ML20140F154

Proposed Tech Specs Re Penn State Breazeale Reactor

ML20140F154
Person / Time
Site:	Pennsylvania State University
Issue date:	04/24/1997
From:	PENNSYLVANIA STATE UNIV., UNIVERSITY PARK, PA
To:
Shared Package
ML20140F117	List: ML20140F113 ML20140F118 ML20140F154
References
NUDOCS 9705020129
Download: ML20140F154 (51)

Text

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TECHNICAL SPECIFICATIONS FOR THE PENN STATE BREAZEALE REACTOR (PSBR) l FACILITY LICENSE NO. R-2 s TABLE OF CONTENTS 1.0 INTROD U CTIO N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 1.1 Defin i tion s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . I 1.1.1 ALARA...................................................................... 1 I

l 1.1.2 Automatic Control . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 1.1.3 Channel...................................................................... 1 -

I 1.1.4 Channel Calibration . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 1.1.5 Channel Check . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 1.1.6 Chann el Tes t . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . I 1.1.7 Cold Critical . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . I 1.1.8 Close Pac ked Array . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 l 1.1.9 Confi ne me n t . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 l 1.1.10 Core Lattice Position . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 1.1.11 Excess R eactivity . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 1.1.12 Experime n t . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 1.1.13 Experimental Facility . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 1.1.14 Instmmen ted Element . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 ;

1.1.15 Limiting Conditions for Operation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 '

1.1.16 Limiting S afety System Setting . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 1.1.17 M an ual Control . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 1.1.18 Maximum Elemental Power Density. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 1.1.19 Maximum Power leve1. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 1.1.20 M easured Val ue . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 1.1.21 Movable Experiment . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 1.1.22 Normalized Power . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 pJ s

1.1.23 Operable.....................................................................

1.1.24 Opera tin g . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

3 3

1.1.25 Pulse M od e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 1.1.26 Reactivity Limits . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 1.1.27 Reactivity Worth of an Experiment . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 1.1.28 Reacter Control S ystem . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 1.1.29 Reactor Interlock . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 1.1.30 Reactor Operating . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 ,

1.1.31 Reactor S ecured . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 l 1.1.32 Reactor S hu tdown . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 1.1.33 Reactor S afety S ystem . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 1.1.34 Reference Com Condition . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 1.1.35 Research Reactor . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 1.1.36 Reportable Occurrence . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 1.1.37 Rod-Tran si en t . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 1.1.38 S afe ty Limit . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 l 1.1.39 S CRAM Time . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 1.1.40 S e cured Experimen t . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 1.1.41 Secured Experiment with Movable Parts . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 1.1.42 S hall, S hould and May . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 l 1.1.43 S him, Regulatin g, S afety Rods . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 l 1.1.44 S hutdo wn Margin . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 1.1.45 Square Wave Mode . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 1.1.46 S teady S tate Power Level . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 1.1.47 TRIGA Fuel Element 6 e .....................................................

7 t 1.1.48 Watc hdog Circuit . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

( ,

yg d Amendment No. 32 Rev.1 (4/24/9'/)

9003 Idb

/ ,

. _ . ~ .-

r l ^

9705020129 970428 '

u ~ ~,

PDR ADOCK 05000005 -

P PDR

ii I

s 2.0 SAFETY LIMIT AND LIMITING SAFETY SYSTEM SETITNG ................... 7 l 2.1 S afety Limit-Fuel Element Temperature . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 l 2.2 Limiting Safe ty System Setting (LS S S) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 i

3.0 LIMITING CONDITIONS FOR OPERATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 i 3.1 Reactor Core Parame ters . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 .)

1 \

l 3.1.1 Constant Power and Square Wave Operation . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 3.1.2 ReactivitB. imitation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 3.1.3 S hu tdown M argin . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10 ,

3.1.4 Pulse Mode Operation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10 <

l 3.1.5 Core Configurate n Limitation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11 3.1.6 TRIGA Fuel Elements . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12

• 3.2 Reac to r Control and Reactor S afety System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 >

3.2.1 Reactor Con trol Rods . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 .

! 3.2.2 Manual Control and Automatic Control . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14 l 3.2.3 Reactor Control System . . . . . . . . . . . . . . . . .'. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14 3.2.4 Reactor Safety System and Reactor Interlocks .......................... 15 3.2.5 Core Loadin g and Unloading Operation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 18 l 3.2.6 S C RAM Time . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 18 i

3.3 Coo lan t S ys tem . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19 1

3.3.1 Coolant Level Limits . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19 3.3.2 Detection of Leak or Loss of Coolant . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19  :

3.3.3 Fission Product Activity . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20 3.3.4 Pool Water Supply for Ixak Protection . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20 3.3.5 Coolan t Conductivity Limits . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21 3.3.6 Coolant Temperature Limits . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21 3.4 C o nfin e m e n t . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 22 l l

3.5 Engineered Safety Features - Facility Exhaust System ,

and Emergency Exhaust System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . 22  ;

l 3.6 Radiation Monitoring System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 23 ,

3.6.1 Radiation M onitoring Inform ation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 23 3.6.2 Evacuation Alarm . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 24 3.6.3 Argon-41 Discharge Limit . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 24

, 3.6.4 ALARA..................................................................... 25 3.7 Limitations of Experiments . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 25 I

4.0 S U RVEILLANCE REQUIREMENTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 28 ,

4.1 Reac to r Parame te rs . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 28 4.1.1 Reacto r Power Calibratio n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 28 4.1.2 Reactor Excess Reactivity . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 28 4.1.3 TRIG A Fuel Elements . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 28  :

i 4.2 Reactor Control and Safety Sys te m . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 29 4

l 4.2.1 Reac tivity Worth . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 29

4.2.2 Reactivity Insertion Rate . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 29  ;

4 j

, l 9003 Idb i Amendment No. 32 Rev.1 (4/24/97 1

l ...

I m i

i O 4.2.3 Reactor S afety S ystem . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 30 b 4.2.4 Reac to r In terlocks . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 31 4.2.5 Overpower SCRAM . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 32 ,

4.2.6 Transient Rod Tes t . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 32 4.3 Coo lant S ystem . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 33 4.3.1 Fire Hose Ins pection . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 33 4.3.2 Pool Water Temperr.ture . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 33 4.3.3 Pool Water Conductivity . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 34 4.3.4 Pool Water Level Alann . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 34 l 4.4 Co nfine m e nt . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 34 t

l 4.5 Facility Exhaust System and Emergency Exhaust System ................ ...... 35 i

l 4.6 Radiation Monitoring System and Effluents . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 35 4.6.1 Radiation Monitoring System and Evacuation Alarm ............ ..... 35 4.6.2 Argon-41................................................................. 36 4.6.3 ALARA....................................:............................... 36 l

4.7 Ex pe rime n ts . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 37 5.0 D ES IG N FEATURE S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 37 5.1 Re ac to r Fuel . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 37 5.2 Reac tor Co re . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ............................. ... 38 5.3 C o n trol R ods . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 38 s 5.4 Fu e l S to ra g e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 38

' 5.5 Reactor B ay and Exhaust Systems . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 38 5.6 Reactor Pool Wate r S ystem s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 39 6.0 ADMINISTRATIVE CONTROLS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 39 6.1 Org anization . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 39 ,

6.1.1 S tru c ture . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 39 6.1.2 Res po ns ib ility . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 39 ,

6.1.3 Staffing...................................................................... 41 6.1.4 Selection and Training of Personnel . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 41 i

41

( 6.2 Re view and Audit . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

41 l 6.2.1 Safeguards Committee Com position . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

6.2.2 Charter and Rules . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 42 l 6.2.3 Revie w Func ti on . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 42 6.2.4 Audit..................................................................... . 42 6.3 Operatin g Proced ures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 43 6.4 Review and Approval of Experiments . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 44 L

6.5 Required Ac tion . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 44 6.5.1 Action To Be Taken in the Event the Safety Limit is Exceeded .. ... .. 44 ,

i 6.5.2 Action To Be Taken in the Event of a Reportable Occurrence ........ 44

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6.6 Reports.............................................................................. 45 t

! 6.6.1 Ope ratin g Rep orts . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 45 l 6.6.2 S peci al Re po rts . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 46 6.7 Records................................................................................. 46 6.7.1 Records To Be Retained for at Least Five Years ..... ...... .... ........ 46 i

6.7.2 Records To Be Retained for at bast One. Training Cycle ......... ... 47 6.7.3 . Records To Be Retained for the Life of the Reactor Facility . ....... 47 I

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4 9003 Idb Amendment No. 32 Rev.1 (4/24/97

1 TECHNICAL SPECIFICATIONS FOR THE PENN STATE BREAZEALE REACTOR (PSBR)

FACILITY LICENSE NO. R-2 V

1.0 INTRODUCTION

Included in this document are the Technical Specifications and the bases for the Technical Specifications. These bases, which provide the technical support for the individual technical specifications, are included for information purposes only. They are not part of the Technical Specifications and they do not constitute limitations or requirements to which the licensee must adhere.

1.1 Definitions 1.1.1 ALARA The ALARA (As Low As Reasonably Achievable) program is a program for maintaining occupational exposures to radiation and release of radioactive effluents to the environs as low as reasonably achievable.

l 1.1.2 Automatic Control .

Automatic control mode operation is when normal reactor operations, including start up, power level change, power regulation, and protective power reductions are performed by the reactor control system without, or with minimal, operator intervention.

1.1.3 Channel O A channel is the combination of sensor, line, amplifier, and output devices O which are connected for the purpose of measuring the value of a parameter.

1.1.4 Channel Calibration A channel calibration is an adjustment of the channel such that its output responds, with acceptable range, and accuracy, to known values of the parameter which the channel measures. Calibration shall encompass the entire channel, including equipment actuation, alarm, or trip, and shall be deemed to include a Channel Test.

1.1.5 Channel Check A channel check is a qualitative verification of acceptable performance by observation of channel behavior. This verification, where possible, shall include comparison of the channel with other independent channels or systems measuring the same variable.

1.1.6 Channel Test A channel test is the introduction of a signal into the channel to verify that it is operable.

1.1.7 Cold Critical Cold criticalis the condition of the reactor when it is critical with the fuel and O

y/

bulk water temperatures both below 100'F (37.8'C).

9003Idb Amendment No. 32 Rev.1 (4/24/97)

2' l.1.8 Close Packed Array Closed packed array is an arrangement of fuel elements wherein no empty Q

V grid positions are completely surrounded by fuel elements.

1.1.9 Confinement Confinement means an enclosure on the overall facility which controls the movement of air into it and out through a controlled path.

1.1.10 Core Lattice Position The core lattice position is that region in the core over a grid plate hole used to position a fuel element. It may be occupied by a fuel element, a control rod, an experiment, an experimental facility, or a reflector element.

1.1.11 Excess Reactivity Excess reactivity is that amount of reactivity that would exist if all control rods (safety, regulating, etc.) were moved to the maximum reactive condition from the point where the reactor is exactly critical (keff=1)in the reference core condition.

1.1.12 Exneriment Experiment shall mean (a) any apparatus, device, or material which is not a normal part of the core or experimental facilities, but which is inserted in these facilities or is in line with a beam of radiation originating from the reactor core; or (b) any operation designed to measure reactor parameters or g characteristics.

1.1.13 Exnerimental Facility Experimental facility shall mean beam port, including extension tube with shields, thermal column with shields, vertical tube, central thimble, in-core irradiation holder, pneumatic transfer system, and in-pool irradiation facility.

1.1.14 Instrumented Element i An instrumented element is a TRIGA fuel element in which sheathed chromel-alumel or equivalent thermocouples are embedded in the fuel.

1.1.15 Limitine Conditions for Oneration Limiting conditions for operation of the reactor are those constraints included in the Technical Specifications that are required for safe operation of the facility. These limiting conditions are applicable only when the reactor is operating unless othenvise specified.

1.1.16 Limiting Safety System Setting i

l A limiting safety system setting (LSSS) is a setting for an automatic protective i device related to a variable having a significant safety function.  ;

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90v3Idb l Amendment No. 32 Rev.1 (4/24/97)

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l 1.1.17 Manual Control Manual control mode is operation of the reactor with the power level controlled 'oy the operator adjusting the control rod positions.

1.1.18 Maximum Elemental Power Density l

The maximum elemental power density (MEPD) is the power density of the element in the core producing more power than any other element in that loading. The power density of an element is the total power of the core  ;

divided by the number of fuel elements in the core multiplied by the normalized power of that element.  ;

l 1.1.19 Maximum Power Level Maximum Power Levelis the maximum power (1.10 MW) to be used for power channel calibrations only. Also, see section 4.2.5.

l 1.1.20 Measured Value The measured value is the value of a parameter as it appears on the output of a channel.

l 1.1.21 Movable Exneriment A movable experiment is one where it is intended that the entire experiment may be l moved in or near the core or into and out of the reactor while the reactor is operating.

l 1.1.22 Normali7.ed Power

(\v# t The normalized power, NP, is the ratio of the power of a fuel element to the average power per fuel element. l l 1.1.23 Onerable l Operable means a component or system is capable of performing its intended function.

l 1.1.24 Onerating Operating means a component or system is performing its intended function.

l 1.1.25 Pulse Mods Pulse mode operation shall mean operation of the reactor allowing the operator to insert preselected reactivity by the ejection of the transient rod.

1.1.26 Reactivity Limits l

The reactivity limits are those limits imposed on reactor core reactivity.

Quantities are referenced to a reference core condition.

l l 1.1.27 Reactivity Worth of an Experiment The reactivity worth of an experiment is the maximum absolute value of the reactivity change that would occur as a result ofinteuded or anticipated changes or credible malfunctions that alter experiment position or configuration.

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'D 9003Idb Amendment No. 32 Rev.1 (4/24/97) l l

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4 l 1.1.28 Reactor Control System The reactor control system is composed of control and operadonal interlocks, reactivity adjustment controls, flow and temperature controls, and display systems which permit the operator to operate the reactor reliably in its allowed modes.

l 1.1.29 Reactor Interlock A reactor interlock is a device which prevents some action, associated with reactor operadon, until certain reactor operation conditions are satisfied.

l 1.1.30 Reactor Operating The reactor is operating whenever it is not secured or shutdown.

l 1.1.31 Reactor Secured The reactor is secured when: i

a. It contains insufficient fissile material or moderator present in the reactor, adjacent experiments, or control rods, to attain criticality under optimum available conditions of moderation, and reflection, or

b. A combination of the following:

1) The minimum number of neutron absorbing control rods are fully inserted or other safety devices are in shutdown positions, as required by technical specifications, and

( i V 2) The console key switch is in the off position and the key is removed from the lock, and

3) No work is in progress involving coce fuel, core structure, installed control rods, or control rod drives unless they are physically decoupled from the control rods, and

4) No experiments in or near the reactor are being moved or serviced that l have, on movement, a reactivity worth exceeding the maximum value l allowed for a single experiment or one dollar whichever is smaller. l l 1.1.32 Reactor Shutdown The reactor is shutdown if it is subcritical by at least one dollar in the reference core condition and the reactivity worth of all experiments is included. ,

1 1.1.33 Reactor Safety System l l

Reactor safety systems are those systems, including their associated input channels, which are designed to initiate automatic reactor protection or to provide information for initiation of manual protective action.

O, 9003Idb Amendment No. 32 Rev.1 (4/24/97)

t l 5 l l 1.1.34 Reference Core Condition l ,- m The condition of the core when it is at ambient temperature (cold) and the reactivity worth of xenon is negligible (<0.21% Ak/k(~$0.30)).

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l 1.1.35 Research Reactor i

A research reactor is defined as a device designed to support a self-sustaining l neutron chain reaction for research, development, educational, training, or  !

experimental purposes, and which may have provisions for the production of I radioisotopes.

l l 1.1.36 Reportable Occurrence l

A reportable occurrence is any of the following which occurs dunng reactor I operation- 1 1

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a. Operation with the safety system setting less conservative than specified in Section 2.2, Limiting Safety System Setting.

b. Operation in violation of a limiting condition for operation.

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c. Failure of a required reactor safety system component which could render  ;

the system incapable of performing its intended safety function.

d. Any unanticipated or uncontrolled change in reactivity greater than one dollar.

O c. An observed inadequacy in the implementation of either administrative or b procedural controls which could result in operation of the reactor outside the limiting conditions for operation.

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f. Release of fission products from a fuel element.

g. Abnormal and significant degradation in reactor fuel, cladding, coolant boundary or contamment boundary that could result in exceeding i 10 CFR Part 20 exposure criteria. l l h. Any other violation of NRC regulations.

l 1.1.37 Rod-Transient I

The transient rod is a control rod with SCRAM capabilities that is capable of  ;

, providing rapid reactivity insertion for use in either pulse or square wave l l mode of operation. j 1.1.38 Safety Limit l

Safety limits are limits on important process variables which are found to be necessary to reasonably protect the integrity of certain physical barriers which guard against the uncontrolled release of radioactivity. The principal physical ,

barrier is the fuel element cladding.

O 9003Idb l Amendment No. 32 Rev.1 (4/24/97) i i

6 l 1.1.39 SCRAM Time l ,, SCRAM time is the elapsed time between reaching a limiting safety system set point and a specified control rod movement.

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l 1.1.40 Secured Exneriment A secured experiment is any experiment, experimental facility, or component of an experiment that is held in a stationary position relative to the reactor by mechanical means. The restraining forces must be substantially greater than those to which the experiment might be subjected to by hydraulic, pneumatic, buoyant, or other forces which are normal to the operating environment of the experiment, or by forces which can arise as a result of credible malfunctions.

I l 1.1.41 Secured Exneriment with Movable Parts i

A secured experiment with movable parts is one that contains parts that are '

intended to be moved while the reactor is operating.

l 1.1.42 Shall. Should. and Mg The word "shall" is used to denote a requirement; the word "should" to denote a recommendation; and the word "may" to denote permission, neither a )

requirement nor a recommendation.

l 1.1.43 Shim. Regulatine. and Safety Rods A shim, regulating, or safety rod is a control rod having an electric motor p drive and SCRAM capabilities. It has a fueled follower section. l I

'l 1.1.44 Shutdown Margin Shutdown margin shall mean the minimum shutdown reactivity necessary to provide confidence that the reactor can be made suberitical by means of the control and safety syst:ms starting from any permissible operating condition although the most reactive rod is in its most reactive position, and that the reactor will remain suberitical without further operator action.

l l 1.1.45 Sguare Wave Mode l Square wave (SW) mode operation shall mean operation of the reactor allowing the operator to insert preselected reactivity by the ejection of the transient rod, and which results in a maximum power within the license limit.

l 1.1.46 Steady State Power Level Steady state power level is the maximum power level (1.0 MW) used in all operations except for power channel calibrations, see 4.2.5 Overpower Scram. ,

l 1.1.47 TRIGA Fuel Element A TRIGA fuel element is a single TRIGA fuel rod of standard type, either 8.5 wt% U-ZrH in stainless steel cladding or 12 wt% U-ZrH in stainless steel cladding enriched to less than 20% uranium-235.

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U 9003Idb Amendment No. 32 Rev.1 (4/24S7) l

! l 1.1.48 Watchdoc Circuit l

! A watchdog circuit is a circuit consisting of a timer and a relay. The timer

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V energizes the relay as long as it is reset prior to the expiration of the timing interval. If it is not reset within the timing interval, the relay will de-energize thereby causing a SCRAM.

1 2.0 SAFETY LIMIT AND LIMITING SAFETY SYSTEM SETTING  !

2.1 Safety Limit-Fuel Element Temnerature Apolicability The safety limit specification applies to the maximum temperature in the reactor fuel.

Obiective The objective is to define the maximum fuel element temperature that can be permitted with confidence that no damage to the fuel element and/or cladding will result.

l Snecification The temperature in a water-cooled TRIGA fuel element shall not exceed 1150'C under any operating condition.

Pasis The important parameter for a TRIGA reactor is the fuel element temperature. This

, parameter is well suited as a single specification especially since it can be measured at (q-} a point within the fuel element and the relationship between the measured and actual temperature is well characterized analytically. A loss in the integrity of the fuel element cladding could arise from a build-up of excessive pressure between the fuel-moderator and the cladding if the maximum fuel temperature exceeds 1150'C. The pressure is caused by the presence of air, fission product gases, and hydrogen from the dissociation of the hydrogen and zirconium in the fuel-moderator. The magnitude of this pressure is determined by the fuel-moderator temperature, the ratio of hydrogen to zirconium in the alloy, and the rate change in the pressure.

The safety limit for the standard TRIGA fuel is based on data, including the h.rge mass of experimental evidence obtained during high performance reactor tests on this fuel.

These data indicate that the stress in the cladding due to the increase in the hydrogen I

pressure from the dissociation of zirconium hydride will remain below the ultimate stress provided that the temperature of the fuel does not exceed 1150 C and the fuel cladding is below 500'C. See Safety Analysis Report, Ref.13 and 30 in Section IX and Simnad, M.T., F.C. Fot.shee, and G.B. West, " Fuel Elements for Pulsed Reactors,"

Nucl. Technology, Vol. 28, p. 31-56 (January 1976).

2.2 Limitine Safety System Settine (LSSS) 1 Annlicability l

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The LSSS specification applies to the SCRAM setting which prevents the safety limit l from being reached.

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I 9003Idb l Amendment No. 32 Rev.1 (4n4/97)

8 Objective The objective is to prevent the safety limit (1150*C) from being reached.

Gl Soecification The limiting safety system setting shall be a maximum of 650*C as measured with an instrumented fuel element ifit is located in a core position representative of the maximum elemental power density (MEPD) in that loading. If it is not practical to locate the instrumented fuel in such a position, the LSSS shall be reduced. The reduction of the LSSS shall be by a ratio based on the calculated linear relationship between the normalized power at the monitored position as compared to normalized power at the core position representative of the MEPD in that loading.

Basis The limiting safety system setting is a temperature which, if reached, shall cause a reactor SCRAM to be initiated preventing the safety limit from being exceeded.

Experiments and analyses described in the Safety Analysis Repon,Section IX - Safety Evaluation, show that the measured fuel temperature at steady state power has a simple linear relationship to the normalized power of a fuel element in the core. Maximum fuel temperature occurs when an instrumented element is in a core position of MEPD.

The actual location of the instrumented element and the associated LSSS shall be chosen by calculation and/or experiment prior to going to maximum reactor operational power level. The measured fuel temperature during steady state operation is close to the maximum fuel temperature in that element. Thus,500*C of safety margin exists before the 1150*C safety limit is reached. This safety margin provides adequate compensation for variations in the temperature profile of depleted and differently loaded fuel elements (i.e. 8.5 wt% vs.12 w % fuel elements). See Safety Analysis Repon,Section IX.

If it is not practical to place an instrumented element in the position representative of MEPD the LSSS shall be reduced to maintain the 500*C safety margin between the 1150*C safety limit and the highest fuel temperature in the core ifit was being measured. The reduction ratio shall be determined by calculation using the accepted techniques used in Safety Analysis Report,Section IX.

In the pulse mode of operation, the same LSSS shall apply. However, the temperature channel will have no effect on limiting the peak power or fuel temperature, generated, because of its relatively long time constant (seconds), compared with the width of the pulse (milliseconds).

3.0 LIMITING CONDITIONS FOR OPERATION The limiting conditions for operation as set forth in this section are applicable only when the reactor is operating. They need not be met when the mactor is shutdown unless specified otherwise.

3.1 Reactor Core Parameters i

1 3.1.1 Constant Power and Square Wave Oneration Applicability n These specifications apply to the maximum power generated during manual control mode, automatic control mode, and square wave mode operations.

(v) 9003Idb Arnendment No. 32 Rev.1 (4/24N7) l

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l Obiective

, The objective is to limit the source term and energy production to that used in t

y/ the Safety Analysis Report. l l Specifications

a. The reactor operational power shall not be intentionally raised above 1 MW except l for pulse operation (see specification 3.1.4) and for power channel calibration.

b. The steady state fuel tem wrature shall be a maximum of 650*C as measured with an instrumented fue element ifit is located in a core position L representative of MEPD in that loading. Ifit is not practical to locate the l instrumented fuel in such a position, the steady state fuel temperature shall be calculated by a ratio based on the calculated linear relationship between the normalized power at the monitored position as compared to normalized j power at the core position representative of the MEPD m that loading. In i this case, the measured steady state fuel temperature shall be limited such that the calculated steady state fuel temperature at the core position representative of the MEPD in that loading shall not exceed 650*C.

c. A maximum power level shall be no greater than 1.10 MW for brief periods to calibrate power channels.

l l Bases l a. Thermal and hydraulic calculations and operational experience indicate that a compact TRIGA reactor core can be safely operated up to power levels of l at least 1.15 megawatts with natural convective cooling. Power operation at l 1.15 megawatts will not produce fuel temperatures which exceed the LSSS from section 2.2 using any allowed core configuration.

Small local variations can occur about the maximum allowed power for a given core loading during normal operation and still provide a large margin of safety in that the maximum fuel temperature remams well below tb safety limit. See Safety Analysis Report, section IX.

b. Limiting the maximum steady state measured fuel temperature of any

! position to 650*C places an upper bound on the fission product release I

fraction to that used in the analysis of a Maximum Hypothetical Accident (MHA). See Safety Analysis Report, section IX.

c. Operation at 1.10 MW is wimin the bounds establshed by the SAR for both ,

steady state operations and pulse operation. See pages IX 25 & 27 of the SAR.

i l 3.1.2 Reactivity Limitation l

l Apnlicability This specification applies to the reactivity condition of the reactor and the i reactivity worth of control rods, experiments, and experimental facilities. It applies to all modes of operation.

Obiective i s

! The oldective is to assure that the reactor is operated within the limits analyzed in the Safety Analysis Report and to assure that the safety limit will not be exceeded.

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I 10 l Specification  :

O The maximum excess reactivity above cold, clean, critical plus samarium poison of the core configuration with experiments and experimental facilities in place shall be 4.9% Ak/k (~$7.00).

j Hasis

! Limiting the excess reactivity of the core to 4.9% Ak/k (~$7.00) prevents the fuel temperature in the core from exceeding 1150*C under any assumed i accident condition as described in the Safety Analysis Repon,Section IX . j 3.1.3 Shutdown Margin i Applicability This specification applies to the reactivity condition of the reactor and the ,

reactivity worth of control rods, experiments, and ex.perimental facilities. It applies to all modes of operation.

l Objective The objective is to assure that the reactor can be shut down at all times and to 1 assure that the safety limit will not be exceeded.

l Specification

. The reactor shall not be operated unless the shutdown margin provided by control rods is greater than 0.175% Ak/k (~$0.25) with

a. All movable experiments, experiments with movable parts, and experimental facilities in their most reactive state.

b. The highest reactivity worth control rod fully withdrawn.

Basis A shutdown margin of 0.175% Ak/k (~$0.25) assures that the reactor can be made suberitical from any operating condition even if the highest worth control j rod should remain in the fully withdrawn position. In addition, the 0.175%

Ak/k (~$0.25) can be easily measured, and therefore verifying that the operation meets this specification. l 3.1.4 Pulse Mode Operation .

Apolicability These specifications apply to the energy generated in the reactor as a result of a i pulse insertion of reactivity.

Obiective I

i The objective is to assure that the safety limit will not be exceeded during pulse j mode operation.

9003Idb Amendment No. 32 Rev.1 (4/24/97)

11 l Specifications

a. The stepped reactivity insenion for pulse operation shall not exceed

a 2.45% Ak/k (~$3.50) and the maximum worth of the poison section of the transient rod shall be limited to 2.45% Ak/k (~$3.50).

b. Pulses shall not be initiated from power levels above 1 kw.

Bases

a. Experiments and analyses described in the Safety Analysis Report,Section IX.C., show that the peak pulse temperatures can be predicted for new 12 wt% fuel placed in any core position. These experiments and analyses show that the maximum allowed pulse reactivity of 2.45% Ak/k (~$3.50), prevents the maximum fuel temperature from reaching the safety limit (1150*C) for any core configuration that meets the requirements of 3.1.5.

The maximum worth of the pulse rod is limited to 2.45% Ak/k (~$3.50) to prevent exceeding the safety limit (1150*C) with an accidental ejection of the transient rod.

b. If a pulse is initiated from power levels below I kw, the maximum allowed full worth of the pulse rod can be used without exceeding the safety limit.

3.1.5 Core Confieuration Limitation Aonlicability t t

't /l These specifications apply to all core configurations.

Obiective The objective is to assure that the safety limit (1150*C) will not be exceeded due to power peaking effects in the various core configurations.

Soecifications

a. The critical core shall be an assembly of either 8.5 wt% U-ZrH stainless steel clad or a mixture of 8.5 wt% and 12 wt% U-ZrH stainless steel clad TRIGA fuel-moderator elements placed in water with a 1.7 inch center line grid spacing.

l b. The maximum calculated MEPD shall be less that 24.7 kw per fuel element.

c. When the keff of the core is less than or equal to 0.99 with all control rods at their upper limit, the fuel may or may not be arranged in a close packed array. The source and detector shall be arranged such that the keff of the subcritical assembly shall always be monitored to assure compliance with keff s 0.99 when all control rods are fully withdrawn.

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d. The NP of any core loading with a maximum allowed pulse worth of 2.45%

Ak/k (~$3.50) shall be limited to 2.2. If the maximum allowed pulse worth is less than 2.45% Ak/k ($~3.50) for any given core loading h] (i. e. the pulse can be limited by the total worth of the transient rod, by the core excess, or administratively), the maximum NP can be increased. The maximum NP can be increased above 2.2 as long as the calculated maximum fuel temperature does not exceed the safety limit with that maximum allowed pulse worth and NP. In addition, the Reactivity Accident 1 in the Safety Analysis Report shall be evaluated to ensure that the safety l limit is not exceeded with the new conditions (See Safety Analysis Report, i Section IX.). l Bases I l

a. The safety analysis is based on an assembly of either 8.5 wt% U-ZrH  !

stainless steel clad or a mixture of 8.5 wt% and 12 wt% U-ZrH stainless l steel clad TRIGA fuel-moderator elements placed in water with a 1.7 inch center line grid spacing.

b. Limiting the MEPD to 24.7 kw per element places an upper bound on the elemental heat production and the source term of the PSBR to that used in the analysis of a Loss Of Coolant Accident (LOCA) and Maximum )

Hypothetical Accident (MHA) respectively. See Safety Analysis Report,Section IX.

c. When the keff of the core is less than or equal to 0.99 with all control rods at their upper limit, the core can not be taken critical. Hence, the requirement for close packed arrays is not necessary to prevent the core from attaining

,3 (d high fuel temperatures.

d. The maximum NP for a given core loading determines the peak pulse temperature with the maximum allowed pulse worth. If the maximum allowed pulse worth is reduced the maximum NP can be increased without .

exceeding the safety limit (1150*C). The amount ofincrease in the  !

maximum NP allowed shall be calculated by an accepted method documented by an administratively approved procedure.

3.1.6 TRIG A Fuel Elements Anolicability l l These specifications apply to the mechanical condition of the fuel.

Objective The objective is to assure that'the reactor is not operated with damaged fuel that might allow release of fission products.

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l O 9003Idb j Amendment No. 32 Rev.1 (4/24/97) l

13 l Soecifications The reactor shall not be operated with damaged fuel except to detect and

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V l identify the fuel element for removal. A TRIGA fuel element shall be considered damaged and shall be removed from the core if:

a. In measuring the transverse bend, the bend exceeds the limit of 0.125 inch over the length of the cladding.

b. In measuring the elongation, its length exceeds its originallength by 0.125 inch.

c. A clad defect exists as indicated by release of fission products.

Bases

a. The limit of transverse bend has been shown to result in no difficulty in disassembling the core. Analysis of the removal of heat from touching fuel elements shows that them will be no hot spots which cause damage to the fuel.

b. Experience wiin iRIGA reactors has shown that fuel element bending that I could result in touching has occurred without deleterious effects. This is because (1) during steady state operation, the maximum fuel temperatures  ;

are at least 500*C degrees Centigrade below the safety limit (1150'C), and 1 (2) during a pulse, the cladding temperatures remain well below their stress limit. The elongation limit has been specified to assure that the cladding  ;

material will not be subjected to strains that could cause a loss of fuel i fm integrity and to assure adequate coolant flow i i l

\ l 3.2 Reactor Control and Reactor Safety System l

3.2.1 Reactor Control Rods l Acolicabilit,y This specification applies to the reactor control rods.

Obiective The objective is to assure that sufficient control rods are operable to maintain the reactor suberitical.

Soecification There shall be a minimum of three operable control rods in the reactor core.

Basis The shutdown margin and excess reactivity specifications require that the reactor can be made suberitical with the most reactive control rod withdrawn.

This specification helps assure it.

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9003Idb Amendment No. 32 Rov.1 (4/24/97)

l 14 l 3.2.2 Manual Control and Automatic Control ]

Arolicability i

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U This specification applies to the maximum reactivity insertion rate associated with movement of a standard control rod out of the core.

Obiective l The objective is to assure that adequate control of the reactor can be maintained i during manual and 1,2, or 3 rod automatic control.

Specification The rate of reactivity insertion associated with movement of either the l regulating, shim, or safety control rod shall be not greater than 0.63% Ak/k l l (~90g) per second when averaged over full rod travel. If the automatic control l uses a combination of more than one rod, the sum of the reactivity of those rods 1 j shall be wt greater than 0.63% Ak/k (~90e) per second when averaged over full l

I travel. .

l Basis The ramp accident analysis (refer to Safety Analysis Report, Chapter IX) i l indicates that the safety limit (1150*C) will not be exceeded if the reactivity addition rate is less than 1.75% Ak/k (~$2.50) per second, when averaged over full travel. This specification of 0.63% Ak/k (~90 ) per second, when averaged l l m over full travel, is well within that analysis.

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3.2.3 Reactor Control System Acolicability  ;

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This specification applies to the information which must be available to the reactor operator during reactor operation.

Obiective The objective is to require that sufficient information is available to the operator l to assure safe operation of the reactor.

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Soecification l The reactor shall not be operated unless the measuring channels listed in Table 1 l are operable. (Note that MN, AU, and SW are abbreviations for manual control mode, automatic control mode, and square wave mode, respectively).

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i, I 9003Idb 1 Amendment No. 32 Rev.1 (4/24/97) l

15 hV Table 1 Measuring Channels t Min. No. Effective Mode Measurine Channel Onerable MN. AU hise EE  :

Fuel Element Temperature 1 X X X Linear Power I X X Percent Power 1 X X r Pulse Peak Power 1 X '

. Count Rate 1 X Log Power .. 1 X X Reactor Period 1 X s

Basis Fuel temperature displayed at the control console gives continuous information ,

on this parameter which has a specified safety limit. The power level monitors assure that the reactor power level is adequately monitored for the manual l control, automatic control, square wave, and pulsing modes of operation. The specifications on reactor power level and reactor period indications are included in this section to provide assurance that the reactor is operated at all times within the limits allowed by these Technical Specifications.

-l 3.2.4 Reactor Safety System and Reactor Interlocks Aoplicability This specification applies to the reactor safety system channels, the reactor interlocks, and the watchdog ciremt.

Objective j The objective is to specify the minimum number of reactor safety system l channels and reactor interlocks that must be operable for safe operation.

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9003Idb I Amendment No. 32 Rev.1 (4/24/97) i l

16 Snecification "

s The reactor shall not be operated unless all of the channels and interlocks  !

described in Table 2a and Table 2b are operable.

Table 2a l Minimum Reactor Safety ,

System Channels Number Effective Mode Channel Doerable Function MN. AU Biig M FuelTemperature 1 SCRAM $ 650"C* X X X l liigh Power . 2 SCRAM s 110% of maximum X X reactor operational power not i to exceed 1.1 MW-Detector Power Supply 1 SCRAM on failure of supply X X l voltage 1

l l

SCRAM Bar on 1 Manual SCRAM X X X

! Console .

l Preset Timer 1 Transient Rod SCRAM 15 X seconds or less after pulse 'I Watchdog Circuit 1 SCRAM on software or self- X X X l check failtee  !

l

• The limit of 650'C may be reduced based on specification 2.2.

Table 2b l Minimum ReactorInterlocks Number Effective Mode Channel Oncrable Function MN. AU Eulse E Source Level 1 Prevent rod withdrawal X .

without a neutron induced-signal on the startup channel Log Power 1 Prevent pulsing fromlevels X above 1 kw 4

l Transient Rod 1 Prevent applications of air X unless cylinder is fully inserted j Shim, Safety, and 1 Prevent movement of any X l Regulating Rod rod except the transient rod i '

Simultaneous Rod 1 Prevent simultaneous manual X X' Withdrawal withdrawal of two rods 1

/ t I U 9003Idb Amendment No. 32 Rev.1 (4/24S7)

17 Bases

a. A temperature SCRAM and two power level SCRAMS assure the reactor is O shutdown before the safety limit on the fuel element temperature is reached.

V' The actual setting of the fuel temperature SCRAM depends on the LSSS for that core loading and the location of the instrumented fuel element (see Technical Specification section 2.2).

b. The maximum reactor operational power may be administratively limited to less than 1 MW depending on section 3.1.5.b of this Technical Specification. The high power SCRAMS shall be set to no more than 110%

of the administratively limited maximum reactor operational power ifit is less than 1 MW.

c. Operation of the reactor is prevented by SCRAM if there is a failure of the detector power supply for the reactor safety system channels.

d. The manual SCRAM allows the operator to shut down the reactor in any mode of operation if an unsafe or abnormal condition occurs.

e. The preset timer ensures that the transient-rod will be inserted and the reactor will remain at low power after pulsing,

f. The watchdog circuit will SCRAM the reactor if the software or the self-checks fail (see Safety Analysis Report,Section VII).

g. The interlock to prevent startup of the reactor withot: a neutron induced signal assures that sufficient neutrons are available for proper startup in all n allowable modes of operation, s

v) h. The interlock to prevent the initiation of a pulse above 1 kw is to assure that fuel temperature is approximately pool temperature when a pulse is performed. This is to assure that the safety limit is not reached.

i. The interlock to prevent application of air to the transient rod unless the cylinder is fully inserted is to prevent pulsing the reactor in the manual control or automatic control mode,

j. In the pulse mode, movement of any rod except the transient rod is l prevented by an interlock. This interlock action prevents the addition of reactivity other than with the transient rod.

k. Simultaneous manual withdrawal of two rods is prevented to assure the reactivity rate of insertion is not exceeded.

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N.)

9003Idb Amendment No. 32 Rev.1 (4/24/97)

l 3.2.5 Core Loadine and Unloadine Oneration Apolicability I c 1 l

Vl This specificadon applies to the source level interlock.

Obiective The objecdve of this specification is to allow bypass of the source level interlock during operations with a subcritical core.

Specification During core loading and unloading operations when the reactor is subcritical, the source levelinterlock may be momentarily defeated using a spring loaded switch in accordance with the fuel loading procedure.

Basis During core loading and unloading, the reactor is subcritical. Thus, momentarily defeating the source levelinterlock is a safe operation. Should the core become inadvertently supercritical, the accidental insertion of reactivity will not allow fuel temperature to exceed the 1150*C safety limit because no single TRIGA fuel element is worth more than 1% Ak/k

(~$1.43) in the most reactive core position.

l 3.2.6 SCRAM Time Q Annlicability (O

This specification applies to the time required to fully insert any control rod to a full down position from a full up position.

Objective The objective is to achieve rapid shutdown of the reactor to prevent fuel damage.

Soecincation The time from SCRAM initiation to the full insertion of any control rod from a full up position shall be less than I second.

l Basis 1

This specification assures that the reactor will be promptly shut down when a SCRAM signalis initiated. Expedence and analysis, Safety Analysis Report,Section IX, have indicated that for the range of transients anticipated for a l l

TRIGA reactor, the specified SCRAM time is adequate to assure the safety of j the reactor. If the SCRAM signalis initiated at 1.10 MW, while the control rod is being withdrawn, and the negative reactivity is not inserted until the end of the one second rod drop time, the maximum fuel temperature does not reach the l safety limit. I h/

9003Idb i Amendment No. 32 Rev. I (4/24/97) I l

l

l 19 3.3 Coolant System j p 3.3.1 Coolant Level Limits

, ,V

! Acolienbility i

This specification applies to operation of the reactor with respect to a required depth of water above the top of the bottom grid plate.

Objective The objective is to assure that water is present to provide adequate personnel shielding and core cooling when the reactor is operated, and during a LOCA.

Snecification The reactor shall not be operated with less than 18 ft. of water above the top of the bottom grid plate.

Basis When the water is more than approximately 18 ft. above the top of the bottom grid plate, the water provides sufficient shielding to protect personnel during operation at 1 MW, and core cooling is achieved with natural circulation of the water through the core. Should the water level drop below approximately 18.25 ft. abt ve the top of the bottom grid plate while operating at 1 MW, a low pool level :larm (see Technical Specifications 3.3.2) will alert the operator who is required by administratively approved procedure to shutdown the reactor. Once

,A this alarm occurs it will take longer than 1300 seconds before the core is

(") completely uncovered because of a break in the 6" pipe connected to the bottom of the pool. Tests and calculations show that, during a LOCA,680 seconds is sufficient decay time after shutdown (see Safety Analysis Report,Section IX )

to prevent the fuel temperature from reaching 950*C. To prevent cladding rupture, the fuel and the cladding temperature must not exceed 950 "C (it is assumed that the fuel and the cladding are the same temperature during air cooling).

3.3.2 Detection of Leak or Loss of Coolant Anolicability This specification applies to detecting a pool water loss.

Objective The objective is to detect the loss of a significant amount of pool water.

Snecification A poollevel alarm shall be activated and corrective action taken when the pool level drops 26 cm from a level where the pool is full. '

l l

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! (u) 9003Idb Amendment No. 32 Rev.1 (4/24/97)

20 Basis The alarm occurs when the water levelis approximately 18.25 ft. above the top of the bottom grid plate. The point at which the poolis fullis approximately 19.1 ft. above the top of the bottom grid plate. The reactor staff shall take action to keep the core covered with water according to existing procedures.

The alarm is also transmitted to the Police Services annunciator panel which is monitored 24 hrs. a day. The alarm provides a signal that occurs at all times (see Safety Analysis Report,Section VII,). Thus, the alarm provides time to initiate corrective action before the radiation from the core poses a serious hazard.

3.3.3 Fission Product Activity Annlicability This specification applies to the detection of fission product activity.

Obiective The objective is to assure that fission products from a leaking fuel element are detected to provide opportunity to take protective action.

Snecification An air particulate monitor shall be operating in the reactor bay whenever the reactor is operating. An alarm on this unit shall activate a building evacuation alarm.

ex Basis ,

(LJ l This unit will be sensitive to airborne radioactive particulate matter containing fission products and fission gases and will alert personnelin time to take protective action.

l 3.3.4 Pool Water Sunnly for Leak Protection Annlicability This specification applies to pool water supplies for the reactor pool for leak protection.

Obiective The objective is to assure that a supply of water is available to replenish reactor l pool water in the event of pool water leakage.

Snecification A source of water of at least 100 GPM shall be available either from the University water supply or by diverting the heat exchanger secondary flow to the pool.

,m l A Q./'

9003Idb Amendment No. 32 Rev.1 (4/24/97)

l l 21 i l

l Provisions for both of these supplies are in place and will supply more than the specified flow rate. This flow rate will be more than sufficient to handle leak rates that have occurred in the past or any anticipated leak that might occur in the future.

3.3.5 Coolant Conductivity Limits

Aoplicability-I This specification applies to the conductivity of the water in the pool.

i  !

l Objectives l The objectives are:

L

a. To prevent activated contaminants from becoming a radiological hazard.  ;

b. To help preclude corrosion of fuel cladding and other primary system f l components. .  ;

!. Specification The reactor shall not be operated if the conductivity of the bulk pool water is greater than 5 microsiemens/cm (5 micromhos/cm).

! l

! Ham  !

l Experience indicates that 5 microsiemens/cm is an acceptable level of water L contaminants in an aluminum / stainless steel system such as that at the PSBR.

L Based on experience, activation at this level does not pose a significant  ;

radiological hazard, and significant corrosion of the stainless steel fuel cladding l will not occur when the conductivity is below 5 microsiemens/cm. '

3.3.6 Coolant Temperature Limits l Aoplicability

! This specification applies to the pool water temperature.

Objective The objective is to maintain the pool water temperature at a level that will not cause damage to the demineralizer resms.

l Snecification An alarm shall annunciate and corrective action shall be taken if during i

operation the bulk pool water temperature reaches 100*F (37.8'C).

l B.am l

l This s xcification is primarily to preserve demineralizer resins. Information 1 availa)le indicates that temperature damage will be minimal up to this

) temperature.

9003Idb Amendment No. 32 Rev.1 (4/24/97)

22 3.4 Confinement n Aoplicability (d

r This specification applies to reactor bay doors.

Obiective The objective is to assure that no large air passages exist to the reactor bay during reactor operation.

Specification The reactor bay tmck door shall be closed when the reactor is operating. Personnel doors to the reactor bay shall not be blocked open and left unattended when the reactor is operating.

Basis This specification helps to assure that the air pressure in the reactor bay is lower than the remainder of the building and the outside air pressure. Controlled air pressure is maintained by the air exhaust system and assures controlled release of any airborne radioactivity.

3.5 Engineered Safety Features - Facility Exhaust System and Emereencv Exhaust System Aonlicability o

() This specification applies to the operation of the facility exhaust system and the emergency exhaust system.

Obiective The objective is to mitigate the consequences of the release of airbome radioactive materials resulting from reactor operation.

Snecification The facility exhaust system shall be operating and the emergency exhaust system shall be maintained in an operable condition when the reactor is operating except for periods of time less than 48 hrs. when it is necessary to permit maintenance and repairs.

l l

3 rb 90031db Amendment No. 32 Rev.1 (4/2467,

23 Elsis i During normal operation, the concentration of airborne radioactivity in unrestricted (Q

(/

areas is below effluent release limits as described in the Safety Analysis Report,Section IX. In the event of a substantial release of airborne radioactivity, an air radiation monitor and/or an area radiation monitor will sound a building evacuation alarm which will automatically cause the facility exhaust system to close and the exhausted air to be passed through the emergency exhaust system filters before release.

This reduces the radiation within the building. The filters will reduce to <10% all of the particulate fission products that escape to the atmosphere. Radiation monitors, Section 3.6.1 of these Technical Specifications and Safety Analysis Report,Section VII, within the building, independent of the exhaust systems, will give waming of high levels of radiation that might occur during operation with the exhaust systems out of service.

3.6 Radiation Monitorine System 3.6.1 Radiation Monitorine Information Annlicability This specification applies to the radiation moititoring information which must be available to the reactor operator during reactor operation.

Obiective The objective is to assure that sufficient radiation monitoring information is available to the operator to assure personnel radiation safety during reactor operation.

V Snecification The reactor shall not be operated unless the radiation monitoring channels listed in Table 3 are operating.

Table 3 Radiation Monitoring Channels Radiation Monitoring Channels Function Number Area Radiation Monitor Monitor radiation levels 1 in the reactor bay.

Continuous Air Monitor radioactive 1 (Radiation) Monitor particulates in the reactor bay air.

Beam Laboratory Monitor Monitor radiation in the 1 l Beam Laboratory required only when the laboratory is in use.

1

'O 9003Idb j Amendment No. 32 Rev.1 (4/24/97)

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t

24 Bases l a. The radiation monitors provide information to operadng personnel of any impending or existing danger from radiation so that there will be sufficient time to evacuate the facility and to take the necessary steps to control the spread of radioacdvity to the surroundings.  ;

l b. He area radiation monitor in the Beam Laboratory provides information to the user and to the reactor operator when this laboratory is in use, j l

3.6.2 Evacuation Alarm Apphcability  ;

This specification applies to the evacuation alarm.

Obiective The objective is to assure that all personnel are alerted to evacuate the PSBR building when a potential radiation hazard exists within this building. t Specification The reactor shall not be operated unless the evacuation alarm is operable and audible to personnel withm the PSBR building when activated by the radiation  !

monitoring channels in Table 3 or a manual switch, i

l BMiLL The evacuation alarm produces a loud pulsating sound throughout the PSBR l building when there is any impending or existmg danger from radiation. The  ;

i sound notifies all personnel within the PSBR building to evacuate the building  ;

as prescribed by the PSBR emergency procedure.

l 3.6.3 Areon-41 Discharge Limit -

Annlicability This specification applies to the concentration of Argon-41 that may be ,

discharged from the PSBR.  !

Obiective L The objective is to insure that the health and safety of the public is not endangered by the discharge of Argon-41 from the PSBR. I Snecification l All Argon-41 concentrations produced by the operation of the reactor shall be

l. below the limits imposed by 10 CFR Part 20 when averaged over a year.

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! 9003Idb Amendment No.32 Rev.1 (4c4s7)

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

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

25 Basis i The maximum allowable concentradon of Argon-41 in air in unrestricted areas as specified in Appendix B, Table 2 of 10 CFR Part 20 is 1.0 x 10-8 Ci/ml.

Measurements of Argon-41 have been made in the reactor bay when the ,

reactor operates at 1 MW. These measurements show that the concentradons l

averaged over a year produce less than 1.0 x 10-8 Ci/mlin an ,

unrestricted area (see EnvironmentalImpact Appraisal, December 12,1996).

3.6.4 As Low As Reasonably Achievable (ALARA)

Anplicability This specification applies to all reactor operations that could result in j occupational exposures to radiation or the release of radioactive effluents to the environs.

Objective The objective is to maintain all exposures to radiation and release of radioacdve effluents to the environs ALARA.

Specification An AL ARA program shall be in effect.

Basis l

Having an ALARA program will assure that occupational exposures to radiation Q and the release of radioacdve effluents to the environs will be ALARA. Having such a formal program will keep the staff cognizant of the importance to minimize radiadon exposures and effluent releases.

3.7 Limitations of Experiments Anolicability These specifications apply to experiments installed in the reactor and its experimental facilities.

Objective The objecdve is to prevent damage to the reactor and to prevent excessive release of radioactive materials in the event of an expedment failure.

Specifications The reactor shall not be operated unless the following condidons governing experiments exist:

l 9003Idb Amendment No. 32 Rev.1 (4/24/97)

l 26

a. The reactivity of a movable experiment and/or movable portions of a secured experiment plus the maximum allowed pulse reactivity shall be less than 2.45% Ak/k

(-$3.50). However, the reactivity of a movable experiment and/or movable portions 4 of a secured experiment shall have a reactivity worth less than 1.4% Ak/k (~$2.00).

When a movable experiment is used, the maximum allowed pulse shall be reduced below the allowed pulse reactivity insertion of 2.45% Ak/k (~$3.50) to assure that the sum is less 2.45% Ak/k (~$3.50).

b. A single secured experiment shall be limited to a maximum of 2.45% Ak/k

(~$3.50). The sum of the reactivity worth of all experiments shall be less than 2.45% Ak/k (~$3.50).

c. When the keff of the core is less than 1 with all control rods at their upper limit and no experiments in or near the core, secured negative reactivity experiments may be added without limit.

d. An experiment may be irradiated or an experimental facility may be used in conjunction with the reactor provided its use does not constitute an unreviewed safety question. The failure mechanisms that shall be analyzed include, but are not limited to corrosion, overheating, impact from projectiles, chemical, and mechanical explosions.

Explosive material sufficient to cause such damage shall not be stored or used in the facility without proper safeguards to prevent release of fission products or loss of reactor shutdown capability.

77 If an experimental failure occurs which could lead to the release of fission products I

() or the loss of reactor shutdown capability, physical inspection shall be performed to determine the consequences and the need for corrective action. The results of the inspection and any corrective action taken shall be reviewed by the Director or a designated alternate and determined to be satisfactory before operation of the reactor is resumed.

e. Experiment materials, except fuel materials, which could off-gas, sublime, volatilize, or produce aerosols under (1) normal operating conditions of the experiment and reactor, (2) credible accident conditions in the reactor, or (3) possible accident conditions in the experiment, shall be limited in activity such that the airbome concentration of radioactivity averaged over a year shall not exceed the limit of Appendix B Table 2 of 10 CFR Part 20.

When calculating activity limits, the following assun;ptions shall be used:

1) If an experiment fails and releases radioactive gases er aerosols to the reactor bay or atmosphere,100% of the gases or aerosols escal:e.

2) If the effluent from an experimental facility exhausts through a holdup tank which closes automatically on high radiation level, at least 10% of the gaseous activity or aerosols produced will escape.

3) If the effluent from an experimental facility exhausts through a filter installation designed for greater than 99% efficiency for 0.3 micron particles, at least 10%

O of these vapors can escape.

i t N,j 9003Idb Amendment No. 32 Rev.1 (4/24/97)

27

4) For materials whose boiling point is above 130*F and where vapors formed by boiling this material can escape only through an undisturbed column of water i ,, above the core, at least 10% of these vapors can escape.

f. Each fueled experiment shall be controlled such that the totalinventory ofiodine isotopes 131 through 135 in the experiment is ne ne ater than 1.5 curies. In addiuon, any fueled experiment which would ger an inventory of more than 5 millicuries (mci) of I-131 through I-135 shall be rev d to assure that in the case of an accident, the total release of iodine will not exceed that postulated for the MHA (see Safety Analysis Report,Section IX,).

l Bases

a. This specification limits the sum of the reactivities of a maximum allowed pulse and a movable experiment to the specified maximum reactivity of the transient rod.

This limits the effects of a pulse simultaneous with the failure of the movable experiment to the effects analyzed for a 2.45% Ak/k (~$3.50) pulse. In i

addition, the maximum power attainable with the ramp insertion of 1.4% Ak/k

! (~$2,00) is less than 500 kW starting from critical.

b. The maximum worth of all experiments is limited to 2.45% Ak/k (~$3.50) so that their inadvertent sudden removal from the cold critical reactor will not result in the i reactor achieving a power level high enough to exceed the temperature safety limit l (1150*C). The worth of a single secured experiment is limited to the allowed pulse l reactivity insertion as an increased measure of safety. Should the 2.45% Ak/k, l

(~$3.50) reactivity be inserted by a ramp increase, the maximum power attainable i is less than 1 MW.

g Since the initial core is suberitical, adding and then inadvertently removing all

() c.

negative reactivity experiments leaves the core in its initial subcritical condition.

l

d. The design basis accident is the MHA (See Safety Analysis Report,Section IX). A l

chemical explosion, such as detonated TNT or a mechanical explosion, such as a steam explosion or a high pressure gas container explosion may release enough energy to cause release of fission products or loss of reactor shutdown capability. A projectile with a large amount of kinetic energy could cause release of fission products or loss of reactor shutdown capability. Accelerated corrosion of the fuel cladding due to material released by a failed experiment could also lead to release of fission products. No experiment shall be conducted that is an unreviewed safety question l

If an experiment failure occurs a special investigation is required to ensure that all l

effects from the failure are known before operation proceeds.

l

e. This specification is intended to reduce the likelihood that airborne activities in l excess of the limits of Appendix B Table 2 of 10 CFR Part 20 will be released to l

the atmosphere outside the facility boundary.

f. The 5 mCilimitation on I-131 through I-135 assures that in the event of failure of a fueled experiment, the exposure dose at the exclusion area boundary will be less l than that postulated for the MHA (See Safety Analysis Report,Section IX) even if the iodine is released in the air.

C l

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9003Idb Amendment No. 32 Rev.1 (4/24/97)

l 28 4.0 SURVELLLANCE REQUIREMENTS

. 4.1 Reactor Parameters V 4.1.1 Reactor Power Calibration Anolicability This specification applies to the surveillance of the reactor power calibration.

Obiective The objective is to verify the perfonnance and operability of the power measuring channel.

l Specification A thermal power channel calibration shall be made on the linear power level monitoring channel annually, not to exceed 15 months.

Basis The thermal power level channel calibration will assure that the reactor is to be operated at the authorized power levels.

4.1.2 Reactor Excess Reactivity Annlicability (l

This specification applies to surveillance of com excess reactivity.

\s i Obiective The objective is to assure that the reactor excess reactivity does not t xceed the Technical Specifications and the limit analyzed in Safety Analysis Report,Section IX.F.

Snecification The excess reactivity of the core shall be measured annually, not to exceed 15 months, and following core or control rod changes equal to or greater than 0.7%

Ak/k (~$1.00).

Bnts Excess reactivity measurements on this schedule assure that no unexpected l changes have occurred in the core and the core configuration does not exceed l

I excess reactivity limits established in the Technical Specifications. l 1

l 4.1.3 TRIGA Fuel Elements

! f Anolicability  !

i This specification applies to the surveillance requirements for the TRIGA fuel j l 3 elements. l N) l 9003Idb Amendment No. 32 Rev.1 (4/24/97)

b 29 Obiective The objective is to verify the continuing integrity of the fuel element cladding. j Specification All fuel elements and control rods with fuel followers shall be inspected visually for damage or deterioration and measured for length and bend before being placed in the core for the first time and at intervals not to exceed the sum .!

of 3,500 dollars in pulse reactivity or two years, not to exceed 30 months, whichever comes first. j EmLi

The frequency ofinspection and measurement schedule is based on tae parameters most like?,y to affect the fuel cladding of a pulsing reactor operated  !

at moderate pulsing levels and utilizing fuel elements whose charac teristics are l

well known.

l 4.2 Reactor Control and Safety System 4.2.1 Reactivity Worth l r

l Anolicability ,

.This specification applies to the reactivity worth of the control rods.  ;

i Obiective .

f  :

The objective is to assure that the control rods are capable of maintaining the _  !

reactor subcritical.

• Soecification  !

l Re reactivity worth of each control rod and the shutdown margin for the core i loading in use shall be determined annually, not to exceed 15 months, or  !

i following core or control rod changes equal to or greater than 0.7% Ak/k i

(~$ 1.00).

HMis The reactivity worth of the control rod is measured to assure that the required shutdown margin is available and to provide an accurate means for determining

, the core excess reactivity, maximum reactivity, insertion rates, and the reactivity l worth of experiments inserted in the core.

i 4.2.2 Reactivity Insertion.Riga 1 l Applicability l

g This specification applies to control rod movement speed.

Objective l The objective is to assure that the reactivity addition rate specification is not

! . violated and that the control rod drives are functioning.

9003Idb

, Amendment No. 32 Rev.1 (4/24/97) i 1

1

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

(

i

! 30 Snecification

,- The rod drive speed both up and down and the time from SCRAM initiation to

( the fullinsertion of any control rod from the full up posidon shall be measured b annually, not to exceed 15 months, or when any significant work is done on the rod drive or the rod.

Basis This specification assures that the reactor will be promptly shut down when a SCRAM signalis initiated. Experience and analysis have indicated that for the range of transients anticipated for a TRIGA reactor, the specified SCRAM time is adequate to assure the safety of the reactor. It also assures that the maximum reactivity addition rate specification will not be exceeded.

l 4.2.3 Reactor Safety System Annlicability The specifications apply to the surveillance requirements for measurements, channel tests, and channel checks of the reactor safety systems and watchdog circuit.

Obiective The objective is to verify the performance and operability of the systems and components that are directly related to reactor safety.

Snecifications V a. A channel test of the SCRAM function of the high power, fuel temperature, manual, and preset timer safety channels shall be made on each day that the reactor is to be operated, or prior to each operation that extends more than one day.

b. A channel test of the detector power supply SCRAM function and the watchdog circuit shall be performed annually, not to exceed 15 months.

c. Channel checks for operability shall be performed daily on fuel element temperature, linear power, count rate, log power and reactor period when the reactor is to be operated, or prior to each operation that extends more than one day.

d. The percent power channel shall be compared with other independent channels for proper channel indication, when appropriate, each time the reactor is operated.

l

e. The pulse peak power channel shall be compared to the fun temperamre each time the reactor is pulsed, to assure proper peak power charmel l

operation.

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v 9003Idb Amendment No. 32 Rev.1 (4/24/97) l

31 Bases Ol V

System components have proven operational reliability.

a. Daily channel tests insure accurate SCRAM functions and insure the detection of possible channel drift or other possible deterioration of

, operating characteristics.

b. An annual channel test of the detector power supply SCRAM will assure that this system works, based on past experience as recorded in the operation j log book. An annual channel test of the watchdog circuit is sufficient to l assure operability.

c. The channel checks will make informadon available to the operator to assure safe operadon on a daily basis or prior to an extended run.

d. Comparison of the percent power channel with other independent power l channels will assure the detection of channel drift or other possible  !

deterioradon of its operadonal characterisdes. I

e. Comparison of the peak pulse power to the fuel temperature for each pulse will assure the detection of possible channel drift or deterioradon of its operadonal characteristics.

l 4.2.4 Reactor Interlocks  ;

i l

Annlicability l l l '

! These specifications apply to the surveillance requirements for the reactor control system interlocks.

l Obiective The objective is to insure performance and operability of the reactor control system interlocks.

Snecifications

a. A channel check of the source interlock shall be performed each day that the reactor is operated or prior to each operadon that extends more than one day.

l b. A channel test shall be performed semi-annually, not to exceed 71/2 months, on the log power interlock which prevents pulsing from power ,

levels higher than one kilowatt.  ;

c. A channel check shall be performed semi-annually, not to exceed 71/2 months, on the transient rod interlock which prevents applicadon of air to the transient rod unless the cylinder is fully inserted.

d. A channel check shall be performed semi-annually, not to exceed 71/2 months, on the rod drive interlock which prevents movement of any rod except the transient rod in pulse mode.

I O

9003Idb Amendment No. 32 Rev. l (4/24/97) l L

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

32

e. A channel check shall be performed semi-a.nnually, not to exceed 71/2

, months, on the rod drive interlock which prevents simultaneous manual i withdrawal of more than one rod.

O m  :

! The channel test and checks will verify operation of the reactor interlock I l system. Experience at the PSBR indicates that the prescribed frequency is l adequate to insure operability.

4.2.5 Overnower SCRAM l Aonlicability

! This specification applies to the high power and fuel temperature SCRAM l channels.

! Objective The objective is to verify that high power and fuel temperature SCRAM channels perform the SCRAM functions. .

Soecification l

The high power and fuel temperature SCRAhis shall be tested annually, not to exceed 15 months.

BMll l Experience with the PSBR for more than a decade, as recorded m  !

l the operation log books, indicates that this interval is adequate to assure operability.  ;

4.2.6 Transient Rod Test Apolicability l These specifications apply to surveillance of the transient rod mechanism.

Objective

! The objective is to assure that the transient rod drive mechanism is maintained i in an operable condition.

l l Soecifications

a. On each day that pulse mode operation of the reactor is planned, a functional performance check of the transierit rod system shall be l performed. The transient rod drive cylinder and the associated air supply l system shall be inspected, cleaned, and lubricated as necessary annually, not to exceed 15 months.

i b. The reactor shall be pulsed annually, not to exceed 15 months, to compare

fuel temperature measurements and peak po,ver levels with those of l q previous pulses of the same reactivity value or the reactor shall not be  !

Q pulsed undl such comparative pulse measurements are performed.

l 9003Idb Amendment No. 32 Rev.1 (4/24.97)

33 Halls

Functional checks along with periodic maintenance assure repeatable performance. The reactor is pulsed at suitable intervals and a comparison made l with previous similar pulses to determine if changes in transient rod drive mechanism, fuel, or core characteristics have taken place.

l 1

4.3 Coolant System 4.3.1 Fire Hose Insnection A_o nlicability l

This specification applies to the dedicated fire hoses used to supply water to the pool in an emergency.

Obiective The objective is to assure that these hoses are operable.

Snecification The two (2) dedicated fire hoses that provide supply water to the pool in an emergency shall be visually inspected for damage and wear annually, not to ,

I l exceed 15 months.

Basis 31is frequency is adequate to assure that significant degradation has not l

( occurred smce the previous mspection. l

%.,))  !

4.3.2 Pool Water Temnerature l Annlicability l

This specification applies to pool water temperature. l Obiective The objective is to limit pool water temperature.

Snecification The pool temperature alarm shall be calibrated annually, not to exceed 15 months.

Basis l

Experience has shown this instrument to be drift-free and that this interval is l adequate to assure operability.

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N) 9003ldb l

Amendment No. 32 Rev.1 (4/24/97)

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

34

'4.3.3 Pool Water Conductivity . ,

i Applicability This specification applies to surveillance of pool water conductivity.

Obiective The objective is to as::ure that pool water mineral content is maintained at an acceptable level.

l Snecification

~

l l

h l

Pool water conductivity shall be measured and recorded daily when the reactor 1

is to be operated, or at monthly intervals when the reactor is shut down for this t

l' time period.

i i Basis, t

Based on experience, observation at these intervals provides acceptable surveillance of limits that assure that fuel clad. corrosion and neutron activadon  ;

of dissolved materials will not occur. l 4.3.4 Pool Water Level Alarm Anolicabili)y .

'Ihis specification applies to the surveillance requirements for the poollevel alarm.

l 1

' ( Obiective ,

The objective is to verify the operability of the pool-water level alarm.  !

i l Soecification i

The pool-water level alarm shall be channel checked monthly, not to exceed 6 weeks, to assure its operability.

Basu l Experience, as exhibited by past periodic checks, has shown that monthly checks of the pool-water level alarm assures operability of the system during the l

month.

4.4 Confinement l' Apolicability This specification applies to reactor bay doors.

Obiective t

The objective is to assure that reactor bay doors are kept closed as per Specification 3.4.

I

• i

?

9003 Idb q I

Amendment No.32 Rev.1 (4/24/97) i

i 35 Snecification Doors to the reactor bay shall be locked or under supervision by an authorized

/ keyholder.

1 V) Basis j l A keyholder is authorized by the Director or his designee.

l 4.5 Facility Exhaust System and Emergency Exhaust System 1

Anolicability These specifications apply to the facility exhaust system and emergency exhaust system.

Obiective The objective is to assure the proper operation of the facility exhaust system and e.mergency exhaust system in controlling releases of radioactive material to the uncontrolled environment. -

l Snecifications

a. It shall be verified monthly, not to exceed 6 weeks, whenever operation is scheduled, that the emergency exhaust system is operable with correct pressure drops across the filters (as specified in procedures).

b. It shall be verified monthly, not to excc ed 6 weeks, whenever operation is l

!q scheduled, that the facility exhaust system securu when the emergency exhaust bj system activates during an evacuation alarm (See Technical Specificationn 3.6.2 i

and 5.5).

1 l

Basis l

l Experience, based on periodic checks performed over years of operation, has demonstrated that a test of the exhaust systems on a monthly basis, not to exceed 6 i weeks,is sufficient to assure the proper operation of the systems. This provides  !

reasonable assurance on the control of the release of radioactive material.

4.6 Radiation Monitoring System and Effluents 4.6.1 Radiation Monitoring System and Evacuation Alarm Acolicability This specification applies to surveillance requirements for the area radiation l monitor, the beam laboratory radiation monitor, the air radiation monitor, and l the evacuation alarm.

l Objective The objective is to assure that the radiation monitors and evacuation alarm are operable and to verify the appropriate alarm settings.

9003Idb Amendment No. 32 Rev.1 (434/97) l

36 -

Specifir2 tina l The area try on monitor, the beam laboratory radiation monitor, the continuou; a . adiation) monitor, and the evacuation alarm system shall be channel tesh tonthly not to exceed 6 weeks. They shall be verified to be operable by a . .annel check daily when the reactor is to be operated, and shall l be calibrated annually, not to exceed 15 months.

Basis Experience has shown this frequency of verification of the radiation monitor set points and operability and the evacuation alarm operability is adequate to correct for any variation in the system due to a change of operating ,

characteristics. Annual channel calibration insures that units are within the  :

specifications defined by procedures.

l 4.6.2 Argon-41 l

Anolicability l This specification applies to surveillance of the Argon-41 produced during ,

l reactor operation.

l Objective i

( To assure that the production of Argon-41 does not exceed the limits specified L by 10 CFR Part 20.

g Soecification 4

i The production of Argon-41 shall be measured and/or calculated for each new l experiment or experimental facility that is estimated to produce a dose greater than 1 mrem at the exclusion boundary.

B. asis l

One (1) mrem dose per experiment or experimental facility represents 1% of the  !

maximum 10 CFR Part 20 annual dose. It is considered prudent to analyze the i Argon-41 production for any experiment or experimental facility that exceeds l

1% of the annuallimit.

l 4.6.3 ALARA I l

l Acolicability l

This specification applies to the surveillance of all reactor operations that could result in occupational exposures to radiation or the release of radioactive er" ants to the environs.

i i s 9003Idb Amendment No. 32 Rev.1 (4/24/97)

37 Obiective l The objective is to provide surveillance of all operations that could lead to p occupational exposures to radiation or the release of radioactive effluents to the environs. ,

Specification As part of the review of all operations, consideration shall be given to .

I altemative operational modes that might reduce staff exposures, release of radioactive materials to the environment, or both.

Bm i

Experience has shown that experiments and operational requirements can,in i l many cases, be satisfied with a variety of combinadons of facility options, core i positions, power levels, time delays, and effluent or staff radiation exposures.

Similarly, overall reactor scheduling achieves significant reductions in staff exposures. Consequently, ALARA must be a part of both overall reactor  ;

scheduling and the detailed experiment planning.  ;

4.7 Exneriments Applicability This specification applies to surveillance requirements for experiments.

Obiective i The objective is to assure that the conditions and restrictions of Specification 3.7 are met. i Snecification I

Those conditions and restrictions listed in Specification 3.7 shall be considered by the PSBR authorized reviewer before signing the irradiation authorization for each experiment.

l BM

, l Authorized reviewers are appointed by the facility director.

5.0 DESIGN FEATURES 5.1 Reactor Fuel i Specifications The individual unirradiated TRIGA fuel elements shall have the following characteristics:

l l a. The total uranium content shall be either 8.5 wt% or 12.0 wt% nominal and i enriched to less than 20% uranium-235. l i

9003Idb Amendment No. 32 Rev.1 (4n.4/97)

I

38

b. The hydrogen-to-zirconium atom ratio (in the ZrHx) shall be a nominal 1.65 H atoms to 1.0 Zr atom.

c. The cladding shall be 304 stainless steel with a nominal 0.020 inch thickness.

5.2 Reactor Core Soecifications

a. The core shall be an arrangement of TRIGA uranium-zirconium hydride fuel-moderator elements positioned in the reactor grid plates.

b. The reflector, excluding experiments and experimental facilities, shall be water, or D 20, or graphite, or any combination of the three moderator materials.

l 5.3 Control Rods l

l l Snecifications 1

l

a. The shim, safety, and regulating control rods shall have SCRAM capability and contain borated graphite, B4C powder, or boron and its compounds in solid form as l a poison in stainless steel or aluminum cladding. These rods may incorporate l fueled followers which have the same characteristics as the fuel region in which they are used.

l

b. The transient control rod shall have SCRAM capability and contain bomted i

graphite, B 4C powder, or boron and its compounds in a solid form as a poison in an aluminum or stainless steel clad. When used as a transient rod, it shall have an adjustable upper limit to allow a variation of reactivity insertions. This rod may b

G' incorporate a voided or a solid aluminum follower.

l 5.4 Fuel Storage Specifications l

a. All fuel elements shall be stored in a geometrical array where the keff is less than 0.8 for all conditions of moderation.

b. Irradiated fuel elements shall be stored in an array which shall permit sufficient natural convection cooling by water such that the fuel element temperature shall not '

reach the safety limit as defined in Section 2.1 of the Technical Specifications.

5.5 Reactor Bay and Exhaust Systems i Soecifications

a. The reactor shall be housed in a room (reactor bay) designed to restrict leakage.

The minimum free volume (total bay volume minus occupied volume) in the reactor bay shall be 1900 m3 .

b. The reactor bay shall be equipped with two exhaust systems. Under normal operating conditions, the facility exhaust system exhausts unfiltered reactor bay air to the environment releasing it at a point at least 24 feet above ground level. Upon r

r 9003Idb Amendment No.32 Rev.1 (4/24/97) l l

39 initiation of a building evacuation alarm, the previously mentioned system is automatically secured and an emergency exhaust system automatically starts. The l emergency exhaust system is also designed to discharge reactor bay air at a point at least 24 feet above ground level.

5.6 Reactor Pool Water Systems l Specification The reactor core shall be cooled by natural convective water flow.

6.0 ADMINISTRATIVE CONTROLS 6.1 Organization 6.1.1 Structure The University Vice President for Research Dean of the Graduate School (level

1) has the responsibility for the reactor facility license. The management of the facility is the responsibility of the Director (Icvel 2), who reports to the Vice President for Research Dean of the Graduate School through the Head of the Nuclear Engineering Department and the Dean of the College of Engineering.

Administrative and fiscal responsibility is within the offices of the Department Head and the Dean.

The minimum qualifications for the position of Director of the PSBR are an advanced degree in science or engineering, and 2 years experience in reactor )

operation. Five years of experience directing reactor operations may be l substituted for an advanced degree.

k_) The Director can at any time temporarily delegate his authority to the Manager of Operations and Training (level 3) who can in-turn further delegate his authority to a qualified Senior Reactor Operator (level 4).

The Operating Staff (level 4) report to the Manager of Operations and Training (level 3) for day-to-day operational matters.

The University Health Physicist reports tMough the Director of Intercollege Programs to the Office of the Vice President for Research Dean of the Graduate School. The qualifications for the University Health Physicist position are the equivalent of a graduate degree in radiation protection,3 to 5 years experience with a broad byproduct material license, and certification by The Amedcan Board of Health Physics or eligibility for certification.

6.1.2 Resoonsibility Responsibility for the safe operation of the reactor facility shall be within the chain of command shown in the organization chart. Individuals at the various .

management levels, in addition to having responsibility for the policies and operation of the reactor facility, shall be responsible for safeguarding the public and facility personnel from undue radiation exposures and for adhering to all requirements of the operating license and technical specifications.

In allinstances, responsibilities of one level may be assumed by designated alternates or by higher levels, conditional upon appropdate qualifications.

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LJ 9003Idb Amendment No. 32 Rev.1 (4/24/97)

._ .. .._ _ _ __ _.._ __ ..__._ ___.._._. _ ___._ _ _ _ _ .__...m . _ ._. _ __

40

, ( 3

! Vice President for Research i I

Dean of the Graduate School l

l 0- ( J l

l r m l i Dean, College 1 l of Engineering l t > I l

l t

r 3 Nuclear Engineering l r 3 Department Head r 3 i University Health k > Penn State Reactor Physicist Safeguards Committee L . J t . J l I i 1

l l l l

l ( 3 I

! I Director l Penn State Breazeale Reactor l

l (I.evel 2)

N ]

l l r 3 l

Manager of Operations and Training (Level 3) j i

Operating Staff l Senior Reactor i Operators and l Reactor Operators (Level 4) i ORGANIZATION CHART i i

i i

9003Idb Amendment No. 32 Rev.1 (4/24S7)

l 41 6.1.3 Staffing j a. The minimum staffing when the reactor is not secured shall be:

1 i

! iN_-) 1) A licensed operator present in the control room, in accordance with  !

applicable regulations.

2) A second person present at the facility able to cany out prescribed wntten instructions.

i l 3) If a senior reactor operator is not present at the facility, one shall be i

available by telephone and able to be at the facility within 30 minutes. l

b. A list of reactor facility personnel by name and telephone number shall be readily available in the control room for use by the operator. The list shall include:

1) Management personnel.

2) Radiation safety personnel. i

. l

3) Other operations personnel.

c. Events requiring the direction of a Senior Reactor Operator shall include:

1) All fuel or control-rod relocations within the reactor core region.

2) Relocation of any in-core experiment with a reactivity worth greater

! (n)

V than one dollar.

3) Recovery from unplanned or unscheduled shutdown (in this instance, l

documented verbal concurrence from a Senior Reactor Operator is required). 1 6.1.4 Selection and Training of Personnel The selection, training, and requalification of operations personnel shall meet or exceed the requirements of all applicable regulations and the American National l Standard for Selection and Training of Personnel for Research Reactors, ANSI /ANS-15.4-1977, Sections 4-6.

6.2 Review and Audit )

l 6.2.1 Safecuards Committee rgmansition A Penn State Reactor Safeguards Committee (PSRSC) shall exist to provide an independent review and audit of the safety aspects of reactor facility operations.

The committee shall have a minimum of 5 members and shall collectively represent a broad spectrum of expertise in reactor technology and other science and engineering fields. The committee shall have at least one member with health physics expertise. The committee shall be appointed by and report to the i l '

Dean of the College of Engineering. The PSBR Director shall be an ex-officio member of the PSRSC.

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9003Idb Amendment Nc. 32 Rev.1 (4/24/97)

. . . _ . . . _ _ _ _ _ _ _ _ _ _ _ _ _ . . . _ . . _ _ _ _ _ _ . _ . . ~ _

42 6.2.2 Charter and Rules r 'Ihe operations of the PSRSC shall be in accordance with a written charter, including provisions for:

a. Meeting frequency - not less than once per calendar year not to exceed 15 months.

b. Quorums - at least one-half of the voting membership shall be present (the i

Director who is ex-officio shall not vote) and no more than one-half of the voting members present shall be members of the reactor staff.

c. Use of Subgroups - the committee chairman can appoint ad-Hoc committees as deemed necessary.

t

d. Minutes of the meetings - shall be recorded, disseminated, reviewed, and l approved in a timely manner.

l ,

! 6.2.3 Review Function  !

The following items shall be reviewed:

l a. 10 CFR Part 50.59 reviews of:

l 1) Proposed changes in equipment, systems, tests, or experiments.

2) All new procedures and major revisions thereto having a significant f effect upon safety.

l

3) All new experiments or classes of experiments that could have a significant effect upon reactivity or upon the release of radioactivity.

l b. Proposed changes in technical specifications, license, or charter.

i- c. Violations of technical specifications, license, or charter. Violations of -

l internal procedures or instructions having safety significance.  :

L l d. Operating abnormalities having safety significance.

l e. Special reports listed in 6.6.2.  ;

l f. Audit reports.  !

l 6.2.4 Audit  ;

The audit function shall be performed annually, not to exceed 15 months, - l l

preferably by a non-member of the reactor staff. The audit function shall be j l performed by a person not directly involved with the function being audited. '

The audit function shall include selective (but comprehensive) examinations of  !

(

operating records, logs, and other documents. Discussions with operating l l personnel and observation of operations should also be used as appropriate.

Deficiencies uncovered that affect reactor safety shall promptly be reported to  :

, the Nuclear Engineering Department Head and the Dean of the College of l

! Engineering. The following items shall be audited: )

9003Idb Amendment No. 32 Rev.1 (4/24s7)

43 l

a. Facility operadons for conformance to Technical Specifications, license, and i procedures (at least once per calendar year with interval not to exceed 15 l months).

c.

l ! k

()

b. The requalification program for the operanng staff (at least once every other calendar year with the interval not to exceed 30 months). l l

c. The results of action taken to correct deficiencies that may occur in the l reactor facility equipment, systems, structures, or methods of operations that affect reactor safety (at least once per calendar year with the interval not to exceed 15 months).

d. The reactor facility emergency plan and implementing procedures (at least I once every other calendar year with the interval not to exceed 30 months). l 6.3 Ooerating Procedurn l Written procedures shall be reviewed and approved prior to the initiation of activities j covered by them in accordance with Section 6.2.3 The procedures in this section preceded by an asterisk shall also be approved and initialed by a representative of the University Health Physics Office. Written procedures shall be adequate to assure the l safe operation of the reactor, but shall not preclude the use ofindependent judgment and action should the situation require such. Operating procedures shall be in effect I and shall be followed for at least the following items:

a. Startup, operation, and shutdown of the reactor.

b. Core loading, unloading, and fuel movement within the reactor.

(s \

c. Routine maintenance of major components of systems that could have an effect (d on reactor safety.

d. Surveillance tests and calibrations required by the technical specifications (including daily checkout procedure).

e. Radiation, evacuation, and alarm checks.

• f. Release ofIrradiated Samples.

• g. Evacuation.

• h. Fire or Explosion.

• i. Gaseous Release.

• j. Medical Emergencies.

'. Civil Disorder.

• l. Bomb Threat.

l

• m. Threat of Theft of Special Nuclear Material.

• n. Theft of Special Nuclear Material.

• o. Industrial Sabotage.

L t 9003Idb l Amendment No. 32 Rev.1 (4/24/97) l

l l

44 i

• p. Experiment Evaluadon and Authorization. l n *q. Reactor Operation Using a Beam Port.

• r. D 20 Handling.

• s. Health Physics Orientation Requirements.

• t. Hot Cell Entry Procedure.

• u. Implementation of emergency and security plans.

v. Radiation instrument calibration I

w. Loss of pool water.

6.4 Review and Anoroval cf Exneriments

a. All new experiments shall be reviewed for Technical Specifications compliance, 10 CFR Part 50.59 analysis, and approved in writing by level 2 management or designated alternate prior to initiation. -

b. Substantive changes to experiments previously reviewed by the PSRSC shall be <

made only after review and approval in writing by level 2 management or t designated alternate.  !

6.5 Reauired Action i 6.5.1 Action to be Taken in the Event the Safety Limit is Exceeded

.O -

In the event the safety limit (1150*C) is exceeded:

a. The reactor shall be shut down and reactor operation shall not be resumed .

until authorized by the U.S. Nuclear Regulatory Commission.

b. The safety limit violation shall be promptly reported to level 2 or designated  ;

altemates. i

c. An immediate report of the occmrence shall be made to the Chairman, ,

l PSRSC and reports shall be made to the USNRC in accordance with )

Specification 6.6.

l j 1

d. A report shall be prepared which shall include an analysis of the causes and extent of possible resultant damage, efficacy of corrective action, and -

recommendations for measures to prevent or reduce the probability of recurrence. This report shall be submitted to the PSRSC for review. I 6.5.2 Action to be Taken in the Event of a Reportable Occurrence l In the event of a reponable occurrence,(1.1.36) the following action shall be

taken:

i

a. The reactor shall be returned to normal or shutdown. Ifit is necessary to j shutdown the reactor to correct the occurrence, operations shall not be resumed unless authorized by level 2 or designated alternates.

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! 9003Idb i Amendment No.32 Rev.1 (434m) t

45 c

b. The Director or a designated alternate shall be notified and correcdve action j taken with respect to the operations mvolved.

c. The Director or a designated alternate shall notify the Nuclear Engmeenng j Department Head who,in turn, will notify the office of the Dean of the College of Engineering and the office of the Vice President for Research Dean of the Graduate School.

d. The Director or a designated alternate shall notify the Chairman of the PSRSC.

e. A report shall be made to the PSRSC which shallinclude an analysis of the cause of the occurrence, efficacy of corrective action, and recommendations for measures to prevent or reduce the probability of recurrence. This report .

- shall be reviewed by the PSRSC at their next meeting.

f. A report shall be made to the USNRC, Document Control Desk, Washington, DC 20555, with a copy to the Regional Administrator.

6.6 Reoorts 6.6.1 Ooerating Reoorts An annual report shall be submitted within 6 months of the end of The Pennsylvania State University fiscal year to the USNRC, Document Control l- Desk, Washington, DC 20555, with a copy to the Regional Administrator, including at least the following items:

a. A narrative summary of reactor operating experience including the energy P produced by the reactor, and the number of pulses 2 $2.00 but less than or equal to $2.50 and the number greater than $2.50.

l

b. The unscheduled shutdowns and reasons for them including, where applicable, corrective action taken to preclude recurrence.

i c. Tabulation of major preventive and corrective maintenance operations

! having safety significance.

d. Tabulation of major changes in the reactor facility and procedures, and tabulation of new tests and experiments, that are significantly different from those performed previously and are not described in the Safety Analysis Report, including a summary of the analyses leading to the conclusions that no unreviewed safety questions were involved and that 10 CFR Part 50.59 was applicable.

e. A summary of the nature and amount of radioactive effluents released or discharged to environs beyond the effective control of the owner-operator as determined at or before the point of such release or discharge. The  !

summary shallinclude to the extent practicable an estimate ofindividual l radionuclides present in the effluent. If the estimated average release after  !

l dilution or diffusion is less than 25 percent of the concentration allowed or j recommended, only a statement to this effect need be presented.

6 l f. A summarized result of environmental surveys performed outside the facility.

L I  :

9003Idb Amendment No. 32 Rev.1 (4/24/97)

l 46 6.6.2 Soecial Renorts

! ,, Special reports are used to report unplanned events as well as planned major facility and administrative changes. These special reports shall contain and (V) shall be communicated as follows:

a. There shall be a report no later than the following working day by telephone and confirmed in writing by telegraph or similar conveyance to the USNRC, l

Operations Center, Washington, DC 20555, to be followed by a written l report to the USNRC, Document Control Desk, Washington, DC 20555, with a copy to the USNRC Regional Administrator, that describes the circumstances of the event within 14 days of any of the following:

1) Violation of safety limits (See 6.5.1)

2) Release of radioactivity from the site above allowed limits (See 6.5.2)

3) A reportable occurrence (Section 1.1.35)

b. A written report shall be made within 30 days to the USNRC, and to the offices given in 6.6.1 for the following: .

1) Permanent changes in the facility organization involving level 1-2 personnel.

2) Significant changes in the transient or accident analysis as described in t the Safety Analysis Report.

l 6.7 Records To fulfill the requirements of applicable regulations, records and logs shall be prepared, and retained for the following items:

6.7.1 Records to be Retained for at Least Five Years

a. Log of reactor operation and summary of energy produced or hours the reactor was critical.

b. Checks and calibrations procedure file.

c. Preventive and corrective electronic maintenance log.

d. Major changes in the reactor facility and procedures.

e. Experiment authorization file including conclusions that no unreviewed safety questions were involved for new tests or experiments.

l

f. Event evaluation forms (iNi6 ding unscheduled shutdowns) and reportable occurrence reports.

g. Preventive and corrective maintenance records of associated reactor

, equipment.

l

h. Facility radiation and contamination surveys.

h V

i. Fuel inventories and transfers.

9003Idb Amendment No. 32 Rev.1 (4/24/97)

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I l 47 i j. Surveillance activities as required by the Technical Specifications.

k. Records of PSRSC reviews and audits.

6.7.2 Records to be Retained for at least One Trainine Cvele

a. Requalification records for licensed reactor operators and senior reactor operators.

6.7.3 Records to be Retained for the Life of the Reactor Facility

a. Radiation exposure for all facility personnel and visitors.

b. Environmental surveys performed outside the facility.

c. Radioactive effluents released to the environs.

j d. Drawings of the reactor facility including changes.

l-I l

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l 9003Idb Amendment No. 32 Rev.1 (4/24/97)

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