ML20012C764

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Amends 82 & 74 to Licenses NPF-2 & NPF-8,respectively, Changing Tech Specs to Incorporate Minor Administrative & Editorial Changes in Listed Areas
ML20012C764
Person / Time
Site: Farley  
Issue date: 03/07/1990
From: Adensam E
Office of Nuclear Reactor Regulation
To:
Shared Package
ML20012C765 List:
References
NUDOCS 9003230190
Download: ML20012C764 (25)
v • d • e


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NUCLEAR REGULATORY COMMISSION

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ALABAMA POWER COMPANY DOCKET NO. 50-348 l

JOSEPH M. FARLEY NUCLEAR PLANT. UNIT 1 z

AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 82 i

License No. hPF-2 1.

The Nuclear Regulatory Comission (the Commission) has found that:

1 A.

The application for amendment by Alabama Power Company (the 1

licensee),datedMarch 20, 1989, as supplemented September 25, 1989, complies with the standards and requirements of the Atomic 1

Energy Act of 1954, es amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Comission; C.

There is reasonable assurance-(i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

The issuance of this license amendment will not be inimical to the comon defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable iequirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifications, as indicated in the attachment to this license amendment; andparagraph2.C.(2)ofFacilityOperatingLicenseNo.NPF-2ishereby

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amended to read as follows:

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s (2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 82, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license arendnent is effective as of its date of issuance and shall

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be inpleiented within 30 days of receipt of the amendment.

FOR THE NUCLEAR REGlLATORY COMMISSION R. Lo for e

I Elinor G. Acensem, Director -

Project Directorate 111 Division of Reactcr Projects. I/II Office of Nuclear P, ear. tor Regulation

Attachment:

Lhances to the Technical E

Specifications P

Date of Issuance: March 7,1990 s

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, ATTACHMENT TO LICENSE AMENDMENT NO. 8?

TO FACILITY CPERATING LICENSE NO. NPF-2 DOCKET NO. 50-348 l

Replace the following pages of the Appendix A Technical Specifications with the enclosed pages.

The revised areas are indicated by marginal lines, b

Remove Pages Insert Pages 3/4 2-6 3/4 2-6 3/4 3-18 3/4 3-18 i

3/4 3-28 3/4 3-28 3/4 3-45 3/4 3-45 3/4 6-19 3/4 6-19 3/4 11-19 3/4 11-19 6-15a 6-15a i

6-19 6-19 6-20 6-20 B 3/4 6-4 8 3/4 6-4 s

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POWER DISTRIBUTION LIMITS I

SURVEILLANCE REQUIREMENTS (Continued) 3.........

I When the F is less than or equal to the FRTP l 2.

appropriatEmeasuredcoreplane,addignalp6ve[imitforthe distribution maps shall be taken and F"#* compared to F"#

and F"Y at least once per 31 EFPD.

The F, lanes containing bank"D" control r6ds and all unrodded coreRTP) sha limit for RATED THERMAL POVER (F e.

core p planes in a Radial Peaking Factor Limit Report per Specification 6.9.1.11.

f.

The F,fegions as measured in percent of core height from the bottom of limits of e, above, are not applicable in the following core plane l

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the fuel 1.

Lower core region from 0 to 15%, inclusive.

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2.

Upper core region from 85 to 100%, inclusive.

3.

Crid plane regions within + 2% of core height around the midpoint i

j of the grids.

I 6

Care riane regions vithin + 2% of core brei ht (+ 2.88 inches) abcrt F

the bank demand position of the tank "D" control ^;ods.

With F exceeding F ' the ei het.n of F on P evaluated to determin7tf F, (Z) is vithiFits atm(Z) shr.ll be g.

ies.

L 4.2.2.3 -When F (Z) is measured for other than F determinations, an overall measured F, (Z),shall be obtained from a power di5lribution map and increased by 3% to account for manufacturing-tolerances and further increased by 5% to account for measurement uncertainty.

l 5

FARLEY-UNIT.1 3/4 2-6 AMENDMENT NO. S/,82

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m y;, y =- j k ie. Q c-TABLE 3,3-3 (Continued)' z ] ENGINEERED SAFETT FEATIME ACTUATION SYSTEM-INSTRUNENTATION MINIMUM- -TOTAL NO. CHAN*IELS CHAtelELS - APPLICABLE' FUNCTIONAL UNIT ~ OF CHIANE! S TO TRIP OPERABLE' MODES ACTION 3. CONTAINMENT IS01ATION a. Phase *A" Isolation

1) Manual 2

1 2 1,2,3,4-18

2) From Safety Injection 2

1 2 1,2,3,4-13 w1 Automatic Actuation Logic wLm b. Phase "B" Isolation

1) Manual 2

1 2 1,2,3,4 18

2) Automatic 2

1 2 1 2,3,4' 13 Actuation Logic

3) Containment Pressure 4

2 3 ' 1, 2, 3 ' 16 High-High-High c. Purge and Exhaust g Isolation m54

1) Manual 2

1-2 - 1, 2, 3, 4 17 'I

2) Automatic 2

1 2 1,2,3,4 .17 l zP Actuation Logic 5 h

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  • f TABLE 3.3-4 ' (constir.ued) y ENGINEERED SAFETT FEATURE ACTUATION SISTDI INSTRUMENTATION TRIP SETPOINTS w

c",', FUNCTIONAL UNIT TRIP SETPOIMT ALLOVABLE VALUES e i 6. AUXILIARY FEEDWATER c 2 a. Automatic Actuation Logic N.A. N.A. H b. Steam Generator ~ Vater Level-low-low > 17% of n:rcov range > 16% of narrow range Instrument span each Instrument span each steam gene ator steam generator c. Undervoltage - RCP > 2680 volts > 2640 volts d. S.I. See 1 alx,ve (all SI Setpcints) w e. Trip of Main Feedvater ?omps N.h. N.A. 7. LOSS OF POWER Y a. 4.16 kv Emergency Bus Undervoltage ~> 32S5 volis bus voltage * > 3222 volts bus voltage * (Loss of Voltage) $ 3418 volts bus voltage

  • b.

4.16 kv Emergency Bus Undervoltage > 3675 volts bus ~oltage* > 3638 volts bus voltage

  • I (Degraded Voltage)

$ 3749 volts bus voltage

  • 8.

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INTERLOCKS a. Pressurizer Pressure, P-11 $ 2000 psig $ 2010 psig b. Low-Low T , P-12 (Increasing) 544[F $546[F 1 (Decreasing) 543 F > 541 F s { c. Steam Generator Level, P-14 (See 5. above) z d. Reactor Trip, P-4 N.A. N.A. h

  • Refer to appropriate relay setting sheet for calibration requiraments.

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.w m Ng 4 7_t._BLE '_4. 3-4 5 [ SEISMIC r!9NITORING INSiRUMEhTi. TION SURVEILLANCE REQUIREMENTS CHANNEL CHANNEL CHANNEL FUNCTIONAL INSTRUMENT CHANNEL CHECK CALIBRATION TEST 1. STRONG MOTION TRIAXIAL ACCELER0 GRAPHS a. lA, IB, 1C M R SA b. 2,3 M R SA w 2. PEAK RECORDING ACCELER0 GRAPES D w a. 5,6,7,8, & 9 H.A. R* N.A. L b. 10 & 11 N.A. N.A. 3. PEAK DEFLECTION RECORDER". a. 4A N.A. R SA P-This unit is a mechanical device and should be opened to perform a visual inspection as outlined 5 in the instruction manual. If, on development of its record, the unit has been shocked in excess [ g of its measurement range,. the unit should be removed and recalibrated. M These units are in high radiation a Mas with difficult access and are removed for recalibration no later than every fourth refueli.,g outage or af ter a seismic event, whichever occurs first. 2 h

h 3 CONTAINMENT SYSTEMS 3/4.6.4 COMBUSTIBLE CAS CONTROL HYDROGEN ANALYZERS 4 9' LIMITING CONDITION FOR OPERATION 3.6.4.1 Two independent containment hydrogen analyzers shall be OPERABLE. APPLICABILITY: MODES 1 and 2. i i ACTION: a. With one hydrogen analyzer. inoperable: l i) restore the inoperable analycer to OPERABLS status within 30 days or be in at least HOT STANDEY vithin the next 6 hours, or ?

11) establish en alternate hydrogen sampling capability, l

b. V!.th botn hydrogen ane.lyztrs inoperable, restore at least one analyzar to 9PERh4LE :tttus within 72 hours or be in at least HOT STANDBY vithin the next 6 hoars. SURVEILLANCE REQUIRE lfFKTS t 3................ 4.6.4.1 ?.ach hydrogen r.nalygtr shall be.lemonst reted OPERABLE at itna; ence pcr 92 days on a STIGGEKED TEFT BA61S by performing a CHANN2L CALIBRf.TI(N using sangle rases containing: a. Ten volume percent hydrogen, balance nitrogen, for rero check. b. Ten volus.e percent hydrogen, balance nitrogen, mixed with compressed air, for span check. s l I l l FARLEY-UNIT 1 3/4 6-19 AMENDMENT NO. (/,82 1 l

4- + 3 s e: . ADMINISTRATIVE CONTROLS 4 6.9 RFPORTING 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 Commission, i pursuant to 10CFR50.4. i 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 niay have significantly altered the nuclear, thermal, or hydraulic performance of the plant. i .h = P F ) 4 s i en t 1 l 8 FARLEY-UNIT 1 6-15a AMENDMENT NO.-5/, 82 ,) 5

l p ADMINISTRATIVE CONTROLS i - Type of container (e.g., LSA Type A, Type B, Large Quantity), and e. f. Solidification agent (e.g., cement, urea formaldehyde). The radioactive effluent release reports shall include unplanned releases from the site to unrestricted areas of radioactive materials in gaseous and liquid effluents on a quarterly basis. The radioactive effluent release reports shall include any changes to the PROCESS CONTROL PROGRAM (PCP) made during the reporting period. i MONTHLY OPERATING REPORT 1 6.9.1.10 Routine reports of operating statistics and shutdown experience, including documentation of all challenges to the PORV's or safety valves, shall be submitted on a monthly basis to the Commission, pursuant to 10CFR50.4, l no later than the 15th of each month following the calendar month covered by the report. Any changes to the OFFSITE DOSE CALCL'LATION MANUAL nhall be submit ted with the Monthly operacing Repcrt within 90 days in which the change (s) was made effective. In e6dition, a tiport of any major changes to the radioactive vaste treatment systems shall be submitted with tne Monthly Ope:ating Report for the period in which the change var implemented. BADIAf FEAKING FACTOR LIMIT RFPOPt 6.9.1.11 The P limit for Rated Thermal Pover ($TC) shall be provided to the Commisslun,pur$,[rsnt to 10CFR50.4, for all ccre plafms containing bank "D" l t control rods and all unrodded core planes at least 60 days prior to cycle initlai criticality. In the event thu the limit vould be submitted at sorte other time during cote life, it vill be robmitted 60 days prior to the date the limit vould become effective unless othervise exempted by the commissisn. Any information needed to support F P will be by requtst frvm the NRC and need { not be included in this report. ANNUAL DIESEL GENERATOR RELIABILITY DATA REPORT 6.9.1.12 The number of tests (valid or invalid) and the number of failures to start on demand for each diesel generator shall be submitted to the NRC annually. This report shall contain the information identified in Regulatory Position C.3.b of NRC Regulatory Guide 1.108, Revision 1, 1977. FARLEY-UNIT 1 6-19 AMENDMENT NO. l E/, /0, 82 1

4. .c ADMINISTRATIVE CONTROLS ANNUAL REACTOR COOLANT SYSTEM SPECIFIC ACTIVITY REPORT 6.9.1.13 This annual report is only required when the results of specific activity analyses of the primary coolant have exceeded the limits of Specification 3.4.9 during the year. The following information shall be included: (1) Reactor power history starting 48 hours prior to the first sample in which the limit was exceeded (in graphic and tabular format); (2) Results of the last isotopic analysis for radioiodine performed prior to exceeding the limit, results of analysis while limit was exceeded and results of one analysis after the radioiodine activity was reduced to less than the limit. Each result i i-should include date and time of sampling and the radiciodine concentrations; (3) Clean-up flov history starting 48 hours prior to the first sample in which the limit was exceeded; (4) Graph of the I-131 concentration (micro Ci/gm) and one other radiofodine isotope concentration (micro C1/gm) as a function of time for the duration of the specific activity above the steady-state level; and (5) The time duration when the specific activity of the primary coolant exceeded the radioiodine limit. ANNUAL SEALED SOURCE LEAKAGE REPORT 6.9.1.14 A report shall be prepared and submitted to the Commission on an annual basis if sealed source or fission detector leakage tests reveal the presence of groater than or equal to 0.005 microcuries of removable contamination. SPECIAL REPORTS 6.0.2 Special reports shall be submitted to the com:aission in accordance with the requirements of 10CFR50.4 within the time pertod specified for each report. Reports should be submitted to the U. S. Nuclear Regelatory Commission, ATIN: Documtnt Control Desk, VashinFton, D.C. 20555, 6.10 RECORD RETENTION In addition to th? applicable ief ord retention requirements of Title 10, Code of Federal hegulatio.is, the fcilosing records shall be retained for at least the minimca raried indicated. 6.10.1 The following records shall be retained for at least five years: Records and logs of unit operation covering time interval at each power a. level. b. Records and logs of principal maintenance activities, inspections, repair and replacement of principal items of equipment related to nuclear safety.e c. ALL REPORTABLE EVENTS submitted to the Commission. d. Records of surveillance activities, inspections and calibrations required by these Technical Specifications. Records of changes made to the procedures required by Specification e. 6.8.1. f. Records of radioactive shipments. g. Records of sealed source and fission detector leak tests and results. FARLEY-UNIT 1 6-20 AMENDMENT NO. 51, 82 i

I CONTAINMENT SYSTEMS BASES 3/4.6.3 CONTAINMENT ISOLATION VALVES The OPERABILITY of the containment isolation valves ensures that the containment atmosphere vill be isolated from the outside environment in the event of a release of radioactive material to the containment atmosphere or pressurization of the containment. Containment isolation within the time limits specified ensures that the release of radioactive material to the environment vill be consistent with the assumptions used in the analyses for a LOCA. 3/4.6.4 COMBUSTIBLE GAS CONTROL The OPERABILITY of the equipment and systems required for the detection and control of hydrogen gas ensures that this equipment vill be available to maintain the hydrogen concentration within containment below its flammable limit during post-LOCA conditions. The containment atmosphere post-necident sampling system can be used as an alternative to a hydrogen analy=er should a hydrogen analyzer become inoperable. Either recombiner unit (or the purga system) 1s capable ot' controlling the expected hydrogeri generation associated with 1) zirconium vatrr reactions, 2) radiolytic decorpor.ition of vater and 3) corrostoa of metals within contalament. Thesc-hydroge1 t control systems are consistent with the recommendations of Regulatory Golde 1.7, " Control of Comb 2stible. Gas Concentiatf o'ss it. Ocntwiement Fe]Iovir.g si LOCA,' March 1971. The hydrogen mixin2 systems are provided to ensure adequate mixing of th-containment atmosphere following a LOCA. This mixing action vill prevent { ' localized accumulations of hydrogea from exceeding the flammable limit. t FARLEY - UNIT 1 B 3/4 6-4 AMENDMENT NO.

16. 82 l

L

's b 'J J UNITED STATES ' E ", c., i NUCLEAR REGULATORY COMMISSION 5 E a! WASHINGTON, D. C. 20555 t o %.... / 4 ALABAMA POWER COMPANY DOCKET NO. 50-364 JOSEPH M. FARLEY NUCLEAR PLANT, UNIT 2 AMENDMENT TO FACILITY OPERATING LICElGE Amendment No. 74 License No. NPF-8 i 1. The Nuclear Regulatory Commission (the Commission) has found that: A. The application for amendment by Alabama Power Company (the licensee),datedMarch 20, 1989, as supplemented September 25, 1989, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules 6cd regulations set fortn it 10 CFR Chapter I; B. The facility 9111 oparate in conformity with the apolicction the provis1 orts of the Act., and the rulv.s and regrit.".ionk of the Commitston; l C. l'hereisrcasenableassurance(1)tht.ttheactivitiesauthorizedby { this a:nendnent can bc conducted wf thout endengering the health and scfety of the public, and (ii) that such activities will be conducted in corapliance with the CommissioCs regulations; D. The issuanc.+ of this license amendment will not te inimical to the comon defanst hnd security or to the health and safety of the public; and E. The issuance of this amendment is in accordtnce with 10 CFR Part 51 of the Ccmission's regulations and all applicable reouirements have been satisfied. 2. Accordingly, the license is amended by changes to the Technical Specifications, as indicated in the attachment to this license amendment; and paragraph 2.C.(2) of Facility Operating License No. NPF-8 is hereby amended to read as follows: 1 4 u j

c_ e g 2 r (2) Technical Specifications i i The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 74, are hereby incorpsrated in the i license. Alabama Power Company shall operate the facility in r accordance with the Technical Specifications. 3. This license amendnent is effective as of its date of issuance and shall be implenented within 30 days of receipt of the amendment. FOR THE NUCLEAR REGULATORY COMMISSION R. Lo for Elinor G. Adensam, Director Project Directorate 11 1 Divisien of Reactor Projects. 1/11 Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specificatiens Date of issuance: March T. 1990 f a h I

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f e ATTAChYENT TO LICENSE AMENDMENT NO.74 TO FACILITY OPERATING LICENSE NO. NPF-8 DOCKET NO. 50 364 Replace the following pages of the Appendix A Technical Specifications with the enclosed pages. The revised areas are indicated by marginal lines. Remove Pages Insert Pages I 3/4 2-6 3/4 2-6 3/4 3-18 3/4 3-18 3/4 3-18 3/4 3-28 3/4 3-45 3/4 3-45 3/4 6-19 3/4 6-19 3/4 11-19 3/4 11-19 6-15a 6-15a 6-19 6-19 6-20 6-20 B 3/4 6-4 8 3/4 6-4 t h t I e l I

, a .g POVER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued) When the F

  • is less than or equal to the FRTP 2.

appropriate *measuredcoreplane,addignalpEve[limitforthe distribution maps shall be taken and F*** compared to F** and F** at least once per 31 EFPD. e. The F limit for RATED THERMAL POVER (FRTP) shall be provided for all core planes containing bank "D" control reds and all unrodded core planes in a Radial Peaking Factor Limit Report per Specification 6.9.1.11. f. The F,fegions as measured in percent of core height from the bottom of limits of e, above, are not applicable in the following core plane l the fuels 1. Lover core region from 0 to 15%, inclusive. 2. Upper core region from 85 to 100%, inclusive. 3. Grid plane regions within : 2% of core height around the midpoint of the grids. 4 Core sinne regions within + 2% of cort height (., 2.!)C inches) about the bank demand position el the bar.k "D" control.ods. E. With F

  • cxceeding P 'theeffectsofF"itsl$m(its.

on F

2) shall be evaluated to de: ermine *it F,-(Z) is within 4.2.2.3 Vhen F (Z) is measured for other than F determinations, an overall measured F, (Z) shall be obtained from a power distribution cap and increased by 3% to accour,t fer manufacturing colerances end further increased by 5% to account 2or measuremr:nt uncertainty.

I FARLEY-UNIT 2 3/4 2-6 AMENDMENT N0. O, 74

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3, =. = - - = ~ -. ~ ~ -.. :.. t z. 3 ~ p Q-l TABLE 3.h-3 (Continued) 9 ' E-. INGINEERED SAFE 7Y FEATURg aC.TUATION SYSTEM INSTRUMENTATION ' ~ to MINIMUM TOTAL No.- CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT GF CHA*INFLS-TO TRIP OPERABLE MODES - ACTION 3. CONTAINMENT ISOLATION a. -Phase "A" Isolation

1) Manual

'2 1 2 1, 2, 3, 4 18

2) From Safety Jajection 2

1 2 1,2,3,4 13 Automatic Actuation Logic b. Phase "B" Isolation Y

1) Manual 2'

1 2 1,2,3,4 18-

2) Automatic 2

-1 2 1,2,3,4 13 Actuatian Logic

3) Containment Fressure 4-2 3

1,2,3 16 High-High-High c. Purge and Exhaust E Isolation mz E '1) Manual 2 1 2 1,2,3,4 17 z*

2) Automatic 2

1 2 1, 2, 3,-4 17- -l ,5 Actuation Logic E ... - -. --. _.: :.. n.,

p;c, w -: p , w w .3 : +c, . TABLE 3.3-4 (Continued) s.,_ ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP'SETPOINTS ~~

c b

FUNCTIONAL UNIT TRIP SETPOINT ALLOVABLE VALUES 8 6. AUXILIARY FEEDUATER C j. a. Automatic Actuation Logic N.A. N.A. c4 b. Steam Generator w Vater Level-low-lov > 17% of narrow range > 16% of narrov range Instrument span each instrument span each steam generator steam generator c. Undervoltage - RCP > 2680 volts' > 2640 volts d. S.I. See 1 above (all SI Setpoints) e. Trip of-Main Feedvater Pumps N.A. N.A. 7. LOSS OF POVER a. 4.16 kv Emergency Bus Undervoltage > 3255 volts bus voltage * > 3222 volts bus voltage * (Loss of Voltage) $ 3418 volts bus voltage

  • b.

4.16 kv Emergency Bus Undervoltage > 3675 volts bus voltage * > 3638 volts bus voltage * (Degraded Voltage) $ 3749 volts bus voltage

  • 8.

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INTERLOCKS a. Pressurizer Pressure, P-11 $ 2000 psig $ 2010 psig e S b. Low-Lov T , P-12 (Increasing)- 544,F $ 546,F (Decreasing) 543 F s -> 541 F o. f c. Steam Generator Level,'P-14 (See 5. above) d. Reactor Trip, P-4 N.A. N.A.

  • Refer to appropriate relay setting sheet for calibration requirements.

l

.A a .;~..- g TABLE 4.3-4 0 l 5 SEISMIC MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMEffrS O CHANNEL CHANNEL CTIANNEL FUNCTIONAL INSTRUMENT CHANNEL CHECK CALIBRATION TEST 1. STRONG MOTION TRIAXIAL ACCELEROGRAPHS a. lAl,1BI,IC' M R SA l b. 2,3 M R SA 2. PEAK RECORDING ACCELEROGRAPHS a. 56,6,7,8, & 9# N.A. R* N.A. l b. 108 & lit N.A. N.A. - l w E v, 3. PEAK DE?LECTION RECORDERS a. 4At N.A. R 'SA l This unit is a mechanical device and should be opened to perform a visual inspection as outlined x in the instruction manual. If, on development of its record,-the unit has been shocked in excess l E of its measurement range, the unit should be removed and recalibrated. e These units are in high radiation areas with difficult access and are removed for recalibration l no later than every fourth refueling outage or after a seismic event, whichever. occurs first. l Sensors located in Farley Unit 1 l E .-n . =

7. ~ e  %~ b s. L ..L" o CONTAINMENT SYSTEMS 3/4.6.4 COMBUSTIBLE GAS CONTROL F . HYDROGEN ANALYZERS LIMITING CONDITION FOR OPERATION 3.6.4.1-Two independent containment hydrogen analyzers shall be OPERABLE. APPLICABILITY: HODES 1 and 2. ACTION: With one hydrogen analyzer inoperable: l a. i) restore the inoperable analyzer to OPERABLE status within 30 days or be in at least HOT STANDBY vithin the next 6 hours, or

11) establish an alternate hydrogen sampling capability l

L b. With both hydrogen analyzers inoperable, restore at least one analyzer to-OPERABLE status within 72 hours or be in at least HOT STANDBY vithin the next 6 hours. SURVEILLANCE REQUIREMENTS i - 4.6.4.1 Each hydrogen. analyzer shall be demonstrated OPERABLE at least once per 92 days on a STAGGERED TEST BASIS by performing a CHANNEL CALIBRATION.using sample gases containing a. Ten volume percent hydrogen, balance nitrogen, for zero check. b. Ten volume percent hydrogen, balance nitrogen, mixed with compressed i air, for span check. l l i i L l FARLEY-UNIT 2 3/4 6-19 AMENDMENT NO. ES, 74 1 li l l~ g

~ l a n. .p .~ a ' i RADIOACTIVE EFFLUENTS 3/4.11.4 TOTAL DOSE .L1HITING CONDITION FOR OPERATION 3.11.4 The dose or dose commitment to any member of the public, due to releases of radioactivity and radiation, from uranium fuel cycle sources shall be limited to less than or equal to 25 mrem to the total body or any organ (except the thyroid, which shall be-limited to less than or equal to 75 mrem) over 4 i consecutive quarters. APPLICABILITY: At all times. -) ACTION: a. Vith the calculated doses from the release of radioactive materials in liquid or gaseous effluents exceeding twice the limits of Specification 3.11.1.2.a. 3.11.1.2.b, 3.11.2.2.a, 3.11.2.2.b, 3.11.2.3.a or 3.11.2.3.b, prepare and submit a Special Report to the Commission, pursuant to Specification 6.9.2, within 30 days, which defines the corrective action to be taken to reduce subsequent releases to prevent recurrence of exceeding the limits of Specification 3.11.4. This Special Report shall include an analysis which estimates the radiation exposure (dose) to a member of the public from uranium fuel cycle sources (including all effluent pathways and direct radiation) for a 4 consecutive quarter period that includes the release (s) covered by this report. If the estimated dose (s) exceeds the limits of Specification 3.11.4, and if the release condition resulting in violation of 40CFR190 has not already been corrected, the Special Report shall include a request for a variance in accordance with the-provisions of 40CFR190 and including the specified information of $ 190.11(b). Submittal of the report is considered a timely request, and a variance is granted until staff action on the request is complete. 'The variance only (' relates to the limits of 40CFR190, and does not apply in any way to the 4 i requirements for dose limitation of 10CFR Part 20, as addressed in l other sections of this technical specification. 'b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. SURVEILLANCE REQUIREMENTS 4.11.4 Dose Calculations Cumulative dose contributions from liquid and gaseous effluents shall be determined in accordance with Specifications 4.11.1.2, 4.11.2.2, and 4.11.2.3, and in accordance with the ODCH. l FARLEY-UNIT 2 3/4 11-19 AMENDHENT NO. (9, 74 L

V }.I .Ns 6 29 9 ADMINISTRATIVE CONTROLS r 6.9 REPORTING REQUIREMENTS H 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 Commission, pursuant to 10CFR50.4. 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 plant. 4 ? 0 FARLEY-UNIT 2 6-15a AMENUMENT NO. $7, 74

-a. M. _,ge ADMINISTRATIVE CONTROLS Type of container (e.g., LSA Type A, Type B, Large Quantity), and e. f. Solidification agent (e.g., cement, urea formaldehyde). Thr. radioactive effluent release reports shall include unplanned releases from the site to unrestricted areas of radioactive materials in gaseous and liquid effluents on a quarterly basis. The radioactive effluent release reports shall include any changes to the PROCESS CONTROL PROGRAM (PCP) made during the reporting period. HONTHLY.0PERATING REPORT 6.9.1.10 Routine reports of operating statistics and shutdown experience, including documentation of all challenges to the PORV's or safety valves, shall be submitted on a monthly-basis to the Commission, pursuant to 10CFR50.4, l l no later than the 15th of each month following the calendar month covered by the report. 7 Any changes to the OFFSITE DOSE CALCULATION HANUAL shall be submitted with the Monthly Operating Report within 90 days in which the change (s) was made effective. In addition, a report of any major changes to the radioactive vaste treatment systems shall be submitted with the Monthly Operating Report for the period in which the change was implemented. RADIAL PEAKING FACTOR LIMIT REPORT 6.9.'l.11 The F limit for Rated Thermal Power (FRTP ) shall be provided to the Commission,-purs,[iant to 10CFR50.4, for all core plafies containing bank "D" l control rods and all unrodded core planes at least 60 days prior to cycle initial criticality. In the event that the limit would be submitted at some other time during core life, it vill be submitted 60 days prior to the date the limit would become effective unless otherwise exempted by the Commission. Any information needed to support FRTP vill be by request from the NRC and need j~ not be included in this report. ANNUAL' DIESEL GENERATOR RELIABILITY DATA REPORT 6.9.1.12 The number of tests (valid or invalid) and the number of failures to ' start on demand for each diesel generator shall be submitted to the NRC annually. This report shall contain the information identified in Regulatory Position C 3.b of NRC Regulatory Guide 1.108, Revision 1, 1977. I l FARLEY-UNIT 2 6-19 AMENDHENT NO. O, $2, 74

k c; r o ADMINISTRATIVE CONTROLS ' ANNUAL REACTOR COOLANT SYSTEM SPECIFIC ACTIVITY REPORT 4 6.9.1.13 This annual report is only required when the results of specific activity analyses of the primary coolant have exceeded the limits of Specification 3.4.9 during the year. The following information shall be included (1) Reactor power history starting 48 hours prior to the first sample in which the limit was exceeded (in graphic and tabular format); (2) Results of the last isotopic analysis for radioiodine performed prior to exceeding the limit, results of analysis while limit was exceeded and results of one analysis after the radiciodine activity was reduced to less than the limit. Each result-( should include date and time of sampling and the radiolodine concentrations; (3)- Clean-up flov history starting 48 hours prior to the first sample in which the limit was exceeded; (4) Graph of the I-131 concentration (micro Ci/gm) and one other radioiodine isotope concentration (micro C1/gm) as a function of time for the duration of the specific activity above the steady-state level; and (5) The time duration when the specific activity of the primary coolant exceeded the radiolodine limit. ANNUAL SEALED SOURCE LEAKAGE REPORT 6.9.1.14 A report shall be prepared and submitted to the Commission on an annual basis if sealed source or fission detector leakage tests reveal the presence of greater than or equal to 0.005 microcuries of removable contamination. SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the Commission in accordance with the requirements of 10CFR50.4 vithin the time period specified for each report. Reports should be submitted to the U. S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Vashington, D.C. 20555. i 6.10 RECORD RETENTION In-addition to the applicable record retention requirements of Title 10, Code of Federal Regulation 7, the following records shall be retained for at least the minimum period indicated. 6.10.1 The following records shall be retained for at least five years: .a. Records and logs of unit operation covering time interval at'each power l

level, b.

Records and logs of principal maintenance activities, inspections, repair and replacement of principal items of equipment related to nuclear safety.8 c. ALL REPORTABLE EVENTS submitted to the Commission. L d. Records of surveillance activities, inspections and calibrations i required by these Technical Specifications. I Records of changes made to the procedures required by Specification e. 6.8.1. .f. Records of radioactive shipments. g. Records of sealed source and fission detector leak tests and results. FARLEY-UNIT 2 6-20 AMENDHENT NO. 99, 74

.e . oe CONTAINMENT SYSTEMS BASES 3/4.6.3 CONTAINMENT ISOLATION VALVES The OPERABILITY of the containment isolation valves ensures that the j containment atmosphere vill be isolated from the outside environment in the i event of a release of radioactive material to the containment atmosphere or pressurization of the containment. Containment isolation within the time limits specified ensures that the release of radioactive ~ material to the environment vill be consistent with the assumptions used in the analyses i for a LOCA. i 3/4.6.4 COMBUSTIBLE GAS CONTROL The OPERABILITY of the equipment and systems required for the detection and control of hydrogen gas ensures that this equipment will be available to maintain the hydrogen concentration within containment belov its flammable limit during post-LOCA conditions. The containment atmosphere post-accident sampling system can be used as an alternative to a hydrogen analyzer should a hydrogen analyzer become inoperable. Either recombiner unit (or the purge system) is capable of controlling the expected hydrogen generation associated with 1) zirconium-vater reactions, 2) radiolytic decomposition of water and 3) corrocion of metals within containment. These hydrogen control systems are consistent with the recommendations of Regulatory Guide 1.7, " Control of Combustible Gas Contentrations in Containment Following a LOCA," March 1971. The hydrogen mixing systems are provided to ensure adequate mixing of the containment atmosphere following a LOCA. This mixing action vill prevent localized accumulations of hydrogen from exceeding the flammable limit. i FARLEY - UNIT 2 B 3/4 6-4 AMENDMENT NO. 74}}

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