ML20198S132
| ML20198S132 | |
| Person / Time | |
|---|---|
| Site: | Pennsylvania State University |
| Issue date: | 01/16/1998 |
| From: | PENNSYLVANIA STATE UNIV., UNIVERSITY PARK, PA |
| To: | |
| Shared Package | |
| ML20198S119 | List: |
| References | |
| NUDOCS 9801260081 | |
| Download: ML20198S132 (50) | |
Text
TECHNICAL SPECIFICATIONS FOR THE PENN STATE BREAZEALE REACTOR (PSBR)
FACILITY LICENSE NO. R 2 TABLE OF CONTENTS 1.0 I NTROD U CTION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . I 1.1 De fi n i tion s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . I 1.1.1 ALARA...................................................................... I 1.1.2 Automatic Con trol . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 l 1.1.3 Channel...................................................................... I 1.1.4 Channel Calibration . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . I 1.1.5 Chan n e l Chec k . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . I 1.1.6 Cha n nel Tes t . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . I 1.1.7 Cold Cri tical . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . I 1.1.8 Con fme me n t . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 1.1.9 E x cess Reacti vi ty . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . - 2 1.1.10 Experime nt . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 1.1. I 1 Experimental Facility . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 1.1.I2 Instrumented Element . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 1.1.13 Limiting Conditions for Operation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 1.1.14 Limiting S afety System Setting . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 1.1.15 Man ual Con trol . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 1,1.16 Maximum Elemental Power Den sity. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 1.1.17 M axiny @ower Level . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 1.1.18 M easuna s lue . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 1.1.19 M ovable Ex perimen t . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 1.1.20 Normalized Po wer . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 1.1.21 Operable..................................................................... 3 1.1.22 Opera ti n g . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 1.1.23 Pu l se M od e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 1.1.24 Reactivity Li mits . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 1,1.25 R eactivity Worth of an Experiment . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 1.1.26 Reactor Con trol System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 1.1.27 Reactor I n terloc k . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 1.1.28 Reac tor Operatin g . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 1.1.29 Reactor S ecured . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 1.1.30 Reactor S hu tdown . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 1.1.31 Reactor S afety System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 1.1.32 Reference Core Condition . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 1.1.33 Researc h Reac tor . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 1.1.34 Reportable Occurrence . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 1.1.35 Rod -Tran s i en t . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 1.1.36 S afe ty Li mi t . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 1.1.37 S CR A M 'n me . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 1.1.38 Secured Experimen t . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 1.1.39 Secured Ex periment with Movable Parts . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 1.1.40 S hall, S hould and May . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 1.1.41 S him, Regulating, S afety Rods . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 1.1.42 S h u tdow n M argi n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 1.1.43 Sq uart Wave M ode . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 NOTARIAL SEAL PAMELA J. STAUFFER. Notary POGc State Conege Bofo. Centre County, PA MY Commeslon Egtres July 2,2001 (h%Y huY /L ,
kNMf( Propowd Amendment No._32 (in6S8)
-- . V $/ f f#
9001260D81 9901197 POR ADOCK 05000005 p PDR
r l
.. 1 u
1.1.44 Steady S tate Power Level . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 1.1.45 TRIGA Fuel Element ..................................................... 7 1.1.46 Watchdog Ci rcui t . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 2.0 SAFETY LIMIT AND LIMITING SAFETY SYSTEM SETTING ................... ?
2.1 S afety Limit-Fuel Element Temperature . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 2.2 Limiting S afety System Setting (LS SS) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 3.0 LIM ITING CONDITIONS FOR OPER ATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 3.1 Reactor Core Parame ters . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 3.1.I Non-Pulse Mode Operation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 3.1.2 Reactivity Limitation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10 3.1.3 S h u tdown M argi n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10 3.1.4 Pul se Mode Operation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . , . . . . . . . . I1 3.1.5 Core Configurauon Limitation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . I1 3.1.6 TR IG A Fuel Elemen ts . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 3.2 Reactor Control and Reactor S afety System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 3.2.1 Reactor Con trol Rods . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 3.2.2 Manual Control and A utomatic Control . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14 3.2.3 Reactor Contml System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14 3.2.4 Reactor Safety System and Reactor Interlocks .. ...................... 15 3.2.5 Core Loadin g and U nloading Operation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17 3.2.6 S CR A M Ti me . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 18 3.3 Coo l a n t S y s t e m . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . I8 3.3.1 Coolant Level Limits . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . I8 3.3.2 Detection of Leak or Loss of Coolant . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19 3.3.3 Fission Prod uct Activity . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20 3.3.4 Pool Water S upply for Leak Protection . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . 20 3.3.5 Coolan t Cond uctivity Li mits . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20 3.3.6 Coolant Temperature Limits . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21 3.4 Co n fi n eme n t . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21 3.5 Engineered Safety Features - Facility Exhaust System and Emergency Ex h au st System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 22 3.6 Radi atk.a M onitorin g System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 23 3.6.1 Radiation Monitoring Information . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 23 3.6.2 Evacuation Alami . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 23 3.6.3 Argon-41 Discharge Limit . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 24 3.6.4 ALARA...................................................................... 24 3.7 Limitations of Ex perime nts . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 25 4.0 S U RVEILLANCE REQUIREM ENTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 27 4.1 Reac tor Parame te rs . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 27 4.1.1 Reactor Power Calibration . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 27 4.1.2 Reactor Excess Reactivity . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 27 4.1.3 TRIG A Fuel Eleme n t s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 28 Pmposed Amendment No. 32 (1/16/98)
ill 4.2 Reactor Control and S afety S ystem . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 29-4.2.1 Reac tivit y Wonh . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . - 29 4.2.2 Reactivity Insenion Rate . . . . . . . . . . . . . . . . . . . . . . . . . . . . , . . . . . . . . . . . . . . . . . . . . . 29 4.2.3 Reactor S afety S yste m . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 30 4.2.4 Reactor interlock s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 31 4.2.5 - Overpower S CR A M . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
. 31 4.2.6 Tran sie nt Rod Te st . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 32 4 . 3 Coola n t S yste m . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 32 4.3.I Fire Hose I n spection . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 32 4.3.2 Pool Water Temperature . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 33 4.3.3 Pool Water Cond uctivity . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 33 4.3.4 Pool Water 12 vel Alarm . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 34
- 4. 4 Confineme nt . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 34 4.5 Facility Exhaust System and Emergency Exhaust System ....................... 34 4.6 Radiation Monitoring System and Efn uents . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 35 4.6.1 Radiation Monitoring System and Evacuation Alarm .................. 35 i 4.6.2 Argon-41.................................................................... 35 4.6.3 ALARA..................................................................... 36 4.7 Ex pe rime n t s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 36 5.0 D ES IG N FE ATU R ES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 37 5.1 R eacto r Fu el . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 37 5.2 R e ac tor Core . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 37
- 5. 3 Con trol Rod s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . -37 5.4 Fuel S to ra ge . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 38 5.5 Reactor Bay and Exhaust Systems . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 38
- 5. 6 Reactor Pool Water Systems . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 38 6.0 ADMINIS'IRATIVE CONTROLS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 38 6.1 Org ani za ti on . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 38 6.1.1 S tru ct ure . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 38 -l 6.1.2 Re spon sibility . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 40 6.1.3 Staffing.................................................................... 40 6.1.4 Selection and Training of Personnel . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 41 6.2 Review and Audit . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 41 6.2.1 S afeguards Committee Composition . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 41 6.2.2 Chart er and R ule s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 41 6.2.3 R evi ew Fu nc t ion . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 41 6.2.4 Audit......................................................................... 42 l 6.3 Opera tin g Proced ures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 42 6.4 Review and A pproval of Experiments . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 43
- 6. 5 Req uired A ction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 43 6.5.1 Action To Be Taken in the Event the Safety Limit is Exceeded ....... 43 6.5.2 Action To Be Taken in the Event of a Reportable Occurrence ........ 44 Proposed Amendment No. 32 (1/16S8)
__ _ _a
. _ . . . - - . . -.~ - -. . .. . ..
iv
- 6.6 Reports................................................................................. .
44 6.6.1 Operating Repon s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . '44 6.6.2 S peci al R epons . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 45 6.7 Records................................................................................. 45 6.7.1 - Records To Be Retained for at least Five Years ....................... . 45 6.7.2 Records To Be Retained for at least One Training Cycle ............. 46 6.7.3 Records To Be Retained for the Lifc of the Reactor Facility ....... .. . 4d-Proposed Amendment No. 32 (1/16/98)
1 TECilNICAL SPECIFICATIONS FOR THE PENN STATE BREAZEALE REACTOR (PSBR)
FACILITY LICENSE NO. R-2
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 Sxcifications and they do not constitute limitations or requirements to which the ..icensee must adhere.
1.1 Definitions 1.1.1 ALARA The ALARA (As Low As Reasonably Achievable) program is a prognm for maintaining occupational exposures to radiation and release of radioactive effluents to the ,:nvirons as low as reasonably achievable.
1.1.2 Automatic Control l 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 A channel is the combination of sensor,line, amplifier, and output devices 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 vanable.
1.1.6 Channel Tes A channel test is the introduction of a signal into the channel to verify that it is operable.
1.1.7 Cold Critical Cold critical is the condition of the reactor when it is critical with the fuel and bulk water !cmperatures both below 100*F (37.8'C).
Pmposed Amendment No.32 (1/16S8)
-._. - . . ~ . . . ~ - - - - - . - - . ,. - -. - -- . ..
i
- 2
.'i 1.1.8 Confinenent Confinement means an enclosure on the overall facility which controls the ;
movement of air into it and out through a controlled path. l
. 1.1.9 ' R== Raetivitv-Excess reactivity is that amount of reactivity that would exic if all control a
- etc.) were moved to the maximum reactive ,
rods condition(safety, regulating,t from the pom where the reactor is exactly critical (keff=1) in the reference core condition.
1.1.10 Experiment l-Experiment shall mean (a) any apparatus, device, or material which is not a normal l urt of the core or ex penmental facilities, but which is inserted in these faci ities or is in line with a beam of radiation originating from the reactor core; or (b) any operation designed to measure reactor parameters _ ,
or characteristics.
1.1.11 Exnerimental Facilltv l 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.
1.1.12 Instrumented Element l An instrumented element is a TRIGA fuel element in which sheathed chromel alumel or equivalent thermocouples are embedded in the fuel. ,
1.1.13 1 imiting conditions for Oneration l ,
Limiting conditions for operation of the reactor are those constraims 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 otherwise specified.
1.1.14 1 imiting Rnfety System Setting l A limiting safety system setting (LSSS)is a setting for an automatic -
- protective device related to a variable having a significant safety function.
1,1,15 Manual Control . l Manual control mode is operation of the reactor with the power level controlled ty the operator adjusting the control rod positions.
4 h
Pmposed Anwndment No.32 (1/16/98) L
. . . . -, -. -- , . a. . . .. -. . --a. .
3.
1.1.16 Marlmum Flemental Power Density l
=
LThe maximum elemental power density (MEPD) is the power density of .
the element in the core producing more zwer than any other element in - ,
that loading. The power density of an c.cment 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. This detinFon is only applicable for _
non-pulse operation.
1.1.17l Mavimum Power 12 vel' l l Maximum Power Level is the maximum measured value of reactor power for non pulse operation.
1.1.18 Measured Value -l ,
The measured value is the value of a parameter as it appears on the output of a channel.
1.1.19 Movable Funeriment l A movable experiment is one where it is intended that the entire experiment may be moved in or near the core or into and out of the reactor while the reactor is operating.
1.1.20 Normalised Power l The normalized power, NP, is the ratio of the power of a fuel element to
. the average power per fuel element.
1.1.21 Onernhle l Operable means a component or system is capable of performing its intended ftnction.
1.1.22 Oneratine l Operating means a component or system is performing its intended function.
-1.1.23 Pulse Mnde -l-Pulse mode operation shall mean operation of the reactor allowing the ,
operator to insert preselected reactivity by the ejection of the transient rod.
1,1.24 Reg-tivity 1.imits l.
> The reactivity limits are those limits imposed on reactor core reactivity.
. - Quantities are referenced to a reference core condition.
1.1.25 Reactivity Worth of an Exneriment ' l The reactivity worth of an experiment is the maximum absolute value of the t
' reactivity change that would occur as a result of intended or anticipated changes
, or credible malfunctions that alter experiment position or configuration.
Proposed Amendment No.32 (1/1688)
- . . . . . - . . , . - . . . -,. -. a. --- -:
4 1.1.26 Eractor Control System l
- The reactor control system is composed of control and operational interlocks, reactivity adjustment controls, flow and temperature controls, and dis lay systems which permit the operator to operate the reactor reliabl in its allowed modes.
1.1.27 Reactor Interlock l A reactor interlock is a device which prevents some action, associated with reactor operation, until certain reactor operation conditions are satisfied.
1.1.28 Reactor onerating l The reactor is operating whenever it is not secured or shutdown.
1.1.29 Reactor Secured l The reactor is secured when:
- a. It contains insuf3cient 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 insened or other safety devices are in shutdown positions, as requimd by technical specifications, and
- 2) The console key switch is in the off position and the key is ren.oved from the lock, and
- 3) No work is in progress involving core fuel, core stnicture, installed control rods, or control rod drives unless they am physically decoupled from the control rods, and
- 4) No experiments in or near the reactor are being moved or serviced that have, on movement, a reactivity worth exceeding the maximum value allowed for a single experiment or one dollar whichever is smaller.
1.1.30 Reactor Shutdown l The reactor is shutdown if it is suberitical by at least one dollar in the reference core condition and the reactivity worth of all experiments is included.
1.1.31 Reactor Safety System l Reactor safety systems are those systems, including their associated input channels, which am designed to initiate automatic reacter protection or to provide information for initiation of manual protective action.
Proposed Amendment No.32 (1/1(d98)
5
' 1.1.32 - Reference ('nre Condition l..
-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)). :
1.1.33! Remarch Reactor l A reseamh reactor is defined as a device designed to support a self -
sustaining neutron chain reaction for researci., development, educational,
- training, or experimental purposes, and which may have provisions for the - -
pmduction of radioisotopes.
l.1.34 Reportable Occurrence l A reportable occurrence is any of the following which occurs during '
reactor operation: ;
- a. Operation with the safety system setting less conservative than ,
t
- specified in Section 2.2, Limiting Safety System Setting.
- b. Operation in violation of a limiting condition for operation.
- c. Failuir 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.
- c. An observed inadequacy in the implementati. n of either administrative or procedural controls which coulcl result in operation of the reactor outside the limiting conditions for operation.
- f. Release of fission products from a fuel element.
, g. Abnormal and significant degradation in reactor fuel, cladding, coolant ,
boundary or containment boundary that could result in exceec,ing 10 CFR Part 20 exposure criteria.
I
~
1.1.35 Rod Transient l 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 mode of operation.
1.1.36 Safety 1 imit l j
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 pri* Nal physical barrier is the fuel element cladding, p
jA. -
Proposed AmendmentNo.32 (1/16S8) c _ _ _, _ __ _ .- _ _ __ . . _ _ _ .
6 -
1.1.37 SCRAM Time l SCRAM time is the elapsed time between reaching a limiting safety 0 system set point and a specified control rod movement.-
~
1,l.38 Secumd Experiment l
-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 expenment might be subjected to by hydraulic, pneumatic, buoyant, or other forces which are normal to the operating envimnment of the experiment, or by forces which can arise as a result of credible malfunctions.
1,1.39 Secured hneriment with Movable Parts l A secured experiment with movsble pans is one that J ntains parts that are intended to % moved while the reactor is operating.
1.1.40 Shall Should and Mav l t-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.
1.1,41 Shim Rennlatina. and Rafety Rnde l A shim, regulating, or safety md is a control rod having an electric motor
' drive and SCRAM capabilities. It has a fueled follower section.
1.1.42 Shutdown Marrin l_
Shutdown margin shall mean the minimum shutdown reactivity necessary to provide confidence that the reactor can be made suberitical by means of the contml and safety systems 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.
1.1.43 Souare Wave ymgi l ScJuare 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 maxirnum power within the license limit.
1.1.44 - Steadv State Power Level l 7 Steady state power level is the nominal measured value of reactor power to which reactor power is being controlled whether by manual or .
- automatic actions. Minor variations about this level may occur due to noise, normal signal: variation, and reactivity adjustments. During manual, y automatic, or square wave modes of operation, some initial, momentary
- overshoot may occur.
Pmposed ." -- t No.32 (Ol6/98) -
7 1.1.45 TRIGA Fuel Element 'l A TRIGA fuel element is a single TRIGA fuel rod of standard type, either 8.5 wt% U-Zrit in stainless steel cladding or 12 wt% U.ZrH in stainless - ':
l steel cladding enriched to less than 20% uranium-235.
1.1.46 Waecuay circuit l A watchdog circuit is a circuit consisting of a timer and a relay. De timer-energizes the relay as long as it is reset prior to the expiration of the timing interval, ifit is not reset within the timmg interval, the relay _will de-energize thereby causing a S? RAM. .
2.0 SAFETY LIMIT AND LIM! TING SAFETY SYSTEM SETTING 2.1 hfety I .imir Fuel Element Temnerntme Annlienhility The safety limit specification applies to the maximum temperature in the reactor fuel.
Obiective
~
%c objective is to define the maximum fuel element temperature that can be permided with confidence that no damage to the fuel element and/or cladding will result.
Specification l The temperatuir in a water-cooled TRIGA fuel element shall not exceed 1150*C under any operating condition.
Basis d
The important parameter for a TRIGA reactor is the fuel element temperature. This parameter is well suited as a single sm:cification especially since it can be racasured
. at a point within the fuel element and the relationship between the measured and actual temperature is well characterited analytically. A loss in the integrity of the fuel element cladding could arise from a build-up of excessive pressure between the
> fuel modera'or and the cladding if the maximum fuel temperature exceeds 1150C.
He pressure is caused by the presence of air, fission product gases, and hydrogen from the dissociation of the hydrogen and zirconium un the fuel-moderator. The -
magnitude of this pressure is determined by the fuel moderator temperature, the ratio of hydrogen to zircontum in the alloy, and the rate change in the pressure.
. The safety limit for the standard TRIGA futA is based on data, including the large mass of ex perimental evidence obtained during high perfonnance reactor tests on this fuel. These data indicate that the stress in the cladding due to the increase in the hydrogen pressure from the dissociation of zirconium hydride will remain below
. the ultimate stress provided that the temperature of the fuel does not exceed ll50'C
' and the fuel cladding is below 500*C, 5cc Safety Analysis Repon, Ref.13 and 30 in Section IX and Simnad, M.T., F.C. Foushee, and G.B. West, " Fuel Elements for .
Pulsed Reactors," Nucl. Technology, Vol. 28, p. 31-56 (January 1976).
o 1%pmed AmendmentNo.32 (1/16/98)
8 2.2 Limiting Saferv System Setting (LSSS)
Apolicability The LSSS specification applies to the SCRAM setting which prevents the safety limit fro n being reached.
Obiective The objective is to prevent the safety limit (1150'C) from being reached.
Soccifkation l The limiting safety system setting shall be a maximum of 650'C as measured with an instrumented fuel element if it is located in a core position representative of the maximum elemental power density (MEPD) in that loading, ifit 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.
DAsi The limiting safety system setting is a temperature which, if reached, shall cause a reactor SCRAM to be initiated preventing the safety limit from bein;, exceeded.
Experiments and analyses described in the Safety Analysis Repon,Section IX -
Safety Evaluation, show that the measured fuel temperature at steady state power Ms a simple linear relationship to the normalized power of a fuel element in the cote. Maximum fuel temperature occurs when an instrumented element is in a core position of MEPD. The actual location of the instrtamented 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 mar;;in provides adequate compensation for variations in the temperature profile of c epleted and differently loaded fuel elements (i.e. 8.5 wt% vs.12 w %
fuel elements). See Safety Analysis Report,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 if it 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. Hcwever, 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).
Pmposed Amendment No.32 (1/t688)
-. -. - - . - - . - -- - .. - . ~ - - - - - - - - - -- - -
9 3.0 LIMITING CONDITIONS FOR OPERATION t The limiting conditions for operation as set forth in this section are applicabic only when .
the reactor is operating. They need not be met when the reactor is shutdown unless s specified otherwise.
3.1 Reactor Core Parameters
-3.1.1 Non-Pulse Made Oneration i Applicability These specifications apply to the power generated during manual contiol mode, automatic control mode, and square wave mode operations.
Obiective The objective is to limit the source term and energy production to that used in the Safety Analysis Report.
Snecifications l
- a. The reactor may be operated at steady state power levels of 1 Mw (thermal) or 1 less.
- b. The maximum power level shall be no greater than 1.1 Mw (thermal). l
- c. The sicJy state fuel temperature shall be a maximum of 650*C as measured with an instrumented fuel element if it is located in a core msition representative of MEPD in that loading, if it is not practical to ocate the instrumented fuel in such a position, the steady state fuel temperature shall be calculated by a ratb 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. In this case, the measured steady state fuel temperature shall be limited such that the calculated steady state fuel tem xrature at the core msition representative of the MEPD in that loar ing shall not excect 650'C.
Bases l-
- a. Thermal and hydraulic calculations and operational experience indicate that a compact TRIG A reactor core can be safely opera J up to power levels of at least 1.15 Mw (thermal) with natural convective cooling,
- b. Operation at 1.1 Mw (thermal)is within the bounds established by the SAR for steady state operations. See Chapter IX, Section C of the SAR.
- c. Limiting the maximum steady state measured fuel temperature of any position to 650*C places an upper bound on the fission product release fraction to that used in the analysis of a Maximum Hypothetical Accident (MHA). See Safety Analysis Report, section IX.
Proposed Amendment No.32 (t/16B8)
-o
. . . . . . , - . . - ~,- . , _ . . - - . _ . . - - . . . , .
10 3.1.2 Reactivity Limitatl0D Aoplicability This specification applies to the reactivity condition of the reactor and the reactivity worth of control rods, experiments, and experimental facilities. It applies to all modes of operation.
Objective The objective is to ensure that the scactor is operated within the limits analyzed in the Safety Analysis Repon and to ensure that the safety limit will not be exceeded.
Soccification 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).
Ilm 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 accident condition as described in the Safety Analysis Repon,Section IX .
3.1.3 Shutdown Margin Apolicability This specification applies to the reactivity condition of the reactor and the reactivity worth of control rods, experiments, and experimental facilities. It applies to all modes of operation.
Objective The objective is to ensure that the reactor can be shut down at all times and to ensure that the safety limit will not be exceeded.
Specification l The mactor shall not be operated unless the shutdown margin provided by control rods is greater than 0.175% Ak/k (~$0.25) with:
- a. All movabic experiments, experiments with inovable parts and experiment 21 facilities in their most reactive state, art
- b. The highest reactivity wonh control rod fully withdrawn.
Basis A shutdown margin of 0.175% Ak/k (~$0.25) ensures that the reactor can be made suberitical from any opera:Ing condition even if the highest wonh control rod should remain in the fully withdrawn position. The shutdown margin requirement may be more restrictive than Specification 3.1.2.
I%;esed Amendment No.32 (1/16/985
11 1
3.1.4 Pulse Mode Operation -
Applicability q
These specifications apply to the energy generated in the reactoi e a result : j of a pulse insertion of reactivity.- 4 Obbctive ;
The objective is to ensure that the safety limit will not be exceeded during' l' pulse mode operation.
sn.cinceinns .
- a. The stepped reactivity insertion for pulse operation shall not exceed 2.45% Ak/k (~$3.50) and the maximum worti; 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 tractivity of 2.45% Ak/k'
(~$3.50), prevents the maximum fuel temperature from reaching the safety lim .t (1150*C) for any core configuratica that meets the rcquirements 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 (ll50*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 L Core Configuration Limitation Applicability These specifications apply to all core configurations except as noted. l 4
Obiective -
The objective is to ensure that the safety limit (1150*C) will not be exceeded due to power peaking effects in the vanous core configurations.
)
Progmed AmendmentNo.32 (1/16#8)
'***'h be.- - p s_. a.a .__
~. -- - . . . . _ - - , - .-. - - -- - -
i 12; snecificniinns i
- a. The critical core shall be an assembly of either 8.5 wt% U 7sH stalrkss steel clad or a mixture of 8.5 wt% and 12 wt% U ZrH stainless steel clad l-TRIGA fuel nuxlerator elements placed in water with a 1.7 inch center:
- line grid spacing.
- b. The maximum calculated MEPD shall be less that 24.7 kw per fuel' element for non pulse operation.
- c. 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 -
(i. c. the pulse can be limited by the total worth of the transient rod, by .l the core excess, or administratively), the maximum NP may be H increased. 'Ihe maximum NP may be increased above 2.2 as long as the 1 calculated maximum fuel temperature does not exceed the safety limit with that maximu'a allowed pulse wonh and NP. In addition, the Reactivity Accident in the Safety Analysis Repon shall be evaluated to ensure that the safety limit is not exceeded with the new conditions (See Safety Analysis Report,Section IX.). ,
Bases
- 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 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 for non-pulse o wration places an upper bound on the elemental heat production anc 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. 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 may be increased without exceeding the safety limit (ll50*C). The amount ofincrease in the maximum NP allowed shall be calculated by an accepted method documented by an admiriistratively approved procedure. ?
I Proposed AnwndnmNo.32 (1/16#8) ,
ww-,-- - , -
13 3.1.6 IRIGA Fuel Flements
'i Ann 11cability
" These specificatlans apply to the mechanical condition of the fuel. -l Oblective The objective is to ensure that the reactor is not operated with damaged fuel l that might allow release vf fission products.-
Specificatiom l s
The reactor shall not 'oe operated with damaged fuel except to detect and identify the fuel element for removal. A TRIG A feel element shall be considered damaged and shall be removed from the core if: l
- 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 4
- 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 touchmg fuel elements shows that there will be no hot spots which cause damage to the fuel,
- b. Experience with TRIGA reactors has shown that fuel element bending that 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 Centigraie below the safety limit (ll50*C), and (2) during a pulse, the cladding temperatures Irmain .
well below their stress lirnit. The elongation limit has been specified to ensure that the cladding material will not be subjected to strams that could cause a loss of fuel integrity and to ensure adequate coolant flow 3.2 Reactar Contml and Reactor Safety System l 3.2.1 Reactor Contml Rods Annlicability This specification applies to the reactor control rods.
Obiective The objective is to ensure that sufficient control rods are operable to l maintam the reactor suberitical.
~
Specification There shall be a minimum of three operable control rods in the reac'or core.
Proposed AmenlmentNo.32 O/16/18)
14E Basia 1
- The shutdown margin and excess reactivity specifications rec ulre that the l reactor cari be made suberitical with the most reactive contro rod 1
. withdrawn. This specification helps ensure it. _- l I
3.2.2 Manual rhntrol and Aninmatic Control .l.
Anolicability --
This specification applies to the maximum reactivity insertion rate :
associated with movement of a standard control rod out of the core.
~
Obiective The objective is to ensure that adequate control of the reactor can be l-maintamed during manual and 1,2, or 3 rod automatic control. ,
i Specification ,
The rate of rewtivity insertion associated with movement of either the regulating, shim, or safety control rod shall be not grea'r than 0.63% Ak/k l
(~$0.90) per second when averaged over full rod travel, if the automatic l control uses a combination of more than one rod, the sum of the reac'.ivity of those rods shall be not greater than 0.63% Ak/k (~$0.90) per seceM when -l averaged c"ci full travel.-
Basis T1.e ramp accident analysis (refer to Safecy Analysis Repon, Chapter IX) 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 (~$0.90) per second, when averaged over full travel, is well within that analysis.
3.2.3 Reactor Contml System
'Anolicability -
- This specification applies to the information which must be available to the reactor operator durmg reactor operation.
- Objective -
The objective is to require that sufficient information is available to the operator to ensure safe operation of the res .or. - l Specification The reactor shall not be operated unless the measuring channels listed in ~
- Table 1 are perable. (Note that MN, AU, and SW are abbreviations for~-
manual control mode, automatic control mode, and square wave mode, .
respectively)._
Pmpneed Anwndment No.3i (in6#8f
~ - -
.\$
Table 1
' Measuring Channels !
Min. No. - , : Effective Mode Menurinn Channel Operable MN. AU & SW Bd3 Fuel Element Temperature 1 X. X <
Wide Rango Instrument !
Linm Power 1 X:
- 1.og Power 1 X Reactor Period /SUR - 1 X -
Power Range Instrument- ,
Linear Power 1 X-Pulse Peak Power 1 X
- i Basis Fuel temperature displayed at the control console gives continuous '
information on this parameter which has a :;pecified safety limit. The power level monitors ensure that the reactor power level is adequately monitored a for the manual 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 assumnce that the reactor-is operated at all times within the limits allowed by these Technical-Specifications.
3.2.4 Reactor Safety System and Reactor Interlocks l Anplicability' -
This specification applies to the reactor safety system channels, the reactor
- interlocks, and the watchdog circuit.
Obiective -
The objective is to specify the minimum number of reactor safety system channels and reactor interlocks that must be operable for safe operation. l Specification The reactor shall not be operated unless all of the channels and interlocks
- de<cribed in Table 2a and Table 2b are operable.
+
~
4 Propmed Amendment No. 32 (1A668) h N g -4y y----a p 9-- '
er 9'-'& ' " " ' '- =r- -+a T- v 'e 4 *- e6 =*n** VWW""1 "u-'----1
16 Trble 2a Minimum Reactor Safety System Channels Number - Effective 11 ode Channel Oncrable Function MN. AU hlg M FuelTemperature 1 SCRAM s 650'C' X X X High Power 2 _ SCRAM s 110% of maximum X X reactor operationalpower not to exceed 1.1 Mw Detector Power Supply 1 SCRAM on failure of supply X X voltage SCRAM Bar on 1 Manual SCRAM X X X Console Preset Timer i Transient Rod SCRAM 15 X seconds orless after pulse Watchdog Circuit i SCRAM on sc ftware or self- X X X-check failure
- The limit of 650*C shall be reduced as required by specification 2.2, l' Table 2b Minimum ReactorInterlocks Number Effective Mode CW Oncrable Funcuon MN. AU hlg M Source Level 1 Prevent rod withdrawal 'X without a neutron induced.
signalon the log power channel Pulse Mode inhibit i Prevent pulsing from levels X above 1 kw TransicalRod 1 Prevent applicotions of air X unless cylinderis fully inserted Shim, Safety, and 1 Prevent movement of any X Regulating Rod rod except the transient rod Simultaneous Rod 1 Prevent simultaneous numual X X Withdrawal withdrawalof tworods a
i Pmpmed Amendment No.32 (1/16S8) - .
l
17E BanCS
, reactor is shutdown before the safety limit on the fuel element temperature is reached. '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).
J
- 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
- S1welfication. The high power SCRAMS shall be set to no more than .
110% of the administatively limited maximum reactor operational t powerifit is less than 1 Mw. . ,
i
. c. Operation of the reactor is prevented by SCRAM if there is a failure of the detector power supply for the reacter safety system channels.
d The manual SCRAM allows the operator to shut down the reactor in any _
moac 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 softwr.re or the self-checks fail (see Safety Analysis Report,Section VII).
- g. The interlock to prevent startup of the reactor without a neutron induced si;taal ensures that sufficient neutrons are available for proper r artup in -
al allowable modes of operation,
- h. The interlock to prevent the initiation of a pulse above 1 kw n to ensure that fuel temperature is appmximately pool temperature when a pulse is performed. This is to ensure that the safety limit is not reached.
- i. The interlock to prevent application of air to the transient aud unless the ,
cylinder is fully mserted is to prevent pulsing the reactor in the manual control or automatic contml mode. .
-J. In the pulse mode, movement of any md except the transient rod is prevented by an interlock. This interlock action prevents the addition of reactivity other than with the transient rod.
- k. Simultaneous muual withdrawal of twc rods is prevented to ensure the reactivity rate ofinserdon is not exceeded.
p 3.2.5 - Core I nading nnd Uninading Goeration -
Apolicabilitv -
This specification applies to the source level interlock. l Obiective -
4 The objective of this specification is to allow bypass of the source level :
interlock during operations with a suberitical core.
.t Proposed Amendment No. 32 (1/16S8) N
, . - .- .n - _ - , ~ . . .- -- -. . ., , .
')
(:
n 18 Specification 1
. During core loading and unloading operations when the retetor is . .
suberitical, the source level interlock 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 suberitical. Thus, ;
momentarily defeating the source level interlock is a safe operation. ;
Should the core become inadvertently supercritical, the accxlental insertion of reactivity will not allow Del 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.
3.2.6 SCRAM Timc Anolicability I.
This specification applies to the time required to fully insert any control rod to a full down position from a full up position.
OLiective The objective is to achieve rapid shutdown of the reactor to prevent fuel damage.
Specification The time from SCRAM initiation to the fullinsertion of any comWaod from a full up position shall be less than I second. .
Balli This specification ensures that the reador will bc xomptly shut down when a SCRAM signal is initiated. Experience and ana;ysis, Safety Analysis Report,Section IX, have indicated that for the range of transientmiticipated for a TRIG A reactor, the specified SCRAM time is adequate to ensuit the
-safety rf the reactor. If the SCRAM signalis initiated at 1.1 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 safety hmit.
- 3.3 Coolant SY.11cm 3.3.1 Coolant Level Limits
' Annlicability This specification applies to operation of the reactor with respect to a required depth of water above the tep of the bottom grid plate.
- Pmpneed E- ~ HtNd.32 (th6#8)
19 j Oldeedve ,
l De objective is to ensure ' that water is > resent to provide adequate i personnel shielding and core cooling w sen the reactor is operated, and l during a LO"A. :
i specificadon !
E ne reactor shall not be operated with less than 18 ft, of water above the top of the bottom grid plate. l When the water is more than approximately 18 ft. above the top of the !
bottom grid plate, the water provides sufficient shielding to protect ' t personnel during operation at 1 Mw, and core cooling is achieved with e natural circulation of the water through the core. Should the water level f drop below appmximately 18.25 ft. a l sove while operating at 1 Mw, a low pool level the top (see Te.hnicalof alarm the bottoml Specifications 3.3.2) will alert the operator who is required by administratively a) proved procedure to shutdown the reactor. Once this alarm occurs it wi .1 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 pol. Tests and calculations show that, during a LOCA,680 seconds is sufhcient decay time after shutdown (see Safety Analysis Report, i Section IX ) to prevent the fuel temperature fror . reaching 950*C, To 'l' 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 i temperature during air cooling). l 3.3.2 Detection of 12ak or Loss of Coolant Aanlicability ,
This specification applies to detecting a pool water loss.
Obiecdve The objective is to detect the loss of a significant amount of pool water. ;
Specification A pool level alarm shall be activated and corrective action taken when th pool level drops 26 cm from a level where the pool is full. ;
Raals
]
The alarm occurs when the water level is appmximately 18.25 ft. above the - i top of the bottom grid plate. De point at which the pool is full is approximately 19.1 ft. above the top of the bottom grid plate. The reactor !
,. staff shall take action to keep the core covered wkh water according to existing procedures. The alarm is also transmitted ta the Police Services -
annunciator panel which is monitored 24 hrs, a day. Tbc alarm provides a signal that occurs at all times (see Safety Analysis Report,Section VII), 1 Thus, the alarm piovides time to initiate corrective action before the ,
_ radiation from the core poses a serious hazard. ,
l i
Propomi Amemiment No.32 (t/16/98) j 7 .j U _.,-,.,.,__.__,;,..,,,_a_ .,,,__._;__ -
20 .;
3.3.3 Fic=lan Prrwiet Activity l g tg This specification applies to the detection of fl2sion product activ' ity.
OIdKb8 f The objective is to ensure that fission products from a leaking fuel element I l
are detected to provide opponunity to take protective action. l s
Specification )
An air paniculate monitor shall be o )erating in the reactor bay whenever the j
reactor is operating. An alarm on th s unit shall activate a building
- evacuation alarm. ,
i Basis ,
. This unit will t,: sensitive to airborne radioactive particulate matter i f
- containing fission pmducts and fission gases and will alert personnelin time to take pmtective action. ;
3.3.4 Pant War Sunniv for i enk Prntection -
Annliemhility This specification applies to pool water supplies for the reactor pool for leak t
protectior' QldcG1XS The objective is to ensure that a supply of water is available to replenish reactor pool water in the event of pool vinter leakage. l c Specification .
A source of water of at least 100 GPM shall be available either fmm the University water supply or by diverting the heat exchanger secondary flow i to the pool.
Basis P Provisions for both of these supplies are in ) lace and will su sply more than
. . the specified flow rate. This flow rate will x more than suflicient to handle leak rates that have occurred in the past or any anticipated leak that might ,
occur in the future.
3.3.5 - ('aatant Canductivity 1 tmits a
Annlicabiliev ,
.. t This specification applies to the conductivity of the water in the pool. ,
4 i
=
f Pmposed ^ - -- M No. 32 (1/16M) {
- . . _a - . . . _ , - . . .a-- .-. . - , -
21 i
, Ohktives l
- The objectives are: _l
- a. To prevent activated contaminants i*.wn becoming a radiological hazard, and
- b. To help preclude corrosion of fuel cladding and other primary system !
components. !
Specification The reactor shall not be operated if the conductivity of the bulk pool water is {
greater than 5 microslemens/cm (5 micrombos/cm). i M
i Experience indicates that 5 microslemens/cm is an acceptable level of water l'
- contaminants in an aluminum / stainless steel system such as that at the f PSBR. - Based 01
- xperience, activation at this level does not pose a significant radiological hazard, and significant corrosion of the stainless steel fuel cladding will not occur when the conductivity is below 5 -
microslemens/em. l 3.3.6 Caalant Temaarminre 1 Imits ;
Annliemhillev This specification applies to the pool water temperature.
. 1 4
Obiective The objective is to maintain the pool water temperature at a level that will not cause damage to the demineralizer resins.
Snecifiention I l
- 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). i Basis This a scification is primarily to preserve demineralizer resins. Information available indicates that temperature damage will be minimal up to this temperature.
3.4 Confinement ,
Applicability
- This specification applies to reactor bay doors. ,
e
' 'Ihe objective is to ensure that no large air passages exist to the reactor bay during l
reactor operation.
Pmruned Amendment No. 32 (1/t638) .
a 7 -hr,6-f1--r-Iaw wirrr-p v y W-y 1r y W im'weer- *-W4q3 47-ee+'--eA-g4-1 >+puiy+*1miuwy +- g- .6i- WWg-'-M-g m .y g qq .., t- 1 gs-Gr-' e
22 Specificadon
%e reactor bay truck door shall be closed when the reactor is operating. Personnel doors to the reactor bey shall not be blocked op:n and left unattended when the reactor is operating.
Bads his specification helps to ensua that the air pressure in the reactor bay is lower - l than the remainder of the buikling and the outside air pressure. Controlled air -
pressure is maintained by the air exhaust system and ensures controlled release of l-any airborne radioactivity.
3.5 - Raelamed Rafetv Features . Facilh_v Ruhmust Svateni and Rmereenev Ruhmunt Applicabilhv Dis specification applies to the operation of the facility exhaust system and the emergency exhaust system.
Ohlective ne objective is to mitigate the consequences of the release of airborne radioactive materials resulting from reactor operation.
Enacmentinne l When the reactoris operating:
- a. The facility exhaust system shall be operating, and
- b. The emergency exhaust system shall be operable except for periods of time less than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> when necessary to permit maintenance and repair.
Banca l-Durin g normal opermion, the concentration of airborne radioactivity in unrestricted areas us 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 autcmatically cause the facility exhaust system to close and the exhausted air to be passed through the emergency exhaust system filters before release. His reduces the radiation within the building. The filters will remove
= 90% all of the particulate fission products that escape to the atmosphere, ne emergency exhaust system activities only during an evacuation where upon all personnel att required to evacuate the building (section 3.6.2), if there is an .
evacuation while the emergency exhaust system ir out of service for malmenance or repair, personnel evacuation is not prevented.
Personnel dose to the public will be equivalent or less whether or not the emergency en haust system functions in the unlikely event an accident occurs during the maintenance or repair.
1 Pmposed Air / nhnent No.32 (t/16#8)
_-. - -- . - .. _-. . - -. -. - . _- __ ~ _ _- - - -
i I
23 3.6 Radinilan Montandng System 3.6.1 Radiarlan Manitoring Information Annlicability ,
This specification applies to the radiation monitoring info < A. we >WA must be available to the = cactor operator during r: actor < e.#,
Obiective I
The objective is to ensure that sufficient radiation monitongg niarnation is available to the operator to ensure personnel radiation safety during reactor operation.
Soccifiention The reactor shall not be operated unless the radiation monitoring channels listed in Table 3 are operating.
Table 3 Radiation Monitoring Channels Rndlation Monitorinn ~
Channch 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.
Beamhole Laboratory Monitor radiation in the 1 Monitor Beamhole Laboratory required only when the -
laboratory is in use.
Bases
- a. The radiation monitors provide information to operating personnel of l ,
any impending or existing danger imm radiation so that there will be -
sufficient time to evacuate the facility and to take the necessary steps to control the spread of radioactivity to the surroundings.
- b. The area radiation monitor in the Beamhole Laboratory provides l information to the user and to the reactor operator when this laboratory -
is in use. -
3.6.2 Evacuation Alarm Aeolicability This specification applies to the evacuation alarm.
Prossed Amendment No.32 (t/108)
24 Objective The objective is to ensure that all personnel are alerted to evacuate the l PSBR building wheu a potential radiation hazard exists within this building.
Snelflentinn The reactor shall not be operated unless the evacuation alarm is operable ,
and udible to personnel within the PSBR building when activated by the radiation monitoring channels in Table 3 or a manual switch. :
HAhls The evacuation alar.n produces a loud pulsating sound throughout the PSbR building when there is any impending or existing dan ger from radiation.
The sound notifies all personnel within the PSBR bul ding to evacuate the building as prescribed by the PSBR emergency procedure.
3.6.3 Argon 41 Discharge l_imit Applienhility This specification applies to the concentration of Argon 41 that may be discharged from the PSBR, Oblective The objective is to ensure that the health and safety of the public is not l endangered by the discharge of Argon 41 from the PSBR.
Soecification All Argon 41 concentrations produced by the operation of the reactor shall l be below the limits imposed by 10 CFR Part 20 when averaged over a year.
Basis The maximum allowable concentration 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 pCi/ml. Measurements of Argon-41 have been made in the reactor bay '
when the reactor operates at 1 Mw. These measurements show that the concentrations averaged over a year produce less than 1.0 x 10 8 pCi/ml in an unrestricted area (see Environmental Impact Appraisal, December 12,1996).
3.6.4 As Irw As Reasonably Achievable (ALARA)
Anolicabilltv This specification applies to all reactor operations that could result in occupational exposures to radiation or the release of radioactive efiluents to the environs.
Obiective The objective is to maintain all exposures to radiation and release of radioactive effluents to the envimns ALARA, Pmpomi Amendment No.32 (t/1638)
25 Specificadon An ALAR A pmgram shall te in effect.
IlA113 llaving an ALARA pmgram will ensure that occupational exposures to radiation and the release of radioactive effluents to the environs will be ALAR A. Ilaving such a formal program will keep the staff cognizant of the importance to minimize radiation exposures and effluent releases.
3.7 Limitations of Exoeriacm3 Applicability
'Ihese specifications apply to experiments installed in the reactor and its experimental facilities.
Objective The objective is to prevent damage to the reactor and to minimize rekase of l radioactive materia s in the event of an experiment failure.
Specifications
%e reactor shall not be operated unless the following conditions governing experiments exist:
- 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). !!owever, the tractivity of a movable experiment and/or movable portions 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 ensure 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 ex xriments in or near the core, secured negative reactivity experiments may be ac ded 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 c ucstion. The failure mechanisms that shall be analyzed include, but are not itm.ted to corrosion, overheating, impact fmm projectiles, chemical, and mechanical explosions.
Explosive material shall not be stored or used in the fa:ility without proper safeguards to prevent release of fission products or loss of reactor shutdown capability.
IWposed Amerdment No.32 (1/16/98)
M ,
if an experimental failure occurs which could lead to the release of fission products or the loss of reactor shutdown capability, physical inspection shall be j performed to determine the consequences and the need for corrective action.
The results of the ins xetion and any conective action taken shall be reviewed .
by the Director or a d esignated alternate and determined to be satisfactory ;
before operation of the reactor is resun,ed. j
- c. Experiment materials, except fuel materials, which could off gas, sublime, f volatilize, or produce aemsols under (1) normal o xratin ; conditions of the i experiment and reactor, (2) credible accident conditions n the reactor, or (3) possib'e accident conditions in the experiment, shall be limited in activity such '
that the airborne 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 assumptions shall be used:
- 1) If an ex xriment falls and releases radioactive gases or acrosois to the t reactor ny or atmosphere,100% of the gases or acrosols escape.
- 2) If the effluent fmm an ex xrimental facility exhausts through a holdup tank which closes automatical y on high radiation level, at least 10% of the gaseous activity or acrosols 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 panicles, at least 10% of those vapors can escape.
- 4) For materials whose boiling point is above 130'F and where vapors fomied by boiling this material can escape only through an undisturbed column of water above the core, at least 10% of these vapors can escape.
- f. Each fueled experiment shall be controlled such that the totalinventory of iodine isotopes 131 through 135 in the experiment is no greater than 1.5 curies.
In addition, any fueled experiment which would ;;enerate an inventory of more than 5 millicunes (mci) of I 131 through I 135 s 1all be reviewed to ensure that in the case of an accident, the total release of iodine will not exceed that -
postulated for the MilA (see Safety Analysis Report,Section IX ).
!!AE1
- a. This specification limits the sum of the reactivities of a maximum allowed pulse and a movat,1c experiment to the specified maximum reactivity of the transient rod. %Is 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 adoltion, the maximum power attainable with the ramp insertion of 1.4% Ak/k
(~$2.00)is less than 500 kw staning from critical,
- b. The maximum worth of all experiments is limited to 2.45% Ak/k (~$3.50) so that their inadvertent sudden removal fmm the cold critical reactor will not result in the reactor achieving a power level high enough to exceed the temperature safety limit (ll50*C). The wonh of a single secured experiment is limited to the allowed pulse reactivity insertion as an increased measure of safety. Should the 2.45% Ak/k,(~$3.50) reactivity be inserted by a ramp increase, the maximum power attainable is less than 1 Mw.
Npneed Amendment No.32 O/16N8)
l 27 l
- c. Since the initial core is suberitical, adding and then inadvenently removing all ,
negative reactivity experiments leaves the core in its initial suberitical condition. l
- d. The design basis accident is the MilA (See Safety Analysis Re ert,Section IX).
A chemical explosion (such as detonated TNT) or a mechanica 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 :o material released by a failed experiment could also lead to release of fission products, if an experiment failure occurs a specialinvestigation is required to ensure that all effects from the failure are known before operation proceeds.
- c. This specification is intended to reduce the likelihood that airbome activities in excess of the limits of Appendix Ts Table 2 of 10 CFR Part 20 will be released to the atmosphere outside the facility boundary,
- f. The 5 mci limitation on 1 131 through I 135 ensuits that in the event of failure l of a fueled experiment, the exposure dose at the exclusion area boundary will be less than that estolated for the MilA (See Safety Analysis Report,Section IX) l cven if the loc ine is released in the air.
4.0 SURVEILLANCE REQUIREMENTS 4.1 Reactor Parameters - ,
4.1.1 Reactor Power Calibration Applicabliity This specification applies to the surveillance of the reactor power calibration.
Obiective The objective is to verify the perfom1ance and operability of the power measuring channel.
Saccification l A thermal power channel calibration shall be made on the linear power level monitoring channel annually, not to exceed 15 months.
Basis The themial power level channel calibration will ensure that the reactor is l operated at the authorized power levels.
4.1.2 Reactor Excess Reactivity Applicability This specification applies to surveillance of core excess reactivity.
Proposed Amendment No. 32 (1/16/98)
28 Dhicctive The objective is to ensure that the reactor excess reactivity does not execed the Technical Specifications and the limit analyzed in Safety Analysis :
Report,Section IX.F. ,
sne.cmenilan
'lhe 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),
11A111 Excess reactivity measurements on this schedule ensure that no unexpected l changes have occurred in the core and the core configuration does not exceed excess reactivity limits established in the Specification 3.1.2. l 4.1.3 TRIGA Fuel Elements Anolicability This specification applies to the surveillance requirements for the TRIGA fuel elements.
i Obiective The objective is to verify the continuing integrity of the fuel chment cladding.
Specification Fuel elements and control rods with fuel followers shall be inspected visually for dama ;c or deterioration and measured for length and bend in accordance with t1e following:
- a. Before being placed in the core for the first time or before return to service,
- b. Every two years, not to exceed 30 months, or at intervals not to exceed the sum of 3,500 dollars in pulse reactivity, whichever comes first, for elements with a NP greater than 1 and for control rods with fueled followers.
- c. Every four years, not to exceed 54 months, for elements with a NP of I or less,
- d. Upon being removed from service. Those removed from service are then exempt from further inspection.
IIAsih The frequency of insxction and measurement schedule is based on the parameters most like y to affect the fuel cladding of a pulsing reactor ograted at moderate aulsing levels and utilizing fuel elements whose c:1aracteristics are we I known. ..
Propaned AmendmentNo.32 (1/16/98)
ii 29 4.2 n..rior conoimt .,wi marciv svaiem -
4.2.1 Itam-tivhv Worth Applicabilliv This specification applies to the reactivity worth of the control rods.
Ohlective The objective is to ensure that the control rods are capable of maintaining l the reactor suberitical.
Snacificatinn The reactivity worth of each contml md and the shutdown margin for the core loading in use shall be detennined annually, not to exceed 15 months, or following core or control rod changes equal to or greater than 0.7% Ak/k
(~$1,00).
Basis The reactivity worth of the control rod is measured to ensure that the .l required shutdown margin is available and to provide an accurate means for determining the core excess reactivity, maximum reactivity, insertion rates, and the reactivity worth of experiments inserted in the core.
4.2.2 Reactivity insertion Rate Annlicabilltv This specification applies to control rod movement speed.
Oblective The objective is to ensure that the reactivity addition rate specification is not l violated and that the control rod drives are functioning.
Specification The rod drive speed both up and down and the time from SCRAM initiation to the full insertion of any control rod from the full up position shall be measured annually, not to exceed 15 months, or when any significant work is done on the rod drive or the rod.
RAlli
- This specification ensures that the reactor will be ) rom ptly shut down when l a SCRAM signal is initiated. Experience and ana' ysis llave indicated that for the range of transients anticipated for a TRIG A reactor, the specified SCRAM time is adequate to ensure the safety of the reactor. It also ensures l that the maximum reactivity addition rate specification will not be exceeded.
Proposed Amendment No 32 (t/1W98)
f M
4.2.3 Remetor Safety System l Annilcahilliv i
The specifications apply to the surveillance requirements for measurements, channel tests, and channel checks of the reactor safety systems and watchdog circuit.
Oblective The objective is to verify the performance and operability of the systems -
and components that are directly related to reactor safety.
Snecifkatiggg .
- a. A channel test of the SCRAM function of the wide range linear, power range linear, fuel temperature, manual, and preset timer safety channels -
sha i 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 functions for both the wide range and the power range and the watchdog circuit shall be performed annuaily, not to exececi 15 months.
- c. Channel checks for operability shall be performed dally on fuel element temperature, wide range linear power, wide range log power, wide range reactor period /SUR, and power ran ge linear power when the reactor is to be operated, or prior to each operat .on that extends more than one day,
- d. The power range channel shall be compared with other independent l ,
channels for pmper channel indication, when appropriate, each time the reactor is operated,
- c. The pulse peak power channel shall be compared to the fuel temperature ,
each time the reactor is pulsed, to ensure pmper peak power channel l operation.
Eases System components have proven operational reliability. l
- a. Daily channel tests ensure accurate SCRAM functions and ensure 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 ;
ensure that this systein works, based on past expenence as recorded in the operation log book.' An annual channel test of the watchdog circuit ;
is sufTicient to enswe operability.
- c. The channel checks will make information available to the operator to ensure safe operation on a daily basis or prior to an extended run, d, Comparison of the percent power channel with other independent power channels will ensure the detection of channel drift or other possible deterioration of its operational characteristics.
Proposed Amendment No.32 (t/16SB) ;
l
31
- c. Comparison of the peak pulse power to the fuel temperature for each pulse will ensure the detection of of its operational characteristics. possible channel drift or de 4.2.4 Reactor Interlocks Appilcability These specifications apply to the surveillance requirements for the reactor control system interlocks.
Obiective The objective is to ensure performance and operability of the reactor control l system interlocks.
Spspifications
- a. A channel check of the source interlock shall be performed each day that the reactor is operated or prior to each operation that extends more than one day except when the neutron signal ls greater than the setpoint when the sourte is removed from the core.
- b. A channel test shall be performed semi annually, nt to exceed 71/2 months, on the pulse mode inhibit interlock which prevents pu! sing from power levels higher than one kilowatt.
- c. A channel check shall be perfomied semi annually, not to exceed /1/2 months, on the tmnsient md interlock which revents application of air to the transient rod unless the cylinder is full inserted.
- d. A channel check shall be performed semi annually, not to exceed 71/2 months, on the rod drive interlock which prevents movement cf any rod except the transient rod in pulse mode, not to exceed 71/2
- e. Amonths, channel check on the shallinterlock rod drive be performed semi annually,imultaneous which prevents s manual withdrawal of more than one rod.
Eases l The channel test and checks will verify operation of the reactor interlock system. Experience at the PSBR indicates that the prescribed frequency is adequate to ensure operability. l Afler extended operation, the photo neutron source strength may be high enough that removing the source may not drop the neutron signal below the setpoint of the sourre interlock. With a large intrinsic source there is no practical way to channel check the source interlock, in which case there is no need for a source interlock.
4.2.5 Ovemower SCRAM Apolicability This specification applies to the high power and fuel temperature SCRAM channels.
Propsed Amendment No.32 (t/lMR)
32 i
<Oblective l 3
The ob,ective is to verify that high power and fuel temperature SCRAM !
channe:s perform the SCRAM functions.
1 Snecificatinn The high power and fuel temperature SCRAM's shall be tested annually, not to exceed 15 months. <
Rasis Experience with tlac PSBR for more than a decade, as recorded in the operation log books, indicates that this interval is adequate to ensure operability.
4.2.6 Trantlent Rnd Test !
r Annlicability These specifications apply to surveillance of the transient rod mechanism. l Obiective ,
The objective is to ensure that the transient rod drive mechanism is l maintained in an operable conditlen.
Snectrientions l l
- a. On each day that pulse mode operation of the reactor is planned, a functional performance check of the transient rod system shall be performed. The transient rod drive cylinder and the associated air supply system shall be inspected, cleaned, and lubricated as necessary l annually, not to exceed 15 months.
I
- b. The reactor shall be pulsed annually, not to exceed 15 months, to compart fuel temperature measurements and peak power levels with those of previous iulses of the same reactivity value or the reactor shall not be pulsed unti such comparative pulse measurements are performed.
Balli Functional checks along with periodic maintenance ensure repeatable l performance. The reactor is pulsed at suitable intervals and a comparison made with previous similar pulses to determine if changes in transient rod drive mechanism, fuel, or core characteristics have taken place.
?
4.3 ' Cao.' Ant System 4.3 Mrc linse Intnection Annlicability This specification applies to the dedicated fire hoses used to supply water to the poolin an emergency.
i Pmpmal Amendmem No.32 (t/lfd98) .
~ - ~ . _ . . _ - ._.- _ _ _ . _ _ _ _
r.._._ _ _ _ _ . _ _ _ _ .. _ _ . _ _ _. - _ . _ ___ . - _
L i
33 -!
Qblecdve ;
i The objective is to ensure that thes,e hoses are operable. l l
[
3and0sadon -l
, The two (2) dedicated fire hows that xovide supply water to the poolin an emergency shall be visually inspectec for damage and wear annually, not to exceed 15 months, t M ,
- This freqcency is adequate to ensure that significant degradation has not .l !
occurred since the previous inspection. l 4 r 4.3.2 Pool Water Temperature ;
i
] Applicability l t
This specification applies to pool water temperature.
Oht cilve ,
The objective is to limit pool water temperature. :l Specificadan :
The pool temperature alarm r. hall be calibrated annually, not to exceed 15 :
mombs. !
M -
t Experience has shown this instrument to be drift free and that this interval is ,
adequate to ensure operability. l 4.3.3 Pool Water Conductivity Annilenhillev -
i This specification applies to surveillance of pool water conductivity.-
MIllEC ]
I The objective is to ensure that pool water mineral content is maintained at l an acceptable levd.
Specification l
_ i
- Pool water conductivity shall be measured and recorded daily when the *
- reactor is to be operated, or at monthly intervals when the reactor is shut
' down for this time period. :
Basis !
~
- Based on experience, observation at these intervals provides acceptable
- C surveillance oflimits that ensure that fuel clad corrosion and neutron l activation of dissolved materials will not occur, Propmed Amendment No.32 (1/lW98) -
,, m - ,, ,_. .- , , . . . - ,..._._,,__,_,,_,,._c~2_,;...._,___ _;____.a.___
34 4.3.4 Paal Wwar 12 vel Alarm Applicabilhv This specification applies to the surveillance requirements for the pool level alarm. !
I The objective is to verify the operability of the pool water level alarm. - -l l Specification The pool water level alarm shall be channel checked monthly, act to exceed 6 weeks, to ensure its operability, j i
Baals Experience, as exhibited by past periodic checks, has shown that monthly ,
checks of the pool water level alami ensures operab!!ity of the system l ,
during the month.
4.4 Confine 19CA1 -;
e Applicabilitv His specification applies to reactor bay doors. l Ohiertive The objective is to ensure that reactor bay doors are kept closed as per Specification l 3.4. ,
Specification Doors to the reactor bay shall be locked or under supervision by an authorized e keyhokler.
Baals :
- A keyholder is authorized by the Director or his designee. l 4.5 Facility Fuhnust System and Fr .Jencv Exhaust System Annliemhility Dese specifications apply to the facility exhaust system and emergency exhaust
. system.
OhicGilEt L %c objective is to ensure the proper operation of the facility exhaust system and l-emergency exhaust system in controlling releases of radioactive matenal to the -i uncontrolled environment.
i 3
I%pased AmendmentNo.32 (tIINWL)-
, co ,0- ,- . ; . _1 _ _ _ _ . - , , , , _ _, . , _ _ . _ - ,
35 Specifications l
- 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 mor,thly, not to exceed 6 weeks, whenever operation is scheduled, that the facility exhaust system is secured when the emergency exhaust system activates during an evacuation alarm (See Technical Specificationn 3.6.2 and 5.5).
BMia Experience, based on periodic checks performed os er years of operation, has l demonstrated that a test of the exhaust systems on a monthly basis, not to exceed 6 weeks,is sufficient to ensure the proper operation of the systems. His provides l reasonable assurance on the contml of the release of radio:ctive ma:erial.
4.6 Itadiatinn Manitoring Svdem and Effluente 4.t5.1 Rndimilnn Manitorine System and Evacuntlan Alarm Annlicahility This specification applies to surveillance requirements for the area radiation monitor sthe Beamhole Laboratory radiation monitor, the air radiation monitet, and the evacuation alarm.
Objective The objective is to ensure that the radiation monitors and evacuation alarm l are operable and to verify the appropriate alarm settings.
Specification The area radiation monitor, the Beamhole Laboratory radiation monitor, the l continuous air (radiation) monitor, and the evacuation alarm system shall be channel tested monthly not to exceed 6 weeks. Hey shall be verified to be operable by a channel check daily when the reactor ls to be operated, and shall be calibrated annually, not to exceed 15 months.
limia Experience has shown this freq uency of verification of the radiation monitor set points and operability and t he evacuation alarm operability is adequate to correct for any variation in the system due to a change of o wrating characteristics. Annual channel calibration ensures that un ts are within the l specifications defined by procedurer,.
4.6.2 Argan 41
'Annliemhility This specification applies to surveillance of the Argon 41 produced during reactor operation.
Proposed Amendment No. 32 (l/16/98)
36 Obbetive To ensure that the production of Argoa41 does not exceed the limits l specified by 10 CFR Part 20.
Snecifientinn The praluction of Argon 41 shall be measured and/or calculated for exh new experinvnt or experimental facility that is estimated to produce a dox greater than 1 mrrm at the exclusion boundary.
Basis 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 Argon 41 production for any experiment or experimental facility that exceeds 1% of the annual limit.
4.6.3 - ALARA Anpliemhility This specification applies to the surveillance of all reactor operations that could result in occupational exposures to radiation or the release of radioactive effluents to the environs.
Ohlective The objective is to pmvide sutveillance of all operations that could lead to occupational exposures to radiation or the release of radioactive effluents to the cuvirons.
Specification As part of the review of all operations, consideration shall be given to att emative operational modes that might reduce staff exposures, release of radioactive materials to the environment, or both.
Basis Experience has shown that experiments and operational requirements can, in many cases, be satisfied with a variety of combinations of facility options, core positions, power levels, time delays, and effluent or staff radiation exposures. Similarly, overall reactor scheduling achieves significant reductions in staff exposures. Consequently, ALARA must x a part of both overall reactor scheduling and the detailed experiment planning.
4.7 hneriments -
Applicability This specification applies to surveillance requirements for experiments.
Obiective
%c objecth " ss to ensure that the conditions and restrictions of Specification 3.7 l
' art mCt.
Proposed Amendmem No.32 (1/1fWR) .
i 37 i sprification l Donc conditions and restrictions listed in S ification 3.7 shall be cw31dered by - i the P3BR authorired reviewer before signin he irradiation authorization for each l expenment. ,
M Authorized revicwers are appointed by the facility director. I 5.0 DE3IGN FEATURES -!
5.1 Rt. actor Fuel l Raarinemetan ;
he individual unirradiated TRIGA fuel elements shall have the following l characteristics:
- 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. ) j
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 Enacifications
- 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 ex wrimental facilities, shall be water, or D 20, or graphite, or any combination of tie three moderator materials.
5.3 Comrol Rods Enacificatinnt l
- a. The shim, safety, and regulating control rods shall have SCRAM capability and contain borated graphite, B4C powder, or sron and its compounds in solid fonn as a poison in stainless steel or aluminum cladding. These rods may incorporate fueled followers which have the same characteristics as the fuel region in which they are used.
- b. The transient contml rod shall have SCRAM capability and contain borated 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 incorporate a voided or a solid aluminum follower.
i Prorwood Amemiment No.32 (t/te/ps)
L
38 5.4 Fuel Storage Specifications
- a. All fuel elements shall te a;ored in a geometrical array where the keffis less than 0.8 for all conditions of moderation,
- b. Irradiated fuel elements shall be stored in an array which shall pennit 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 Dav und Exhaust Systems Soccifications
- 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 19(X) m3 .
- b. The reactor bay shall be equipped with two exhaust systems. Under nonnal o yrating conditions, the facility exhaust system exhausts unfiltered reactor bay a r to the envimnment releasing it at a point at least 24 feet above ground level.
Upon 'nitiation of a bul;dlng evacuation alann, the previously mentioned system is auwmatically secured and an emergency exhaust system automatically stans.
The 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 Soccincation l 1 hc reactor coie shall be cooled by natural convective water flow.
6.0 ADMINISTRATIVE CONTROL.S 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 (level 2),
who reports to the Vice President for Research Dean of the Graduate School through the llead of the Nuclear Engineering Department and the Dean of '
the College of Engineering. Administrative and fiscal responsibility is within the offices of the Department llead and the Dean.
The minimum qualifications for the position of Director of tue PSBR are an l advanced degree in science or engineering, and 2 years experience in reactor i operation. Five years of experience directing reactor operations may be l substituted for an advanced degire. l The Manager of Radiation Pmtection reports through the Director of I Environmental llealth and Safety, the assistant Vice President for Safety and Environmental Services, and to the Senior Vice President for Finance and llusiness/firasurer. The quali0 cations for the Manager of Radiation i l
Pmpwd Ameralment No,32 (1/16,%)
t
I 39 >
r 3 r 3 i Senior N Present for Finance Vke Present for Research ,
and Business / Treasurer . Dean of the Graduate School :
L' J L J -l 2 i I
r 3 Assistant Vice President for Safety
- and Environmental Services r 3 L J Dean, College ofEngineering j l.
r .
3 Director of Environmental :'
Health and Safety L 'J r 3 .
Nuclear Engineering !
Department Head t J l r 3 r 3 ,
Manager of _ Penn State Reactor . zi Safeguards Committee :
Radiation Protection '
J L J L
I ' l l l Mww i
---- I Penn State Breareale Reactor l
(Level 2) <
L 'J ,
1 Manager of Operations and Training ,!
(Level3)
L _J .
t r- 7 Operating Staff . j
- Senior Reactor Operators and i Reactor Operators 4 (Level 4) !
t J 5 !
i ORGANIZATION CHART . 1 a:r :
i
'I i
V Proposai Amendment No. 32 (1/1668) .
i .
- _ ..m. . . ~ . - ,,,-.e__, -, . . , . . . ~,--,-,.,_,-m.- .--._.-..~,m-,.---r,..~,-
40 Protection protection,pos! tion are the equivalent of a graduate degreel in ra and certification by ne American Boani of Health Physics or eligibility for certification.
6.1.2 Re p dhilitv 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 all instances, res mnsibilities of one level may be assumed by designated alternates or by his ier levels, conditional upon appropriate qualifications.
6.1.3 Staffing
- a. The minimum staffing when the reactor is not secured shall be:
- 1) A lleensed operator present in the control room,in accordance with applicable regulations.
- 2) A second perr.on present at the facility kble to carry out prescribed i written instructions.
- 3) If a senior reactor operator is not present at the facility, one shall be available by telephone and able to be at the facility within 30 minutes,
- t. A list of reactor facility personnel by name and telephone number shall be readily available in the control room for use by the operator. De list shallinclude:
- 1) Management personnel.
- 2) Radiation safety personnel.
- 3) Otheroperations personnel,
- c. Events requiring the direction of a Senior Reactor Operator shall include: -
- 1) All fuel or control md relocations within the scactor core region.
- 2) Relocation of any in core experiment with a reactivity worth greater
, than one dollar.
- 3) Recovery from unplanned or unscheduled shutdown (in this instance, -
- documented verbal concurrence from a Senior Reactor Operator is required).
, Pmposed Ammkum No.32 (1/1@8) -
e
41 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 Standard for Selection and Training of Personnel for Research Reactors, ANSI /ANS 15.41988, Sections 4 6. l 6.2 Review and Audit 6.2.1 Safeguards Committee Comoosition A Penn State Reactor Safeguards Committee (PSRSC) shall exist to pmvide an independent review and addit of the safety aspects c,f reactor facihty 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 engincenng fields. The committee shall have at least one neember with health physics expenise. The committee shall be appointed by and te ) ort to the Dean of the College of Engineering. The PSBR Director shal be an ex officio member of the PSRSC.
6.2.2 Charter and Rules The operations of the PSRSC shall be in accordance with a written chaner, 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 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 Subgmups the committee chainnan can appoint ad lioc committees as deemed necessary,
- d. Minutes of the meetings shall le recorded, disseminated, reviewed, and appmved in a timely manner.
6.2,3 Review Function The following items shall be reviewed:
- a. 10 CFR Part 50.59 reviews of: l
- 1) Proposed changes in equipment, systems, tests, or experiments. l
- 2) All new procedures and major revisions thereto having a significant effect upon safety.
- 3) All new experiments or classes of experiments that could have a significant effect upcm reactivity or upon the release of radioactivity,
- b. Proposed changes in technical specifications, license, or charter. l
- c. Violations of technical specifications, license, or chaner. Violations of internal pnxedures or instructions having safety significance.
Propned Amendment No.32 (t/16S8)
42
- d. - Operating abnormalities having safety significance. l
- e. Special reports listed in 6.6.2. l
- f. Audit reports. l 6.2.4 Andh The audit function shall be performed annually, not to exceed 15 months, preferabl y a non member of the reactor staff. The audit function shall be rio a person not direct! involved with the function being audited.
audit function shallinclude lective (but cu, whensive) examinations of operating records, logs, and other documents. kIncussions with operating -
wrsonnel and observation of operations should also be used as ate.
Aficiencies uncovered that affect reactor safety shall promptly re mrted to the Nuclear Engineering Department Head and the Dean of the Co lege of Engineering. The following items shall be audited:
- a. Facility ations for conformance to Technical Specifications, license, and ures (at least once per calendar year with interval not to exceed 15 months).
- b. The requalification program for the operating staff (at least'once every other calendar year with the interval not to exceed 30 months).
- c. The results of action taken to correct deficiencies that may occur in the reactor facility equipment, systems, structures, or methods of operations that not to affect exceed 15 reactor months safety)(at
. least once per calendar year with the
- d. The reactor facility emer gency plan and implementing procedures (at least once every other ca' endar year with the interval not to exceed 30 months). ,
6.3 Onerating Pmcadures Written pmcedures shall bs .eviewed and approved prior to the initiation of activities covered by them in accordance with Section 6.2.3. Written procedures shall be unte to ensure the safe operation of the reactor, but shall not preclude the une ofI pendent ud gment and action should the situation require such, ting procedures s all be in effect and shall be followed for at least the lowing items:
- a. Startup, operation, and shutdown of the reactor.
' b. . - Core loading, unloading, and fuel rnovement within the reactor.
- c. Routine maintenance of major components of systems that could have an effect 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. l Pmpneed Amendment No.32 (1/16#s)
_ - _ _ _ . . - _ , , - _ _ 2__
43
- g. Evacuation.
- h. Fire or Explosion.
- i. Gascous Release.
J. Medical Emergencies.
- k. Civil Disorder.
- 1. Bomb Threat.
- m. Threa 4 Theft of Special Nuclear Material.
- n. Theft of Special Nuclear Material.
- o. . industrial Sabotage,
- p. - Experiment Evaluation and Authorization.
_ q. Reactor Operation Using a Beam Port.
- r. .D 2011andling.
- s. licalth Physics Orientation Requirements.
- t. Hot Cell Entry Procedure, i
- u. Implementation of emergency and security plans.
- v. Radiation instrument calibration
- w. Loss of pool water.
6.4 - Review and Annmval of Erneriments
- a. All new experiments shall be reviewed for Technical Specifications compliance,10 CFR Part 50.59 analysis, and aproved in writing by level 2 management or designated allemate prior to in t ation.
- b. Substantive changes to experiments previously reviewed by the PSRSC shall be made only after review and approval in wnting by level 2 management or designated allemate.
6.5 Ramired Action 6.5.1 Actinn to be Taken in the Event the Safety Limit is Eww 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. "he safety limit violation shall be promptly reported to level 2 or .
- designated alternates.-
~
. Propmed Amendment No.32 (1/1688)
44
- c. An immediate report of the occunence shall be made to the Chairman, PSRSC and reports shall be made to the USNRC in accordance with Specification 6.6. l
- d. A report shall be prepared which shall include an analysis of the causes and extent of possible resultant damage, cincacy of corrective action, ,
and recommendations for measures to prevent or reduce the probability I of recurrence. This report shall be submitted to the PSRSC for review. ,
, 6.5.2 Acelan to be Taken in the Event of a Raaneenkle Occurrence !
In the event of a reportable occurrence, (1.134) the following action shall be l taken:
4
- a. The reactor shall be returned to normal or shutdown. If it is necessary to l shutdown the reactor to correct the occurrence, operatie.is shall not be !
sumed unless authorized by level 2 or designated altemates,
- b. The Director or a designated alternate shall be notified and conective action taken with respect to the operations involved, i c. The Director or a designated alternate shall notify the Nuclear ;
Engineering Department Head who, in turn, will notify the office of the ,
Deim of the Col ege 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 t PSRSC.
- c. A report shall be made to the PSRSC which shall include an andlysis of the cause of the occur 7ence, 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 re wrt shall be made to the Document Control Desk, USNRC '
- .Wasiington, DC 20555.
- 1 6.6 Re a rts 6.6.1 Oneratine Renorts An annual report shall be submitted within 6 months of the end of The i
Pennsylvania State University fiscal year to the Document Control I.ksk, "
USNRC, Washington, DC 20555, including at least the following items:
- a. A narrative summary of reactor operating exixtience including the energy produced by the reactor, and the num er of pulses 2 $2.00 but
. less than or equal to $2.50 and the number greater than $2.50.
bi - The unscheduled shutdowns and reasons for them including, where
- applicable, corrective action taken to preclude recurrence.
- c. ' Tabulation of major preventive and corrective maintenance operations 1 having safety significance.
Proposed Amomiment No.32 (t/IN98)'
.a . . , . . ~ . _ . ~..w., -.
-. .--. -,-...-,-.-w. c.-.~c . ., -,_ --- . w. . . w .-
l 45 !
- d. Tabulation of major changes in the reactor facility and procedures, and !
tabulation of new tests and experiments, that are significantly different i from those performed previously and are not described in the Safety ' l Analysis Report, including a summary of the analyses leading to the l' conclusions that no unreviewed safety questions were involved and that 10 CFR Part 50.59 was applicable. .,
f e. A summary of the nature and amount of radioactive emuents released or discharged to environs beyond the effective control of the owner-1 o >erator as determined at or before the point of such release or i d scharge. 'the summary shallinclude to the extent practicable an
+
estimate ofindividual radionuclides present in the emuent. If the !
estimated average release after dilution or diffusion is less than 25 i percent of the concentration allowed or recommended, only a statement to this effect need be pn:sented. l
- f. A summarized result of environmental surveys performed outside the
- facility. (
. 6.6.2 Special Reports 2 Special reports ait used to report unplanned events as well as >lanr.ed major facility and administrative changes. These special reports sha i contain and shall be communicated as follows:- l 1
- a. There shall be a report no later than the following working day by !
telephone to the Operations Center USNRC, Washington, DC 20555, to - !
be followed by a written report to the Document Control Desk, USNRC, L Washington, DC 20555, that describes the circumstances of the event ,
within 14 days of any of the following.
- 1) Violation of safety limits (Sec 6.5.1)
- 2) Release of radioactivity from the site above allowed limits (See 6.5.2)
L: -
- 3) - A reportable occurrence (Section 1.1.34) l ;
- 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 12 personnel. ;
- 2) Significant changes in the transient or accident analysis as described in the Safety Analysis Report.
6.7 Reconk
- To fulfill the requirements of applicable regulations, records and logs shall be prepared, and retained for the fol lowing items:
6.7,1 Reards to he Retnined for nr t #nt Five Years -
l a leg of reactor operation and summary of energy produced or hours the reactor was critical.
-' b. Checks and calibrations proceds:e file. _ _
o Proposed Ar-
- nt No. 32 (1/16h8) i
~ .
- - . - . ~ , . . . - - - - , - . . . - - - . . . , - - _ _ . . - ~ ~ . .
46
- c. Preventive and corrective electronic maintenance log.
- d. Major changes in the reactor facility and procedures.
- c. Experiment authorization file including conclusions that no unreviewed safety questions were involved for new tests or experimeats. ;
- f. Event evaluation famis (including unscheduled shutdowns) and reportable occurrence reports. ,
- g. Preventive and corrective maintenance records of associated reactor equipment,
- h. Facility radiation and contamination surveys.
- 1. Fuelinventories and transfers.
J. Surveillance activities as required by the Technical Specifications.
- k. Records of PSRSC re /lews and audits.
6.7.2 Records to be Retnined for at I ramt One Training 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 envimns.
- d. Drawings of the reactor facility including changes.
J f'
Nposed Amendment No.32 (1/16N8) -
(
" - - - - , - - --.m..
,_ ,