. JJ. EA,

~

6 1HDQ Q

idSTOFTABLESIContinued)

LAHLE MGE 4.3.4-1 AiWS RECIRCULATION PUMP TRIP AC10AT10N INSTRUMENTATION SURVEILLANCE REQUIREMENTS.......

3/4 3-35 3.3.5-1 REACTOR CORE ISOLATION COOLING 5YSTEM ACTUATION INSTRUMENTATION.................................

3/4 3-37 3.3.5-2 REACTOR CORE ISOLATION COOLING SYSTEti ACTUATION INSTRUMENTATION SETP0INTS.......................

3/4 3-39 4.3.5.1-1 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS.......

3/4 3-40 3.3.6-1 CONTROL ROD BLOCK INSTRUMENTATION...............

3/4 3 42 3.3.6-2 CONTROL ROD BLOCK INSTRUMENTATION SETP0INTS.....

3/4 3-44 4.3.6-1 CONTROL R00 BLOCK INSTRUMENTATION SURVEILLANCE REQUIREGENTS....................................

3/4 3-45 3.3.7.1-1 RADIATION MONITORING INSTRUMENTATION............

3/4 3-48 4.3.7.1-1 RADIATION MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS.......................

3/4 3 50 0-3.3.7.2-1 SEISMIC MONITORING INSTRUMENTATION..............

3/4 3-52 4.3.7.2-1 SEISMIC MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS.......................

3/4 3-53 3.3.7.3-1 METEOROLOGICAL MONITORING INSTRUMENTATION.......

3/4 3-55

-4.3.7.3-1 METEOROLOGICAL MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS.......................

3/4 3-56 3.3.7.4-1 REM TE SHUTDOWN SYSTEM INSTRUMENTATION..........

3/4 3-58 4.3.7.4-1 REMLTE SHUTOOWN MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS.......................

3/4 3-59 3.3.7.5-1 ACCIDENT MONITORING INSTRUMENTATION.............

3/4 3-61

-4.3.7.5-1 ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS.......................

3/4 3-63 O

FERMI - UNIT 2 xxiii

RfDIX

.Q LIST OF TABLES (Continued)

_. TABLE Eg 3.3.7.9-1 DELETED..........................................

3/4 3-68 i

I i

3.3.7.12 1 EXPLOSIVE GAS MONITORING INSTRUMENTATION........

3/4 3-77 4.3.7.12-1 EXPLOSIVE GAS MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS......................

3/4 3 81 3.3.9-1 FEEDWATER/ MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION.................................

3/4 3-87 3.3.9-2 FEEDWATER/ MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION SETPOINTS.......................

3/4 3-88 4.3.9.1-1 FEEDWATFR/ MAIN TURBINE TRIP SYSTEM AClVATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS.......

3/4 3-89 3.3.11-1 APPENDIX R ALTERNATIVE SHUTDOWN INSTRUMENTATION.,

3/4 3-91 4.3.11.1-1 APPENDIX R ALTERNATIVE SHUTDOWN INSTRUMENTATION SURVEILLANCE REQUIREMENTS........

3/4 3-92 3.4.3.2-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION vAtvES..........................................

3/4 4-12

-3.4.3.2-2 REACTOR COOLANT SYSTEM INTERFACE VALVES LEAKAGE PRESSURE MONITORS...............................

3/4 4-12 1

0 3.4.4-1 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS.........

3/4 4-15 4.4.5-1 PRIMARY COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM................................

3/4 4-18 4.4.6.1.3-1 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM--

WITHDRAWAL SCHEDULE.............................

3/4 4-22 4.6.1.1-1 PRIMARY CONTAINMENT ISOLATION VALVES / FLANGES LOCATED IN LOCKED HIGH RADIATION AREAS..........

3/4 6-lb 3.6.3-1 PR'IMARY CONTAINMENT ISOLATION VALVES............

3/4 6-22 3.6.5.2-1 SECONDARY CONTAINMENT VENTILATION SYSTEM AUTOMATIC ISOLATION DAMPERS.....................

3/4 6-53 3.7.3-1 SURVEY POINTS FOR SHORE BARRIER.................

3/4 7-12 3.7.7.5-1 DELETED..........................................

3/4 7-32 3.7.7.6-1 DELETED..........................................

3/4 7-37 i

4.8.1.1.2-1 DIESEL GENERATOR TEST SCHEDULE..................

3/4 8-8 l

l l

1 O FERMI - UNIT 2 xxiv Amendment No. Q, M, E2, i

l!111El Q

L{gTJF T_ ABLES,,,,[Continuedj_

IMIE ME 4.8.2.1-1 BATTERY SURVEILLANCE REQUIREMEfils.............

3/4 8-12 3,8.4.2-1 PRIMARY CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES................

3/4 8-19 3.8.4.3-1 MOTOR-0PERATED VALVES THERMAL OVERLOAD PROTECTION....................................

3/4 8-21 3.8.4.5-1 STANDBY LIQUID CONTROL SYSTEM ASSOCIATED ISOLATION DEVICES 480 V MOTOR CONTRCL CENTERS..

3/4 8-27 i

B 3/4.4.6-1 REACTOR VESSEL TOUGHNESS......................

B 3/4 4-6 O

5.7.1-1 COMPONENT CYCLIC OR TRANSIENT LIMITS..........

5-7 6.2.2-1 MINIMUM SHIFT CREW COMPOSITION...............,

6-5 m

O FERMI - UtilT 2 xxv Amendment No. Ef,

. - - ~. -.

DEFINITIONS

(~Y lliEC10ENCY N01AT10N v

1.14 The FREQUENCY NOTATION specified for the performance of Surveillance Regt..rements shall correspond to the intervals defined in Table 1.1.

c

-10ENTiflE0 LEAKACE 1.15 IDENTIFIED LEAKAGE shall be:

a.

Leakage into collection systems, such as pump seal or valve packing leaks, that is captured and conducted to a sump or collecting tank, or b.

Leakage into.the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of the leakage detection systems or not to be PRESSURE B0UNDARY LEAKAGE.

J30LATJON SYSTEM RESPONSE TIME 1.15 ihe ISOLATION SYSTEM RESPONSE TIME shall be that time interval from when the monitored parameter exceeds Rs isolation actuation setpoint at the channel sensor until the isolation valves travel to their required positions.

Times shall include diesel generator starting and sequence loading delays where applicaole.

The response time may be measured by any series of sequential, overlapping or total steps such that the entire response time is measured.

LIMITING CONTROL R00 PATTERN 1.17 A LIMITING CONTROL R00 PATTERN shall be a pattern which results in the core being on a thermal hydraulic limit, i.e., operating on a limiting value for APLHGR, LHGR, or MCPR.

LINEAR HEAT GENERATION RATE O

1.18 LINEAR HEAT GENERATION RATE (LHGR) sh 11 be the heat generation per unit length of fuel rod.

It is the integral of the heat flux over the heat transfer area associated with the unit length.

LOGIC SYSTEM FUNCTIONAL TEST 1.19 A LOGIC SYSTEM FUNCTIONAL TEST shall be a test of all logic components, i.e.,

all relays and contacts, all trip units, solid state logic elements, etc., of a logic circuit, from sensor through and including the actuated device, to verify OPERABILITY. The LOGIC SYSTEM FUNCTIONAL TEST may be performed by any series of sequential, overlapping or. total system steps such that the entire logic system is tested.

MAXIMUM FRACTION OF LIMITING POWER DENSITY l.20.The MAXIMUM FRACTION OF L.MITING POWER DENSITY (MFLPD) shall be the highest L

value of the FLPD which exists in the core.

MEMBER (S) 0F THE PUBLIC 1.21 MEMBER (S) 0F THE PUBLIC shall include all perscas who are not occupationally associated with the plant.

This category does not include employees of the utility, its contractors or vendors.

Also excluded from this category are l

persons who enter the site to service equipment or to make deliveries.

This category does include persons who use portions of the site for recreational, occupational or other purposes not associated with the plant.

O l

l FERMI - UNIT 2 1-3 I

1 l

~ - - -

1 I

MF_IfDT10NS-

.O tRNIMUM CRITICAL POWER RATIO V.

1.22 The MINIMUM CRITICAL POWER RATIO (MCPR) shall be the smallest CPR which exists in the core.

OFF-GAS TREATMENT SYST W l.23 An 0FF-GAS TREATMENT SYSTEM is any system designed and installed to reduce l

radioactive gaseous effluents by collecting reactor coolant system offgases 1

from the reactor coolant and providing for delay or holdup for the purpose of redadng the total radioactivity prior to release to the environment.

OFFSITE DOSE CALCULATION MANUAL 1.24 The OFFSITE DOSE CALCULATION MANUAL (00CM) shall contain the methodology and

)

parameters used in the calculation of offsite doses resulting from radioactive

)

gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring Alarm / Trip Setpoints, and in the conduct of the Radiological Environmental-Monitoring Program.

The ODCM shall also contain (1) the Radioactive Effluent Controls and Radiological Environmental Monitoring Programs required by Section 6.8.5 and (2) descriptions of the information that should be included in the Annual Radiological Environmental Operating and Semiannual Radioactive Effluent Release Reports required by Specifications 6.9.1.7 and 6.9.1.8.

OPERABLE - OPERABILITY 1.25 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 Q

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 CONDITION - CONDITION 3

1.26.

An OPERATIONAL CONDITION, i.e.,

CONDITION, shall be any one inclusive combination of mode switch position and average reactor coolant temperature as specified in Table 1.2.

PHYSICS TESTS 1.27 PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instruinentation and (1) described in Chapter 14 of the FSAR, (2) authorized under the provisions of 10 CFR 50.59, or (3) otherwise approved by the Commission.

PRESSURE B0yNDARY LEAKAGE 1.28 PRESSURE B0UNDARY LEAKAGE shall be leakage through a nonisolable fault in a reactor coolant system component body, pipe wall, or vessel wall.

PRIMARY CONTAINMENT INTEGRITY 1.29 PRIMARY CONTAINMENT INTEGRITY shall exist when:

l a.

All primary containment penetrations required to be closed during accident conditions are either:

1.

Capable of being closed by an OPERABLE primary containment Q

automatic isolation system, or FERMI - UNIT 2 1-4 Amendment No.

0EFINITIONS

{.

v 2.

Closed by at least one manual valve, blank flange, or deactivated automatic valve secured in its closed position, except as provided in Ta'.9 3.6.3-1 or Specification 3.6.3.

b.

All primary containment equipment hatches are closed and sealed.

c.

Each primary containment air lock is in compliance with the requirements of Specification 3.6.1.3.

I d.

The primary containment leakage rates are within the limits of Specification 3.6.1,2.

e.

The suppression chamber is in compliance with the requirement of Specification 3.6.2.1.

f.

The sealing mechanism associated with each primary containment penetration, e.g., welds, bellows, or 0-rings, is OPERABLE.

g, The suppression chamber to reactor building vacuum breakers are in compliance with Specification 3.6.4.2.

THE PROCESS CONTROL PROGRAM 1.30 The PROCESS CONTROL PROGRAM (PCP) shall cortain the current formulas, sampling, analyses, test, and determinations to be made to ensure that processing and packaging of solid radioactive wastes based on Q-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, State regulations, burial ground reqairements, and other requirements governing the disposal of solid radioactive waste.

PURGE - PURGING 1.31 PURGE or PURGING is the controlled process of discharging air or gas from l

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.

RATED THERMAL POWER i

1.32 RATED THERMAL POWER shall be-a total reactor core heat transfer rate to the reactor coolant of 3293 MWf.

l I

O FERMI - UNIT 2 1-5 Amendment No. //,

l REDNJ110NS-REACTOR N@lRlIDII_SYSTITitCSPONSE 1FME 1.33 REACTOR PROTECTION SYSTEM RESPONSE TIME shall be the time interval from when the monitored parameter exceeds its trip setpoint at the channel sensor until de-energization of the scram pilot valve solenoids.

The response time may be measured by any series of sequential, overlapping or total' steps such that the entire response time is measured.

REPORTABtE EVENT 1.34 A REPORTAllLE EVENT shall be any of those conditions specified in Section 50.73 te 10 CFR Part 50.

R00 DENSITY 1.35 R0D DENSITY shall be the number of control rod notches inserted as a fraction of the total number of control rod notches.

All rods fully inserted is equivalent to 100% R00 DENSITY.

SECONDARY CONTAINMENT INTEGRill 1.36 SECONDARY CONTAINMENT INTEGRITY shall exist when:

a.

All secondary containment penetrations required to be closed during accident conditions are either:

1.

Capable of being closed by an OPERABLE secondary containment automatic isolation system, or 2.

Closed by at least one manual valve, blind flange, or deactivated automatic danper secured in its closed position, except as provided in Table 3.6.5.2-1 of Specification 3.6.5.2.

b.

All secondary containment hatches and blowout panels are closed and sealed.

O c.

The standby gas treatment system is in compliance with the requirements of Specification 3.6.5.3.

d.

At least one door in each access to the secondary containment is closed (except as noted in item g below).

j e.

The sealing mechanism associated with each secondary containment penetration, e.g., welds, bellows or 0-rings, is OPERABLE.

f.

The pressure within the secondary containment is less than or equal to the value required by Specification 4.6.5.1.a.

g.

Both railroad bay access doors are OPERABLE and closed except for j

ingress and egress or testing as specified-by Specification 3.6.5.1.

SHUTDOWN MARGIN 1.37-SHUTDOWN MARGIN shall be the amount of reactivity by which the reactor is subtritical or would be subcritical assuming all control rods are fully inserted except for the single control rod of highest reactivity worth which is assumed to be fully withdrawn and the reactor is in the shutdown condition; cold, i.e., 68 F; and xenon free.

SITE BOUNDARY 1.38 The SITE B0UNDARY shall be that line beyond which the land is neither owned, nor leased, nor otherwise controlled, by the licensee.

FERMI - UNIT 2 1-6 Amendment No. 34

i DfflN1110NS 1.39 DELETED SOURCE CHfCK 1.40 A SOURCE CilECK shall be the qualitative assessment of channel response when the channel sensor is exposed to a radioactive source.

STACjlfRED TEST BA_ SIS 1.41 A STAGGERED TEST BASIS shall consist of:

a.

A test schedule for n systems, subsystcms, trains or other designated components obtained by dividing the specified test interval into n equal subintervals, b.

The testing of one system, subsystem, train or other designated component at the beginning of each subinterval.

THERMAL POWER 1.42 TilERMAl. POWER shall be the total reactor core heat transfer rate to the reactor coolant.

TURBINE BYPASS SlSTEM RESPONSE TIME (3

1.43 The TURBINE BYPASS SYSTEM RESPONSE TIME shall be that time interval from when the turbine bypass control unit generates a turbine bypass valve flow signal until the turbine bypass valves travel to their required positions. The response time may be measured by any series of sequential, overlapping or total steps such that the entire response time is measured.

UNIDENTIFIED 1.EAVKtE 1,44 UNIDENTiflED LEAKAGE shall be all leakage which is not IDENTlflED LEAKAGE.

UNRESTRICT_(D ARE6 1.4S 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.

FERMI - UNIT 2 17 Amendment No.

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j i

Q DEFINITIONS 1,46 DELETED

. VENTING I

1.47 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 provided or required during VENTING. Vent, used in system namas, does not imply a VENTING process.

O:

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FERMI UNIT 2 1-8 Amendment No.

[

O O

ot TABLE 3 7.5-1 m

%3 ACCIDENT MONITORI.s INSTRUMENTATION 4

C3 MINIMUM APPLIC.ABLE REQUIRED NUMBER CHANNELS OPERATIONAL INSTRUMENT OF CHANNELS OPERABLE CONDITIONS ACTION 1.

Reactor Vessel Pressure 2

1 1, 2 80-2 Reactor Vessel Water Level' a.

Fuel Zone 2

'I I, 2

.80 b.

Wide Range 2

1 1, 2 80

3. ' Suppression Chamber Water Level 2

1 1, 2 80 4.' Suppression Chamber-Water Temperature 2

I 1, 2 80 m

x

[

5.' Suppression Chamber Air Temperature 2

I 1, 2 80' l

S 6.

Suppression Chamber Pressure 2

1 1, 2.

80 7.

Drywell Pressurc, Wide Range 2

I

. I, 2.

80

[

8.

Drywell Air Temperature 2

1 1, 2 80 9.

Drywell Oxygen Concentration 2

1-1, 2 80 f

10.

Drywell Hydrogen Concentration 2

1 1, 2 80 11.

Safety / Relief Valve Position Indicators 1*/ valve 1*/ valve I, 2 80 l

12..

Containment High Range Radiation Monitor 2

2 I, 2, 3 81 E

0

Pressure switch

o LO O

TABLE 3.3.7.5-1 (Continued)-

9' 25

' ACCIDENT MONITORING INSTRUMENTATION t

MINIMUM APPLICABLE g:

REQUIRED NUMBER-CHANNELS OPERATIONAL'

[

INSTRUMENT OF CHANNELS OPERABLE CONDITIONS ACTION.

13. Standby Gas Treatment System Radiation Monitors SGTS - Noble Gas- (Low-range)#

1/0PERABLE 1/0FERABLE i, 2, 3 81 a.

SGTS subsystem 5GTS subsystem b.

SGTS - Noble Gas (Mid-range) 1/0PERABLE 1/0PERABLE 1, 2, 3 81 SGTS subsystem SGTS subsystem c.

'SGTS - AXM-Noble Gas (Mid-range) l/0PERABLE 1/0PERABLE 1, 2, 3 81 SGTS subsystem. 53TS subsystem w1 d.

SGTS - AXM-Noble' Gas (High-range) 1/0PERABLE 1/0PERABLE 1, 2, 3

81 co s,

SGTS subsystem SGTS subsystem 14.

Neutron Flux 2

1 1, 2

.80 15.

Deleted 16.

Primary Containment Isolation Valve Position' 1/ valve 1/ valve 1, 2, 3 82 N

3 S-E n

EF T

Also included'in'the OFFSITE DOSE CALCULATION MANUAL.

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PAGES 3/4 3-72 THROUGH 3/4 3-75 HAVE BEEN DELETED O

O FERMI - UNIT 2 3/4 3-71 Amendment flo. 44,

llLSTEliD11AU014 f XPLQSlyl GAS M0fillRillf!G_lNSTRUMENTA110tj j

l.lM111NG CONDITION FOR OPERAlION 3.3.7.12 The explosive gas monitoring instrumentation channel shown in Table I

3.3.7.12-1 shall be OPERABLE with its alarm setpoint set to ensure that the limits of Specification 3.11.2.6 are not exceeded.

APPL 1[ABILill: As shown in Table 3.3.7.12-1 ACTION:

a.

With an explosive gas monitoring instrumentation channel alarm j

setpoint less conservative than required by the above Specification, declare the channel inoperable and take ACTION shown in Table 3.3.7.12-1; or change the setpoint so it is acceptably conservative.

b.

With less than the minimum number of explosive gas monitoring j

instrumentation channels OPERABLE, take the ACTION shown in Table 3.3.7.12-1.

Restore the inoperable instrumentation to OPERABLE status within 30 days and, if unsuccessful, prepare and submit a I

special report to the Commission pursuant to Specification 6.9.2 to explain why this inoperability was not corrected in a timely manner.

c.

The provisions c.' Specification 3.0.3 and 3.0.4 are not applicable.

O SURVElllANCE REQUIREMENTS i

4.3.7.12 Each explosive gas monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, CHANNEL CAllBRATION l

and CHANNEL FUNCTIONAL TEST operations at the frequencies shown in Table 4.3.7.12-1.

O FERMI - UNIT 2 3/4 3-76 Amendment No. 7,4,

O O-0; TABLE 3.3.7.12-1 95 EXPLOSIVE GAS MONITORING INSTRUMENTATION i

C5 MINIMUM CHANNELS INSTRUMENT OPERABLE APPLICABILITY ACTION 1.

OFFGAS MONITORING SYSTEM (At the 2.2 minute delay piping)-

124 i

a.

Hydrogen Monitdr 1

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TABLE 3.3.7.12-1 (Continued)

E EXPLOSIVE GAS MONITORING INSTRUMENTATION Ea

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--.. -.. ~..

TABLE 3.3.7.J2-i (Continuedj.

TABLE _ NOTATIONS l

- During main _ condenser offgas treatment system operation.

ACTION STATEME![Tji l

l l

1 I

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ACTION 124 -

With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, operation of main condenser offgas treatment system may continue provided grab samples are collected at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and analyzed within the following 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

Otherwise, suspend release of radioactive effluents via this pathway.

I I

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FERMI - UNIT 2 3/4 3-80 Amendment No. E,4, 7),

O O

O TABLE 4.3.7.12-1 b

EXPLOSIVE GAS MONITORING INSTRUMENTATION SURVEILLANCE REOUIREMENTS r,

c CHANNEL MODES Ill WHICH 5

CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE CHECK _

CALIBRATION TEST RE0VIRED INSTRUMENT 1.

OFFGAS MONITORING SYSTEM (At the 2.2 minute delay pipi:'g) 0 Q(3)

M Hydrogen Monitor a.

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UNIT 2 3/4 3-82 Amendment No.

. -.. = _ - - - _ -, - -, -.. -.. _, -..-.-, _

TABLE 4.3.7.12-1 (Contirwndl O

If # E N O T A 1.LQ1 6 I

t During main condenser offgas treatment system operation.

1 i

4 i

O (3). The CHANNEL CALICRATION shall include the use of standard gas san

>s containing a. nominal:

1.

One volume percent hydrogen, balance nitrogen, and

.l 2.-

Four volume percent hydrogen, b61ance nitrogen, s

f k

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kh FERMI - UNIT 2 3/4 3 84 Amendment No, i

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PAGES 3/4 11 3 TliROUGH 3/4 11-6 i

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FERMI - UNIT 2 3/4 11-1 Amendment No.

......... 2

.._..._._....__.._.._.._._,._....._.2

ILAD10AC11vE EFFLVIRIS O

LLQV101101DVP 1N1KS LIMITING CONDITION FOR OPfRA110h

3. '.1. ;. 4 1he quantity of radioactive material contained in any outside

'emponry tank shall be limited to less than or equal to 10 curies, excluding teit4 m and dissolved or entrained noble gases.

APPLICA0lLITX:

At all times.

ACTION:

a.

With the quantity of radioactive material in any of the above tanks exceeding the above limit, immediately suspend all additions of radioactive material to the tank, within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months 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 0.9.1.8.

b.

The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVElltANCE REQUIREMENTS 4.11.1.4 The quantity of radioactive material contained in each of the above tanks shall be determined to be within the above limit by analyzing a representative sample of the tank's contents at least once per 7 days when 1

radioactive materials are being added to the tank.

5 ll O

FERMI - UNIT 2 3/4 11 L l

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O FERMI - Utill 2 3/4 11-15 Amendment tio.

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31.idl RADj0 ACTIVE ffftUi!RS O

l 3/4.11.2 GAS 10lS EfftVI!as EXPLQSlyLDAS MIXTURE LIMITING CONDITION FOR OPEMTION 3.11.2.6 The concentration of hydrogen in the main condenser offgas treatment system shall be limited to less than or equal to 47, by volume.

APPLICABILITY: Whenever the main condenser steam Jet air ejectors are in operation.

ACTION:

l a.

With the concentration of hydrogen in the main condenser offgas treatment system exceeding the limit, restore the concentration to within the limit within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , b.

The provisions or Specifications 3.0.3 and 3.0.4 are not applicabic.

SURVElltANCE REQUIREMENTS O

4.ii.2.6 The concentration or hydro 9en in the main condenser otrois treatment system shall be determined to be within the above limits by continuously monitoring the waste gases in the main condenser oftgas treatment system with the hydrogen monitors required OPERABLE by Table 3.3.7.12-1 of Specification

_3.3.7.12.

l' l :.

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O FERMI - UNIT 2 3/4 11-16 Amendment No.

1..

i 1 AD10AC11]If_fifLVifdh O

tiAIN CONDfRSLB 1IM111NG CONOLT10N FOR OPRAll0N 3.11.2.7 1he gross radioactivity rate of noble gases measured at the discharge of the 2.2 minute delay piping shall be limited to less than or equal to 340 mil 11 curies /sec af ter 30 minute decay.

APPLICABIllTY_: OPERAil0NAL CONDITIONS 1, 2*,

and 35 ACTION:

With the gross radioactivity rate of noble gases at the the discharge of the 2.2 minute delay piping exceeding 340 millicuries /sec after 30 minute decay, restore the gross radioactivity rate to within its limit within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least STAR 1UP, with all main steam lines isolated, within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

SURVEllLANCE Rf00lRfMENTS 4.11.2.7.1 The radioactivity rate of r.oble gases at the discharge of the 2.2 minute delay piping shall be continuously monitored in accordance with the ODCH.

4.11.2.7.2 The gross radioactivity rate of noble gases from the main O

condenser steam jet air ejector shall be determined to be within the limits of Specification 3.11.2.7 at the following frequencies by performing an isotopic analysis of a rep.*esentative sample of gases taken at the discharge of the 2.2 minute delay piping:

a.

At least once per 31 days, b.

Within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months following an increase, as indicated by the Offgas Radiation Monitor, af greater than 50%, after factoring out increases due to changes in THERMAL POWER 1cvel, in the nominal steady-state fission gas release from the primary coolant.

c.

The provisions of Specification 4.0.4 are not applicable.

l

When the main condenser air ejector is in operation.

TERMI - UNIT 2 3/4 11-17 Amendment flo. #,

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FERMI-- UNIT 2 3/4 12-1 Amendmcnt No j

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INSlRUMU41AllDH Q-BASES MONITORING INSlRufi[Flell0B (Continued) 3/4.3.7.8 Clll0 RIBE DElECTION SYSTDiS The OPERABillTY of the chlorine and detection system ensures that an accidental chlorine release will be detected promptly and the necessary protective actions wi'1 be automaticaily initiated to provide protection for control room personnel Upon detection of a high concentration of chlorine, the control room emergency ventilation system will automatically be placed in the chlorine mode of operation to provide the required protection.

The detection system required by this specification is consistent with the recommendations of Regulatory Guide 1.95 " Protection of Nuclear Power Plant Control _ Room Operators against an Accidental Chlorine Release", Revision 1, January, 1977.

3/4.3.7.9 DEl1Ij~J1 3/4.3.7.10 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 primary system and avoid or mitigate damage to primary system components. The allowable out-of-service times and surveillance requirements are consistent with the recommendations of Regulatory Guide 1.133, " Loose-Part Detection O

Program for the Primary System of Light-Water-Cooled Reactors," May 1981.

l.-

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O FERMI - UNIT 2 B 3/4 3-5 Amendment No. S, E/,

l 114 STRUM [N1AT10N O

f1Asrs MON I TOR i tlG_11[SJ RUM E N T AT 10N ( Con t i nued )

3/4.3.7,12 fXPLOSIVE GAS MONITORING INSTRUMENTAT10ft l

The explosive gas monitoring instrumentation is provided to monitor the concentrations of potentially explosive gas mixtures in the main condenser offgas treatment system. The OPERABillTY and use of this instrumentation is 3

consistent with the requirements of General Design Critorion 60 of Appendix A to 10 CFR Part 50.

3/4.3.9 TEEDWATER/ MAIN TURBINE TRIP SYSTEM ACTUATION INSTRVli(1(lal10B The feedwater/ main turbine trip system actuation instrumentation is provided to initiate action of the feedwater system / main turbine trip system in the event of a high reactor vessel water level due to failure of the feedwater controller under maximum demand.

3/4.3.11 APPENDIX R ALTERNATIVE SHUTDOWN INSTRUMENTATION The OPERABILITY of the alternative shutdown system ensures that a fire will not preclude achieving safe shutdown.

The alternative shutdown system instrumentation is independent of areas where a fke could damage systems normally used to shutdown the reactor.

Thus, the system capability is consistent with General Design Criterion 3 and Appendix R to 10 CFR 50.

1 l

O FERMI - Utili 2 B 3/4 3-6 Amendment No. f@, /J,

_ _, - _ _. _ _.. _. _... _ _.. _ _. =. _ _... _ _ _ _ _ _.. _. _. - _ _.. _ -.

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FERMI - UNIT 2 B 3/4 11-1 Amendment No.

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i RADI0 ACTIVE fffLV G LS O

t a,scs 3/4.11.1 LLQUID EfflVL!ilS l

3/4.11.1.4 LIQUID HOLDUP tat 1KS The tanks listed in this specification include all those outdoor radwaste tanks that are not surrounded by liners, dikes, or walls capable of holding O

the tank 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-tanks' contents, the resulting concentrations would be less than the limits of 10 CFR Part 20, Appendix B, Table 11, Column 2, at the nearest potable water supply and the nearest surface water supply in an Uf1 RESTRICTED AREA.

l l

l O FERMI - UtilT 2 B 3/4 11-2 Araendment 110.

i RADI0 ACTIVE EffL0Ef411 O'-

i BASES 3/4.11.2 GAsf0!JS ErftVLtES i

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fBD10 ACTIVE EFFLUENTS

.O nos~~$a coadeaser arr9 s system is maintained below the flammability limits of hydrogen and oxygen.~~

^

Maintaining the concentration of hydrogen below the flammability limit 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.

3/4.11.2.7 MAIN CONDENSER Restricting the gross radioactivity rate of noble gases from the main condenser provides reasonable assuranie that the total body exposure to an individual at the exclusion area boundary will not exceed a small fraction of

~~-the limits of 10 CFR Part 100 in the event this effluent is inadvertently discharged directly to the environment without treatment.~~

~~This specification implements the requirements of General Design Criteria 60 and 64 of Appendix A to 10 CFR Part 50.~~

\\

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~~B 3/4 12-1 Amendment'llo.~~

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~
ADMINISIRATIVf CON 1R01,5 O
~~_J_MSE0!ElBIL111 6.1.1 The Plant Manager shall be responsible for overall unit safe operation and shall delegate in writing the succession to this responsibility during his absence.~~
The Plant Manager shall have control over those onsite activities necessary for safe operation and maintenance of the plant.
6.1.2 The Nuclear Shift Supervisor or, during his absence from the control room, a designated individual shall be responsible for the control room command function. A management directive to this effect, signed by the Assistant Vice President and Manage Nuclear Production shall be reissued to all station personnel on an annual basis.
522 ORGANIZATION 522,1.. 0FFSITE AND ONSITE 0_R MNIZAT10f[$
j Onsite and offsite organizations shall be established for unit operation and corporate management, respectively. The onsite and offsite organizations shall include the positions for activities affecting the safety of the nuclear power plant.
a.
Lines of authority, responsibility, and communication shall be established and defined for the highest management levels through intermediate levels to and including all operating organization positions.
These relationships shall be documented and updated, as appropriate, in the form of organization charts, functional O
descriptions of departmental responsibilities and relationships, and job descriptions for key personnel positions, or in equivalent forms of documentation. These requirements shall be documented in the Updated final Safety Analysis Report.
b.-
The Senior Vice President shall have corporate responsibility for overall plant nuclear safety and shall take any measures needed to ensure acceptable performance of the staff in operating, maintaining, and providing technical support of the plant to 3
-ensure nuclear safety.-
c.
The individuals who train the operating staff and those who carry out health physics and quality assurance functions may report to the appropriate onsite manager; however, they shall have-sufficient organizational freedom to ensure their independence from operating pressures.
5 4 L JLf0T staff a.
Each on duty shif t shall be composed of at least the minimum shift crew composition shown in Table 6.2.2 1; b.
At least one licensed Operator shall be in the control room when fuel is in the reactor, in addition, while the unit is in A
OPERA 110NAL_CONDil10N 1, 2 or 5, at least one licensed Senior V
Operator shall be in the control room; FERMI - UNIT 2 6-1 Amendment No. 11, /p, E3,
1 ADM STRAT:Vf C0l41R0lS
[IIII"![lT4Eliontinued)
S c.
A flealth Physics Technician shall be on site when fuel is in the reactor.
I The llealth Physics Technician position may be unfilled for a period of I
time not to exceed 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , in order to accommodate unexpected absence, provided immediate action is taken to fill the required positions; d.
All CORE ALTERAT10f1S shall be observed and directly supervised by either a licensed Senior Operator or licensed Senior Operator Limited to fuel llandling who has no other concurrent responsibilities during this operation; c.
DELETED l
f.
Administrative procedures shall be developed and implemented to limit the working hours of unit staff who perform safcty-related functions (e.g.,
licensed Senior Operators, licensed Operators, health physics personnel, auxiliary operators, and key maintenance personnel).
Adequate shift coverage shall be maintained without routine heavy use of overtime.
The objective shall be to have operating personnel work a nominal 40 hour4.62963e-4 days 0.0111 hours 6.613757e-5 weeks 1.522e-5 months week while the unit is operating.
llowever, in the event that unforeseen problems require substantial amounts of overtime to be used, or during extended periods of shutdown for refueling major maintenance, or major unit modifications, on a temporary 6a, sis the following guidelines shall be followed:
1.
An individual should not be permitted to work more than 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months O.
straight, excluding shift turnover time.
2.
An individual should not be permitted to work more than 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months in any 24-hour period, nor more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months in any 48 hour5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months period, nor more than 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months in any 7 day period, all excluding shift turnover time.
3.
A break of at least 8 hcurs should be allowed between work periods, including shift turnover time.
4 Exc4.t during extended shutdown periods, the use of overtime should be considered on an individual basis and not for the entire staff on a shift.
Any deviation from the above guidelines shall be authorized by the plant Manager or a' Section Superintendent or higher levels of management, in accordance with established procedures and with documentation of the basis for granting the deviation. Controls shall be included in the procedures such 3M individual overtime shall be reviewed monthly by the Plant Man gor. ' a Section Superintendent to assure that excessive hours h ve nct Lew ssigned.
Routine deviation from the above guidelines is not aut b W d.
O FERMI - Utill 2 62 Amendment tio. JJ, #, M, 62
ADMINISTRATIVE CONTROLS I
LROCEDMRES AND PROGRAMS (Continued) 6.8.2 Each plant administrative procedure, and changes thereto, shall be i
reviewed in accordance with Specification 6.5.1.6, and approved by the Plant Manager prior to implementation, and shall be reviewed periodically thereafter as set forth in administrati"e procedures.
6.8.3 Each plant procedure required by Specification 6.8.1, other than administrative procedures, and changes thereto, shall be reviewed in accordance with 6.5.3, and approved by the Plant Manager prior to implementation and shall be reviewed periodically thereaf ter as set forth in administrative procedures. The Plant Manager may delegate approval authority in writing for specific types of procedures to a Section Superintendent.
l l
6.8.4 Temporary changes to procedures of Specification 6.8.1 may be made provided:
a.
.The intent of the original procedure is not altered; b.
The change is approved by two members of the unit management staff, at least one of whom holds a Senior Operator license on Fermt 2; and c.
The change is documented, and reviewed and approved in accordance with either 6.8.2 or 6.8.3 above, as appropriate, within 14 days of implementation.
6.8.5 The following programs shall be established, implemented, and t
maintained:
a.
Primary Coolant Sources Outside Containment A pro vam to reduce leakage from those portions of systems outside containment that could contain highly radioactive fluids during a serious transient or accident to as low as practical levels.
The systems include the HPCI, CS, RHR, RCIC, reactor water sampling, containment sampling, reactor water cleanup, combustible gas control, control rod drive discharge headers, and standby gas treatment systems.
The program shall include the following:
1.
Preventive maintenance and periodic visual inspection requirements, and 2.
Integrated leak test requirements for each system at refueling cycle intervals or less, b.
In-Plant Radiation Monitorino A program which will ensure the capability to accurately determine the airborne iodine concentration in vital areas under accident conditions.
This program shall include the following:
O FERMI - UNIT 2 6-15 Amendment No. JJ, 30
@MittlSTRATIVE C0t11ROLS O
macrot ra ^"o enoca^"s (co" tim"ed) e 1.
Training of personnel, 2.
Procedures for monitoring, and 3.
Provisions for maintenance of sampling and analysis equipment.
c.
Post-accident Sampjing A program which will ensure the capability to obtain and analyze reactor coolant, radioactive lodines and pa)ticulates in plant gaseous effluents, and containment atmosphere samples under accident conditions.
The program shall include the followirg:
1.
Training of personnel, 2.
Procedures for sampling and analysis, and 3.
Previsions for maintenance cf sampling and analysis equipment, d.
Hiah Density Spent fuel Racks A program which will assure that any unanticipated degradation of the high density spent fuel rocks will be detected and will not compromise the integrity of the racks.
O Radioactive Effluent Controis eroaram e.
A program shall be provided conforming with 10 CFR S0.36a for the control of radioactive effluents and for maintaining the doses to i
MEMBERS OF THE PUBLIC from radioactive effluents as low as reasonably achievable. The program (1) shall be contained in the ODCM, (2) shall be implemented by operating procedures, and (3) shall include remedial actions to be taken whenever the program limits are exceeded.
The program shall include the following elements:
I 1)
Limitations on the operability of radioactive liquid and gaseous monitoring instrumentation including surveillance tests and setpoint determination in accordance with the methodology in the ODCM, 2)
Limitations on the concentrations of radicactive material released in liquid effluents to UNRESTRICTED AREAS 1
conforming to 10 CFR Part 20, Appendix 0 Table 11, Column 2,
I 3)
Monitoring, sampling, and analysis of radioactive liquid and I
gaseous effluents in accordance with 10 CFR 20.106 and with the methodology and parameters in the ODCM.
O FERMI - UNIT 2 6-16 Amendment No f;, JJ, 70,
it ADMINISIRAT1Vf CONTROLS O
eROC E oVRE S 2 tin _tacaA113 (Con t i nued )
4)
Limitations on the annual and quarterly doses or dose commitment to a M[MBER Of THE PUBLIC from radioactive materials in liquid effluents released from each unit to UNRESTRICIED AREAS conforming to Appendix ! to 10 CFR Part 50, 5)
Determination of cumulative and arojected dose contributions from radioactive effluents for t1e current calendar quarter and current calendar year in accordance with the methodology and parameters in the ODCM at least every 31 days, 6)
Limitations on the operability and use of the liquid and gaseous effluent treatment systems to ensure that the appropriate portions of these systems are used to reduce releases of radioactivity when the projected doses in a 31-day-period would exceed 2 percent of the guidelines for the annual dose or dose commitment conforming to Appendix 1 to 10 CfR Part 50, 7)
Limitations or the dose rate resulting from radioactive material role, sed in gaseous effluents to areas beyond the SITE BOUNDARY conforming to the doses associated with 10 CFR Part 20 Appendix B, Table 11, Column 1, 8)
Limitations on the annual and quarterly air doses resulting from noDie gases released in gaseous effluents from each unit to areas beyond the SITE BOUNDARY conforming to Appendix ! to 10 CfR Part 50, 9)
Limitations on the annual and quarterly doses to a MEMBER Of
-THE PUBLIC from lodine 131, lodine 133, tritium, and all radionuclides in particulate form with. half-lives greater than 8 days in gaseous effluents released from each unit to areas beyond the SITE BOUNDARY conforming to Appendix 1 to 10 CFR Part 50, 10)
Limitations on venting and purging of the Mark I containment through the Standby Gas Treatment System or the Reactor Building Ventilation System to maintain releases as low as reasonably achievable, 11)'
Limitations on the annual dose or dose commitment to any MEMBER Of THE PUBLIC due to releases of radioactivity and to radiation from uranium fuel cycle sources conforming to 40 CFR Part 190.
FERMI - UNIT 2 6-16a Amer.dment No.
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ADMINISTRATIVf CONTR0lS l
O a m 0VRo mD PR = NS (Continued) f.
ILadj9haial EnvironmentAljipJtjinr_ jag Proorpj A program shall be provided to monitor the radiation and radionuclides in the environs of the plant.
1he program shall arovide (1) representative measurements of radioactivity in the lighest potential exposure pathways, and (2) verification of the accuracy of the of fluent monitoring program and modeling of environmental exposure pathways.
The program shall (1) be contained in the ODCM, (2) conform to the guidance of Appendix !
to 10 CFR Part 50, and (3) include the following:
1)
Monitoring, sampling, analysis, and reporting of radiation and radionuclides in the environment in accordance with the methodology and parameters in the ODCM, 2)
A Land Use Census to ensure that changes in the use of areas at and beyond-the SITE B0VHDARY are identified and that modifications to the monitoring program are made if required by the results of this census, and, 3)
Participation in an Interlaboratory Comparison Program to ensure that independent checks on the precision and accuracy of the measurements of radioactive materials in environmental sample matrices are performed as part of the Q
quality assurance program for environmental monitoring.
6.9 REPORTING RE0VIREMENTS ROUTINE REPORTS 6.9.1 In addition to the applicable reporting requirements of Title 10, Code l
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 supplier, and-(4) modifications that may have significantly altered the nuclear, thermal, or hydraulic performance of the unit.
L 6.9.1.2 The startup report shall address each of the tests. identified in Subsection 14.1.4,8 of the final Safety Analysis Report and shall include a description of the measured values of the operating conditions or characteristics obtained during the test program and a comparison of these values with design predictions and specifications.
Any corrective actions
-O
' hat " re r*a"'r d t FEPJ11 - UNIT 2 6-16b Amendment No.
@MJtdS1RATIVf C0f41R0js O
amuum (Conunued) obtain satisf actory operation shall also be described.
Any additional specific details required in license conditions based on othar commitments shall be included in this report.
6.9.1.3 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 f ollowing 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.
A!! Lit 1Al. rep 0R]S 6.9.1.4 Annual reports covering the activities of the unit 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.
6.9.1.5 Reports required on an annual basis shall include:
a.
A tabulation on an annual basis of the number of plant, utility, and other personnel (including contractors) receiving exposures greater than 100 mrems/yr and their associated man rem exposure according to O
work and job functions,* (e.g., reactor operationi and surveillance, inservice inspection, routine maintenance, special maintenance
[ describe maintenance), waste processing, and refueling). The dose assignments to various duty functions may be estimated based on pocket or thermoluminescent dosimeters (ll0) dosimeters or film badge measurements. Small exposures totalling less than 20% of the individual total dose need not be accounted for.
In the aggregate, at least 80% of the total whole-body dose received from external sources should be assigned to specific major work functions; and b.
Documentation of all challenges to main steam line safety / relief valves, and c.
A summary of ECCS outage data including:
1.
ECCS outage dates and duration of outages, 2.
Cause of each ECCS outage, 3.
ECCS systems and components in the outage, and 4.
Corrective action taken.
d.
The reports shall also include the results of specific activity analysis in which the primary coolant exceeded the limits of Specification 3.4.5.
The following information shall be included:
~
O
" ' ' '"'" '
~~t ' " '" o " "" " t ' t '" r o o" i r o" " t '~~
f 62 4 7 "'
cr" " rt 2 -
FERMI - tmli 2 6-17 Amendment flo, 11
ADMINISMATIVLCO,iHROLS O
strit;At RtP0RiS (Continued)
(1) reactor power history starting 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months prior to the first sample in which the limit was exceeded; (2) results of the last isotopic analysis for radiciodine performed prior to exceeding the limit, results of analysis while limit was exceeded and results of one analysis after the radiciodine activity was reduced to less than limit (each result should include date and time of sampling and the radiotodine concentrations);
(3) clean-up system flow history starting 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months prior to the first sample in which the limit was exceeded; (4) graph of the 1-131 and one other radiolodine isotope concentrations in microcuries per gram as a function of time for the duration of the specific activity above the steady-state level; and (5) the time duration when the specific activity of the primary coolant exceeded the radioiodine limit.
MONTHLY OPERATING REPORTS 6.9.1.6 Routine reports of operating statistics and shutdown experience shall pd be 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 Administrator of the Regional Office of the NRC, no later than the 15th of each month following the calendar month covered by the report.
ANNVM RADIOLOGICAL ENVIRONMENTAL OPERATING REP 0BI 6.9.1.7 The Annual-Radiological Environmental Operating Report covering the operation of the unit during the previous calendar year shall be submitted before May 1 of each year. The report shall include summaries, interpretations, and analysis of trends of the results of the Radiological Environmental Monitoring Program for the reporting period.
The material provided shall be consistent with the objectives outlined in (1) the ODCM and (2) Sections IV B.2, IV.B.3, and IV.C of Appendix 1 to 10 CFR Part 50.
SEMIANNUAL RADIDACTIVE EFFLUENT RELEASE REPORT 6.9.1.8 The Semiannual Radioactive Effluent Release Report covering the operation of the unit during the previous 6 months of operation shall be submitted within 60 days after January 1 and July 1 of each year.
The report
'i shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid wasta released from the unit, The material provided shall be (1) consistent with the objectives outlined in the ODCH and PCP and (2) in conformance with 10 CFR 50.36a anc' action IV.B.1 of Appendix 1 to 10 CFR Part O
5-FERMI - UNIT 2 6-18 Amendmer,t No, JJ.
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-t ADMINISUlATIVE CONTROLS O
t THIS PAGE HAS BEEN INTENTIONALLY DELETED O
.4 l'
lL 1
i l-i FERMI - UNIT.2 6-19 Amendment No. J/,
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~ -. -.... ~.. -
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ADMINISTRAllVE CONTROLS pq v
i THIS PAGE HAS BEEN INTENTIONALLY DELETED I
l(
T.
_ FERMI - UNIT 2 6-20 Amendment No. JJ, /f,
ADMINISTRAllVE CONTR0lS O
SeEC a nePoR15 6.9.? Special reports shall be submitted to the Regional Administrator of the Regioaal Office of the' NRC within the time period specified for each report.
CORE OPERATING _llMITS REPORT 8
6.9.3 Selected cycle specific core operating limitt shall be established and documented in the CORE OPERATING LIMITS REPORT (COLM) before each reload cycle or any-remaining part of a reload cycle. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC in Gen al Electric Company reports NEDE-240ll-P-A and NEDE-20566-P.
The core operating limits shall be determined so that all applicable limits (e.g., fuel thermal mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, and transier.t and accident analysis limits) of the safety analysis are met.
The COLR, including any mid-cycle revisions or supplement thereto, shall be submitted upon issuance to the NRC Document Control Desk, with copies to the Regional Administrator and Resident inspector prior to use.
-6.10 RECORD RETENTION 5.10.1 In addition to the applicable record retention requirements of Title 10, Code of Federal _ Regulations, the following records shall be retained for at least the minimum period indicated.
Q 6.10.2 ' The following records shall be retained for at least 5 years:
a.
Records and logs of unit operat_ ion covering time interval at each power level.
b.
Records and logs of principal maintenance activities, inspections, repair, and replacement of principal items of equipment related to nuclear safety, c.
ALL REPORTABLE EVENTS.
d.
Records of surveillance activities, inspections, and calibrations required by these Technical Specifications, e.
Records of changes made to the procatures required by Specification 6.8.1.
f.
Records of radioactive shipments.
g.
Records of scaled source and fission detector leak tests and results.
h.
Records of annual physical inventory of all sealed source material of record.
O FERMI - UNIT 2 6-21 Amendment No. IJ, 64
ADMINISTRATIVE CONTROLS-
~~O e.10 Rtc0Ro RETENTION - gontinuedi 6.10.3 - The following records shall be retained for the duration of the unit Operating License:~~
a.
Records and drawing changes reflecting unit design modifications made to systems and equipment described in the final Safety Analysis Report.
b.
Records of new and irradiated fuel inventory, fuel transfers, and assembly burnup histories.
c.
Records of radiation exposure for all individuals entering radiation control areas.
d.
Records of gaseous and liquid radioactive material released to the environs',
e.
Records of transient or operational cycles for those unit components identified in Table 5.7.1-1.
f.
Records of reactor tests and experiments.
g.
Records of training and qualification for current members of the unit staff.
Q.
h.
Records of insery".e inspections perfarmed pursuant to these l
Technical Specificaticns, i.
Records of quality assurance activities required by the Operational Quality Assurance Manual.
J.
Records of reviews performed for changes made to procedures or equipment or reviews of tests and experiments pursuant to 10 CFR 50.59, l
l k.-
Records-of meetings of the OSR0 and the NSRG, L
1.
Records of the service lives of all hydraulic and mechanical snubbers required by Specification 3.7.5 including the date at which the service life commences end associated installation and maintenance records.
m.
Records of analyses required by the radiological environmental monitoring program that would permit evaluation of the accuracy of the analysis at a later date.
This should include precedures effective at specified times and QA records showing that these procedures were followed.
n.
Records.of reviews performed for changes to the OFFSITE DOSE CALCULATION MANUAL and PROCESS CONTROL PROGRAM.
i l
FERMI-- UNIT 2 6-22 Amendment No. JJ, E/,
--=
-6DMjNISTRATIVECONTROLS-
!3 V
HIGH RADIATION AREA (Continued) a.
A radiation monitoring device which continuously indicates the radiation dose rate in the area, b.
A radiation monitoring device which continuously integrates the radiation dose rate in the area and alarms when a preset integrated dose is received.
Entry into such areas with this monitoring device may be made after the dose rate levels in the area have been established and personnel have been made knowledgeable of them.
c.
A health physics qualified individual (i.e., qualified in radiation protection procedures) with a radiation dose rate monitoring device who is responsible for providing positive control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified by the Health Physicist in the RWP.
6.12.2 In addition to the requirements of Specification 6.12.1, areas accessible to personnel with radiation levels such that a major portion of the body could receive in I hour a dose greater than 1000 mrems shall be provided with locked doors to prevent unauthorized entry, and the keys shall be maintained under the administrative control of the Nuclear Shift Supervisor on duty and/ or the health physic supervision. Doors shall remain locked except during periods of access by personnel under an approved RWP which shall
]'
specify the dose rate levels in the immediate work area and the maximum allowable stay time for individuals in that area.
For individual areas accessible to personnel with radiation levels such that a major portion of the body could receive in I hour a dose in excess of 1000 mrems* that are located within large areas, such as the containment, where no enclosure exists for purposes-of locking, and no enclosure can be reasonably constructed around the individual-areas, then that area shall be roped off, conspicuously posted, and a flashing-light shall be activated as a warning device.
In lieu of the stay time specification of the RWP, continuous surveillance, direct or remoto (such as use of closed circuit IV cameras) may be made by personnel qualified in radiation-protection procedures to provide positive exposure control over the activities within the area.
6.13 PR9 CESS CONTROL PROGRAM (PCP) 6.13.1 The PCP shall be approved by the Commission prior to implementation.
6.13.2 Changes to the PCP:
a.
Shall be documented and records of reviews performed shall be retained as required by Specification 6.10.3n.
This documentation shall contain:
{
' Measurement made at 18 inches from source of radioactivity.
FERM1 - UNIT 2 6-23 Amendment No. JJ,
I ADMINIS1RAliVE CONTROL 5 O
u3 eR0cfSS mnR0ummumunnunum 1)
Sufficient information to support the change together with j
the appropriate analyses or evaluations justifying the change (s) and 2)
A determination that the change will maintain the overall conformance of the solidified waste product to existing requirements of federal, State, or other applicable regulations.
b.
Shall become effective after review and acceptance by the OSR0 and the approval of the Plant Manager.
6.14 OffSITE DOSE CALCVM110N MANUAL (00CM) 6.14.1 The ODCM shall be approved by the Commission prior to implementation.
6.14.2 Changes to the ODCH:
a.
Shall be documented and records of reviews performed shall be retained as required by Specification 6.10.3n.
This documentation shall contain:
1)
Sufficient information to support the change together with the appropriate analyses or evaluations justifying the Q
changes (s) and 2)
A determination that the change will maintain the level of radioactive effluent control required by 10 CfR 20.106, 40 CFR Part 190, 10 CFR 50.36a, and Appendix I to 10 CFR Part 50 and not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations.
b.
Shall become effective after review and acceptance by the OSR0 and the approval of the Plant Manager, c.
Shall be submitted to the Commission in the form of a complete, legible copy of the entire ODCM as part of or concurrent with the l
Semiannual Radioactive Effluent Release Report for the period of the report in which any change to the ODCM was made.
Each change shall be identified by markings in the margin of the affected pages, clearly indicating tiie area of the page that was changed, and shall indicate the date (e.g., month / year) the change was implemented.
O FERMI - UNIT 2 6-24 Amendment No. JJ,
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ADMINISTRATIVE CONTROLS
O Tills PAGE HAS BEEN INTENT 10NAllY DELETED O
O FERMI - UNIT 2 6-25 Amendment No. JJ,
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DiCLOSURE 6 DRAFT OFFSITE DOSE CALCULATION MANUAL O
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ML20086G130

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