ML20097B188

Proposed TS Deleting Sections 2.9.1 Re Liquid & Gaseous Effluents & 2.9.2 Re Solid Radwaste

ML20097B188
Person / Time
Site:	Fort Calhoun
Issue date:	06/01/1992
From:	OMAHA PUBLIC POWER DISTRICT
To:
Shared Package
ML20097B178	List: ML20097B176 ML20097B181 ML20097B188 ML20097B192 ML20097B202
References
NUDOCS 9206040193
Download: ML20097B188 (102)

Text

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TECHNICAL SPECIFICATIONS '

TABLE OF CONTENTS

_? ace DEFINITIONS............................................................ 1 1.0 SAFETY LIMITS RID LIMITING SAFETY SYSTEM SETTINGS. . . . . . . . . . . . . . . . 1-1 l 1.1 S afe ty Limit s - Re ac to r Co re . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1- 1 1.2 Safe ty Limit , Reactor Coolant System Press ure . . . . . . . . . . . . . . . 1L 1.3 Limiting Safety System Settings , Reactor ProtectiVO Sy3 tem.................................. .....1-6 2.0 L D1IT I!iG CO N DITION S FO R O P ERATIO N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2- 3 2.0.1 General Requirements................. .............. 2-0 2.1 Re ac to r Coolant Sys t em. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2- 1 2.1.1, Operable Components . . . . . . . . ,4 .......................2-1 2.1.2 He atup an d Cooldo' n Rate . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2- 3 2.1.3 Re acto r Coolant Radioactivity . . . . . . . . . . . . . . . . . . . . . . . 2- 8 2.1.4 Reactor Coolant System Le akage Limits . . . . . . . . . . . . . . . 2-11 2.1.5 Maximum Reactor Coolant Oxygen and Halogens Co n c e n t rat i o ns . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 - 13 2.1.6 ~ Pressurizer and Stes System Safe ty Valves . . . . . . . . . . 2-15

_h 2.1.7 Press uri z e r Ope rab ility . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-16 a 2.1.8 Re ac to r Coolant Sys tem Vents . . . . . . . . . . . . . . . . . . . . . . . . 2- 16b 2.2 Chemical and Voltee Control System. . . . . . . . . . . . . . . . . . . . . . . . . 2-l~

2.3 Ene rEency Co re Cooling Sys tem . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-20 2.L Containment Cooling........................................ 2-2k 2.5 S t e am and Fee dva te r Sys t ems . . . . . . . . . . . . . . . . . . . . . . . . . ...... 2-23 2.6- Conte.inment System......................................... 2-30 2.7 . Electrical Systems......................................... 2-22 2.3 Re f ue lin g O pe rat ions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2- 3 ~

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2 . '.C Reactor Core................................................

2.10.1 Minim = Conditions for Criticality . . . . . . . . . . . . . . . . . 2- 3

- 2.10.2- Reactivity Control Systems and Core Physics

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.m o3................................... cn 2.10.'3 In-Co re Inst rument at ion. . . . . . . . . . . . . . . . . . . . . ....... 2 2.13 . h Fwe r Di st rib ution _ Limi ts . . . . . . . . . . . . . . . . . . . . . . . . . . 2- F 2.11 COntain.cnt 3uilding.and Fuel .3torage 3'illing Jr:ne........ ' 53 i Amendmen. 3o. 32, 33, 32, 34, fT, 57, I u , ;o' i

e na, v 9206040193.920601 PDR ADOCK 05000285 P pop L . _-

TABLE OF CONTENTS (Continued) l pace 2.12 Control Room-Systems.....................................

2.13 2-59 Nuclear Detector Cooling System. . . . . . . . . . . . . . . . . . . . . . . . . . . 60 2.14 Engineered Safety Features System Initiation Instrumentation Settings............................... 2-61 2.15 Instrumentation and Control S 2.16 River Level..................ystems......................

2-65 2-71 2.17 2.18 Miscell aneous Radioactive Material Sources. . . . . . . . . . . . . .

Shock Suppressors 2-72 2.19 Fire Protection (Snubbers).............................

System...................................

2-73 2.20 2-89 Steam Generator Coolant Radioactivity........ ........... 2-96 2.21 Post-Accident Monitoring Instrumentation...

2.22 2-97 Toxic Gas Monitors....................................... ............ 2-99 3.0' SURVEILLANCE REQUIREMENTS...................................... 3-Oa 3.1 3.2 Instrumentation and Control.................. ........ 3-1 3.3 Equipment and Sampling Tests.................. ........ 3-17 Reactor Coolant System and Other Components Subject to ASME XI Boiler and Pressure Vessel Code Inspection and-Testing Surveillance...............................

3.4 3-21 Reactor Coolant System Integrity Testin 3.5 C o n t a i nme n t Te s t . . . . . . . . . . . . . . . . . . . . . . . g . . . . . . . . . . . . . 3. -. 3. 6.

'3.6 ................... 3-37

. Safety Injection and Containment Cooling Systems Tests... 3-54 3.7 Emergency Power System Periodic Tests......... 3-58 3.8 Main Steam. Isolation Valves................ .. .......... ........ . 3-61 L3.9 Auxiliary Feedwater System.............................

3.10 Reactor Core 3 62 Parameters...................................

3.11 RediMcgic:1 Emircement:1 Meriteing orogramss

.. 3-63 3.12 R' d i ol og i c4M& S t0 5 2mpl i ng 3d-McWOP4ng . . . . . . . . . . . . . . .

. . 3-64 W _ M dAttioA.c.Twe %=iTe DiS potM SyJTewt 3 -6 9 3.12.1 Liquid Gnd-Ca3aCO3--Effluent 3................

3.-12.2 ... 3-69 SCl i d - S &d404014Ve48tt c . . . . . . . . . . . . . . . . . . . . - 3 4t-3.13 3.14 Radioactive Material Sources Surveillance..... ... ......

Shock Suppressors 3-76 3.15 (Snubbers)...................... ....

Fire Protection System................................... 3-77 3.16 ... 3-80 Residual Heat Removal _ System Inte . 3-84 3.17 Steam Generator Tubes............grity Testing.........

........................ 3-86

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4 . 0 D E S I Gil F E ATU R E S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4. -.1. . . . . . . . . .

4.1 Site.......

4.2 .......... ...... ........................... 4-1 Containment Design Features..................... ........ 4-1

-4.2.1 Containment Structure.......................... 4-1 4.2.2. Penetrations...... ..... ...... ............... 4-1 l 4.2.3 Containment Structure Cooling Systems. . . . . . . . . 4-2 i

i ii Amendment No. E/,25,93,JDA,J22, 136

TABLE OF CONTENTS (Contmued)

Butt 4.3 - Nuclear Steam Supply System (NSSS) , . . . . . . . . . . . , . . . . . . . . . . . . . . . . . . . 4 3 4.3.1 Reactor Coolant System . . . . . . . . . . . . . . . . . . . . . . .. . .. . .... .. 4-3 4.3.2 Reactor Core and Control . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-3 4.3.3 Emergency Core Cooling . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-3 4.4 Fuet Storage ..................................... ........ 4-4 4.4.1 N ew Fue t S tara ge . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 -4 4.4.2 Spent Fuel Storage ........................ .............4-4 4.5 Seismic Design for Class I Systems .......... ......................4-5 5.0 ADMINISTRATIVE CONTROLS 5-1 5.1 Responsibility . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ... .............51 5.2 Organization ................ .... ...... ....... .........51 5.3 Facility Staff Qualifications .................... ................ 5-la 5.4 Trainin g . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ........53 5.5 Review and Audit . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-3 5.5.1 Plant Review Committee (PRC) .. .. . . . . . . . . . .. . . . . . . . . . . .. . . . 5-3 5.5.2 Safety Audit sod Review Committee (SARC) . . . . . . . . . . . . . . . . . . . . . . 5-5 5.5.3 Fire Protection Inspection . . . . ..... .. ... ............... 5-Sa 5.6 Reportable Event Action . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-9 5.7 Safety Limit Violation ........... . ... . .....................59 5.8 Procedures ................................ . .. .. .. ... . ... 5-9 5.9 Reporting Requirements ........ ..............................5-10 5.9.1 Routine Reports . . . .... .............................. 5-10 5.9.2 Reportable Events . . . . . . . . . . . . . . . . . . . . . . . . ............. 5-12 5.9.3 Special Reports ...........,..................... ..... 5-15 5.9.4 Unique Reporting Requirements . . . . . . . . . . . . ........ ....... 5-15 5.9.5 Core Operating Limits Report .............. .. ...... ..... . 5-17a 5.10 Records Retention . . . . . . . . . . . . . . . . . . . . . . .. ...... ......... 5-18 5.11 Radiation Protection Program . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-19 5.12 DELETED 5.13 Secondary Water Chemistry . . . . . . . . . . . . . . . . . . ... ....... . . . . . . 5-20 5.14 Systems Integrity . . . . ..................... .. ............. . 5-21 5.15 Post-Accident Radiological Sampling and Monitoring .....................5-21 4tTachwenT 1-*

6.0 INTERIM SPECIAL TECHNICAL SPECIFICATIONS 1 6.1 Limits on Reactor Coolant Pump Operation ... .............. .. ... . .. . 6-1 6.2 Use of a Spent Fuel Shipping Cask . . . . . . . . . . . .. ............ . .. .. 6-1 6.3 Auxiliary Feedwater Automatic Initiation Setpoint .......... . ......... . 6-1 6.4 Operation With Less Than 75% ofincore Detector Strings Operable ... ......, . ... . .. .. . .... ,. .. . 6-1 iii Amendment No. M,34,43.54.4547 73,80,84,0h90,141 I

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I ATTACHMENT 1 l

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5.16- Radiological ~ Effluents and Environmental Monitoring Programs . . 5-22 l 5.15.1 Radioactive Effluent Controls Program ....... . . 5-22 5.16.2 Radiological Environmental Monitoring Program .... . 5-23 j

5.17 Of fsite Dose Calculation Manual (00CM) . . . . . . . . . . . . 5-25 5.18 Process Control Program {PCP) .................525  ;

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l TABLE OF CONTENTS - TABLES i l

TABLE OF CONTENTS l TABLE DESCRWIION PAGE 3-3 Mirimum Frequencies for Checks, Calibrations, and Testing of Miscellaneous Instrumentation and Controls . . . . . . . . . . . . . . . . . . . . ..... 3 13

..... ........ ........ .. . 3 14

........ .....,............315

. .........................316

................... . . . . . 3-16a

..................... . ... 3-16b

.......................... 316c 3-3 a Minimum Frequency for Checks, Calibrations and Functional Testing of Alterna*.c Rhuswt Panels (Al-185 and AI.212) and Emergency AutLn .ceedwater Panel (Al-179) Instrumentation and Control circuits . . . . . ...................... ..... .... 316d

............ .......... . ..... ..... 3-16e 3-4 Minimum Frequencies for kmpling Tests . . . . . . . . . . . . . . . ........ .... 3 18

.......... .. ..... ..... .. . . . . 3-19 35 Minitru:n Frequencies for Equipment Tests . . . . . . . . . . . .. ............ 3-20

............... ..... . . . . . . . . . 3 20s

......... ..... .. . . . . . . . . . . . 3 -20b

...... . ............. . .... . 3-20c

. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 20d 3-6 Reactor Coolant Pump Surveillance . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-27 30 R: die!ogiea'-Enviroan::

, ' "=itoringProgm . . }-66

........ . ....... ... ~ 7---

H R2dienetive-1:iquid-Weste-Sampling-andenalysh . . .... . ... . r i- .. 3 , ,,

... . .. ... . . .. ... c,,

3-12 R&-':ve Gre=: "lzte Sampling-and-Analysk . . . , . .. 3 ... . . ... ... .... .. .. s-?s, 3-13 Steam Generator Tube Inspection . . . . . . . . . . . . . . ... .. .... . 3-90 5.2-1 Minimum Shift Crew Composition .............. . .. . ......... .5-2 l

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l v Amendment No. H6A25,142 l

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TECHNICAL SPECIFICATIONS - TABLES TABLE OF CONTEtlTS ( AL.P.H.AB.ETICAL O_RDER)

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Cont,inue(-

TABLE pESCRIPTI0tt PAGE 3-5 Minimum Frequencies for Equipment Tests . . . . . . . . . . . 3-20 3-20a 3-20b 3-20c' j 3-20d j 3-4 Ninimum Frequencies for Sampling Tests. . . . . . . . . . . . 3-18 3-19 5.2-1 Minimum Shif t Crew Composition. . . . . . . . . . . . . . . . 5-2 ,

-?-10 Post-Accident !;onitoring Ir.strumentation Operating Limits . .  ?-98 2-98a 2-98b 43-12 C;di:Jctive C;;ccus Waste San 9 4 -ir.g and Analysis . . . . . . .

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11 " dicactive Liquid "aste Sampling 'nd ^nalyrir.

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3-0 ce dib'egical Er"4-cr e-t "critering oregrar . . . . . . . .

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? RCS PressureLisolation Valves . ........... . . .. 2-?d

, 3-6 Reactor _ Coolant Pump Surveillance . . . . . . . . . . . . . . 3-?7 >

l-1 'RPS LSSS. . . . . . . . . . . . . . .-. . . . . . . . . . . . 10 1-10a 3

3-!3 SteamfGenerator Tube Inspection . .............. 3-90 j-

'2-11 Toxic Gas Monitoring Operating Limits . . . . . . . . . . . .  ?-100 1 t

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vii Amendmen.. f10 116 1-. .

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- DEFINITIONS Azimuthal Eower Tilt - T; Azimuthal Power Tilt shall oe the maximum difference between the power generated in any core quadrant (upper or lower) and the average power of all quadrants in that axial half (upper or lower) of the core divided by the average power of all quadrants in that axial half (upper ce lower) of the core.

Unrodded Planer Radia! Peakine Factor - F_.y The Unrodded Planar Radial Peaking Factor is the maximum ratio of the peak to average power density of the individual fuel rods in any of the unrodded horizontal planes, excluding azimuthal tilt, T,. The maximum F,7 limit is provided in the Core Operating Limits Report. l Unrodded Intecrated Radial Peaking Factor - F.

The Unrodded Integrated Radial Peaking Factor is the ratio of the peak pin power to the average pin power in an unrodded core, excluding azim' thal tilt, T,. The maximum Fa limit is provided in the Core Operating Limits Report.

Fire Suopression Water System The fire suppression water system consists of fire pumps and distribution piping with associated sectionalizing control or isolation valves. Such valves include yard hydrant curb valves, and the first valve ahead of the water flow alarm device on each sprinkler, hose standpipe or spray system riser.

-Procest-GontrobProemm-(PGFW

-A-manual-or-set-of-operating-procedures-detailing-the-program-of-samplingr-analysisr,-end--

-evaluatiom AtthchroersT 2

• Dose Equivalent I-131 That concentration ofI-131 (pCi/gtr) which alone would produce the same thyroid dase as the quantity and isotopic mixture of I-131, I-132, I-133, I-134 and I-135 actually present. In other words, 7 Amendment No. 32,38,6h86;141

DEFINITIONS Dose Equivalent I-131 (pCi/gm) = gCi/gm of I-131

+ 0.0361 x pCi/gm of I-132

+ 0.270 x Ci/gm of I-133

+ 0.0169 x pCi/gm of I-134

+ 0.0838 x pCi/gm of I-135  !

$ - Average DM:gration. Energy 1

E is the average (weighted in proportion to the concentration of each radianuclide in the reactor i coolant at the time of sampling) of the sum of the average beta and gamma energies per disintegration, in MEV, for isotopes, other than iodines, with half lives greater than 15 minutes l making up at least 95% of the total non-iodine radioactivity in the coolant.

MtachmenT 3+

Offsite Dc;c CalculatiwManual (ODCMI l

A man taining the' methodology and parameters to be used in the: 1) jeal1 on of doses in the unrestrict due to radioactive liquid and gaseous effluents realculation of liquid and gaseous effluent mottitoring instrumentation setpoints, and 3) ific details pertinent to the radiological environmental mo pro, m.

Purge-Purging A means for the removal eplacement of gases within the atainment building, Ventine-meansfor-the-redstionef-pressure-greater-than-atmospherio-within4huontainment-structureL Core Ooerating Limits Reoort (COLR)

The Core Operating Limits Report (COLR) is a Fort Calhoun Station Unit No I specific document that provides core operating limits for the current operating cycle. These cycle-specific core operating limits shall be determined for each reload cycle in accordance with Section 5.9.5. Plant operation within these operating limits is addressed in the individual specifications.

References i

(1) USAR, Section 7.2 i

! (2) USAR, Section 7.3 l

i 8 Amendment No. 6h86,141 l.

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bac h FY1Nt ~ d Process CQatrol Program (PCP)

The document (s) that contains the current formulas,' sampling, analyses, tests, 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 20, 61, 71, State Regulations, burial ground requirements, and other requirements governing the disposal of solid radioactive waste.

k h clvvleni 1 3 Offsite Dose Calculation Manual (ODCM)

The document (s) that contain the methodology and parameters used in the calculations of offsite

[ doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent radiation monitoring Warn /High (trio) Alarm setpoints, and in the conduct of the Environmental Radiological Monitoring Program. The ODCM shall also contain:

1) . The Radiological Effluent Controls and the Radiological Environmental Monitoring Program required by Specification 5.16. ,

2) . Descriptions of the information that should be included in the Annual Radiological

- E vironmental Operating Report and Semiannual Radioactive Effluent Release Reports required by Speci'ications 5.9.4.a and 5.9.4.b.

Unrestricted Area Any area ator bevond the site boundary access to which is not controlled by the licensee for purposes of protection ofindividuals from exposure to radiation and radioactive materials.

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+L[i 12.0- LIMITING CONDITIONS FOR OPERATION 2.,1 - Reactor Coolant System (Continued) 2.1.3 Reactor Coolant Radioactivity Applicability-Applies to the radioactivity of the reactor coolant. ,

Objective To ensure that the reactor cot ant radioactivity is maintained at a level comnensurate with the occupational and public safety.

Speci fication (1) The radioactivity of the reactor coolant shall be limited to:

a. 1 1.0 uCi/gm DOSE EQUIVALENT I-131, and

b. 1 100/f uCi/gm  ;

(2) With the radioactivity of the reactor coolant >1.0 uCi/cm DOSE i-EQUIVALENT I-131 for more than 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> during one continuous time interval or exceeding 60 uCi/gm, be in at least HOT SHUTDOWN with T avg <536*F withi.n 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

(

(3) With the radioactivity of the reactor coolant > 100/E uci/gm, be i in Lat least HOT . SHUTDOWN with T3yg < 536"F within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

(4) ' With the radioactivity of the reactor coolant >1.0 uCi/gm DOSE I EQUIVALENT I-131, perform the sampling and analysis requirements of items 1.(a)(2)(ii) and 1.(b)(2)(i) of Tabla 3-4 until the radioactivity of the reactor coolant is restored to Within its limits. Data pursuant to Specification S.9.4.bw for the Annual Report shall be compiled as follows:

a. Reactor power history starting 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> prior to the first sample in which the limit was exceeded

b. -Purification System flow history starting 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> prior to the first sample in which the limit was exceeded.

c. The -time duration when the radioactivity of- the reactor coolant exceeded 1.0 uCi/gm DOSE EQUIVALENT I-131.

d. Results of the last isotopic analysis for radiciodine perforned 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 contain the date and tine of sampling and the radiciodine concentrations.

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Amendment No. 23.67, 102 2-8

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2.0 LIMITING CONDITIONS FOR OPERATION f 2.8 Refuelino Ooerations Apolicability-Applies to operating limitations during refueling operations.

Obiective i To minimize the possibility of an accident occurring during refueling operations that could affect public health and safety.

Specifications The following conditions shall be satisfied during any refueling orerations: gecoas nacdvm muTc<

(1) The equipment hatch and one door in the air lock shall be properly closed. In addition, all automa tic containment isolation valves shall be operable or at least one vals e in each line shall be closed.

% Exk,wst Shek (2) -TM)4n containment atmosphere < and hene. &x%y Lw3nt-vent 44et4on-duct r monitorythat initiate c'.1sure of the containment pressure relief, air sample, and purge system valves shall be tested and verified to be operable immediately prior to refueling operations. The f4ge monitors shall employ one-out-of-f+ve logic from separate contact at tputs for

, VIAS. %e -%o t

I' ' -(3) Radiation levels in the containment and spent fuel storage areas shall be monitored continuously.

(4) Whenever core geometry is being changed, neutron flux st,all be continuously monitored by at least two source range neutron monitors, with each monitor providing continuous visual indication in the control room. When core geometry is not being changed. at least one source range neutron monitor shall.be in service.

-(5) - At least one shutdown cooling pump and heat exchanger shall be in operation. However, the pump and heat exchanger may be removed from operation for.up to one hour per 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> period during the performance of core alterations in the vicinity of the reactor coolant hot leg loops or during manipulation of a source.

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2-37 Amendment No. /E,EB ,133 i-L _

2.0 LIMITING CONDITVONS FOR OPERSTIONS 2.8 Refuelina 00erations (Continued)

( (6) Direct communication between personnel in the control room and at the refueling machine shall be available whenever changes in core I

. geometry are taking place.

(7) 'When irradiated fuel is being handled in the auxiliary building, I the exhaust ventilation from the spent fuel pool area will be diverted through the charcoal. filter.

(8) Prior to initial core loading and prior to refueling operations, I a complete check out, including a load test, shall be conducted on fuel handling cranes that will be required during the refueling operation to handle spent fuel assemblies.

(9) A minimum of 23_ feet of water above the top of the core shall be I maintained whenever irradiated fuel is being handled.

(10) Storage in Region 1 and Region 2 of.the spent fuel racks shall be I restricted to fuel assemblies having initial enrichment less than or equal to 4.0 weight percent of U-235.

(11) Storage in Region 2 of the spent fuel racks shall be restricted to I those assemblies whose parameters fall within the " acceptable" region of Figure 2-10.

If any of the above conditions are not met, all refueling operations shall f

cease immediately, work shall be initiated to satisfy the required conditions, and no operations that may change the reactivity of the core shall be made. Hewever, refueling-operations-may-commence-and-cont 4nue with 1ess-than-6-centa4nment-atmosphere-and-plant-venttiat4on-duct- ,

radi4t4en-mon 4ters-pro V4ded-that-g FO&&y-pa rtiC Nat44 Rd-4 Od ine-modto r&-4 Fe sonttor4ng-the stack-effluent. These three-plant-ventilation-duct-xadiation monitors-will init4.ta-closure -of the-conta4nment-pressure relief, & sample-and-purge-system-valves-end- shall employ a en: Out4

-three-logic-for the in4t44 tics of WAS# --

A -The- spent fuel assembly m.y be transferred directly from the reactor core l to -the spent fuel pool Region 2 provided the independent verification of l assembly burnups as-def4ned-in Special Precedure SP SUPNUP 1 has been I completed and the assembly burnup meets the acceptance criteria identified I in Technical Specification Figure 2-10. l Crem %e rescTm crire

' Movenest cQtradiated fuel movement shall not be initiated before the reactor core has decayed for a minimum of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> if the reactor har been operated at power levels in excess of 2% rated power.

L.-h wn subc/Mced The equipment and general procedures to be utilized during refueling operations are discussed in the USAR. Detailed instructions, the above specifications, and the design of the fuel handling equipment incorporating built-in interlocks and safety features provide assurance that no 2-38 Amendment No. J,2f,JE,f),7J,133

kepctee Wb NONI

-l') .0 LIMITING CONDITIONS FOR OPERAT10t:S 2o Radioactive f.ffluents j 2.9. Licuid and Gaseous Effluents I plicability l

App ies to the controlled release of radicactive materials in lic d and gase s effluents from the facility. The provision: )f Technica Speci-ficati n 2.0.1 for Limiting Condition for Operation are not ap icable.

Objectiv To define i.e limits and conditions for the controlled re ease of radio-active mater is in licuid and gasecus effluents to the nvirons to ensure that these re ases are as low as is reasonably achievy le in confon-ance with 10 CFR Par 50.34a and-50.36a, and t> ensure th t these releases result in concen ations cf radioactive materials i liquid and gaseous effluents release to unrestricted areas that are ithin the _ limits l specified in 10 CFR art 20.

To ensure that the rel ases of radioactive ma rials above background in unrestricted areas are low as is reasonab achievable, the followino design objectives apply.

A. Licuid Effluents I (1) The dose or dose co. itment a member of the public during any calendar year sho Id n < exceed 3_ millirems to the total body.

(2) The dose or dose comittle + to a member of the public during any calendar year she ld no exceed 10 millirems to any organ.

B. Gaseous Effluents

-(1) The calculated unual air dose to gama radiation at any location which could be occupied . individuals _in ut. restricted areas should ot exceed 10 millira  ;

-(2) The calcul ted annual air dose due to eta radiation at any location hich could be occupieo by in viduals in unrestricted areas s ould not exceed 20 millirads; ar i k (3) The alculated annual-total quantity of iod'ne-131, tritium, anc} all radioactive material in particulate orm with half-1 3es greater than 8 days should not result i san annual dose L ;r dose comitment to any organ of an ind'vidud. in an unrestricted aree frno all pathways of exposure' excess of 15 millirems.

l

]

r L

/ 2-40 Amendment No. U ,_Il3 l-f- ~. _

h%OckyTRd1 Y 2.0 LIMITING CONDITIONS FOR OPERATION 1 2.9 Radioactive Waste Disoosal System Aoplicability Applies to the transfer of waste gases to the waste gas decay tanks. The provisions of Technical Specification 2.0.1 for Limiting Condition for Operation are not applicable.

i Obiective .

To ensure compliance with General Design Criterion 60 of Appendix A to 10 CFR 50.

Specification (1) The concentration of hydrogen and oxygen in the waste gas decay tanks shall be limited to below flammability concentrations. With hydrogen and oxygen concentrations above flammability concen*. rations, restore the concentrations to below flammability limits within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

(2) The hydrogen and oxygen momtors shall be monitoring the inservice gas decay tank during the transfer of waste gases to the waste gas decay tank. Whenever the monitors are inoperable, transfer of waste gases to a gas decay tank may continue provided grab samples are taken from the gas decay tank and analyzed:

a. Every eight hours during degassing operations, and

b. Daily during other operations.

Basis Specification 2.9 ensures that the concentration of potentially explosive gas mixtures entrained in the gas decay tank (s) will be maintained below the flammability limits of hydrogen and oxygen. Maintaining the concentration of hydrogen and oxygen below their flammability limits with a measurement program 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.

2-40 Amendment No. 86;l13 (Next Page is 2-48)

l I

b .O LIMITING CONDITIONS FOR OPEP,ATIONS

2. Radioactive Effluents (Continued) i 2,9,3 liauid and Gaseous Effluents (Continued) l l

l (1 Specifications for Liquid Vaste Effluents (i) The release rate of radioactive material in li id effluents shall be centrolled such that the i tar-taneous concentrations for radionuclides, o r than dissolved or entrained noble gases, do not sceed the '

values specified in 10 CFp, Part 20, Appe x B. for unrestricted a-eas. For dissolved or e9 rained r.ctle casess the concentration shall be limiltd to 2.0 E-04 inicroCi/mi total activity. / I (ii) W th the concentration of radioact raterial released to nrestricted areas exceeding i e above limits, appr riate corrective actions all be taken imediately to re/ ore concentrations with ti the above limits.

h. The cun:ulative te contributions f ort radioactive rM erials in liquid effluents eleased to unre ricted areas sba!! Se deter-mined, in accordan g with the , on a quarterly basis. If thedosecentributi5tp,dueto he cumulative release of liquid

.) effluents averaged ov r a ccl ddar quarter, excccd one-half of the design objectives he f ilowing course of actions shall be taken:

(1) Make an investiga 10 to identify the causes fer such relcases. /

(ii)

Define and in/tiate a pr gran of action to reduce such releases to the design lev ls.

(iii) Submit a recial report, pur ant to Specification S.9.3, within ' days from the end o the quarter during which releas occurred, identifying t e causer and describing the p oposed program of action t reduce such release to the esign levels.

1

c. The ecuig; ment or subsystem (s) of the liquid adwaste treatment systen As identified in the ODCP shall be ope ted prior to the disc ge of radioactive materials in licuid w. tes. If the rad active liquid wastes were discharged withou treatrent by on or mnre of the pieces _of-equipment or subsyst i(s) idertified i the ODCM and it appears that one-half of the anr al objective ill be exceeded during the calendar quarter, a spec cl report,

.oursuant to Specification 5.9.3, shcil' 3e prepared anAsubmitted to the Comission within 30 days. Thisreportshallinqudethe

- follcwing information:

) Identification-cf equipment or subsysters not operab (i) and reasons for inoperability.

l- 2-41 Amcncrent No. 77, PE, II3 \

._._ _._ _ __..~..-._m . _ .- _ _.

1

.0 LIMITING CONDITIONS FOR OPERATIONS - j7' i

-:2 Q1 Padioactive Ef fluents - (Continued) 4

-]. :

12.9.. Licuid and _ Gaseous Effluents (Continued)

(ii) ' Action (s)'-taken to restore the inoperable equipment to status.

(iii) Su= mary description of action (s) taken to prevent /a recurrence.

d. During release of radioactive liquid vaste excludin releases

. rc= the steam generators , the -following conditione shall be met:

(i) At least one circulating veter purp shal' be in operation to provide a- dilution flow of approxima* ely 120,000 gpm in the discharge tunnel.

(ii) a e overboard header effluent radi- ion monitor shall be see in accordance with the ODCM t alarm and automatically clos the discharge valve prior o exceeding the 1D:its

speci ied in 2.9.1(1)a.(i) abo ..

(iii) The gros ' liquid vaste acti ty and flow rate shall be continun' ly monitored and ecorded during the- release.

If the eff ' ent radiation . onitor is inoperable,- ef fluent releases ma continue pr vided that prior .to initiating a release:

)- 1. At least tv- in pendent samples are analyzed in

-accordance vi Specification 3.12.1(1).

2. At least tv qu ' i fied individuals independently verify the ele as ra' calculations .

If the flow ate indica r is inoperable , effluent releases = / continue pro ided the flow rate is -deter-

-minel at east once per h . curs during actual release.

If' the radioactivity cannot -

recorded automatically, eff1"_nt releases may continue rovided the grcss radio-act ' vity -level is recorded manu, ly at least ence per k he rs during ' actual release. .

e._ Whenev r steam generator liquid is- being r leased to the. dis-char tunnel 1)_ the steam generator blevdos radiation monitors sha'. be set to alarm and automatically c1cse the blowdown isola-ti n valves prior to exceeding the limits spec fled in

^ . 9.1(1)a(i) above, and 2) the ' gross activity f each blevdcun -

line shall be monitored and . recorded by the blevc =vn radiation nonitors . _ If one of the two -radiation- monitors is inoperable ,

the activity for both -blevdovn lines shall be monit ed by the 2 L2 Amendment No. 86

-r-wr+------ -v- * * * - 1---r4--

l T~

2.0 LIMITING CCSDITIONS FOR OPERATIONS 2.9 "adicactive Effluents (Continued) 2.9.1 geuid and Gaseous Effluents (Continued) operable radiation monitor. If both radiation monitors re incperable, steam generator liquid release may continu9 provided appropriate grab samples are analyzed for principal gpfra ccitters at a sensitivity of 5.0E-07 xCi/t1 and reco/c6c i.t least caily wher. the specific activity cf the sarp 4 is less an or equal to 0.01 pCi/ gram dose equivalent I .1 and at 1 st once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the specific activi y of the sec ndary coolant is greater than 0.01 pCi/grar dose ecuivalent 1-13 If the radicactivity cannot be reccrd autoratically, efflue t releeses may continue provided the ross radioactivity _

level i A recorded nanually at least crice p t four hours during actual rblease.

(2) Specifications r Gasecus Waste Effluents

a. (i) The rel ase rate of radioactiye r.aterials in gaseous effluent shall be controllef such that the instan- .

taneous c centrations cf p dionuclides do not exceed the values ecified ir 1 CFR Part 20, Appendix P.

Table ? for grestricted areas. Unrestricted area concentrations shall be calculated based on the annual

} average Chi /Q.

(ii) Kith the ccncentra 'on cf radioactive material released to unrestricted ar s exceedinn the above limits, appro-priate correctiv etsinns shall be taker. irrediately to restore concent ation 1 thin the above linits.

b. The radiation dose ntributiona frer racioactive raterials l in gaseous effluen s shall be det rmined, in accordar.ce with the ODCM, on a cuarte ly basis. If th dose contributiers, due tc luents everaged over a the cumulative calendar quart r lease exceed ofone-half gaseousofe (t e design objectives, the follwino to se of actinrs shall be t ken:

(i) Pak an investigation to identify te cause for such re ease rates.

(ii) iefine and ir.itiate a prcgram of actic to reduce such releases to design levels.

Submit a special report, pursuant to Spe icaticr E.9.3.

(if) j within 30 days from the end of the quarter ring which release cccurred, identifying the causes and escriting the prrpcsed prograr of action to reduce dose ontribu-tions.

. The ecuipment of subsysten(s) of the gaseous radwaste tre rent systen as identified in the GDCM shall be operated prior t the discharce of radiotetive materials in gastcus wastes. !f t e radioactivegaseouswastesweredischargedwithcuttreatment'h{

2 43 Amen dent No. Pf, 113

. - - - - . - .. .-.- .- -.. _-- - -~-

)

~2.0(LIMITING CMDIi!ONS FOR OPER ATIONS '

2.9 Vaotoap 3ve Eff' vents (Continuec) 2.9.1 t oulo ano Gaseous Effluents (Continuec) one'or more of the equipment or subsystem (s) identifica in th ODCM, a special recort, purs_uant to Specification 5.9.3, sha i be repared ano suDmitted to the Commission within 30 days. T is r ort snali include the following information: '

(i) Identification of equipment or subsystem (s)' - t operable and reason for inoperability.

3 (i Action (s) taken to restore the inoperabl equipment to operable status.

(iii) ummary description of action (s) take to prevent a urrence.

d.

Thehydro,nandoxygenmonitorsshallpemonitoringthein- -

service gas decay tank duri_ng the tra fer of waste gases to the gas ceca tank and_the concentra on of hydrogen and oxygen shall u limited to below f1 ability concentrations. Whenever the moni+ rs are inoperable, transfer of wast gases to a gas ecay tank may continue provided analyzeo: (1) eve grab sam les are taken rom the gas decay tank and 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> d ing degassing operations, l and '(2) caily durin other op ations.-

e. The Auxiliary Building ex h st Stack gaseous, particulate, and iodine activity con +9 s may be inoperable provided that

j= 1) releases from a gas _d ay tank, containment pressure relief line, and the ce ta'nment. purge line are secured, and

2) whenever the Auxili ry B ilding Exhaust Stack gas or particulate samples will activity be tak .onitor is inoperable, appropriate grab and ana zed once per eight (8) hours.

f.

During power oper tion, the con nser air ejector discharge l shall be monitor d for gross radi ctivity. If this monitor is inoperable, grab samples-shall taken-and analyzed daily for pri cipal gamma emitters,

g. During rel _se of gaseous radioactive astes from the. I gaseous w ste: discharge header or durin, containment venting to the A xiliary Building Exhaust Stack, 'he following l conciti ns shall-be met:

(i). The gas, iodine, and particulate monit s-shall be monitoring the Auxiliary Building Exhau- Stack. l i) At least one exhaust fan shall'be in opera ion.

L (iii) The effluent control radiation monitors shal be set in accordance with the ODCM to alarm and auto.. tically terminate the releases prior to exceeding the 'mits specified in 2.9.1(2)a(i) above.

-) (iv) The activity shall be monitored and recorded. The L

l ~'

flow rate shall be monitored and recorded, or determined by calculation.

2-44 Amendment No. I2//86.137

2.0 L?MITING CONDITIONS r0R OPERATIONS

2. 2.9

.' l- Racioact we Ef fluents = (Continuec)

Liculo anc Gaseous Effluents (Continued) s- (v)

During the release of gaseous r,astes from the cent in-ment purge line, a-containment gas monitor and a particulate monitor shall monitor the con ainme. ,, in addition to conforming with (i) through (;iv) a ove,

h. During releases ' rom the Laboratory and Radioacti Waste Processing Building Exhaust Stack, the following conditions shall be. met: *

(i) The Laboratory and Radioactive Waste P cessing /

Building (LRWPB) Exhaust Stack gas, ' ine, and particulate monitors shall be monit ing the LRWPB Exhaust Stack. The effluent contrp radiation monitors shall be set in accordange with the 00CM to larm prior to exceeding the li (ts-specified in 9.1(2)a(i) above. The gas a .ivity monitor may-be in perable provided that appr riate grab samples be tak n and analyzed once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. The particulate and peine activity monitor 4 may be inoperable proviP ' that samples are entinuously collected as requ 'n Table 3-12, (ii) The ef, A qt flow rat shall be monitored and recorded, o determi ed by calculation.

Basis

) t Releases of radioa u vity in li id wastes wittin the design objective levels provide reasonable assylan e that the resulting annual exposure fromlicuideffluentswillntexchdthelimitsspecifiedinAppendix I to 10 CFR Part 50. Tbes specific tions provide-reasonable assurencethattheresultygexposurewillnotexceed3mremtototal ,

bocy or: 10 mrem to any o gan. At the ame time, these specifications ,

permit the flexibility f operation, co atible with considerations of

-health and safety. to- ssure that the pu lic.i; provided a dependable spurte of power unde unusual operating c ditions:which may temporarily result /n-releases higher than he design objective. levels butstillwithinyecencentrationlimitssp ifieo in 10 CFR Part 20.

The design obj ives have been developed base son operating experience,ca[e.culation procedures based on mode s and data set forth in Regulate:/ Guide 1.109, and the evaluation of- rt Calhoun' facility I-in accordan(e with Appendix ! of 10 CFR Part 50 do 9 design

-objectives. The design objectives take into account a comoination of variable including fuel failures, primary system _ lea age.,

primary o-secondary system leakage and the performanc of various radica tive; waste treatment systems.

! .Spe fication 2.9.1(1)a reouires the licensee'to limit the co centration of radioactive materials in liquid effluents- leased f om the site to levels specified in 10 CFR Part 20, Appendi B, for H -

, nrestricted areas. This specification provides assurance th no l

- member of the general public will be exposed at any time to liq id

[ 2-45 Amendment No. 12,86, 137

2. -[IMITING CONDITIONS FOR OPEDATIONS-2.9 Racioactive Effluents (Continuec) 2.9.1 hicuia ano Gaseous Effluents (Continuec)

Ba s (Continued)-

conta' ina radioactive materials in excess of limits cons'idered permis ible under the Commission's Regulations.

Specifica ion 2.9.1(1)b establishes the frequency of dose c culations in accoraa ce with the ODCM. This specification also estaWlishes the reporting r nuirements in accordance with Section IV.A of Appencix 1

-these Technich Specifications.to 10 CFR Pa

• 50, in addition to the requireme Specification 2.

1(1)c requires the operatien of th equipment or i subsystem (s) of t -

radioactive liquid waste systen[ as identified in the ODCM, to reduc the release of racioactive ma f> rials in liquid j effluents to as low s reasonably acnievable, cc istent with the requirements of 10 CF Part 50.36a, and General esign Criterion 60 of Appendix A-to 10 CFR P t 50, Normal use of t e equipment or i subsystem (s) in the raa cactive licuid waste ystem provides l reasonable assurance that the quantity rele ed will not exceed the  !

design objectives.  !

I '

)

1 2-45a Amendment No.137

9.0 LIMITING CONMTIONS FOR OF'GATIONS

.9 Eadioactive Ef fluents (Centinued) 2.9b licuid and Baseous Effluents (Ccntinued) .

)

nasis (Continued)

S; cificatica 2.9.l(l)d, consistent with the requirements of general Design Criaria 60 and 6h of 10 CFR Fart , Appendix A, requires peratien of suita le equipment to dilute , centrol, and tonitor the rel ases of radio-active paterials in liquid vastes , ether than stear. gener .tcr liquid, fron>

the ove card header during any pericd when releases ar, taking place.

. ,e.c .w ...e .,a n_ x+ nr +. s. c+e -. . n_, c c .n.e .r a+ n.

e p, c ar

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, 4. s, u i d "h.. e n.

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- .u di.e s *.. .~ ad m +,,a- +".e a" .- 4..--^ r_- e n + . '. #,~ ~. b'.'

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• c . . ~ n.~ i * . '% -

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. . .=o. * ~g ~ e~, ., e . .. ' . . ' . - .

radiation monito's are found inoperable ar.d ifj eteam cenerator liquid is being released to the environment ,-the speciffed sampling frequency pre-

.a v4 a e s a.c .e., .u - an w * +.". . ', . m, . .,. a gi a .- w"*.'.".".g

. .'s" .v a # a ". ed- 4 ". r d... [',

7 a

m

. . . 4. +. a.ad ~.."."s.

f cf time when repairs are being made. /

i.

The release of radioact(ve materials in jaseous vas te e f fluents to unre.

a

.a.

.. on e s.

a a ,. n. sa. ., , , .., ,. . '.mo,-m. i. n.

..n. s...m_..-

. . - , . .x..,.. a...ee- .4.4+, .

a ,

specified in 10 CTR Part -O at an', t ' .e and shoul; be as low as is reason i ably achievable in accorda.'e with de requiretent; of 10 CFE Parts $0. N '

and 50.36a. These specifications ;rovide reasonable assurance that the resulting annual air dese due to tanna radiation vill not exceed 10 mrad and that the resulting annual Jr dose te beta radiation vill not exceed J ^O ..ad. f . - "s.. e e t" . e ^s. "ms "- a m- +. ' ' .' .. u . .+ e _ .' r .-. *. ". . e e .' e .n. + .".".ee s n - . m m .4..N.e'.

tiens also provide rens cnabl as s ;ran ce th at no individual in an unrestric a *e .a. "41' *

. . .ece4".e 'n. -. . " . " ' ' de e *. +"..a. .

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compatible with co. miierations of health a$4 s afety , to assure '. hat the {

public is provide / vith a dependable source .f power even under unusua. l 4.. r .ce um e o,. .w y e ,.a+e.n .

o. c ,.r..a . /t m..o

< ,o

- h,4 w. . .. a.y . .. . r.. g . a. .:y .. .~u,.. s i

cuch numerical guides for design objectives 1 at\stil3 within levels that 1

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n -w - m. ~u :. ,,,~~a.

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of doses fr'd natural background radiation. \

e s. c <a<<

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~ ,' ~n <- . .w. . . n. _- y. .

+ .,...;.u.... _ e.

radioac /ive caterials in gaseous e f fluents iron t'.e ststion to levels speci \

<<e, a , o m.

. . c. r. .. . . ." ,. , g n%n . e. . ~a <. .v. n, ^-

... .-.- . . .< *. ^.. d su1 r.m . ~.S. i. e rm ad.."."a. .

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.w. . g g' 4 c. v .' '. k a_

4 -

expf. sed at any time to cas es con t .ining radioactive mater hls in excess of '

l' ..its specified in the Comission's reculations. -

\  !

i Specificatien 2.9 l( 2';b establishes the frequen^y c: dose calwlations in )

w - .o a 9;s.e e v.i a.. .

.w. m.c v

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+ w. u . .a. 4 . . n c \ .a.g .m .m. .s 4. . . ,

m 2

2-M teendment no. 12 , k

\

l f.0 LIMITING CONDITIONS FOR OPERATIONS

t. 9 Rao,cact,ve Effluents (Continvec) 2.9, i

Lioulo ano-Gaseous Effluents (Continued) asis (Continuec) re irements in accordance with Section IV.A of Appendix I to JO CFR Par 50, in aedition to the requirements of Section 5.9 of th se Techn' cal-Specifications. 1 Specifit tion 2.9.1(2)c recuires the operation of eauipme.t or

-subsyste the ODCM,

5) of the radioactive gaseous waste system, ary identified in effluents t oreducethereleaseofradioactivemateria)tingaseous as low as reasonably achievacle, consist,ent with the requirements f 10 CFR Part 50.36a, and General Oes'.n Criterion 60 of Appendix A to 0 CFR Part 50. Normal use of the e.Jipment or subsystem (s) in the radioactive gaseous waste sys em provides reasonable assur ce that the quantity released will not exceed the design objectives.

Specification 2.9.1( d ensures that the co entration of potentially entrained.in the ga decay tank (s) will be main-explosivegasmixture(ailitylimitsof tained below the flamm regen and oxygen.

Maintaining the concentr tion of hydrog and oxygen below their flammability limits with measurementA ragram provides assurance that the releases of radioactiv smateriais kill be controlled in conformancewiththerequireKentsofGeneralDesignCriterion60of Appendix A to 10 CFR Part 50.

I Specification 2.9.1(2)e provides assurance that releases from gas decay tank, Auxiliary Buildingji:x ust Stack, containment pressure l relief line, and containment furg- Ne are not made whenever the stack gas, particulate and 'cdine ' ors are inoperable. l Specification 2.9.1(2)f sures that gross radioactivity, during power operation, is mon' cred from the-cQndenser air ejector

[

discharge.

\

Specificat on 2.9.1 c)g requires coeration \ suitable equipment to- l l-dilute, control, apd monitor in orcer to pro de assurance that radioactive mater /als-released in the gaseous ffluents are properly-controlled and 2 nitored in accordance with the eouirements of t

General D.esign7 riteria 60 and 64 of 10 CFR Part 50, Appendix A.

<- Specificati l . Radioactive'ph 2.9.1(2)h Waste provides Processing Buildingfor(LRWPB) releaseswhenev from he Laboratory and the LRWPB Exhausts inoperatyle.

/ackgas,particulateoriodineactivitynoniorsare l

2-47 Amendment No.86,137

. ~ - . ~ . - . ~ . . . . . , - - - . ~ - - ~ ~. - -~ . . - . . . _. . . . . , . ... . .

. 0 LIMITING CONDITIONS FOR OPERATICt:S

2. -Radioactive Effluents _ - ( Continued } /

1) 2 9 2 Solid Radioactive Waste A_ plicability1 Thi specification s

applies to the processing and packaging o ' solid and compa ted raivaste.

Objecti -

To ensure *cnformance with 10 CFR Part 20 and 10 CFR art 71' prior to ,

shipment of olidified radwaste from the facili+.y. . e prov:slons of Technical Sp ification 2.0.1 for Limiting Conditi ..s for Operation are not applic-ble.

Specification The equipment or su system (s) of the solid v dvaste system, as identified in the-Process Contr Progra= ( PCF) , sh al - be operated to provide for 1the solidification. of et solid vastes ar - the compaction of compressible vast es . ified by requiremerts specified JWaste solidifi ation vill be vy[s to meet the above " objec in the PCP. If solidifi radvaste fa a .

regulations or the accept .ce criter of the PCP, no offsite shipments shall- be made of the non-ec.-forming aterials.

Basis

}: The solid radvaste systen is - ngrally operated on a batch basis, and is

-available to perform abnorma' or hgergency functions. The proper opera- ,

tion of the solid radwaste fstem-e sures that the pertinent requirements of 10 CFR Part 20 and 10 . Part-71 vill be implemented. This specifica-i tion also complies with + me requireme_ ts; of 10 CFR -Part 50.36a and General Design Criterion 60_of .ppendix A to 1 CFR Part 50. The operatirg procedures, process p' ameters and the a ceptance criteria, included in the Process Control ".cgram, vill provide compliance with these require-ments.

e i

l V

i i l-l-

l

' L7a . Amendnent 3o, 86 L

1 l l' '

l

, -nc,, , , ,.a ,, ,, .- ,n-. . c.-,. . . . . - - - - ...-nw- . - - .

l t

2.0 LIMITING CONDITIONS FOR OPERATION,,

f 2.14 Engineered Safety Features System initiation Instrumentation Settings Applicability ,

Applies to the engineered safety faatures system initiation instrumentation settings.

Objective-To provide for automatic initiation of the engineered safety features in the event that principal process variable limits are exceeded.

Specifications The engineered safety features system initiation instrumentation setting limits shall be as stated in Table 2-1. l Basis (1) High Containment Pressure The basis for the 5 psig set point for the high pressure signal is to establish a setting which would be exceeded quickly in Oe event of a DBA, cover a spectrum of break sizes, and yet be far enough above normal operation maximum saternal pressure to prevent spurtws ir,'tiation.

(

High containment pressure initiates the steam generator isolation signal which will close the main steam isolation and bypass valves and the main feedwater isolation and bypass valves.

(2) Pressurizer Low Pressure The pressurizer low pressure safety injection signal is a diverse signal to the high containment prcssure safety injection signal.

The 1600 psia setting incluN an uncert41pty of 1 22 psia and is ,

the setting used in the safety analysis.lli (3) Containment High Radiation (Air Monitoring) 4e-eont44%nt-air-monitoring-cys. tam comarites a mov.ing-waar 44+temt4c4e-monitor (channe_1_h0 s) and A umnlo chambag 7eS-90fti tot % ha nnel--S.M-051-)- ins t a ll e d in a entenn hnutiD21.'l 4pt40na&1y r-th+mp14ng-point-Jor-channers-RWO50 ad W-151 can

-te--swi-tchedJcom-thScont&4cment-te the vent 44at4on- discharge duct-t t

4he-ven444 44on-44&oharge moni. toning _sys tem - eons s.ts nf a mnging

!' taper 4-44hr-pa r4141+ moni te r-- ( PC 061 ) a nd 2 - s amp l e c h mh e r gas wnstw-(E062-)- ins 4tt !! ed i n 2 - comon housdng. An indi" s nitnr for !-lMdRM-- 050) a! se moni tors thesa--ralsases,.

'Tb wnTain ment reWaGn h h sjnal den be inlTATed by a CMTA M tnt otm*S rnSATim modo

^

^E

$^5 ecm$ ftdh% nu%ge An Awxil thcy bigin % sT  %& QcQuv5 l 2-61 Amendment No. 108

+- - u,-w-- -v.-- e -e- y. ,, ,gr--w.--- 4'm , . . , - - . + p ys =y-*-wwv7- .-w wwwwe--ry

... ____- -... - -. . - _ _ - - - . - .=. ._. -- - _- --

2.0 LIMITING CONDITIONS FOR OPERATION ,

2.14 Engineered Safety Features System Initiation Instrumentation Settings (Continued) '

(3) Containment High Radiation (Air Monitorine (Continued)

The setpoints for the isola' ion function will t>e calculated in accordance with the ODCM.

Each channekin4upplied from-a separate 4nstrumentA CAusand each auxiliary relay requires

-power to-operate On. failure-of-a-singlo A C.-supply,4he-A and-Ibmatrices-will-assume-a

-one-out of.two. logic.

(4) Low Steam Generator Pressure A signal is provided upon sensing a low pressure in a steam generator to close the main steam isolation valves in order to minimize the temperature reduction in the reactor coolant system with resultant loss of water level and possible addition of reactivity. The setting of 500 psia includes a 122 psi uncertainty end was the setting used in the safety analysis.m Closure of the MSIVs (and the bypass valves, along with main feedwater isolation and bypass valves) is accomplished by the steam generator isolation signal which is a logical combination of low steam generator pressure or high containment pressure.

As part of the AFW actuation logic, a separate signal is provided to terminate Dow to a steam generator upon sensing a low pressure in that steam generator if the other steam generator pressure is greater than the pressure setting. This is done to minimize the temperature reduction in the reactor coolant system in the event of a main steam line break. The setting of 466.7 psia includes a +31.7 psi uncertainty; therefore, a setting of 435 psia was used in the safety analysis.

(5) SIRW Tank Low Level .

Level switches are provided on the SIRW tank to actuate the valves in the safety injection pump suction lines in such a mantier so as to switch the wer supply from the SIRW tank to the containment sump for a recirculation mode of operation after a period of approximately 24 minutes following a safety injection signal. The switch-over point of 16 inches above tank bottom is set to prevent the pumps from running dry during the 10 seconds required to stroke the valves and to hold in reserve approximately 28,000 gallons of water of at least the refueling boron concantration. The FSAR loss of coolant accident analysis

• assumed the recirculation started when the minimum usaole volume of 283,000 gallons had been pumped from the tank. <

l i

i 2 62 Amendment No. 5,32,43,65,66,403,408d33,141 i

i

. - . . . - , - . - - . , , . . ~ . . . _ . - , - , . .-., - -

2.0 LIMITING CONDITIONS FOR OPERATIONS 2 . '. 4 Engineered Safety Features System Initiation Instrur,entation

( ht,_t,i_n,Lqs, (Continued)

~

(6) Low Steam Generator Water Level As part of the AFW actuation logic, a signal is provided to initiate AFW flow to one or two steam generators upon sensing a low water level in the steam generator (s) if the absolute steam generator pressure criteria are satisfied. This function ensures adequate steam generator water level is maintained in the event of a failure to deliver main feedwater to either steam generator. The setting of 28.2% of wide range tap e,oan includes a +13.2% uncertainty; therefore a setting of 15% of wide range tap span was used in the safety analysis.

(1) High Steam Generator Delta Pressure As part of the AFW logic, a high steam generator differential pressure signal is generated to provide AFW to the higher pressure steam generator with a concurrent low level signal if botn steam generator pressures are less thart 466.7 psia.

If the differential pressure between steam generators is less

- than the setting. neither steam generator is supplied with AFW in the presence of a low level signal. The setting of 119.7 psid includes a -15.3 psi uncertainty; therefore a setting of 135 psid was usec in the AFW safe', analysis.

(

References (1) USAR. Section 14.1.3 i

(2) USAR. Section 44 carM-- 7. 3. Z . i .

(3) USAR. Section 14.12 (4) USAR. Section 14.15 (5) USAR, Section 7.4.6 (6) USAR, Section 7.5.2.5 (7) USAR. Section 14.4.1 i

2-63 Amendment No. f.5. 108

2.0 LlHITING CONDITIONS FOR OPERATION 1

( 2.15 instrumentation ano _ contral systems (Continued)

(5) In the event that any of the following Emergency Auxiliary feedvater l Panel instrumentation or control cir;uits become inoperable, either restore the inoperable component (s) to operable status within seven days, or be in hot shutdown within the next twelve hours. This (

specification is applicable in Modes 1 and 2.

SteamGeneratorLevel,WideRange(Al-179)

SteamGeneratorLevel,NarrowRange(AI-179)

SteamGeneratorPressure(Al-179)

Pressurizer pressure (AI-179)

Basis During plant operation, the complete instrumentation systems will normally be in service. Reactor safety is provided by the reacter protection system, which automatically initiates appropriate action to prevent exceed-ing established limits. Safety is not compromised, however, by continuing operation with certain instrumentation channels out of service since provisions were made for this in the plant design. This specification cut 11nes limiting conditiens for operation necessary to preserve the effectiveness of the reactor control and protection system when any one or more of the ch nnels are out of service.

( All reactor protection and almost all engineered safety feature channels are supplied with sufficient redundancy to provide the capability for channel test at power, except for backup channels such as derived circuits in engineered safeguards coV Jl system.

When one of the four channels is taken out of service for maintenance, the protective system logic can be changed to a two-out-of-three coincidence for a reactor trip by bypassing the removed channel. If the bypass is not effected, the out-of-service channel (Power Removed) assumes a tripped condition (except high i pressurizerpressure),gte-of-changeofpower,highpowerlevelandhigh which results in a one-out-of-three channel logic.

-1f in the 2 of 4 logic system of the reactor- protective system one channel -

is bypassed and a second channel manually placed in a tripped condition, j

the resulthg logic is 1 of 2. At rated power, the minimum operable high-power level channel is 3 in order to provide adequate power tilt-detection.

If only 2 channels are operable, the reactor power level is reduced to '

70% rated _ power which protects the reactor from possibly exceeding design peaking factors due to undetected flux tilts and from exceeding dropped CEA peaking factors.

All engineered safety features are initiated by 2-out-of-4 logic matrices exceptcontainmenthighradiationwhichoperatesona1-out-of-g-basis, 9

MT G"M M ev1[t Mcd dM h na( bWg ihe CcmTrM nc.,J References pre m ce ce 4 e 0, c c,s % eg (1) FSAR; Section 7.2.7.1 UMR l 2-66a Amendment No. E,20,25,32,#3,88, 125 a_- _ . - . _ . - - . . - _ . . - - _ - - - . , - . - _ . . - _ - - . . -.

i TABLE 2 4 INSTRUMENT OPERATING CONDITIONS FOR ISOLATION FlGCTIONS Test Maintenance >*

Minimum Minimum Persissible and Operable Degree of Bypass Inoperable e Pedundancy Condition Bycass Functional Unit Channels 1 Containment isolation A Manual 1 None Hone N/A B Containment High Pressure A 1 During Leak (f)

B 2fa)e2 a) e 1 Test Pressurizer Low / Low A 1 Reactor Coolant (f)

B 2((a)(e) 2 a)(e) 1 PressuregsThan 1700 psia l

2 Steam Generator Isolation A Manual 1 None None N/A B Steam Generator Isolation 1 None Hone N/A (i) Steam Generator Steam Generator (f)

Low Pressure A 2/Ste'ag 1/ Steam g

Gen l s Gen Pressure (gessThan 550 psia l B 2/Ste'ag 1/ Steam Gen l 1 Gen (ii)ContainmentHigh Pressure A 1 During Leak (f)

B 2((a) 2 a) e e i rest 3 yentilation isolation A Manual 1 None None N/A B Containment High Radiation A None If Containment Vent 44ation Re hef (f)- N/A B 2((d) 2 d) None

. ord Pu)elsolatlon' Are Closed Valves a A and B circuits each have 4 channels.

b Auto removal of bypast above 1700 psia.

c Auto removal of bypass above 550 psia.

2-69 Amendment No. E8. S3, 108

/ f i

-l l

l TABLE 2-4 I

/ (Continued) ecther %e d A and B circuits are both actuated by.any-one of the five-VIAS initiating '

channels; WO60rRM 0517RM-060rRM*061ror-RM4062rhoweverronly-P&O90 and-RM.05i are required for containment ventilation-isolation.

e If minimum operable channel conditions are reached, one inoperable channel-must be placed in the tripped condition within eight hours from the time of discovery of loss of operability. The remaining inoperable channel may be bypassed for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> from the time of discovery of loss of operability and, if an inoperable channel is not returned to operable status within this time frame, a unit shutdorm must be initiated (see Specification 2.15(2)). l f If one channel becomes inoperable, that channel must be placed in the tripped or bypassed condition within eight hours from the time of discovery of loss of operability. If bypassed and that channel is not returned to operable status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> from the time of discovery of loss of operability, that channel must be placed in the tripped condition within the following eight hours. (See Specification 2.15(1) and exception l associated with maintenance.)

I 2-69a knendment Nc 88,108

\

l 3.0 SJ)RVElllaNCE PEOUIMMNTS l BASIS Specifications 3.0.1 through 3.0.4 establish the general requirements applicable to Surveillance Requirements. These requirements are based on the Surveillance Reouirements stated in the Code of federal 4+qutramants. 10 CFR 50.35(c)(3):

by%.TU:nu

" Surveillance reauirements are requirements relating to test, calibration or inspection to ensure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting condition of operation will be met."

Specification 3.0.1 establishes the limit for which the specified time interval for Surveillance Requirements may be extended. It permits an -

allowable extension of the normal surveillance interval to facilitate surveillance scheduling and consideration of plant operating conditions that may not be suitable for conducting the surveillance: e.g.,

transient conditions or other ongoing surveillance or maintenance activities. It also provides flexibility to accommodate the length of a fuel cycle for surveillances that are performed at each refueling outage and are specified with an 18-month surveillance interval. It is not intended that this Lcovision be used repeatedly as a convenience to extend surveillance intervals beyond that specified for surveillance that are not perforned during refueling outages. The limitation of Specification 3.0.1 is based on engineering judgment and the recognition that the most probable result of any particular surveillance being performed is the verification of conformance with the Surveillance Requirements. This provision is sufficient to ensure that the reliability ensured through surveillance activities is not significantly degraded beyond that obtained from the specified surveillance interval.

The provisions of Specification 3.0.2 define the surveillance intervals .

. for use in the Technical Specifications. This clarification is I

provided to ensure consistency in surveillance intervals throughout tha Technical Soecifications. A few surveillance reouirements have uncommon interval s. for-maga- TMa-M-r4wtrn_4ampi ing of EiA

-onca_ car._.saasrA In such a case the surveillance interval shall be performed as defined by the individual specifications.

Specification 3.0.3 extends the testing interval required by codes and standards referenced by the Technical Specifications This clarification is provided to remove any amoiguities relative to the frequencies for performing the required inservice inspection and testing activities. Under the terms of this specification. the more restrictive recuirements of the Technical Specifications take precedence over the coces and standarcs referenced therein.

Specification 3.0.4 establishes the failure to perform a Surveillance Recuirement within tho allowed surveillance interval, as defined by the provisions of Specifications 3.0.1 and 3.0.2, as a conaition that constitutes a f ailure to meet the OPERABILIT( re;utrements 'or the corresponoing Limiting Concition for Operation, inder the provisions 3-Ob amencment No. 122, 129

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TABLE 3-2 (continued) i.

MINIMUM FREQUENCIES FOR CHECKS; CALIBRATIONS AND TESTING OF l,!

ENGINfERED SAFETY FEATURES, INSTRUMENTATION AND CONTROL'

+1)

= Surveillance Channel Description Function Frequency Surveillance Method is. D h tn knnun ovfornal

6. (continued) b. Ca'ibrate F v p g g e_irg 1, radiatica se"eca --

c. r it

" c. c-ete escrated em4;2+4nn i ntegr>',ma esmet e n. re m ..

in un f t* in t et t nt,Iinn

% Q@

cra ane! at a t en, 2nd i is01 3* inn i nck nitt relay fur.cti^^2! chect Manual Safety injection a. Test R a. Manual initiation.

7.

! iation 4

a. Test R a. Manual initiation.

, 8. Manual Containment Isol-ation Initiation b. Observe isolation valves

b. Check R 4

closure.

a. Test R a. Manual switch operation;

9. Manual Initiation Con- pumps and valves tested tainment Spray separately.

Automatic Load Sequencers a .. Test Q a. Proper operation will be

! 10. verified during safety feature actuation test of Iten 3(a) above.

Diesel Testing See Technical Specification 3.7 11.

t

- - + - - + . . - - . , &

r N Cit h M & vt T b

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

6. b. Test M b. . Detector exposed to remote operated radiation check (continued) source or test signal to verify instnmentation, one chanrd at a time, and isolation :ockc-ut relay functional check.

c. Calibrate R c. Secondary and Electronic Calibration performed at refueling frequency. Primary calibration performed with exposure to radioactive sources only when required by the secondary and electronic calibration.

i i

i l

[

t

TABI!M]

!!IIH!!81HLDJJOICIES FOR GIEUS,_GI.IIMPIOtt; AllD 'ImritF, QF_HISCEIJ AllEFF; it!GI]Mfil!TATIR{ A!!i}.gntmp_IS Sutvei11arce Onnnel Ibscription Dtncticn l'ruloency Sinveillance M. tin!

1. Primury CEA Ibsition a. Cieck S Inlication System a.

Q.zqurison of uitput data with seconisty gal'IS.

b. 'Ibst M b. Test of pu2r depenlent insertion limits, devia-tion, ani seq 2cnm monitorin3 systems.

2

c. Calibrate R c.

Ihysically measurul CEIII gnsition used to verity '

system accuracy. Calibrate CEA position inter-locks.

2. Seconlaty CEA lbsition a. Oneck S

Inlication System

a. Orparison or outp2t data with prinury CAPIS.

b. 'Ibst b.

j 11

'Ibst of power depenlent insertion Iimit, devia-tion, cnit-of-sequence, an1 overlap monitorinJ systems.

, w C

c. Calibrate R c. Calibrate seconlary CEA position inlication system an1 CEA interlock alarms.

y MmaI%-Amident

-Process -armi r r a ,-.--c ed 9 i_ n_e-C im i i n v: a-m wd apt int e .1 te g i g RadiaElen41onitors sigruls usal to verify instrumen* p6taticn. I o_ Except Eftlu Test j

b. M b. Detector expomi to rpacte operatal radiation j Radiation Monitors

, a e.

check sourus er tcSt signal.

7 8g g . Calibrate R c. 141- 63 ,, M, an] 11 anl 101-064 - One tina factory o

gg libration is acceptable prinidati linearity a solid souruzs are used to check the intaJrity of

', the detectors. Ict-091A ani B - In situ calibra-l

, ,-- tion by electronic sigrul substitution is accept-able for all rarne decades above 10 Iyhr. In y s1Qlibration for at .lcast one decade below w

10 IVhr nu}Nol saarce. A11 be by

> reans of r:cnitors calil; ratedtoradiation

- Expostare kimn

~

radiation source.

h (1) ne Eiutveillance rayiirements for ef fluent radiation monitors are described unter Specit Effluent radiation monitors are: Ici-041,104-042,104-043, int-054 A,101-054B, ICt-055, IM-055A, ion is 3.12.1.

idi-060, ili-GGi, siw! 10' '"2,-.-H4.n5n mi rmos t ;u g "7, I the Auxiliaty intildity 11tuust Stack. mnsiderol ef fluent radiation ronitors tihen anonit Q't

(. y . y-

~

. TABLE 3-3 (continued)

MTillinM FREQUEllCIES F0H CIIECKS. CALIBRATIONS AND TESTING OF MISCELIAflEOUS If!STRUMENTATION AND CONTROLS Surveillance Channel Description Function Frequency Surveillance Method 14 Emergency Plan Radiation a. Calibrate A a. Exposure to known radiation Instrutnents source.

1

b. Test M b. Battery check.

4

, 92 -- hel.ca cr.tal Monitor; s. L",c ch "

. Cy ratien11 Seck.

[tWachmeld (CD b. celibrate a

b. "ority =1 -fl~ indhator-

6. Pressurizer Level In- a. Check S a. Comparison of independent struments level readings.

b. Calibrate P. b. Known difterential pressure applied to sensor.

c. Test M c. Signal to alarm meter relay

! adjusted with test device to verify setting.

7 CEA Drive System a. Test R a. Verify proper operation of Interlocks all CEDM system interlock ,

using simulated signals where necessary.

b. Test P b. If haven't been checked ior

, three months and plant is shut-i down.

i t

G I

4

h'MC h Mervk h

3. Area and Post Accident a. Check D a. Normal readings observed and internal test signals used to verify instrument

, Radiation Monitors") operation.

b. Test M b. Detector exposed to remote operated radiation check source or test H;aal.

, c. Calibrate R c. Secondary and Electronic calibration performed at refuel:ng frequency. Primary calibration with exposure to radioactive sources only when required by the secondary and electronic calibration. RM-091 A/B - Calibration by electronic signal substitution is acceptable for all range decades above 10 R/hr. Calibration for at least ene decade below I- R/hr. shall be by means of calibrated radiation source.

"8 Post Accident Radiation Monitors are: RM-0631]M/H, RM-064, and RM-091 A/B. Area Radiation Monitors are: RM-070 thru RM-082, RM4)84 thru RM-089, and RM-095 thru RM-098.

A11aok men GS j

5. Primary to Secondary a. Check D a. Normal readings observed and internal test signals leak-Rate Detection used to verify instrument operation.

Radiation Monitors (RM-054A/B, RM-057)

b. Test M b. Detector exposed to remote operated radir. tion check sources or test sigral.

c. Calibrate R c. Secondary and Electronic calibration performed at refueling frequency. Primary Calibration performed with exposure to radioactive sources only when required by the secondary and electronic calibration.

TABLE 3 4 (Continued)

MINIMUM FRE0VENCIES FOR SAMPLING TFSV

( Type of Heasurement and Analysis

1. Reactor Coolant (Continued)

(c) Cold Shutdown (1) Chloride 1 per 3 days (Operating Mode 4)

(d) Refueling Shutdown (1) ChloriGe 1 per 3 days (3)

(Operating Mode 5) (2) Boron Concentration 1 per 3 days (3)

(e) Refueling Operation (1) Chloride 1 per 3 days (3) /

(2) Boron Concentration 1 per shift (3) /

2. SIRW Tank Boron Concentration 1 per 31 days

3. Concentrated Boric Acid Boron Concentration 1 per 31 days Tanks

4. S1 Tanks Boron Concentration 1 per 31 days

5. Spent fuel Pool Boron Concentration 1 per 31 days s

6. (Sitam Is<rt.pc 4,ntY5bfccDese gPy t4 5 i M opnai$ Gene %Tcc thdes lad 2) h&wn Qdv.de>n -1st (1) Until the radioactivity of the reactor coolant is restored to 1 lu C1/gm /

DOSE EQUIVALENT l-131.

(2) Sample to be taken after a minimum of 2 EFPD and 20 days of power operation have elapsed since reactor was suberitical for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or longer.

(3) Boron and Chloride sampling / analyses-are not required when the core has been off-loaded. Reinitiate boron and chloride sampling / analyses one /

shift prior to reloading fuel into the cavity to assure adequate shutdown /

margin and allowable chloride levels are met. /

(& LD%en GTtss Qec e<cJet Dose &pivalent T-\bt 1: Meeds Go ferCed of %e tim;Ts in s p:,e c S e c tic d 2 . 2 c)

{c c eq skqu be i,screasect % A vublimen the suptin) of 6 Tiaimesd ct per a d y s is toeek, mb Shnw p w acc he Quibritee I- tai acee4 75 percent oP %s lind$ Oe & *P "1 M Analtjsis ffEpocy li siwtl k inc m ed % a. minimu m J cne. pec c{n ,

Amendmc .. No. 28, E7, EE, J2/,133 3-19

3.0 SURVE!LLANCE REQUIREMENTS

.(- 3.10- Reactor Core Parameters (Continued)

(6) Azimuthal Power Tilt (Tq)

Whenever the core power is above 70% of rated power, the azimuthal power tilt shall be determined to be within its limits by calculating the tilt at least once every day using either:

a. The excore detectors with at least four safety channels I

operable, or

b. The incore detectors with at least two strings of three rhodium detectors per full core height quadrat, operable.

(7) DNS Parameters

a. The cold leg temperature, pressurizer pressure, and axial shape index shall be verified to be within the limits of Section 2.10.4(5) at least once per shif t,

b. The reactor vessel coolant total flow rate shall be deter-mined to be within its limit by measurement at least once per month.

('

l Amendment No. 32, 76, 92 3-63b I (

hT A6 i$ h

= - .-. . = -- _ _, .- .- -

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3h SURVEilLA!!CE RECUIREMT21TS

(- 3.11 Ea'diologien1 Environrtental Monitoring Program pplicability App ies to radiological monitcring of plant envi.ons.

,C,_bJSet\ te i i

To estab sh a radiological monitoring program adequate to me < ure changes 1. the levels of environmental radioactivity due to lant effluents.

Srecificatien.

(1) The radiolo ical envircnmental monitoring progrr shall be cenducted according to 'T'able 3.9 Additional details of

• ie radiological environmental onitoring program are in the CDi .!. No cha.ncen shall be made to the 'DCM which might reduce the e; ectiveness of the program. Analyt cal results of this prograr and deviations from the sampling schedule -hall be reported to the .'raicsion pursuant to Specification 5 9. b.

(2) If the level of radio: etivity in an en .rcnmental sampling medium exceeds the reporting . . vel specified in the CDC!!, a ron-routine report shall be prepare and cubmitt d to the Commission pursuant to Specification 5.9.!..b.'.

(3). A land use survey shall be ndu .ed once per 2h menths between the dates of June 1 and October . This survey ahall identify the loca-tion of the nearest milk ani.m, and the nea rest cosidence in each of the 16 cardinal sectors with' ' distance : f five miles . The results of the land use survey shal' be ., 'bmitted te the Com.ission pursuant to Specification S.9.h.h. .he su ey shal be conducted under ue following conditicns :

a. Within a one-mile .tdius from the ,lan: site , enumeration by door-to-door or e aivalent c:unting technique.

b. Within a five ile radius , enumeratic: :r using re ferenced information f m county agricultursi ag nts or other reliable sources.

I i

!f it is lea. ad from this survey that milk ani. als are present at a locatien whir . yields a calculated thyroid dcce reater than frem I previously ampled animais , the new -locati:n shal be added to the monitoring program. '"ne s ampling locatic:. ^.avin; . w inves t :21:u-  ;

l at ed doe . may then be dropped frem the monitoring p 'ogram at the j end of

• e- grazing seuen- durin6 vhi:h *:he survey was co n duc t e =1.  ?

Also, 77 locaticn(s)- from which milk can nc lancer be sbtained may ~

j ',

be- ir pped and replaced if pra:ti:1ble 2rr the ;ni .ori's progre.

ar.d .he Cc mission shall be notified pursu;: .t to Specific tion 5.9,b.b. .t

,: o s

3- 6 :-.  : .en dne: - 30. H , ?d, 56

. _ . . . - _ _ _ _ _ _ _ _ . . _ _ . . _ _ = . _ . _ _ . _ _ _ _ . ~ . . . _ _ . - _ . . . _ . . _ _ . _ _ _

l

3. SURVEILLANCE REQUIRD4U1TS 3.11 kiological Fnvironmental Monitoring Program (Continued) 3-(h) ,'alyses shall be performed on radioactive materials as pay. of an terlaboratory Comparison Progren that has been apprpfed by the h ". The results of these analyses shall be inclu ed in the Annual . diological Environmental Operating Report.

pgis . .

The radiological envfr retental monitoring progra. required by this ,

specification provides asurements of radiatic and of radinactive materials in those exposugpathways and for . hose radienuclides which lead to the hichest potenti R radiation ex sures of individuals ,

attributable to the operationNr Fort C houn Station.

The specification for land use sur 9- is provided to ensure that chances in the use of- unrestricted areas ge 'dentified and that modification 2 i to the monitoring program are er e if quired by the results of this ensus. ~~his cencus satisfie3 the requi r.ents of Section IV.b. 3 of  !

Appendix I to 10 CFR Part 5 '.

The requirement for par + cipation in an Interla.+ratory Comparisen Program-is provided t. ensure that independent che ks on the precision and accuracy of the ...easurements of ra:iioactive mate tal in environ-mental media are erformed in order to demonstrate the validity of I, results. -

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TAltLE 3-9' (Continued) f Rn.11ological Fnvi ron:nen tal Ibnit'oring Program (Co"tinued) '

f Exposur 1%thuay Types of Analysis 2

- und/or L pie Collection Sitel ,4 requency- .i Y

. h. Milks ei . trearest family cow when ' avail- ' Carraa Isotopic and I-131 1. Semimonthly l able, or one (1) Dairy farm grazing season within 6 kilometers. (May to October) i

, b. On 1) ' Dairy farm between l- 8 kil ters and 30 kilo-meters AAckground).

1

( 5. Fish- a. Four fish samp -- within Car:n sotopic Once per season I vicinity of plant 's charge (May to October) i

b. One (1) background sarpt

[<go, upstream of plant discharg .

d a

i (f . Sediment One sample'from downst am area arraa Isotopic Semi-annually l on the Statian' side f the Hisvouri River.

4

, 1 See Table 10 of'the'0DCit.

s\

Tise lwer 11rait, of dete lon (LLD) for analysia is defined in the CDCM in kcordance with the wording of c ImHEG-Oh i2,1:ev. 2. \-x

.j a;!

s 1*etails of th- nercency TLD stat. ions are contained in Enercency Preparedness Implementing Procedures.

< [a s 6 N l F. 'ituen u aus betu count indicates radioactivity greater than lE-12 pC1/ml or 1 pCi/m 3, a\ gamma spectral mal'.is will be perferned. \

^

%s

j ' nen ud I L r c;> iec, 'are u. ,t ' uv allable , a broad 1 eat' vecetation sa

/:.ple sha11 be col 1ected monthly vhe

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t l  :$

t a

, ._ . . _ . -- _,_. . _- _ _ . __ . . a

A1mchment ~1 3.0\SURVE!LLANCE REQUIREl:Ef TS , 3.12 hadioloaical Weste Sarnplino and Fonitoring 3.12.1 gquigandGaseousEffluents App)icability Appliekto the sampling, ronitoring, and testirg used for liquid ap'd

/[

gasecus ffluents.

/  !

0)Jective To ensure that radioactive liquid and gaseous releases frem Ahe facility are m'aintained as Icw as redsonably achievable and'within the limits speiVied by Specifications 2.9.1(1) and 2.9.lft).

Sjecificationt (2) Lioutd Effluer.th

a. Radioactive guid waste sampling end apIivity analyses shall be performed i accordance with Table J-11. The results of these analyses ., all be used with the/celculational rethods in l the ODCil to assu q that the concentr6 tion at the point of release is limited' to the values i Specification 2.9.1(1)a.

b. Prior to release of ch batch c5 liquid effluent, the batch

, and ar)41yzed for principal gamm;

$- shallbemixed,sampid(ionalofotherlimitationspreclude emitters. When operat specific gamma radionuclide nalysis of each batch, gross radioactivity measurementh ehall be made to estimate the quantity and concer;tratic \of radioactive materials released in the batch, and a week) simple composited from proportional aliquots from each batc)( relehsed during the week shall be analyzed for the prin pal gam -emitting radionuclides.

c. The overboard heade radiation to itor shall have a:

(i) Source c ck prior to any rb materiasfromthemonitoror}easeofradioactive e hotel waste tanks.

(ii) Quar rly channel functional te (iii) Ch nel calibration at refueling f'requency.

d. The steash generator blowdown radiction monit s shall have:

(1) Daily channel checks.

(1 Monthly source checks.

) 3-69 Amendment flo. 28, l'f, 4

w , ---..,m , -

hD2d iM&M 3.0 SURVEILLANCE REOUIREMENTS 3.12 Radioactive Waste Disposal System Applicability Applies to the instrumentation used to determine hydrogen and oxygen concentrations in the waste gas decay tanks.

Objective To ensure the concentrations of hydrogen and oxygen in the gaseous radioactive waste system are maintained below their Dammability concentrations as required by Specification 2.9.

Snecifications The hydrogen and oxygen monitoring system for the waste gas decay tanks shall have a:

a. daily channel check (when in senice)

b. monthly cross comparison with a grab sample

c. quarterly channel calibratioit using a gas mixture with concentrations in the range of interest Basis The speci5 cation ensures that instrumentation med to determine the concentration of potentially explosive gas mixtures entrained in the gas decay tank (s) will be maintained in an operable condition. Maintaining the instrumentation used to determine hydrogen and oxygen concentration with a surveillance program 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 50.

3-69 Amendment No. 28, 86, 443 (Next Page is 3-76)

3.0 SURVEILLANCE :E001REMENTS 3.12 Pacioleoscei waste Sampi,no anc MonitoM no (Continuec)

-3.12.1 Licuic ano Geseous Effluents (Continuec)

(iii) Quarterly channel functional tests.

(iv) Channel calibration at refueling frequency.

e. /

The steam generator bloadawn effluent flow rate wi}4 be  !

calibrated at refueling frequency and visually de@rmined operable daily. /

f.

Records shall be maintained of the radioactive concentrations and volume before dilution of acn batch of licuid effluent released and of the average ilution flow anc length of time over which each disenar . occurred.

Analytical results shall be suomitted to +.e Commission in accordant e witt. Section 5.9.4.a of these specifications.

(2) Gaseo .s Effluents

a. Ra cactive gaseous waste sampling d activity analyses

' sha M ba performed in accordance w h Table 3-12. The results of these analyses shall b used with the calcu1(tionalmethodsinthe00C to assure that the concente is limit- tion of radioactive maJerials in unrestricted areas to the values i'n Sp4cification 2.9.1(2)a.

b. (i) An Au iliary Building,txhaust Stack monitor shall have [

materi(als from a gaa sourtank decay e check or theprior to any release containment. A of radi monthly S urce che will be performed during refueling tagesj'if a purge or gas decay tank release is not don dur,iAg that month.

(ii) The Auxiliaryj uilding Exhaust Stack gaseous, particulate,/an4 iodine monitors and the Laboratory andRadioacJivaKasteProcessingBuildingExhaust Stackgasedus, have 6 frterly park.ticulate,andiodinemonitorsshall c nel functional test.-

xiliary Buildin (iii) ThepartiAu'ulate,

~ and iodin Exhaust Stack gaseous, and monitors and the Laboratory adioactive Waste P ocessing Building Exhaust Sta k gaseous, particulatt , and iodine monitors shall L

b calibrated at refueling frequency.

j. (iv)- The Auxil_iary Building E haust and the Laboratory and Radioactive Waste Pro ssing Building Exhaust

/- stack flow rates will be e librated s and functionally tested at refue ing frequency. The AuxiliaryBuildingExhausta(ndtheLaboratoryand Radioactive Waste Processing B ilding Exhaust stack radiation monitors flow

\

3-70 Amendment No.86,122,137 l

/

rates will be calibrated ana functionally tested i at refueling frequency. The stack flow rates ano i radiation monitor flow rates will be ceterminec l operable by visual inspection daily.

(v) The Laboratory and Radioactive Waste Process ng Building Exhaust Stack gaseous, particulat and iodine activity monitors shall have a dai.

channel check and a monthly source check

c. The condenser air ejector monitor shall have a:

) Daily channel check.

(ii Montnly source check.

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j 3-70a

, Amendment No.86//I22,137 l e m e 9 n

,,, - . , ,.,.,-.,w-y--y,.-,, . , - -- ----r -#-, p. -w- , 2.-,-.-,,

3. , 50RVElLLPHCE PEOUIREMENTS 3.lc Radiological Waste Sampliric and honitorire (Continued) 3.12.1 Liquid ano Gaseous Effluents (Contirutc) f' (iii) Quarterly channel functional ttst. /

\ (iv) Channel calibration at refueling f requency. / l The hydrogen and exygen renitoring syster for the gas cecp

\tanksshallhavea: /

(i) Daily channel check (when in service). ,/

( 1V.

Honthlycrosscomparisonwithagrabsamp/

(iii) Quarter'y channel calibration using gas mixtures with concentrations in the range of int 6 rest.

\ /

e. Records shall be maintained and reports of thc serpling and results of analyses

\ shall be submitted to the Cernmission in accordance With Section 5.9.4.a of these 'pecifica tions.

Basis N \

\ j Thesurveillancerequireme\tsgivenunderSpecification3.12.1(2) provide assurance that radioactive gaseous effluents f/ rom the station are properly controlled and monitored overgthe life of the station in conformance with the requirements of General Degign Criterid 60 and 64 of 10 CFR Part 50, Appendix A. These surveillance \requiremphts provide the data for the

, licensee and the Commission to ebluate/the performance of the station relative to radioactive gaseous wettes/ released to the environment. The existing minimum sensitivity of airb.orne effluent monitor RM-062 is SE-06 mC1/cc/100 cpm and this minimum sensitivity shall be maintained if the tr.ot tor is replaced. Reports on t,h'e Mantities of the radioactive materials released in gaseous effluentSyshall be furnished to the Comis-sion on the basis of Section 5.9.4.a of these Technical Specifications.

On the basis of such reports e/d I any additional information the Comission may obtain from the license'e j 6r others, thkComission may from time to time require the licensee to take such actido as the Commission deems appropriate.

The surveillance requirements given under Specifdcation 3.12.1(1) provide assurance that licuid dastes are properly contro1{ed and monitored in conformance with theff equirements of General Design Criteria 60 and 64 of 10 CFR Part 50, Appendix A, during any planned release of radioactive materials in liquid effluents. These surveillance r'equirements provide the data for the/1icensee and the Commission to evaluste the station's performance rela'tive to radioactive liquid wastes reledsed to the environ-ment. Reporty'on the quantities of radioactive materials < released in e basis of liquid effluphts Section 5.9/4.a ofshc11 thesebe furnished Technical Specificaticns. to the Comission On the on th' basis of such reports ard any additieral information the Comission may otitain from the licenseej/or others, the Cormission may from time to time requ' ire the licensee to take cuch action as the Comission deems appropriate.

/ 3-71

\

AmendmentNo.Ef,NE,,122

/

,' \\

\

\

.,__ .-h____----_____.,-- -

3.0 SURVEILLANCE REOUIREMENTS I 3.12 ediological ~ te Sampling and Monitorino (Continued) 3.12.2 So d Radioactive Waste Appl bility Applies the sampling, testing, and analysis of the wet radica[tive waste.

Objective  !

To ensure that the solid radioactive wastes meet the limit specified in Section 2.9.2 o these Specifications. 4 Speci fica tions_

/

(1) The Process Cont p1 Program (PCP) shall be used to verify the "

-solidification of 4t least one representative est specimen (drum) from at least every twelf th batch of . t radioactive waste (e.g., evapora r concentrates). 1 A. If any test specime fails to verify solidification, the following actions sh be taken:

(i) Verify solidificati n of all ther drums from the batch under test,

')

(ii)-Review the adequacy of th solidification parameters defined in the PCP and >velop/ verify alternative solidi ficati'. paramet rs, if required, in accordance with the PCP.

m -.

In tha event the s idificati o parameters are altered:

.(a) Select one epresentative dh m from each consecutive batch to rify solidificati until at least 3 consecut e-drums verify soli lancesheduledefinedinSpec\fication. .The surveil- i ' cation 3.12.2(1),

above may be resumed afte,r 3 con ecuti.'e drum; verify soli fication.

(b) - Mo ify the PCP as required and report the changes to

!e NRC in accordance with Specificati n 5.9.4,a. '

. Basis

/

This specifica ion was_ developed .to ensur.e the requirements of 0 CFR Parts 20 and ' fo.r' solid radioactive waste are met. The purpos of placing wet adioactive wastes in a solid, dry form is to limit d1%persion of radioact disposal ntainer've material (drum) before, to thec'uring environs or afinterthe event disposal. of failure of \

These re uire-ments pr ide periodic documentation that solidified wet radioactive sie

-)-

. materia s are in suitable form for transportation and disposal.

x

\

3-71a Amendment No. Ef, 91 \

/.

- , . - , _ - _ - - - ~ ~ - - , - - - - - ~ - ~ ' ~ ~ ~ ~ ~ ' ~ ' " ~ ~ ~ ~ ' ~ ~ "

- _ _ _ - . . - . . - . .- . . . . ~ . -

\ TABLE 3-11 RA010 ACTIVE LIQUID WASTE SAMPLING AND ANALYSIS A. Monit c & Hotel Waste Tanks Releases l l

\ lype of Lt. s . Limit /of

,Samoling Frekvency j tLLD) l i Activity Analysis Detection!pCi (4) ( /ml ) - l Each Batch Principal G6mma Emitters (2)(5) l 5 E-07 j 131(2) l /1.0E-06 Monthly From One Dissolved Noble Gases 1.0 E-05 Batch (Gamma Emitters)

Monthly Composite (1) 4- 3  ! 1.0 E-05 Gr ss a  !, 1.0 E-07 iQuarterly Composite (1) Sr-8 Sr-90 .! l 5.0 E-08 B. Steam Generator Blowdown-

_ Lower Limit of j Type ofs Detection (LLD)

Sampling Frequency Activity Analy'/ sis (4) (pCi/ml)

Veekly Composite (1) Principal Gamm Emitters (5) 5.0 E-07 I-131(6) [ \ 1.0 E-06 Weekly ( ) Dose Equi lent 1-131\ 1.0 E-06

/ _

(GamAa Emitters)

Monthly Dissol/ed Noble Gases \ 1.0 E-05 Monthly Composite (1) H-3!

/

\ \ '

1.0 E-05 G/oss a s 1.0 E-07 Quarterly Composite (1) Sr 89, Sr-90

\ 5.0 E-08

\

NOTES:

\

(1) To betrepresentati e of the avC? age quantities and concentration 9f radioactive mater /als in liquid ef fluents, samples should be collected in proportion to e_ rate of flow of the effluent stream. Prior to ana'i ses, all samples t n for the composite should be mixed in order for the composite sa le to be representative of the average effluent release. s

/ 'y 3-72 Amendment No. 28, 26,122 ,y

\

\

1

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.=. .- - _ - . .-- .-_-- _. .. .- - - _ . __ . - .

TABLE 3-11 .

PAD 10ACT1yE L10010 WASTE SAMPlit:G AND ANALYSIS (Continued) -

NOTES:

(2) Or gross radioactivity as described in Specification 3.12.1(1)b.

(3) When steam generater iodine activity exceeds 50 percent of limitt in Specificatioh 2.20, the sampling and analysis frequency shall ye increased to a minimum b five times per week. When the steam generator iodine activity excee ti 75 percent of this limit, the sarnpling and Analysis frequer.cy shall be increased to e minimum of once per day.

(4) The lower limit c etection (LLD) is defined in the ODC) based en t!UPEG 0472, Rev. 3.

(5) The principal garnma erdtters for which the LLD spec 'ication applies exclusively are the fol'10 wing radionuclides: Hn-54 Fe-59, Co-58, Co-60, 2n-65, Mo-99, Cs-134, Cs1 7. Ce-141 and Ce-144

.(6) A weekly grab sample and analyses program includ ng gem.a isotopic identification will be init1'Ated for the turbipe building sump effluent etite analysis indicates the when 1-131 the steam generator concentration is greater blow'down water comp / microcurie / milliliter.

\than 1.0 E-06 7

t l

\

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3-73 Amendment No. 28, Pf) 122

/ x

/ \

.~ > -

TABLE 3-12 RADIOACTIVE GaSE0tf5 WASTE SAMPt!NG AND ANALYSIS

\

$ampling and Lener Lim 1 of Type of Activity Detectio (LLD)

G6seous , urte Analvsis Ter "ency Analvsis (4) I, i/ mil A. Gas Dec Prior Principal Gamma (5) 1.0 Ep 4(1) i TankRelehses to each release Emitters /

B. Containment Prior Principal Gamma 56) 1 E-04(I)

Purge Release to each release Emitters ./

or Cantainment Prior Pressure Relief to each release H-3

_t.ineDeJ, pates 1.0 E-06 Monthlv (3) Tritium (H-3) 1.0 E-06 C. Condenser air Montnly Principa Gary;a(5)

Eiector De feases \ Emitters / 1.0 E-04(1)

D. ContinuousI2)

Auxiliary Weekly harcoal Building t-d lamplel a 1-131 Laboratory &

1.0 E-12 Radioactive Weste Weekly (2 PrinopalGamma(6) 1.0 E-11 Processing (Particulat ) Emitters 1-131

, Buil;ing Exhaust & rticulates Stack i< leases w th half-lives freater than 8 days _,

Monthlv Composite \ / Gross a 1.0 E-Il Quarterly Composit'a $r-89, Sr-90 1.0 E-11 "

(Particulates) /~ s\ -

EQT[1:

(1) ~ For cer tain mixtures of g a emitters, it may not be possible to measure radionuclides at eyels near the other nuclides are pres t in the sample (a sen3itivity limits when levels. Under these circumstances, i willbemoreappro\muchhigher priate g to calculete the levels of such racion lides using observed eatios with those ,

radionuclides which 're measurable.

N 3-74 Amendment No.86,137 '\

s w -,+ fe -y- , w , w, .pr,.-,,-,A-,g, ,-,,,,,.pn,n,, ,.,-,,-,,,,,,,,,,w,,- 'v..,-,

.- - . ... . . . . - . . -. . . - .-. . . ~ - .~ - - . - . - _ . . - _ ~ ~

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-TABLE 3-12'

- 3 i' I RADI0 ACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS (Continued) -

NOTES:

-i L(2) To be rep. sentative of the average quantities and concentrat! is of. l',

radioactive materials in particulate form released in 6aseous effluents, -!

sample shoul ate of the effluent stre'beand collected the des 16nin proportion f,' rateto the vill design be used flow in /.imating.

ee rele as es . -

l i

. ( 3) Tseyaired cnly whe. steam Generator blovdevn radioactiv y for tritium (Table 3-11 Sectic B) exceeds 3.0E-03 microcurie /mi liliter.

(k) Se- Lover Limit of De etion (LLD) is defined in t e ODCM based on NUPEG Oh72, Eev. 3 .

.\'

(5) Se principal samma emitt rs for which the LLD specification applies f.

exclusively are the follow- 6 radionuclides: Kr-87, Kr-88, Xe-133,  ;

Xe-133m, Xe-135, and Xe-136 or Saseous emi .sions and Mn-5h , Fe-59, Co-56, j

.Co-60, On-65, Mo-99, Cs-13h, s-137, Ce-1k . Ce-1kh for particulate emissions.  %.

7

.F l

/ 3-75 Amendment no. 36

\ ,

5.9.3 loecial Reports s Specialr reports shall ;be_ submitted to the Regional- Administrator of the  ;

1 appropriate' NRC Regional- Office within the time period. specified -for '

each report. These reports shall be submitted covering the activities ,

identified below pursuant to. the requirements of the applicable reference specification.where appropriate: *

a. In-service' inspection report, reference 3.3.

b. Tendon _ surveillance, reference 3.5.

~

c. -Containment structurel tests, reference 3.5.

d. Special maintenance reports.

e.. Containment leak rate tests, reference 3.5.

f. Rad 4 east 4ve effluent-r+1easelb--referenc+-h9e O g. Materials radiation surveillance specimens reports, reference 3.3.

I g -hr Fire' protection equipment outage, referer.ce 2.19.

l

- h-L. Post-accident monitoring instrumentation, reference 2.21. [

5.9.4 Unioue Reportino Recuirements a,

' Radioactive Effluent Release Reoort . ,

~A eport Lovering the operation of the Fort Calhoun Station durJng/

the evious six months shall be submitted within 60 days after M anuar 1 and July 1 of each year per the requirements of O'CFR

@bckveM50.36a.

k LThe radioacti effluent release report shall includ .a summary of the quantities o radioactive liquid and gaseou ffluents and solid waste release from the. plant as outli in Regulatory Guide 1.21, Revision 1. ,

l. The radioective.efflucnt r ease repo hall include a summary of L -tha meteorologica1' condition oncup nt with the release of

!- gaseous ieffluents during each q grter as outlined in Regulatory

Guide 1.21, Revision 1.

L 1

- The- radioactive effluent ease repor hall include an assessment L :of. radiation doses fro he radioactive 11 uid and gaseous

[- . effluents released f m the unit during eac alendar quarter as outlined .in Regul ory Guide 1.21, Revision 1. In-addition, the

-unrestricted a boundary maximum noble gas gamm(air and beta air doses shall evaluated. The meteorological condit4ons -concurrent

=with the

> x l 5-15 Amer.dment No. 9,25,7E,fE,EE,J ,JJE j/ 133 .

' m.a

.,m . . . _ , - . . . _

5.N 9, Unique Reporting Requirements (Continued)

I .q. Radioactive Effluent Release Report (Continued) releases of effluents shall be used for determining the gaseo s

/

pathway doses. The assessment of radiation doses shall be p rformed accordance with the Offsite Dose Calculation Manual (0D ).

U)

Thkradicactive effluent release report shall include any' changes to t e Process Control Program (PCP) or to the Offsite pose Calcula-tion 'nual (ODCM) made during the reporting period. /will ' level of detail ommensuratetothesignificanceofthechang7 be provided,

b. Radiological Environmental Operating Reports

1. Annual Eeport /

N /

Anannualkeportcontainingthedatatakenintheradiological d

environmentgl monitoring program, in a9 ordance with the ODCM, for the previous calendar year of opepation shall be submitted prior to May 4 of each year. The co ' tent of the report shall include:

(a) Summarized a d tabulated re its of the radiological environmntal surveillance activities following the format of Reguiqtory Guid. 4.8, Table 1. In the event that some results( are n t available, the report shall be submitted notihg an explaining the reasons for the I missing results, ige,issingdatashallbesubmitted as soon as possible a supplementary report.

(b) Interpretatior.s a sthtistical evaluation of the results, including -an asgessment of the obterved impacts of the 'lant oper ion on the environment.

(c) The results o particioatio in the Interlaboratory Comparison ogram.

(d) The results of land use survey eqquired by Specifica-tion 3.}((3) .

(e) The r uits of specific activity analysis in which the pri- Specification ry coolant exceeded The following the limits information sha b{l' be includeo-2.1.3.

) Reactor power history starting 48 hyrs prior to the first sample in which the limit was er eeded;

\

(1) Thes ' changes can be initiated either by the licensee (imp \pmenta-ti  : subject to review by the PRC) or by the Commissica ( unple-m tation: subject to their applicability to the Fort Calhoun\

tation design, review by the PRC and followed by a revico by 'the

") -

ARC).

l' / 5-16 Amendment No. P6, 102 \,

\

/

5 V 5. 9. Unioue Reporting Requirements (Continued)

-l'  ;

, Radiological Environmental Operating Reports (Continued)

1. Annual' Report (Continued)

(2) Results of the last isotopic analysis for radioiodine performed prior to exceeding tt i limit, results of analysis while limit wa l exceeded and results of one analysis aft r the radiciodine activity was reduced t less than the limit. Each result should i lude date and time of sampling and the ra ioiodine concentrations;

( Purification system flow history starting 48

- hours prior to the first sample in which the limit was exceeded; (4) G ph of the I-131 concentr ion and one other ra 'oiodine isotope concen ration in microcuries per am as a function of time for the. duration of th specific activit abcVe the steady-state level; nd.

(5) The time ration w n the specific activity of the pr' 1r coolan exceeded the radiciodine

.)'

limit.

2. Non-Routine Report If a confirmed measure radlonuclide concentration in an environmental samplin mediu. average over any calendar quarter sampling pe od excee the reporting level'  !

referenced in Tabi 3-9, footno e 2, and if the radio-activity is attri)r table to plan operation, a written report shall be ubmitted to the emission within 30 days from the end of the quarter.

The report s all-include an evaluatio of any release conditions,/ environmental factors, or b her aspects necessa ry o explain the anomalous resui .

/

l l-if / \

5-17 Anenament No. 9,19,Vf,

\\

-,>1 - . , . e- . .

TCd\MCvd

a. Semiannual Radioactive Effluent Release Report The Semiannual Radioactive Effluent Release Report covering the operation of the unit during the previous 6 months of operation shall be sutaitted 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) consis-tent with the objectives outlined in the 00CM and PCP, and 2) in conformance with 10 CFR 50.36a. and Section III.B.1 of Appendix I to 10 CFR 50.

b. Annual Radiolooical Environmental Ooeratino Reoort 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 c"."sistent with the objectives outlined in (1) the 00CM and (i, Ser on IV.B.2, IV.B.3, and.IV.C of the Appendix I to 10 CFR 50.

5.0 ADMINISTRATIVE CONTROLS _

- 5._10.2 _ The following records shall be retajned for _ the duration of the Facility Operating License:

a. Records of drawing changes reflecting facility 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 facility radiation and contamination surveys.

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

e. Records _ of gaseous and liquid radioactive material released to the environs.

f. Records of transient or operational cycles for those facility com-ponents designed for a limited number of transients or cycles,

g. Records _ of training and qualification for current members of the plant staff.

h. Records of in-service inspections perfomed pursuant to these Technical Specifications.

i. Records of Quality Assurance activitios required by the QA Manual.

j. Records.of reviews perfomed for changes made to procedures or equip-

-) ment or reviews .of tests and experiments pursuant to 10 CFR 50.59.

k. Records of meetings of the Plant Review Comittee and-the Safety Audit

.and Review Comittee.

1. -Records of Environmental- Qualification of Electric Equipment pursuant to 10'CFR-50.49.

m. Records of tN service lives of all hydraulic and mechanical snubbers which are covered under the provisions of Section-2.18 of the Technical Specifications, including the date at which the service life commences

-and associated installation and maintenance records.

n. . Records of analyses required by'the :.adiological Environmental Monitoring Program.

D-5.10.3 A complete-record of the analysis employed in the selection of any fuel assembly to be placed in Region 2-of the spent fuel racks will be retained as long as that bundle remains in Region 2 (reitrence Technical Specifica-tions 2.8(12) and 4.8.4). '

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

o. 9ecwth o4 redem wcGemec4 k eM eg vvrde to %e O&iTe bne CAleMak -19WW cmY 'b mces CcrWTrol Pro re.

Ordet 70/24/E0, Amendment No. EP,Et ,93g,5 9,1 a:

l 20,Ilcol @s)b)[2)(N r-- cmCA 4 5 #t o "lDIVd* M' 5.0 40MINISTPATIVE r wip0LS a(IcctvCd ancl K do I(C l'It )

i 5.11.1 in lieu of the  : o n t ro l d e v i c e " oe-4 a re--s49nal " r e qu i red by paragraph &203M)tWof 10 CFR 20,Teacn high radiation area (as defined in @ 20-2p2MW-of 10 CFR 20) in which the intensity of

,~20. 4 ra iation is 1000 mrem /hr or less shall be barricaded and e sp1cucusly posted as a high radiation area 2r<1 antrance theretc shall be controlled by required issuance of a Radiation Work Permit.*

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

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 cevice may be made after the dose rate level in the area has been established and personnel have been made knowledgeable of them.

c. An individual qualifieo in radiation protection crocecures who is equipped with a raciation dose rate monitoring device.

Individual shall be responsible for providing positive control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified by the Supervisor-Radiation Protection in the Radiation Work Permit. 1 Red'et6ed t-. e nvy 5.11.2 The requicementd of 5.11.1. abcve, shall als6 apply to each high radiationarea{nwnichtheintensityofrad iation is greater than 1000 mrem /hr (Very-High Radiation Area). In addition, locked doors shall be provided to prevent unauthorized enter-into such areas and the keys shall be maintained under the administrative control of the Shift Supervisor on duty and/or the Supervisor-Radiation Protection i with the following exception:

y(esici6ect

3. In lieu of the aoove, for accessible localizedJery High Raciation areas located in large areas such as containment, where no lockable enclosure exists in the immediate vicinity of-4he-

"ery Hsh-Rad 4at-4cn-area. to control access to the -Very- High Radiation area and no such enclosure can be readily :cnstructed, then the -Vefy High Radiation area shall be:

L--- asTricT<d - ReGrded

1. roped off such that an individdal at the rope boundary is exposed to 1000 mrem /hr or less, ii. conspicuously posted, and iii. a flashing light shall be activated as a warning device.

"Rac:ation Protection cersonnel shall be exempt from the RWP issuance recuirement ':uring the performance of their assigreo radiation orotection dut:es. provided they comply with approved rac:ation orctection procecures 'cr entry into nign radiation areas.

i-19a caencment No. 22. 61. 22 1

5.0 ADMINISTRATIVE CONTROLS 5.16 Radiological Effluents and Environmental Monitorine Programs The following programs shall be established, implemented, and maintained.

5.16.1 Radioactive Effluent Controls Program A program shall be provided conforming with 10 CFR 50.36a for the control of radioactive effluents and for maintaining the doses to individuals in unrestricted areas from radioactive efnuents 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:

a. Limitations on the operability of radioactive liquid and gaseous radiation monitoring instrumentation including operability tests and setpoint determination in accordance with the methodology in the ODCM.

b. Limitations on the concentration of radioective material released in liquid effluents to unrestricted areas conforming to 10 CFR Part 20.

c. MUnitoring, sampling, and analysis of radioactive liquid and gaseous effluents in accordance with 10 CFR Part 20 and with the methodology and parameters in the ODCM.

d. Limitations on the annual and quarterly doses or dose commitment to individuals in unrestricted areas from radioactive materials in liquid effluents released to unrestricted areas conforming to Appendix I to 10 CFR Part 50.

e. Determination of cumulative doses from radioactive effluents for the current calendar quarter and current calendar year in accordance with the ODCM on a quarterly basis.

5-22

.a

5.0 ADMINISTRATIVE CONTROLS 5.16 Radiological Efnuents and Environmental Monitoring Programs (continued)

f. 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 in plant effluents.

g. Limitatiom on the concentration resulting from radioactive material released in gaseous effluents to unrestricted areas conforming to 10 CFR ,

Part 20.

h. Limitations on the annual and quarterly air doses resulting from noble gases released in gaseous efauents to unrestricted areas conforming to Appendix I to 10 CFR Part 50.

i. Limitations on the annual and quaderly doses beyond the site boundary from Iodine-131, tritium, and all radionuclides in ptrticulate form with half lives greater than 8 days in gaseous effluents released to unrestricted areas conforming to Appendix I to 10 CFR Pr.rt 50.

5.16.2 Radiological Environmental Monitoring Frogram A program shall be provided to monitor the radiation and radionuclides in the environs of the plant. The program shall provide (1) representative measurements of radioactivity in the highest potential exposure pathways, and (2) verification of the accuracy of the effluent runitoring 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:

a. Monitoring, sampling, analysis, and reporting of radiation and radionuclides in the environment in accordance with the methodology and parameters in the ODCM.

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

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

5-23

l 5.0 ADMINISTRATIVE CONTROLS 5.17 Q{f3jte Dose Calculation Manual (ODCM)

Changes to the ODCM:

a. Shall be documented and records of reviews performed shall be retained as required by Specification 5.10.2.o. This documentation shall contain:

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

2. A determination that the change will maintain the level of radioactive effluent control requirted by 10 CFR Part 20,10 CFR Part 50.36a, and Appendix I to 10 CFR Part 50 and not adversely impact the accuracy or reliability or effluent, dose, or setpoint calculations.

b. Shall become effective after review by the Plant Review Committee and the Manager - Fort Calhoun Station,

c. Temporary changes to the ODCM may be made in accordance with Technical Specification 5.8.3.

d. Shall be submitted to the Nuclear Regulatory Commission in the form of a complete, legible copy of the entire ODCM as a part of o; concurrent with the 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 the area of the page that was changed and shall indicate the date (e.g., month / year) the change was implemented.

5-24

4 l

5.0 . ADMINISTRATIVE CONTROLS 5.18 Process Control Procram (PCP)

Changes to the PCP:

a. Shall be documented and records of reviews performed shall be retained as required by Specification 5.10.2.o. This documentation shall contair..

l

1. Suf0cient informr. tion to support the change together with the appropriate i analyses or evaluations justifying the change (s) and  !

l

2. A determination that tne change will mainta'.. the overall conformance of l the solidified waste program to existing requirements of federal, state, or other applicable regulations.

b. Shall become effective after review and acceptance by the PRC and the approval of the Plant Manager.

c. Temporary changes to the PCP may be made in accordance with Technical Specification 5.8.3.

d. Shall be submitted to the Nuclear Regulatory Commission ' the form of a cornplete, legible copy of the entire PCP as a part of or concurrent with the

, Semiannual Radioactive Effluent Release Report for he period of the report in

! which any change to the PCP was made. Each chaiige shall be identified by j markings in the margin of the affected pages, clearly indicating the area of the l page that was chaliged and shall indicate the date (e.g., month / year) the change l

was implemented.

L l

(

l l

l 1

y i

5-25 l

V

, . . . , - - . . . . ~ .-- - . ~ - . ~ . _

TECHNICAL SPECIFICATIONS .. ,

TABLE OF CONTENTS Page DEFINITIONS . . . . .. .............................. ......... ........ 1 1.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS . . .. . . . . . . . 1-1

-- 1.1 -. - Safety Limits - Reactor Core .. .................... ..............11 1.2 Safety Limit, Reator Coolant System Pressure . . . . . . . . . . . . . . . . , . . . . . . . 1 -4 1.3 - Limiting Safety System Settings, Reactor Protective System ......... ,.. . . . . 1 -6 2.0. LIMITING CONDITIONS FOR OPERATION 2-0 2.0.1 General Requirements ...........,................... . . . 2-0 2.1 - Reactor Coolant. System ... ........... ......... , ......... .. '!.1

-2.1.1 Operable Components ................. .. ....... . ., . .2-1 2.1.2 Heatup and Cooldown Rate . . . . . . . . . ............... . . . . . . . 2-3 2.1.3 Reactor Coolant Radioactivity . . . . . . . ......... . ............ 2-8 2.1.4 Reactor Coolan ::ystem I rabge Limits . . ... .............. . 2-11 2.1.5_ Maximum Reactor Coolant Oxygen and Halogens Concentrations . . . . . . . . . 213 2.1.o Pressurizer and Steam System Safety Valves ............. .... .. . 2-15 2.1.7 Pressurizer Operability . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . 2-16a 2.1.8 ; Reactor Coolant System Vents . . . . . .. .. . . ...... . . . 2-16b 2.2 Chemical and Volume Contrcl Systera . . . . . . ......... . . . . . . . . . . . . . 2 17 2.3 Emergency Core Conting System . . . . ................ . . . . . . . . . . . . 2-20 2.4 - . Containment Cooling . . . . . . . . . . . ....... ................ . . . 2-24

- 2.5 Steam and Feedwater Systems . . . . . . . . . ............. ........... . 2-28 2.6 ' Containment System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-30 2,7: Electrical Systems . . . . . ....................................2-32 2.S. Refueling Operations ................. .... . . . . . . . . . . . . . . . . . . 2-3 7 2.9 'tadioactive Waste Disposal System ................................

2-40 l 2.10_ Reactor Core . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 -4 8 .

2.10.1 . Minimum Conditions for Criticality . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-48

.2.10.2 Reactivity Control Systems and Core Physics Parameter Limits . . . . . . . . . . . . , . . . , . . . . . . . . . . , . . . . 2-50 J 2.10.3 - In-Core Instrumentation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 54 2.10.41 Power Distribution Limits . . . . . . . . . ............... ...... 2 56 2.11- Containment Building and Fuel Storage Building Crane . . . . . . . . . . . . . . . . . . . . 2-58 i

i Amendment No. 42.38,52,64,47,67,  ;

80,84,86

=

TABLE OF CONTENTS (Contmued) -.

b5:

2.12 Controt Room Systems . i . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 5 9 2.13 l Nuclear Detector Cooling System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-60 2.14. Engineered Safety Features System Initiation -

~ Instrumentation Settings . . . . . . . . . . . . . . ... . . . . . . . . . . . . . . . . . . . . 2 -61 2.15 Instrumentation and Control Systems . . . . . . . . . . . . . ........... . . . . . . 2-65 2.16 Rive r Level . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 -71 2.17 - Miscellaneous Radioactive' Material Sources . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-72

'2.18 Shock Suppressors (Snubbers) . . . . . . .............................2-73 2.19- _ Fire Protection System . . . . . . . . . . . . . . . . . . . . . . . . . . . . ........... 2-89 2.20 Steam Generator Coolant Radioactivity . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-96 2,21- _ Post Accident Monitoring Instrumentation .................. . . . . . . . . 2 97 2.22 -- Toxic Gas Monitors . i . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .- 2 -99 '

3.0 SURVEILLANCE REQUIREMENTS 3 0a

3. l' , Instmeentation and Control . . . . . . . . . . . .................... ....31 3.2 = Equipment and Sampling Tests . . . . . . . .............. .. .. ....... 3-17 3.3 Reactor Coolant System and Other Components Subject to)SME XI Boiler and Pre:;sure Vessel Code Inspection and Testing Surveillance . . . . . . . . . . . . . . . . .. .. 3 21 3.4 Reactor Coolant System Integrity Testing . . . . . . . . . . . ......... ..... .. 3-36 3.5 Containment Test ... ..... ........................ .. . . . . 3 37 3.6 . Safety injection and Containment Cooling Systenu Tests ' . . . . . . . . . . ..... . . . 3-54 3.7 Emergency Power System Periodic Tests . . ... . . . .. . . . . . . . . . . . . . 3 -5 8 3.8 Main Steam Isolation Valves ........................ . . . . . . . . . . 3 -61 3.9- Auxiliary Feedwater System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .... . 3-62 3.10- Reactor Core Parameters - . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 63 l 3.11 DELETED 3.12 Radioactive Waste Disposal System ........... .......... ......... 3 -69 3.13 Radioactive Material Sources Surveillance ............,...............376 3.14- Shock Supprewors (Snubbers) . . . . . . . . . . . . ............. . . . . . . . . . 3 77

+

3.15 1 Fire Protection Systems . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 80 -

3.16 . Recirculation Hen Removal System Integrity Testing . . .................. 3-84 3.17 Steam Generator Tubes . . . . . . ............ . . . . . . . . . . . . . . . . . . . 3 86

-4.0 DESIGN FEATURES 41 q

4. I ' Si te . . . . . . . . . . .- ... ........ ,,....... .... .............41

-- Containment Design Features . . . . . . . . . . . . . . . . . . . . . . . . . . . . ......

^

_ 4.2 . 41-4.2.1 Centainnnat Structure . . . . . . . . . . . . . . . . . ' . . . . . . . . . . . . . . . . . . . 4 1 4.2.2 Penetrations ........ ... ............. ........ ....41 .

,4.2.3 Containment Structure Cooling Systems .. .. ... .. .. ... ... ........ 4-2 f

is Amendment No. 84,%,93,M4-G2,136

__ . .--_ .- _ u. .u_ _ - _ . . .,

TABLE OF CONTENTS (Continued)

Butt 4.3 Nuclear Steam Supply System (NSSS) . ... .. ........ ..... , . . . . . .4-3 4.3.1 Reactor Coolant System . . . . . . . . . . . . . ........ .. . . . . . .43

.4.3.2 Reactor Core and Control . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .4-3 4.3.3 Emergency Core Cooling .. ... . ... . .......... . . . . .43 4.4 Fuel Storage ........... ........ . ............ . . . . . 44 4.4.1 New Fuel Stor.ge . ..... . . .... ................... . . 4-4 4.4.2 Spent Fuel Storage .... . . .. .. ..... . . . . . . . . . . . . 4-4 4.5 Seismic Design for Class I Systems . . ... ........ . . . . . . . . .45 5.0 AD5flNISTRATIVE CONTROI.S 51 5.1 Responsibility . . . . . . . . . . . . . . . . . . . . ..... . . . ... . . .. . 51 5.2 Organization ... . . ........ .... .. ... ..... . . . . . 5-1 5.3 Facility Staff Qualifications ....... ..................... . . . . . . 5 la 5.4 Training . . . . . ....... .. .... . . . . ....... ... . . . . . 5-3 5.5 Review and Audit .... . . ... . . . . . . . . . 5-3 5.5.1 Plant Review Committee (PRC) .... . . .. .. ....... . 5-3 5.5.2 Safety Audit and Review Committee (SARC) . .. . . . . . .5-5 5.5.3 Fire Protection Inspection . . . . ..... .... . .. ........ . . 5-Ba 5.6 Reportable Event Action ...... ... .. . .... . .. .. . . . . .. . 59 5.7 Safety Limit Violation .. .... .. . ........... ... . . . . . . 5-9 5.8 Procedures .............. .... .... ........ ... . . . . . . . 5-9

. 5.9 Reporting Requirements .. ... .. .. ... ... .. ... . . . . . . . 5-10 5.9.1. Routine Reports . . . ........ ... ..... .. .... . . . . . . . 5-10 5.9.2 Reportable Events . . .......... ...... ....... .. . . . . 5-12 5.9.3 Special Reports ........... ... ... ....... ... . . . . . . .5-15 5.9.4 Unique Reporting Requirements . . . . . . . . . . . ...... . . . . . . . . . 5-15 5.9.5 Core Operating Limits Report . .. ..... ........ . . . . . . . 5-17a 5.10 Records Retention . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ........ .5-18 5.11 Radiation Protection Program . . . . . . . . ...... ........ .. . . . . . . . . 5 19 5.12 DELETED

^ 13

. Secondary Water Chemistry . . . . . ... ......... ... .... . . . . . 5-20 5.14 Systems Integrity . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-21 5.15 - Post-Accident Radiological Sampling and hionitoring . ...................521 5.16 Radiological Effluents and Environmental Monitoring Programs . . . . ..... . . . 5-22 5.16.1 Radioactive Effluent Controls Program .... . .. . . . . . . . . . . . . . . . 5 -2 2 5.16.2 Radiological Environmental Monitoring Program ... ...... . . . . . . . . 5-23 5.17 Offsite Dose Calculation Manual (ODCM) . ........... .... . . . . . 5-25 5.18 Process Control Program (PCP) .......... .......... ........ . . . 5-26 6.0 INTERIM SPECIAL TEC11NICAL SPECIFICATIONS . . . . .... ...... . . . . . . 6-1 6.1 Limits on Reactor Coolant Pump Operation .. .... .. .... . . . . . . . 6-1 6.2 Use of a Spent Fuel Shipping Cask . . . . ......... .... .. . . , , . .. . 6-1 6.3 - Auxiliary Feedwater Automatic Initiation Setpoint . ... .. . . . . . . 6-1 6.4 Operation With Less Than 75% ofIncore Detector Strings Operable . .. .. .... . .. . . ... . . . . 6-1 iii Amendment No. 12,21,??,51,55,57, "f-- 1~ C

- . A. , o f ,n1. ,n. n, ,1aAv&

1

_ . . .. . - . s TECllSlCA14 SPEflEICATIONS TABLES TAllt.E OF CONTENTS

'IABLE DESCRll'rION PAGE 3 Minimurn Frequenci.s for Checks, Calibrations, and Testing of Miscellaneous Instrumentation and Controls . ..... . .... . .. . . . 3-13

... ...... . .. .. .. .. 3-14

.. . .. , .. .. . 3-15

........... ..... . ....... 3-16

........... .. . -. . . 3-16a

... ... . .. .. . . . . 3-16b

. .. . .. . ,, ... . 3-16c 3-3a Minimum Frequency for Checks, Calibrations and Functional Testing of Alternate Shutdown Panels (A1.l?5 and Al-212) and Emergency Auxiliary Feedwater Panel (A1 179) Instrumentation and Control Circuits . . ..... .... .. ........ ... . .... . . . 3 16d

.... ...... ..... .. .. . . .... . 3-16e 3-4 Minimum Frequencies for Sampling Tests .......... ,..... ., .. . . . . 3 18

.... .. .... . . . . . 3-19 3-5 Minimum Frequencies for Equipment Tests ...... ..... ... .... . . . 3-20

.. .. .. . .. . ,.. . . 3-20a

. . . ... . . . .. . . 3-20b

. .. .. . . .. .. .... . . 3-20e

........... .. ..... ... ... 3-20d 3-6 Reactor Coolant Pump Surveillance . .. . . ... . . . 3-27 3-13 Steam Generator Tube Inspection ,, . . . . .. . .. . 3-90 5.2-1 Minimum Shift Crew Coraposition .. ...... ... . .. .. .. .. 5-2 v Amendment No. MGd25,142

DEFINITIONS Azimuthal Power Tilt - T, Azimuthal Power Tiit shall be the maximum difference between the power generated in any core quadrant (upper or lower) and the average power of all quadrants in that axial half (upper or lower) of the core divided by the average power of all quadrants in that axial half (upper or lower) of the core.

Unrodded Planer Radial Pejtking Factor - F.,

The Untodded Planar Radial Peaking Factor is se maximum ratio of the peak to average power density of the individual fuel rods in any of the unrodded horizontal planes, excluding azimuthal tilt, T, The maximum F,, limit is provided in the Core Operating Limits Report. l Unrodded Integrated Radial Peaking Factor - Fa The Unrodded Integrated Radial Perimg Factor is the ratio of the peak pin power to the average pin power in an unrodded core, excluding azimuthal tilt, T,. The maximum F, limit is provided in the Core Operating Limits Report.

Fire Suppression Water System The fire suppression water system consists of fire pumps and distribution piping with associated sectionalizing control or isolation va'ves. Such valves include yard hydrant curb valves, and the first valve ahead of the water flow alarm device on each sprinkler, hose standpipe or spray system riser.

Process Control Program (PCP)

The document (s) that contains the current formulas, sampling, analyses, tests, 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 20, 61, 71, State Regulations, burial ground requirements, and other requirements governing the disposal of solid radioactive waste.

Dose Equivalent I-131 That concentration of I-131 ( Ci/gm) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132,1-133, I-134 and I-135 actually present. In other

words, 7 Amendment No. 3h,3&S746,141

DEFINITIONS Dose Equivalent 1-131 (pCi/gm) = pCi/gm of I-131

+ 0.0361 x pCilgm of I-132

+ 0.270 x pCi/gm of I-133 4 0.0169 x Ci/gm of I-134

+ 0.0838 x Ci/gm of I-135 h - Average Disintegration Energy 5 is the average (weighted in proportion to the concentration of each radionuclide in the reactor coolant at the time of sampling) of the sum of the average beta and gamma energies per disintegration, in MEV, for isotopes, other than iodines, with half lives greater than 15 minutes making up at least 95% of the total non-iodine radioactivity in the coolant.

Offsite Dose Calculation Manual (ODCM)

The document (s) that contain the methodology and parameters used in the calculations of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous andg liquid effluent radiation monitoring Warn /High W darm) setpoints, and in the conduct of the Environmental Radiological Monitoring Program. ihe ODCM shall also contain: '

1) The Radiological Effluent Controls and the Radiological Environmental Monitoring Program required by Specification 5.16.

2) Descriptions of the information that should be included in the Annual Radiological Environmental Operating Report and Semiannual Radioactive Effluent Rc! ease Reports required by Specifications 5.9.4.a and 5.9.4.b.

Unrestricted Are.a Any area a_t_or beyond thq2i_Lc boundary access to which is not controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials.

Core Operating Limits Report (COLR)

The Core Operating . Limits Report (COLR) is a-Fort Callioun Station Unit No.1 specific document that provides core operating limits for die current operating cycle. These cycle-specific core operating limits shall be determined for each reload cycle in accordance with Section 5.9.5. Plani operation within these operating limits is addressed in the individual specifications.

References l -- (1) USAR, Section 7.2 l {2) .USAR, Section 7.3 8 Amendment .No. 6h86,141 l

.m . _ ._ . _ _ .. __._ ,_ _. . _ ._. . - _ -. - _ _ _ ._._ _ . _ _ _ _ _

. 2.0 - : LIMITING CONDITIONS FOR OPERAUSN -

f2.1; ' Reactor Coolant System (Con:Inued) 12 L3 Reactor Ccolant Padicactivity Apolicability Applies to the radioactivity of the reactnr coolant.

Dbs;1in To ensure that the reactor coolant radioactivity is maintained at a level commensurate with the occupational and public safety.

Snecification

-(1) The radioactivity of the reactor coolant shall be limited to:

a. I 1.0 Ci/gm DOSE EQUIVALENT I-131, and

b. I 100/s pCi/gm (2) _ With the rac'ioactivity of the reactor coolant > 1.0 pCi/gm DOSE EQUIVALENT -

I-131 for mare than 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> during one continuous time interval or exceeding 60 pCi/gm, be in at least HOT SHUTDOWN with T ,, <536 F within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

t (3) With the radioactivity of the reactor coolant > 100/5 Ci/gm, be in at least HOT

. SHUTDOWN.with T,,,- < 536 F within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

-(4). With the radioactivity of the reactor coolant > 1.0 pCi/b m DOSE EQUIVALENT

I-131, perform the sampling and analysis requirements of items-1.(a)(2)(ii) and 1.(b)(2)(i) of Table 3-4 until the radioactivity of the reactor coolant is restored to within its limits. Data pursuant to Specification 5.9.4.h for the Annual Report l;

[

o lshall be compiled as follows:

a. Reactor power history starting 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> prior to the first sample in which y the limit was exceeded.-

b. Purification System flow history starting 48' hours prior to the first sample L

-in which the limit was exceeded.

L c.- The time duration when the radioactivity of the reactor coolant exceeded -

1.0.pCi/gm DOSE EQUIVALENT I-131, p d. . Results of the last isotopic analysis for radiciodine performed prior to L 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 should contain the date and time of sampling and the radiciodine concentrations.-

l~

2-8 AmerAment No. 28,67,102-a .- ... ,

2.0 ~ -LIMITING CONDITIONS FOR OPERATION 2.8- -Refueling Ooerations

- Applicability -  ;

Applies to operating limitations during refueling operations.

Obiective To minimize the possibility of an accident occurring during refueling operations that

__could affect public health and safety.

Soecifications ,

The following conditions sha!! be satisfied during any refueling operations: -

(1) The equipment hatch and one doar in the air lock shall be properly closed. _ In _

addition, all automatic contamment isolation valves shall be operable or at least one valve in each line shall be closed.

(2) One containment atmosphere gaseous radiation: monitor and =one Auxiliary Building Exhaust-Stack gasrus radiation monitor that initiate closure of the contain. ment pressure relief, .or sample, and purge system valves shall be tested and verified to be operable immediately prior to refueling operations. The two monitors shall employ one-out-of-two logic from separate contact outputs for VIAS.

1_, (3) Radiation levels- in the containment and ~ spent fuel strage areas shall _be -

- monitored continuously.

F (4); Whenever core geometry is being changed, neutron flux shall be continuously _

monitored by at least two source range neutron monitors, with each monitor-providing continuous visual indication in the control room. When core geometry is not being changed, at least one source range neutron monitor shall be in service; o

(5): At least one shutdown cooling pump and heat exchanger shall be in operation.

-However, the pump and heat exchanger may be removed from operation for up

~ to one hour _ per 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> period during the perforPance of core alterations in the

. vicinity of the reactor coolant hot leg loops or during' manipulation of a source.

p _

d

e N

l 2 Amendment No. 25,56,133

-- , -N>,Ac, -- ,,..e. . , , - , L mn , , - w - -- -. ,-n-~ --------~ ,-.,t~e s n L-

2.0 IJMITING CONDITIONS FOR OPERATION 2.8 Refueline Operations (Continued)

(6) Direct communication between personnel in the control room and at the refueling machine shall be available wl enever changes in core geome'g are ta.king place.

(7) When irradiated fuel is being handled in the auxiliary building, the exhaust ventilation from the spent fuel pool area will be diverted through the charcoal filter.

(8) Prior to initial con loading and prior to refueling operations, a complete check out, including a load test, shall be conducted on fuel handling cranes that will be required during the refueling operation to handle spent fuel assemblies.

(9) A minimum of 23 feet of water above the top of the reactor core shall be maintained whenever irradiated fuel is being handled.

(10) Storage in Region I and Region 2 of the spent fuct racks shall be restricted to fuel assemblies having initial enrichment less or equal to 4.0 weight percent of U-235.

(11) Storage in Region 2 of the spent fuel racks shall be restricted to those assemblics whose parameters fall within the " acceptable" region of Figure 2-10.

If any of the above conditions are not met, all refueling operations shall cease immediately, work shall be initiated to satisfy the required conditions, and no operations that may change the reactivity of the core shall be made.

A spent fuel assembly may be transferred directly from the reactor core to the spent fuel pool Region 2 provided the independent verification of assembly burnups has been completed and the assembly burnup meets the acceptance criteria identified in Technical Specification Figure 2-10.

Movement ofirradiated fuel from the reactor core shall not be initiated before the reactor core has been suberitical for a minimum of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> if the reactor has been operated at ,

power levels in excess of 2% rated power.

Bases The equipment and general procedures to be utilized during refueling operations are discussed in the USAR. Detailed instructions, the abu e .,pecifications, and the design of the fuel handling equipment incorporating built-in interlocks and safety features provide assurance that no l

2-38 Amendment No. 5,24,25,43,-75,133

_~. _ . ._

2.0 LIMITING CONDITIONS FOR OPERATION 2.9 Radioactive Waste Discosal System Applicability Applies to the transfer of waste gases to the waste gas decay tanks. The provisions of Technical Specification 2.G.1 for Limiting Condition for Operation are not applicable.

Dbiective To ensure compliance with General Design Criterion 60 of Appendix A to 10 CFR 50.

Soecification (1) The concentration of hydrogen and oxygen in the waste gas decay tanks shall be limited to below flammability concentrations. With hydrogen and oxygen concentrations above flammability concentrations, restore the concentrations to below flammability limits within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

(2) The hydrogen and oxygen monitors shall be mocitoring the inservice gas decay tank during the transfer of waste gases to the waste gas decay tank. Whenever the monitors are inoperable, transfer of waste gases to a gas decay tank may continue provided grab samples are taken from the gas decay tank and analyzed:

a. Every eight hours during degassing operations, and

b. Daily during other operations.

hasb Specification 2.9 ensures that the concentration of potentially explosive gas mixtures entrained in the gas decay tank (s) will be maintained below the flammability limits of hydrogen and oxygen. Maintaining the concentration of hyd' gen and oxygen below their flammability limits with a measurement program 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.

2-40 Amendment No. 86;l13 (Next Page is 2-48)

2.0 LIMrrING CONDITIONS FOR OPERATION I 2.14 Engineered Safety Features System Initiation Instrumentation Settings Apolicability Applies to the engineered safety features system initiation instrumentation settings.

Objective To provide for automatic initiation of the engineered safety features in the event that principal process variable limits are exceeded.

Soccifications The engineered safety features system initiation instrumentation setting limits shall be as stated in Table 2-1 B1Si3 (1) Hich Containment Pressure The basis for the 5 psig setpoint for the high pressure signal is to establish a setting which would be exceeded quickly in the event of a DBA, cover a spectrum of break sizes, and yet be far enough above normal operation maximum internal pressure to prevent spurious initiation.

High containment pressure initiates the steam generator isolation signal which will close the main steam isolation and bypass valves and the main ft.edwater isolation and bypass valves.

(2) Pressurizer Low Pressure The pressurizer low pressure safety injection signal is a diverse signal to the high containment pressure safety injection signal. The 1600 psia setting includes an uncertainty ofi 22 psia and is the setting used in the safety analysis.*

l (3) Containment High Radiation (Air Monitoring)

The containment radiation high signal can be initiated by a containment atmosphere gaseous radiation monitor or an Auxiliary Building Exhaust Stack gaseous radiatioti monitor.m l

l 2-61 Amendment No.108 i

2.0 LIMITING CONDITIONS FOR OPERATION 2.14 Engineered Safety Features System Initiation Instrumentation Settings (Continued)

(3) Containment Hich Radiation (Air Monitoring)(Continued)

The setpoints for the isolation function will be calculated in accordance with the ODCM.

(4) Low Steam Generator Pressus A signal is provided upon sensing a low pressure in a steam generator to close the main steam isolation valves in order to minimize the temperature reduction in the reactor coolant system with resultant loss of water level and possible addition of reactivity. The setting of 500 psia includes a 122 psi uncertainty and was the setting used in the safety analysis.*

Closure of the MSIVs (and the bypass valves, rJong with main feedwater isolation and bypass valves) is accomplished by the steam generator isolation signal which is a logical combination of low steam generator pressure or high containment pressure.

As part of the AFW actuation logic, a separate signal is provided to terminate flow to a steam generator upon sensing a low pressure in that steam generator if the other steam generator pressure is greater than the pressure setting. This is done to minimize the temperature reduction in the reactor coolant system in the event of a main steam-line break. The setting of 466.7 psia includes a +31.7 psi uncertainty; therefore, a setting of 435 psia was used in the safety analysis.

(5) SIRW Tank Low Level .

Level switches are provided on the SIRW tank to actuate the valves in the safety injection pump suction lines in such a manner so as to switch the water supply from the SIRW tank to the containment sump for a recirculation mode of operation after a period of approximately 24 minutes following a safety injection signal. The switch-over point of 16 inches above tank bottom is set to prevent the pumps from running dry during the 10 seconds required to stroke the valves and to hold in reserve approximately 28,000 gallons of water of at least the refueling boron concentration. The FSAR loss of coolant accident analysis

• assumed the recirculation started when the minimum usable volume of 283,000 gallons had been pumped from the tank.

2-62 Amendment No. 5,32,43,65,86,403dOBr133,141

~ .

2.0 LIMITING CONDITIONS FOR OPERATIQN 2.14 Engineered Safety Features System initiationJnstrumentation Settings (Continued)

(6) Low Steam Generator Water Level As part of the AFW actuation logic, a signal is provided to initiate AFW flow to one or two steam generators upon sensing a low water level in the steam generator (s) if the absolute steam generator pressure criteria are satisfied. This function ensures adequate steam generator water level is maintained in the event of a failure to deliver main feedwater to either steam generator. The setting of 28.2% of wide range tap span includes a + 13.2% uncertaimy; therefore, a setting of 15 % of wide range tap span was used in the safety analysis.

(7) Hich Steam Generator Delta Pressure As part of the AFW logic, a high steam generator differential pressure signal is generated to provide AFW to the higher pressure steam generator with a concurrent low level signal if both steam generator pressures c '

less than 466.7 psia. If the differential pressure between stmm genere .s is less than the setting, neither steam generator is supplied with AFW in the presence of a low level signal. The setting of 119.7 psid includes a -

15.3 psi uncertainty; therefore, a setting of 135 psid was used in the AIM safety analysis.

References (1) USAR, Section 14.1.3 (2) US AR, Section 7.3.2.6 l (3) USAR, Section 14.12 (4) USAR, Section 14.15 (5) USAR, Section 7.4.6 (6) USAR, Section 7.5.2.5 (7) USAR, Section 14.4.1 2-63 Amendment No. 65,108

2.0 LIMITING CONDIILQNS FOR OPERATION 2.15 Instrumentation and Control Systems (Continued)

(5) In the event that any of the following Emergency Auxiliary Feedwater Panel instrumentation or control circuits become inoperable, either restore the inoperable component (s) to operable status within seven days, or be in hot shutdown within the next twelve hours. This specification is applicable in Modes 1 and 2.

Steam Generator Level, Wide Range (AI 179)

Steam Generator Level, Narrow Range (AI-179)

Steam Generator Pressure (AI-179)

Pressurizer Pressure (AI-179)

Balls During plant operation, the complete instrumentation systems will normally be in service. Reactor safety is provided by the reactor protection system, which automatically initiates appropriate action to prevent exceeding established limits. Safety is not compromised, however, by continuing operating with certain instrumentation channels out of service since provisions were made for this in the plant design. This speci'ication outlines limiting conditions for operation necessary to preser,'e the effectiveness of the reactor control and protection system when any one or more of the channels are out of service.

All reactor protection and almost all engineered safety feature channels are supplied with sufficient redundancy to provide the capability for channel test at power, except for 'oackup channels such as derived circuits in engineered safeguards control system.

When one of the four channels is taken out of service for maintenance, the protective system logic can be changed to a two-out-of-three coincidence for a reactor trip by bypassing the removed channel. If the bypass is not effected, the out-of-service channel (Power Removew assumes a tripped condition (except high rate-of-change of power, high power level and high pressurizer pressure),m which results in a one-out-of-three channel logic. Ifin the 2 of 4 logic system of the reactor protective system one channel is bypassed and a second channel manually placed in a tripped condition, the resulting logic is I of 2. At rated power, the minimum operable high-power level channel is 3 in order to provide adequate power tilt detection. If only 2 channels are operable, the reactor power level is reduced to 70% rated power which protects the reactor from possibly exceeding design peaking factors due to undetected flux tilts and from exceeding dropped CEA peaking factors.

All enginected safety features are initiated by 2-out-of-4 logic matrices except containment high radiation which operates on a 1-out-of-2 basis. The containment radiation high signal isolates the containment pressure relief, air sample and purge system valves.

References (1) USAR, Section 7.2.7.1 l

2-66a Amendment No. 8,20,26,32,+3,88,125

TABLE 3d INSTRUMENT OPERATING CONDITIONS FOR ISOLATION FUNCTIONS Test, Maintenance Minimum Minimum Permissible and Operable Degree of Bypass Inoperable

& Functional Unit Channels Redundancy Condition Rypns 1 Containment Isolation A- Manual 1 None None N/A B Containment High-Pressure A 2('X" 1 During Leak (0 B 2('Xd 1 Test Pressurizer Low / Low A 2 ('"" 1 Reactor Coolant (0 B 2(*** 1 Pressure Less Than 1700 psia

• 2 Steam Generator Isolation A Manual I None None N/A B Steam Generator Iso!*. tion 1 None None N/A (i) Steam Generator Low Pressure A 2/ Steam 1/ Steam Steam Generator (O Gen

• Gen Pressure Less Than 550 psia

• B 2/ Steam 1/ Steam Gen (" Gen (ii) Contain' ment High Pressure A 2('"O I During Leak Test (0 3' '2('X" 1 3 Ventilation Isolati.on A Manual 1 None None N/A B Containment High Radiation -A 2* None If Containn e.t N/A y B 2* None Relief and Purge Valves Are Closed

, a- A and B circuits each have 4 channels.

b Auto removal of bypass above 1700 psia.

c Auto removal of bypass above 500 psia.

l l 2-69 Amendment No. 88;93108 1 .

IAIlLE 2-4 (Continued) d A and B circuits are both actuated by either one of the two initiating channels.

e If minimum operable channel conditions are reached, one inoperable channel must be placed in the tripped condition within eight hours from the time of discovery of loss of operability The remaining inoperable channel may be bypassed for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> from the time of discovery of loss of operability and, if an inoperable channel is not returned to operable status within this time frame, a unit shutdown must be initiated (see Specification 2.15(2)).

f If one channel becomes inoperable, that channel must be placed in the tripped or bypassed condition within eight hours from the time of discovery of loss of operability.

If bypassed and that channel is not returned to operable status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> from the time r. discovery of loss of operability, that channel must be placed in the tripped condition within the following eight hours. (See Specification 2.15(1) and exception associated with maintenance,)

2-69a Amendment No. 88,108 l

l 1

l'

3.0 SURVEILLANCE REOUIREMENTS BASIS Specifications 3.0.1 through 3. ' establish the general requirements applicable to Surveillance Requirements. The~ requireraents are bned on the Surveillance Requirements stated in the Code of Federal Regulations,10 CFR 50.36(c)(3): l

" Surveillance requirements are requirements relating to test, calibration, or inspection to ensure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting condition of operation will be met."

Specification 3.0.1 establishes the limit for which the specified time interval for Surveillance Requirements may be extended. It permits an allowable extension of the normal surveillance interval to facilitate surveillance scheduling and consideration of plant operating conditions that may not be suitable for conducting the suiveillance; e.g.,

transient conditions or other ongoing surveillance or mairtenance activities. It also provides flexibility to accommodate the length of a fuel cycle foi surveillance that are performed at each refueling outage and are specified with an 18-month surveillance interval. It is not mtended that this provision be used repeatedly as a convenience to extend surseillance intervals beyond that specified for surveillance that are not performed during refueling outages. The limitation of Specification 3.0.1 is based on engineering judgement and the recognition that the most probable result of any particular surveillance being performed is the verification o rconformance with the Surveillance Requirements.

This provision is sufficient to ensure that the reliability ensured thrcagh surveillance activities is not significantly degraded beyond that obtained from the specified surveillance interval.

The provisions of Specification 3.0.2 define the surveillance intervals for use in the Technical Specifications. This clarification is provided to ensure consistency in surveillance intervals. throughout the Technical Specifications. A few surveillance requirements have uncommon inteivals. In such a case the surveillance interval shall be l performed as defined by the individual specifications.

Specification 3.0.3 extends the testing interval required by codes and standards referenced by the Technical Specifications. This clarification is provided to remove any ambiguities relative to the frequencies for performing the required inservice inspection and testing activities. Under the terms of this specification, the more restrictive requirements of the Technical Specifications take precedence over the codes 'and standards referenced therein.

Specification 3.0.4 establishes the failure to perform a Surveillance Requirement within the allowed surveillance interval, as defined by the provisions of Specifications 3.0.1 and 3.0.2, as a condkion that constitutes a failure to meet the OPERABILITY rear ts for the corresponding Limiting Condition for Operation. Under the provisio.

3-Ob Amendment No. m,129

M -._ __

TABLE 3-2 (continued)

MINIMUM FREOUENCIES FOR CHECKS. CALIBRATIONS AND TESTING OF ENGINEEREC SAFETY FEATURES. INSTRUMENTATION AST CONTROLS Survei!Iance Channel Description Function _ Frecuency Surveillance Method Containment Pressure High a. Calibrate R a. Known pressure applied to sensors and 4.

CPHS actuation logic verified.

Signal

b. Tcst M b Pressure switch operation simulated one circuit at a time.

a. Test M a. Simulation of FPLS and CHPS 2/4 logic using built-in

5. Containment Spray Logic testing system. Both " standby power" and "no standby power

• circuits will be tested for A and B channels. Test will verify functioning of initiation circuits of all equipment normally operated by safety feature actuation signals.

b. Test R b. Complete at:tomatic test initiated sensor eperation (item 1(b) and 4(b)) and including all normal automatic operations.

a. Chect D a. Normal readings observed and internal test

6. Containment Radiation signal used to verify instrueent operation.

High Signal">

CRHS monitors are the containment atmosphere gaseous radiation monitor and (1) the maxiliary Building Exhaust Stack gaseous radiation monitor.

l l 3-8 ,

TABLE 3-2 (continued)

MINIMUM FREOUENCIES FOR CHECKS. CALIBRATIONS AND TESTING OF ENGINEERED SAFETY FEATURES. INSTRUMENTATION AND CONTROLS Surveillance Channel Desciiotion Function - Frecuency Surveillance Method

6. (continued) b. Test M b. Detector expemi to remote operated radiation check source or test signal to verify instrumentation, one channel at a time, and isolation lockout relay

• functional check.

c. Calibrate R c. Secondary and Electronic Calibration performed at refueling frequency. Primary calibration performed with exposure to radioactive sources only when required by the secondary and electronic calibration.

7. Manual Safety In,iection a. Test R a. Manual initiation.

Initiation

8. Manual Containment a. Test R a. Manual initiation.

Isolation Initiation

b. Check R b. Observe isolation valves closure.

9. Manual Initiation a. Test R a. Manual switch operation; pumr and valves Containment Spray tested separately.

10. Automatic Load Sequenc:rs a. Test Q a. Proper operation will be verified during safety feature actuation test of Item 3(a) above.

11. Diesel Testing See Technical Specification 3.7 3-9 Amendment No. 84,1I1

3 TABLE 3-3 (cominued)

MINIMUM FREOUENCIES FOR CHECKS. CALIBRATIONS AND TESTING OF MISCELLANEOUS INSTRUMENTATION AND CONTROLS Surveillance Channel Description Function Frecuency Surveillance Method

. 1. Primary CEA Position a. Check S a. Comparison of output data with secendary CEAPIS.

Indication System

b. Test M b. Test of power dependent insertion limits, deviation, and sequence monitoring systems.

, c. Calibrate R c. Physically measured CEDM position used to verify system accmycy. Calibraic

CEA position interlocks.

i 2. Secondary CEA Position a. Check S a. Comparison of output data with primary CEAPIS.

Indication System

b. Test M b. Test of power dependent insertion limit, deviation, out-of-sequence, and overlap monitoring systems.

c. Calibrate R c. Calibrate secondary CEA position indication system and CEA interlock alarms.

l 3. Area and Post Accident a. Check D a. Normal readings observed and internal test signals used to verify instrument

! Radiation Monitors

• operation. -

I b. Test M b. Detector exposed to remote operated radiation check source or test signal.

c. Calibrate ' R c. Secondary and Electronic calibration performed at refuc!ing frequency. Primary

calibration with exposure to radioactive sources only when required by the secondary and electronic calibration. RM-091 A/B - Calibration by electronic signal substitution is acceptable for all range decades above 10 R/hr. Calibration for at least one decade below 1- R/hr. shall be by means of calibrated radiation source.

• Post Accident Radiation Monitors are: RM-053UhUH, RM-064, and RM-091 A/B. Area Radiation Monitors are: RM-070 thru RM-082, RM-084

! thru RM-089, and RM-095 thru RM-098.

3-13 Amendment No. !-!,8!,S6;93137

TA.BLE 3-3 (continued)

MINIMUM FREOUENCIES FOR CHECKS. CALIBRATIONS A,'JD TESTING OF MISCELLANEOUS INSTRUMENTATION AND COY [#LS Surveillance Channel Description Function Frecuency Surveillance Method

4. Emergency Plan Radiation a. Calibrate A a. Exposure to known radiation source.

Instruments

b. Test M b. Battery check.

i

5. Primary to Secondary a. Check D a. Normal readings observed and internal test signals Ixak-Rate Detection used to verify instrument operation.

Radiation Monitors (RM-054A/B, RM-057)

b. Test M b. Detector exposed to remote operated radiation check sources I or test signal.

I

c. Calibrate R c. Secondary and Electronic calibration performed at refueling frequency. Primary Calibration performed with exposure to radioactive sources only when required by the secondary and ,

cIcctronic calibration.

i

6. Pressurizer Level a. Check S a. Comparison cf independent level readmgs.

Instruments

b. Calibrate R b. Known differential pressum applied to sensor.

c. Test M c. Signal to alarm meter relay adjusted with test device to verify setting.

! 7. CEA Drive System - a. Test R a. Verify proper operation of all CEDM systeni interlocks, using

! Interlocks simulated signals where necessary.

I i

a 3-14

I i

TABLE 3-.1 (Continued)

I MIN 151UM FREOUENCIES FOR SAMPLING TESTS Type of Measurement Ettd_Anillnit._

, 1. Reactor Coolant i (Continued)

(c) Cold Shutdown (1) Chloride 1 per 3 days (Operaung Mode 4)

(d) Refueling Shutdown (1) Chloride 1 per 3 days *

(Operating Mode 5) (2) Boron Concentration 1 per 3 days *

(e) Refueling Operation (1) Chloride 1 per 3 days *

(2) Boron Concentration 1 per shift *

2. SIRW Tank ikiron Concentration 1 per 31 days

3. Concentmted Boric Boron Concentration 1 per 31 days Acid Tanks

4. Si Tanks Boro'i Concentration 1 per 31 days

5. Spent Fuel Pool Boron Concentration 1 per 31 days

6. Steam Generator Blowdown isotopic / nalysis for 1 per 7 days *

(Operating Modes 1 and 2) Dose Equivalent 1-131 (1) Until the radioactivity of the reactor coolant is restored to .s. I pCi/gm DOSE EQUIVALENT l-131.

(2) Sample to be taken after a minimum of 2 EFpD and 20 days of power operation have elapsed since reactor was saberitical for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or longer.

(3) Baron and chloride sampling / analyses are not required when the core has been off-loaded. Reinitiate boron and chloride sampling / analyses one shift prior to reloading fuel into the cavity to assure adequate shutdown margin and allowable chloride levels are met, (4) When Steam Generator Dose Equivalent I-131 exceeds 50 percent of the limits in Specification 2.20, the sampling and analysis frequency shall be increased to a minimum of 5 times per week. When Steam Generator Dose Equivalent I-131 exceeds 75 percent of this limit, the sampling and analysis frequency shall be increased to a minimum of once per day.

L 3-19 Amendment No. 28,67,86A24,133

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3.0 SURVEILLANCE REOUIREMENTS 3.10 Reactor Core Parametets (Continued)

(6) Azimuthal Power Tilt (Tq)

Whenever the core power is above 70% of rated power, the azimuthal power tilt i shall be determined to be within its limits by calculating the tilt at least once every day using either:

a. The encore detectors with at least four safety shannels operable, or

b. The incore detectors with at least two strings of three rhodiura detectors ,

per full core height quadrant operable.

(7) DNB Parameters

a. The cold leg temperature, pressurizer pressure, and axial shape index shall be verified to be within the limits of Section 2.10.4(5) at least once per shift.

b. The reactor vessel coolant total flow rate shall be determined to be within '

its limit by measurement at least once per month.

3-63b Amendment No. 42,%,92 (Next Page is 3-69)  !

3.0 SURVEILLANCE REOUIREMENTS 3.12 Radioactivs Waste Disposal System AnpJisability Applies to the instrumentation used to determine hydrogen and oxygen concentrations ,

in the waste gas decay tanks.

Oblective To ensure the concentrations of hydrogen and oxygen in the gaseous radioactive waste system are maintained below their flammability concentrations as required by Specification 2.9.

Soccifications The hydrogen and oxygen monitoring system for the waste gas decay tanks shall have a:

a. daily channel check (when in service)

b. monthly cross comparison with a grab sample

c. quarterly channel calibration using a gas mixture with concentrations in the range of interest Basis The specification ensures that instrumentation used to determine the concentration of potentially cxplosive gas mixtures entrained in the gas decay tank (s) will be maintained in an operable condition. Maintaining the instrumentation used to determine hydrogen and oxygen concentration with a surveillance program 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 50.

3-69 Amendment No. 28,86, G3 (Next Page is 3-76)

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5.9.3 Special Reports Special reports shall be submitted to the Regional Administrator of the appropriate NRC Regional Of0cc within the time period speciGed for each report. These reports shall be submitted covering the activities identified below pursuant to the requirements of the applicable reference specincation where appropriate:

a. In-service inspection report, reference 3.3.

b. Tendon surveillance, reference 3.5.

c. Containment structural tests, reference 3.5.

d. Special maintenance reports,

e. Containment leak rate tests, reference 3.5.

f. Materials radiation surveillance specimens reports, reference 3.3.

g. Fire protection equipment outage, reference 2.19.

h. Post accident monitoring instrumentation, reference 2.21 5.9.4 Unique Reportine Reauirements

a. Semiannual Radicactive Effluent Release Repen The Semiannual Radioactive Efnuent 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 efnuents and solid waste released from the unit. The material provided shall be 1) consistent with the objectives outlines in the ODCM and PCP, and 2) in conformance with 10 CFR 50.36a. and Section Ill B.1 of Appendix ! to 10 CFR Part 50.

b. Annual Radiological Environmental O PHaliDEESDQu 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)Section IV.B.2, IV.B.3, and IV.C of Appendix ) to 10 CFR Part 50.

5-15 Amendment No. 9,24,38,46,86, 440,443,133'

__ . . _ _ _ _ . - ~ _ _ _ _ _ . _ _ _ . . _

5.0 ADh11NISTRATIVFmCONTROLS i 5.10.2 The following records shall be retained for the duration of the Facility Operating License *

a. Records of drawing changes reflecting facility design modifications made to l systems and equipment described in the Final Safety Analysis Report.

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

c. Records of facility radiation and contamination surveys.

d. Reccrdt of radiation exposure for all individuals entering radiation control areas.

c. Records of gaseous and liquid radioactive material released to the environs,

f. Records of transient or operational cycles for those facility components designed for a limited number of transients or cycles.

g. Records of training and qualification for current members of the plant staff,

h. Records of in-service inspections performed pursuant to these Technical Speci0 cations.

i. Records of Quality Assurance activities required by the QA 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.

k. . Records of meetings of the Plant Review Committee and the Safety Audit and Review Committee.

l 1. Records of Environmental Qualification of Electric Equipment pursuant to 10 CFR 50.49.

m. Records of the service lives of all hydraulic and mechanical snubbers which are covered under the provisions of Section 2.18 of the Technical Specifications, including the date at which the service life commences and associated installation and maintenance records.

n. Records of analyses required by the Radiological Environmental Monitoring Program.

o. Records of reviews performed for changes made to the Offsite Dose Calculation Manual and the Process Control Program.

5.10.3 A complete record of the analysis employed in the selection of any fuel assembly to be placed in Region 2 of the spent fuel racks will be retained as long as that bundle remains in Region 2 (reference Technical Specifications 2.8(12) and 4.8.4).

5.11 Radiation Protection Program  ;

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

5-19 Grdfr 70/24/80 Anuximent No.59,86,93,99,105

5.0 ADMINISTRATfVE CONTROLS 5.11.1 In lieu of the " control device" required by paragraph 20.1601(a)(1)(2)(3) of 10 CFR

20. and as an alternative method allowed under 20.1601(c) each high radiation area (as defined in i 20.1601 of 10 CFR 20)in which tne intensity of radiation is 1000 mrem /hr or less shall be barricaded and conspicuously posted as a high radiation area and entrance thereto shall be controlled by required issuance of a Radiation Work Permit.*

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

a. A radiation monitoring device which continuously indicates the rediation 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 level in the area has been established end personnel have been made knowledgeable of them.

c. An individual qualified in radiation protection procedures who is equipped with a radiation dose rate monitoring device. This individual shall be responsible for providing positive control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified by the facility Supervisor Radiation Protection in the Radiation Work Permit.

5. I1.2 The requirements of 5.11.1, above, shall also apply to each high radiation area in which the intensity of radiation is greater than 1000 mrem /hr (Restricted High Radiation Area).

In addition, locked doors shall be provided to prevent unauthorized entry into such areas and the keys shall be maintained under the administrative control of the Shift Supervisor oa duty and/or the Supervisor-Radiation Protection with the following exception:

a. . In lieu of the above, for accessible localized Restricted High Radiation areas located in large areas such as containment, where no lockable enclosure exists in t a immediate vicinity to control access to the Terricted High Radiation area, and ao such enclosure can be readily constructed, tNn the Restricted High Radiation area shall be:

1. roped off such that an individual at the rope boundary is exposed to 1000

. mrem /nr or less, il conspicuously posted, and iii a flashing light shall be activated as a warning device.

,

• Radiation Protection personnel shall be exempt from the RWP issuance requirement during the l performance of their assigned radiation protection duties, provided they comply with approved i

radiation protection procedures for entry into high radiation areas, i

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5-19a Amendment No. 28,M,132

5.0 ADhif NISIRATIVE CONTROLS 5.16 Badiological Effluents and Environmental hionitorine Procrams The following programs shall be established, implemented, and maintained.

5.16.1 Radioactive Ef0uent Controls Program A program shall be provided conformir.g with 10 CFR 50.36a for the control of radioactive effluents and for maintaining the doses to individuals in unrestricted areas from radioactive efnuents as Icw v reasonably achievable.

The program (1) shall be contained in the ODChi, (2) shall be implemented by ,

operating procedures, and (3) shall include remedial actions to be taken  !

whenever the program limits at t exceeded. The program shall include the  !

following elements *  !

a. Limitations on the operability of r-dioactive liquid and gaseous radiation l monitoring instrumentation including operability tests and setpoint  ;

determination in accordance with the methodology in the ODChi.

b. Limitations on the concentration of radioactive material released in liquid efGuents to unrestricted areas conforming to 10 CFR Part 20.

c. hionitoring, sampling, and analysis of radioacdve liquid and gaseous effluents in accordance with 10 CFR Part 20 and with the methodology and parameters in the ODChi,

d. Limitations on the annual and quarterly doses or dose commitment to individuals in unrestricted areas from radioactive materials in liquid efauents released to unrestricted areas conforming to Appendix I to 10 CFR Part 50.

e. Determination of cumulative doses from radioactive efGuents for the current calendar quarter and current calendar year in accordance with the ODChi on a quarterly basis.

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5.0 ADh11FISTRATIVE CONTROLS 5.16 Radiological Efnuents and Environmental hionitoring Programs (continued)

f. Limitations on the operability and use of the liquid and gaseous efDuent treatment systems to ensure that me appropriate portions of these systems are used to reduce releases of radioactivity in plant ef0uents.

g. Limitations on the concentration resulting from radioactive material released in gaseous ef0uents to unrestricted areas conforming to 10 CFR Part 20.

h. Limitations on the annual and quarterly air doses resulting from noble gases released in gaseous effluents to unrestricted areas conforming to Appendix I to 10 CFR Part 50.

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i. Limitations on the annual and quarterly doses beyond the site boundary from Iodine-131, tritiurn, and all radionuclides in particulate form with half lives greater than 8 days in gaseous ef0uents released to unrestricted areas conforming to Appendix I to 10 CFR Part 50.

5.16.2 Radiological Environmental hionitoring Program A program shall be provided to monitor the radiation and radionuclides in the environs of the plant. The program shall provide (1) representative measurements of radioactivity in the highest potential exposure pathways, and (2) verification of the accuracy of the ef0uent monitoring program and modeling of environmenal exposure pathways The program shall (1) be contained in the ODChi, (2) conform to the guidance of Appendix I to 10 CFR Part 50, and (3) include 'he following:

a. hionitoring, san.ging, analysis, and reporting of radiation and radionuclides in the environment in accordance with the methodology and parameters in the ODChi,

b. A Land Use Census to ensure that changes in the use of areas at and beyond the site boundary are identified and that modi 0 cations to the monitoring program are made if required by the results of this census,

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

5-23

- _ . . - .- . - . - . - . -. - -- a.

5.0 ADMINIS.TRATIVE CONTR_OLS 5.I7 Q(hite.Du Calculkliqu.Jdanual (ODDD Changes to the ODCht:

a. Shall be documented and records of reviews performed shall be retained as required by Specification 5.10.2.o. This documentation shall contain:

1. Sufficient information to support the change together wit the appropriate analyses or evaluations justifying the change (s) and l

2. A determincion that the change will maintain the level of radioactive effluent control required by 10 CFR Part 20,10 CFR Part 50.36a, and Appendix I to 10 CFR Part 50 and not adversely impact the accuracy or reliability or effluent, dose, or s(tpoint calculations,

b. Shall becona; cffective after review by the Plant Review Committee and the Manager - Fort CMhoun Station,

c. Temporary changes to the ODCM may be made in accordance with Technical Specification 5.8.3.

d. Shall be submitted to the Nuclear Regulatory Commission in the form of a complete, legible copy of the entire ODCM as a part of or concurrent with the Semiannual Radioactive Effluent Release Report for the period of the report in which any change to the OLCM was made. Each change shall be identified by markings in the margin of the affected pages, clearly indicating the nea of the page that was changed and shall indicate the date (e.g., month / year) the change was implemented.

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5.0 ADMINISTRATIVE CONTRQLS 5.18 Process _Conitol.flearam (PCE)

Changes to the PCP:

a. Shall be documented and records of reviews performed shall be retained as required by Specification 5.10.2.0. This documentation shall contain:

1. Sufficient information to support the change together with 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 program to existing requirements of federal, state, or other applicable regulations.

b. Shall become effective after review and acceptance by the PRC and the approval of the Plant Manager,

c. Temporary changes to the PCP may be made in accordance with Technical Specification 5.8.3.

d. Shall be submitted to the Nuclear llegulatory Commission in the form of a complete, legible copy of the entire PCP as a part of or concurrent with the Semiannual Radioactive Effluent Release Report for the period of the report in which any change to the PCP was made. Each change shall be identified by markings in the margin of the affected pages, clearly indicating the area of the page that was changed and shall indicate the date (e.g., month / year) the change was implemented.

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5-25 1

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1 ATTACHMENT B '

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DISCUS $10N, JUSTIFICATION, AND NO $1GNIFICANT HAZARDS CONSIDERATIONS DISCUSSION AND JUSTIFICATION The Omaha Public Power District (OPPD) proposes to revise the fort Calhoun Station l! nit No.1 Technical Specifications to implement Generic letter 89-01 concerning the Radiological Effluent Technical Specifications (RETS), and to revise the requirements for the Containment Radiation High Signal following the guidance of NUREG-0133.

ProDosed chanaes to RETS As specified in Generic letter 89 01, it is proposed to relocated RETS from the Technical Specifications to the Offsite Dose Calculation Manual (00CM). The following is a description of the proposed changes:

1. The definitions for the ODCM and Process Control Program (PCP) were revised to agree with the definitions in Generic letter 89-01.

Definitions for purge-purging and venting were moved to the ODCH.

2. Technical Specifications 2.9 under Limiting Condition for Operations Section and 3.11 and 3.12 under Surveillance Requirements Section were relocated from the Technical Specifications and incorporated into the ODCH.

3. ' ables 3-2, 33, and 34 were revised to ensure that LCO

.equirements, which are currently controlled by Specification 3.12, u,e retained. The surveillance function and frequency for Containtnent Radiation High Signal, Area and Post Accident Radiation Monitors, and Primary-co-Secondary leak rate detection radiation monitors in Table 3-2 are updated for consistenc/. Sampling requirements for steam generatar blowdcwn are bein Table 3-11, which is being deleted, to Table 3 4. g Operating relocatedmodes from are bein added to the sampling of steam generator blewdown consisten with the actions required by Specification 2.20.

4. Tables 3-9. 3-11 and 3-12 were relocated from the Technical SpecificaticasandincorporatedintotheODCH.

l S. Specifications 2.0 and 3.12 are proposed to retain requirements in l the Technical Specifications that are related ta explosive gases.

L This war a specific renuirement of the Generic Letter. The grab l sample rovisions are in agreement with the present Specifica{ ion l 2 d. A 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> Limiting Condition for Operation has been 2.9.l(f added r the time allowed for hydrogen and oxygen concentratiens to l De out of specification. This is consistent with the Standard RE1S, l 6. Reportir.g and records requirements were revised in Section S.9.4 and Sections 5.16, 5.17, aad 5.18 under Administrative Controls were added to the Technical Specifications to define administrative details of the ODCM and PCP. Records retention for RETS will be covered under existing Technical Specification 5.10.1.

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Proposed Chanaes to the Containment Radiation Hiah Sianti l Consistent with NUREG-0133, it is proposed to remove the Ventilation Isolation functions of the particulate and iodine monitors Actuation sampling the Signal LVIAS) Building Exhaust Stack and Containment.

A'Jxiliary The proposal woull remove the VIAS functions of three effluent monitors. These are VIAS signals for Stack lodine (RM-060), Stack Particulate (RM-061) and Containment Particulate (RM-050).

The particulate and iodine monitors will continue to sample effluents, but will not initiate VIAS. NUREG-0133 recommends the use of gas monitors for VIAS initiation because it is considered to be impractical to apply instantaneous alarm / trip setpoints to integrating radiation monitors sensitive i to radioiodines or radioactive materials other than noble gases. 1 The gas monitors will continue to isolate releases so that 10 CFR Part 20 instantaneou; limits and 10 CFR Part 50 annual limits will be complied with.

The following pages are being revised in the Technical Specifications to update the VIAS f unctions so that only gas monitors initiate VIAS; pages 2-37, 2-38, 2 61,-2-66a, 2-69, and 2-69a.

Chanacs_to Continuous Samolina Reauirements Technical Specification 2.9.l(2)h and Table 3-12 currently require that continuous samples be taken during continuous releases, it is proposed to revise this requirement to provide 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to initiate the auxiliary sampling required during continuous releases. The problems associated with meetin requirements to continuously sample were reported to the NRC in LER's 91 g028, the and 92-001. As described in Regulatory Guide 1.21 Appendix A, the intent of continuous sampling is to identify radionuclides being released and to predict their concentrations. Two hours is proposed as a reasonable amount of time which will allow investigation of trouble alarms on equipment and to complete calibrations, filter changes and testing requirements without significantly affecting the results of the sampling. Current requirements allow periodic grab samples be taken when monitors are inoperable for releases which have a greater impact on the total effluents released from the plant. These changes have been included in the draft ODCH.

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Admittillrgtivtchanaes

a. The Table of Contents is being revised to reflect proposed changes,

b. Page 2-8 is being revised to correct a reference that was changed as a result of the Generic letter changes.

c. Page 2-38 is being revised to delete a reference to the specific procedure which implements the independent verification of fuel burnup which is required to move fuel directly to Region 2 of the spent fuel pool. The requirements remain, only the name of the procedure is being deleted. Additionally, wording which reflects the CE Standard Technical Specifications are being added to clarify movement of irradiated fuel.

d. Reference 2 on Page 2-63 is being revised to reflect a Section of the Updated Safety Analysis Report the Containment Radiation High Signa l, (USAR) which better describes

a. Table 2 4, item 3B on page 2 69 is being revised to reflect the requirements of CE Standard Technical Specifications. As described in USAR Section 7.3.2.6 and Specification 2.8(2), the Containment Radiation High Signal (CRHS) isolates the containment pressure relief, air sample, and purge system valves. As currently written, it could be implied that all valves which receive a VIAS are required to be closed in order to place the CRHS channels in bypass. This change clarifies the requirements that onl y the containment vent and purge valves are required to be closed. A description of these valves is being added to the basis of Specification 2.15 on page 2-66a.

f. Page 2-91 is being reiised to reflect a change in Section 5.9.3.

g. Page 3 Ob is being revised to delete an exanple of a surveillance test table that will be relocated to sc ODCM.

h. Page 3-63b is-being revised to include a statement on the bottom of the page indicating that the next page will be page 3-69.

i. Page 5-19 is being revised to add records retention requirements for the ODCM and PCP.

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NO SIGNIFICANT HAZARDS CONSIDERATIONS  !

The' proposed changes do not involve a significant hazards consideration because operation of Fort Calhoun Station Unit No. 1, in accordance with these changes, would not:

1. Involve a significant increase in the probability or consequences of-an accident previously evaluated.

The removal of the present Radiological Effluent Technical Specifications to the Offsite Oose Calcuhtion Manual in '

tccordance with the guidelines presented in Generic Letter 89-01 will not cause any increase in the probability or consequences of an accident. Only the procedural details have been transferred to the ODCM. Programmatic controls have been retained or added to ensure continued compliance with federal requirements.

The removal of VIAS signals from the particulate and iodine radiation monitors will not cause an increase in the probability ,

or consequences of an accident. -Initial indications for radioactive release will occur promptly on the noble gas radiation monitors. VIAS will be initiated as-soon as the Auxiliary Building Exhaust Stack or the Containment noble gas monitor reaches its alarm setpoint. =The removal: of the particulate and iodine monitors will not alter the initiation of VIAS in any postulated accident.

2. Create the possibility of a new or different kind of accident from

. any previously analyzed.

The relocation of RE7S from Technical Specifications to the ODCM

is an administrativa change. Present tests, calibrations, or inspections nececsary to ensure th quality of systems and components will continue to-be perrormed, and this-change will not i create a new or different kind of accident. -The removal of the VIAS input from the particulate and iodine radiation monitors is e the result of the recognition, by NRC documents, of the impracticality of applying instantaneous alarm setpoints to integrating radiation monitors. The primary and fastest indications of-an actual radioactive release are noble gases. The radiation monitors for noble gases will continue t; provide inputs for VIAS. Therefore,-no new or different kind of accident been created.

3. Involve a significant reduction in the margin of safety.

The administrative changes made will not cause a reduction in the margin of safety. The present RETS requirements are retained in the ODCM with Programmatic Controls in the Technical Specifications. This is consistent with the guidance of Generic Letter 89-01.

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D The removal of VIAS inputs from the particulate and iodine radiation monitors will not cause a significant reduction in the margin of safety. The primary indicators for radioactive releases, noble gases, will still be monitored and will initiate VIAS upon reaching their alarm setpoints. Spurious alarms, due to the incorrect application of instantaneous limits to integrating monitors will be eliminated, reducing the functional requirements that the Engineered Safeguard Features must comply with and enhance the reliability of VIAS. Noble gas concentrations are approximately 10,C00 times greater than the particulate and iodine concentrations, therefore the margin of safety ir detecting radioactive releases will be retained.

Therefore based on the above considerations, it is OPPD's position that this proposed amendment does not involve a significant hazards considfration as defined by 10 CFR 50.92, and the proposed changes will not result in a condition which significantly alters the impact of the station on the environment. The proposed changes meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(e)(g) and, pursuant to 10 CFR 51.22(b) no environmental assessment need be prepared.

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COMPARISON OF PARAGRAPH NUMBERS FROM TECH. SPECS TO ODCM Tech. Spec. QQCy 1 2.9.1 Objective 1.0 1 2.9.1.A.(1) 1.1.2 l 2.9.1.A.(2) 1.1.3 l 2.9.1.B.(1) 1.2.2 2.9.1.B.(2) 1.2.3 2.9.1.B.(3) 1.2.4

2. 9.1. (1) a. (i) 1.1.1 l 2.9.1.(1)a.(ii) 1.1 4.1.1 2.9.1.(1)b.

2. 9.1. (1) b. (i) 4.1.1.A 2.9.1.(1)b.(ii) 4.1.1.B 2.9.1.(1)b.(iii) 4.1.1.C R2. 9.1. (1) c . 4.2.1

2. 9.1. (1) c. (i) 4.2.1.A

2. 9.1. (1) c. (ii) 4.2.1.B 2.9.1. (1) c. (iii) 4.2.1.C

.9.1.(1)d. 2.1.1

.. 9.1. (1) d . (i) 2.1.1.1

2. 9.1. (1)d. (ii) 2.1.1.2

2. 9.1. (1) d. (lii) 2.1.1.4

2. 9.1. (1) d . (iii) 1. 2.1.1.4.A 2.9.1.(1)d.(iii)2. 2.1.1.4.B 2.9.1. (1)d. (lii) (Notes) 2.1.1.6 2.1.1.5

2. 9.1. (3 ) e.1) 2.1.2.2 2.9.1.(1)e.2) 2.1.2.3 2.9.1.(1)e. 2.1.2.3.A 2.1.2.3.B 2.1.2.4

2. 9.1. (2 ) a. (i) 1.2.1 2.9.1. (2) a. (ii) 1.2 2.9.1.(2)b. 4.2.1

12. 9.1. (2 ) b. (i) 4.2.1.A

2. 9.1. (2 ) b. (ii) 4.2.1.B 2.9.1. (2) b. (iii) 4.2.1.C 2.9.1.(2)c. 4.2.2 2.9.1. (2) c. (i) 4.2.2.A

2. 9.1. (2) c. (ii) 4.2.2.B

2. 9.1. (2) c. (lii) 4.2.2.C 2 9.1.(2)d. Retained in Tech. Specs.

2J0il.(2)e. . 2.2.1.4 2.9.1. (2 ) e.1) 2.2.1.4.A

2. 9.1. ( 2) e. 2) 2.2.1.4.B 2.9.1.(2)f. 2.2.2.1 2.9.1.(2)g. 1.2.1.1

2. 9.1. (2 ) g. (i) 2.2.1.1.A

2. 9.1. (2 ) g. (ii) 2.2.1.1.C 2.9.1.(2)g.(iii) 2.2.1.1.B

2. 9.1. (2 ) g. (iv) 2.2.1.1.D 2.2.1.1.E y y + ---m--r-- ,y--- ., e-- - e + - - - , - - - y .- - , , - - +

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1 COMPARISON OF PARAGRAPH NUMBERS FROM TECH. SPECS TO ODCM Page Two Tech. Spec. QDgg

2. 9.1. (2) g. (v) 2.2.1.2.A/B

2. 9.1. (2) h. (1) 2.2.3.1 2 .1 .. (2)h. (i) 2.2.3.1.A 2.2.3.1.B ]

2.9.1.(2)h.(ii) 2.2.3.1.C l

3.11 Objective 5.0 3.11(1) 5.1.1 3.11(2) 5.1.3 l

3.11(3) 5.1.4 1

3.11(3)a. 5.1.4.A 3.11(3)b. 5.1.4.B 3.11(3)b.(Note) 5.1.4.B 3.11(4) 5.1.5 3.12.1 Objective 3.0 3 .12 .1 ( 1) a . 3.1.1 ,

l l 3.12.1 ( 1) b. 3.1.2 )

3.1.2.1 j 3.1.2.2 )

3.12.1(1) c. (i) (ii) (lii) Table 3, Part-A 3.12.1(1) d. (1) (ii) (iii) (iv) Table 3, Part A Table 3, Note #4 3.12.1(1)e.

l 3.12.1 (1) f . 3.1.3 i 3.12.1(2)a. 3.2.1 l l

3.12.1(2 ) b. (i) (ii) (lii) (iv) Table 3,-Part D Table 3, Part B l

3.12.1(2) b. (v) 3.12.1(2 ) c. (i) (ii) Table 3, Part B Table 3-11, Note (4) 3.3 j

! Table 3-12, Note (4) 3.3 l 5.9.3.a. 4.3 l 5.9.4.b.1. 4.4 l 5.9.4.b.1.a. 4.4.1 i 5.9.4.b.1.b. 4.4.2 l' 4.4.3 5.9.4.b.1.c.

5.9.4.b.1.d. 4.4.4 5.9.4.b.1.e. 4.4.5 5.9.4.b.1.e.(1) 4.4.5.1 l 5.9.4.b.1.(2) 4.4.5.1.A 5.9.4.b.1.(3) 4.4.5.1.B 5.9.4.b.1.(4) 4.4.5.1.C

5.9.4 b.1.(5) 4.4.5.1.D

! 5.9.4.b.2. 4.5 l-l i