ML20203L317

Proposed TS Changes to Amend 32 Dtd 980223.Revision Includes Changes Confined to Sections 3.4 & 3.5 & Repagination of Pages 12-22 & 42-44

ML20203L317
Person / Time
Site:	Pennsylvania State University
Issue date:	02/27/1998
From:	PENNSYLVANIA STATE UNIV., UNIVERSITY PARK, PA
To:
Shared Package
ML20203L306	List: ML20203L302 ML20203L317
References
NUDOCS 9803050442
Download: ML20203L317 (50)

Text

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TECIINICAL SPECIFICATIONS FOR TIIE PENN STATE BREAZEALE I!EACTOR (PSBR)

FACILITY LICENSE NO. R-2 TABLE OF CONTE!4TS 1.0 I NT R O D U CTI O N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .I . .

1.1 D e fi n i ti o n s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .I. . .

1.1.1 A 1. A R A ............................... ............................... ..... I 1.1.2 A uto matic Con trol . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 l 1.1.3 Channel..................................................................... I 1.1,4 Ch an nel Calibration . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . , . . . . . . . . . I 1.1.5 C h an ne l Ch ec k . . . . . . . . . . . . , , . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . I 1.1.6 C h a r. nc l Te s t . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . , . . . . . . . I 1.1.7 Col d Cri t i c al . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . I 1.1.8 Co n fi ne m e n t . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.

1.1.9 Ex c e ss Peac t i vi ty . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 1.1.10 Ex pe ri me n t . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 1.1.11 Ex perimental Facility . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . , . . . . . . . . . . . . . . . . . . 2 1.1.12 Instmmen te d Eleme nt . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 1.1.13 Limiti ng Corditions for Operation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 1.1.14 Limiting Safety System Setting . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 1.1.15 Man u a l Con tro! . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 1.1.16 Maximum Elemental Power Density. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 1.1.I7 M axi mum Power Le vel . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 1.1.18 Meas ured Val ue . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . , 3 1.1.19 Movable Ex peri me r t . . . . . . . . . . . . . . . . . . . _ . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 1.1.20 N ormali ze d Powe r . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 1.1.21 Operable.................................................................... 3 1.1.22 O pe rati n g . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 1.1.23 Pube M ode . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 1.1.24 R eacti vi ty Li mi t s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 1.1.25 Reacti vity Worth of an Experiment . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 1.1.26 Reactor Control Syste m . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 1.1.27 Reac t or In t e rloc k . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 1.1.28 Reac tor Op: rat i n g . . . . . . . . . . . . . . , . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 1.1.29 R eac t or Sec ured . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 1.1.30 Re ac to r S hu t dow n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 1.1.31 Reac tor S afety S ystem . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 1.1.32 Reference Core Condition . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 1.1.33 Researc h Re ac t or . . . . . . . . . . . . . . . . . . . . . . . , . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 1.1.34 Reponable Occurrence ....... ........ ................................... 5 1.1.35 Rod-Trans i e n t . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 1.1.36 S a fe t y Li mi t . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 1.1.37 S C R AM Ti me . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 1.1.38 Se c u red Ex pe ri me n t . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 1.1.39 Secured Experiment with Movable Parts ... .................. ...... . 6 1.1.40 S h all , Should and May . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 1.1.41 Shim, Regulating, Safety Rode ........... ......................... 6 1.1.42 Shutdown Margin . . . . . . . . . . . ...... ... .......................... 6 1.1.43 Square Wave Mode . . . . . . . . . . . .. ...... .. .. ............. ........ 6 4

NOTAFUAL SEAL PAMELA J. STAUFFER. Notary PubAc State College Boro, Centre 5ty PA My Commission Expires J.dy 2. 001

1. ,3

-- Propmed Amendment No. 32 (2/23/98) fit 3

- l 9803050442 990227 PDR ADOCK 05000005 P PDR -

il 1.1,44 Steady State Power Level . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 1.1.45 TRIGA Fuel Element ..................................................... 7 1.1.46 Watc hd og Ci rc u i t . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 2.0 SAFETY LIMIT AND LIMITING SAFETY SYSTEM SETrlNG ................... 7 2.1 Safety Limit-Fuel Element Temperature . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 2.2 Limiting Safety System Setting (LSS S) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 3.0 LIMITING CONDITIONS FOR OPERATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 l 3.1 Reactor Core Parameters . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 ........

3.1.1 Non Pulse M ode Operation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9

] 3.1.2 3.1.3 Reactivity Li mi tation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

S h u t dow n M argi n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

10 10 3.1.4 Pulse Mode Operation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ........... ...... I1 3.1.5 Core Configuration Limitation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . I1 3.1.6 TRI G A Fuel Elemen ts . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .12 .....

3.2 Reactor Control and Reactor Safety System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 3.2.1 Reac tor Control Rods . . . . . . . . . . . . . . . . . . . . . . . . . . . . , , . . . . . . . . . . . . . . . . . . . . . . 13 3.2.2 Manual Control and A utomatic Control . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14 3.2.3 Reactor Con trol Sy ste m . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14 3.2.4 Reactor Safety System and Reactor Interlocks . . . . . . . . . . . . . . . . . . . . . . . . . 15 3.2.5 Core Loadi ng and Unloading Operation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17 3.2.6 S C R AM Ti me . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 18 .......

3.3 Cool ant S ys te m . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . I8 3.3.1 Coolant Level Li mits . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .18 .....

3.3.2 Detection of Leak or Loss of Coolant . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .19 3.3.3 Fission Product Activity . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19 3.3.4 Pool Water Supply for Leak Protection ................................ 20 3.3.5 Coolant Conductivity Limits . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20 3.3.6 Coolant Temperature Limits . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21 3.4 Co n fi n e m e n t . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ........ . . . . . . . . . 21 3.5 Engineered Safety Features - Facility Exhaust System and Eme rge ncy Exhau st Syste m . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 22 3.6 Radiation Monitoring System ..... ... .. .............. .. .. ....... .... ...... 23 3.6.I Radiation Monitoring Information .... ... ... . ..... . . ... . ... 23 3.6.2 Evacuation Alami . . . . . . . . . . . . . . . . . . . . . . . . . . . .. ...... ............ 23 3.6.3 Argon-41 Discharge Limit .. . ..... .... . .... . ... ....... ... 24 3.6.4 A L A R A .. ..... ..... ...... ........ .. .............. . .. . .......... 24 3.7 Limitations c f Experiments . . . . . . . . . . . . . . . . . . . . . . . . . . .... .... ... . ... 25 4.0 SURVEILLANCE REQUIREMENTS .... . .. ..... ......... . . .. ..... 27 4.1 Reactor Parameters . . . . . . . . . . . . . . ... ... .. ........... ..... . ...... .... 27 4.1.1 Reactor Power Calibration. . . . . . . . . . . . . . . .. . . . . . . .... ... . 27 4.1.2 Reactor Excess Reactivity .... . . . . .. .... . .. .. 27 4.1.3 TRIGA Fuel Elements . . . . . . . . . . . . ..... . ... . ......... 28 Proposed Amendment No. 32 (2/23/98)

111 4

4.2 P eactor Control and S afety System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 29 4.2.1 R e acti vi ty Worth . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 29 4.2.2 Reactivity Insertion Rate . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 29 4.2.3 Reac tor Safe ty Sy ste m . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 30 4.2.4 R eac tor I n te rl oc ks . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 31 4.2.5 O ve rpowe r S CR A M . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 31 4.2.6 Tran sie n t R od Test . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 32 4.3 Cool an t S y s t e m . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 32

$ 4.3.1 Fi re Ho se I n s pec t i o n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 32 4.3.2 Pool Water Te mperature . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 33 4.3.3 Pool Wate r Conductivity . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 33 4.3.4 Pool Water Ix vel A larm . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 34 4.4 Co n fm' e me n t . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 34 4.5 Facility Exhaust System and Emergency Exhaust System ....................... 34 4.6 Radiation Monitoring System and Effluents . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 35 4.6.1 Radiation Monitering System and Evacuation Alarm .................. 35 4.6.2 Argon-41................................................................... 35 4.6.3 ALARA...................................................................... 36 4.7 Ex pe ri m e n t s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 36 5.0 D ES I G N FEATU R ES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 37 5.1 R e ac t o r Fu e l . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 37 5.2 R e ac t o r Co re . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 37 5.3 Co n t rol R o d s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 37 5.4 Fu e l S t o ra g e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 38 5.5 Reactor Bay and Exhaust Systems . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 38 5.6 React or Pool Water Systems . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 38 6.0 ADMINISTRATI V E CONTROLS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 38 6.1 O rg a nizati o n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 38 6.1.1 S t ru e t u re . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 38 6.1.2 R e spon sibili ty . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 40 6.1.3 Staffing..................................................................... 40 6.1.4 Selection and Training of Personnel . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 41 6.2 Re vi e w and A udi t . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 41 6.2.1 Safeguards Committee Composition . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 41 6.2.2 Ch arte r a nd R ul es . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 41 6.2.3 Re vi ew Func ti on . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 41 6.2.4 A udi t . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .................................. 42 6.3 Ope rati n g Proce du res . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 42 6.4 Review and Approval of Experiments , . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 43 6.5 Required Action ..... ..... . ................ ... ..... .......... . . .. .. .... 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 (2/23/98)

1 iv 6.6 Repons............................................................................... 44 l 6.6.1 Ope rati n g Re pons . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 44 6.6.2 S peci al R e po ns . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

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 5.7.3 Records To Be Retained for the Life of the Reactor Facility .......... 46 Proposed Amendment No. 32 (2/23/98) l I

1 TECilNICAL SPECIFICATIONS FOR Tile 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 S wcifications 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) progrc .'s a program for maintaining occupational caposures to radiation and reic.ise of radioactive effluents to the environs as low as reasonably achievable.

1.1.2 Automatic Control l

Automatic control mode operation is when normal reactor operations, l 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 Gannd 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 channei calibration h an adjustment of the channel such that its output responds, with acceptable range, and accuracy, to known values of the parameter which the channel measures. Calibration shall encompass the entire channel, including equipment actuation, alarm, or trip, and shall be deemed to include a Channel Test.

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

1.1.6 GanD:1 Test A channel test is the introduction of a signal into the channel to verify that it is operable.

1.1.7 Cold Critiqal Cold critical is the condition of the reactor when it is critical with the fuel and bulk water temperatures both below 100*F (37.8'C).

Pmposed Amendment No. 32 (2/23/98)

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2 1.1.8 Confinctnc.nl Confinement means an enclosure on the overall facility which controls the movement of air into it and out through a controlled path.

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

1.1.10 Exneriment l Expeiiment shall mean (a) any apparatus, device, or material which is not a normal , art of the core or expenmental facilities, but which is inserted in i these faci ities or is in line with a beam of radiation originating from the reactor core; or (b) any operation designed to mea.sure reactor parameters or characteristics.

1.1.11 Exnerimental Facility 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 instmmented element is a TRIGA fuel element in which sheathed chromel-alumel or equivalent thermocouples are embedded in the fuel.

1.1.13 Limiting Conditions for Opgrallon - l Limiting conditions for operation of the reactor are those constraints included in the Technical Specifications that are required for safe operation of the facility. These limiting conditions are applicable only-when the reactor is operating unless otherwise specified.

1.1.14 Limiting Safety 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 by the operator adjusting the control rod positions.

Proposed Amendment No. 32 (2/23/98)

3 1.1.16 Maximum Hemental Power Density l The maximum elemental power density (MEPD)is the power density of the element in the core producing more power than any other element in that loading. The power density of an e:ement 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 definition is only applicable for non. pulse operation.

1.1.17 Maximum Power Level l Maximum Power Lesel is the meximum measured value of reactor power for non-pulse operation.

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

1 1.1.19 Movable Experiment 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 Nonnalized Pown 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 Operable l Operable means a component or system is capable of performing its intended function.

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

1.1.23 Pulse Mode 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 Reactivity ! [ its l The reactiv;ty limits are those limits imposed on reactor core reactivity.

Quantities are referenced to a reference core condition.

1.1.25 Esactivity Worth of an Exneriment l The reactivity worth of an experiment is the maximum absolute value of the 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 (2/23/98)

4 1.1.26 Reactor Control System l The reactor control systen. 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 a: lowed modes.

1.1.27 Reactor 1nterlock l A reactor interlock is a device which prevents some action, associated with reactor operation, until cenain reactor operation conditions are satisfied.

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

1.1.29 Reactor Secured l

l The reactor is secured when:

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

b. A combination of the following:

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

2) The console key switch is in the off pcsition and the key is removed from the lock, and

3) No work is in progress involving core fuel, core structure, installed control rods, or control rod drives unless they are pbgcally 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 recetcr 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 are designed to initiate automatic reactor protection or to proside information for initiation of manual protective action.

Proposed Amendment No. 32 (2/23/98)

J

I i

5 l

j l.1.32 Reference Core Condition l 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.l.33 kicarch Reactat l A research reactor is deni,ed as a device designed to support a self-sustaining neutron chain reaction for research development, educational, training, or experimental purposes, and whict, may have provisions for the production of radioisotopes.

t 1.1.34 Repx2b!cOct.utrcnce l s

A reportable o.currence is any of the following which occurs during reactor operation:

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

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

c. Failure of a required reactor safety system component which could render the system incapable of perfoiming its intended safety function.

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

c. An observed inadequacy in the implementation of cliher administrative

, or procedural controls which could 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 reacttr fuel, cladding, coolant boundary or containment boundary that could result in exceeding 10 CFR Part 20 exposure criteria.

l

1. t.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.l.36 Safety Limit l

Safety limits are limits on important process variables which are found to be necessary to reasonably protect the integritp of certain physical barricts which guard against the uncontrolled release cf radioactivity. The principal physical banier is the fuel element cladding.

Pro;med Amendment No.32 (2/2N98)

6 1.1.37 SCR Ahi Time l

SCRAM time is the clapsed time betweca reaching a limiting safety system set point and a specified control rod movement.

1.1.38 Secured Exneriment l

A secured ex 3eriment is any experiment, experanentrJ ^icility, or component of an experiment that is held in a satiot.arj s or,ition relative to the reactor by mechanical means. The restrainitts forces .aust tv l substantially greater than those to which the experiment might be '

subjected to by hydraulic, pneumatic, buoyent, or other forces which are normal to the operating environmer t of the experiment, or by forces which can arise as a result of credible malfunctions.

i . l .39 Secured Exneriment with hiovable Pans l A secured experiment with movable parts is one that contains parts that are intended to be moved while the reactor is operating.

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1.1.40 Shall. Should. and hiny l 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. Regulating. and Safety Rods l

A shim, regulating, or safety rod is a control rod having an electric motor drive and SCRAh! capabilities, it has a fueled fo!!awcr section.

1.1.42 Shutdown hiargin l Shutdown margin shall mean the minimum shutdown reactivity necessary to provide confidence that the reactor can be made suberitical by means of the control and safety 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 Sgitare Wave hiode l

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

1.1.44 Steadv State Power Level l

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. hiinor variations about this level may occur due to noise, normal signal variation, and reactivity adjustments. During manual, automatic, or square wave modes of operation, some hitial, momentary overshoot may occur.

lYo;ued Amendment No. 32 '2/23/98)

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 Zrilin stainless steel cladding or 12 wi% U Zrliin stainten steel cladding enriched to less than 20% uranium 235, 1.1.46 Watchdeg Circuit l

A watchdog circuit is a circuit consisting of a timer and a relay. The timer energizes the relay as long as it is reset prior to the expiration of the timing interval if it is not reset within the timmg interval, the relay will de-energize thereby causing a SCRAM.

2.0 SAFI;I Y I.lMIT AND I.lMITING SAFETY SYSTEM SislTING 2.1 Safety Limit Fuel Element Temnerature Applicability The safety limit specification applies to the maximum temperature in the reactor fuel.

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

Snecification l 3 The temperature in a water-cooled TRIGA fuel element shall not exceed i150'C under any operating condition.

Basis The imponant parameter for a TRIGA reactor is the fuel element ten'perature. This parameter is well suited as a single s cification especially since it can be measured at a point within the fuel element anc. the relationship between the measured and actual temperature is wcll characterized analytically. A loss in the integrity of the fuel element claddin ; could arise fr<.m a build up of excessive pressure between the

{

fuel moderator and :ie cladding if the maximum fuel temperature exceeds 1150*C. l The pressure is caused by the presence of air, fission product gases, and hydrogen f rom the dissociation of the hydrogen and zirconium m the fuel modert. tor. The magnitude of this pressure is determined by the fuel moderator temperature, the ratio of hydrogen to zirconium in the alloy, and the rate change in the pressure.

The safety limit for the standard TRIGA fuel is based on data, including the large mass of experimer.tal evidence obtained during high performance reactor tests on this fuel. These data indicate that the stress in the claddin;; due to the increase in the hydrogen pressure from the dissociation of zircon um lydride will remain below the ultimam stress provided that the temperature of the Onci does not exceed il50*C and the fuel cladding is below 500*C. See Safety Analysis Report, Ref.13 and 30 in Section IX and Simnad, M.T., F.C. Foushee, and G.B. Wee " Fuel Elements for Pulsed Reactors," Nucl. Technology, Vol. 28, p. 31 56 Ganuary 1976).

Pmposed Amendment No. 32 (2/23/95) s

1 l

l 8 2.2 L{mitingjgety System Setting (LSSS)

Applicability The !.SSS specification applies to the SCR AM setting which prevents the safety limit from being reached.

OldErikc The objective is to prevent the safety limit (ll50*C) fium being reached.

Specification l

. The limiting safety system setting shall be a maxhnum of 650*C as measured with an instrumented fuel element if it in located in a core position representative of the ,

maximum elemental power density (h1EPD) in that loadin p If it is not practical to j

, locate the instrumented fuel in such a position, the LSSS s 1all be reduced. The l reduction of the LSSS shall be by a ratio based on the calculated linear relationship '

between the normalized pc,wer at the monitored position as compared to normalized power at the core position representative of the h1EPD in that loading.

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

Ex periments and analyses described in the Safety Analysis Report, Sectiort IX -

Salety Evaluation, show that the measured fuel temperature at steady state power has a simple linear relationship to the normalized power of a fuel element in the core. Maximum fuel temperature occurs when un lastrumented element is in a core ,

position of MEPD. The actual location of the instrumented element and the associated LSSS shall be chosen by calculation and/or experiment prior to going to maximum reactor operational power level. The measured fuel temperature during steady state operation is close to the maximum fuel temperature in that element.

Thus,500'C of safety margin exists before the i150'C safety limit is reached. This safety margin provides adequate compensation for variations in the temperature profile of depleted and differently loaded fuel elements (i.e. 8.5 wt% vs.12 w %

fuel elements). See Safety Analysis Report Section IX.

Ir it is not practical to place un 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 ths Nighest 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. Ilowever, the temperature channel will have no elfeet on limiting the peak power or fuel temperature, gencruted, because of its relatively long time constant (seconds),

compared with the width of the pulse (milliseconds).

hopmed Anendnwnt No. 32 (2/23M8) -

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9 3.0 I IMITING CONDITIONS FOR OPERATION

'lhe limiting conditions for operation as set forth in this section are upplicable only when the reactor : s operating. They need not be met when the reactor is shutdown unless specified otherwise.

3.1 Reactor Core ParatDclCD 3.1.1 Non. Pulse hiode Ooeration l Applicability These specifications apply to the power generated during manual control mode, automatic control mode, and square wave mode operations.

Olticclite The objective is to limit the source term and energy production to that used in the Safety Analyt.is Report.

Soccifientions l

a. The reactor may be operated at steady state power levels of 1 Mw (thermal) or less,

b. The maximum power level shall be no greater than 1.1 Mw (thermal). l

c. The steady state fuel temperature shall be a maximum of 650*C as measured with an instrumented fuel element if it is located in a core losition 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 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. In this case, the measured steady state fuel temperature shall be limited such that the calculated steady state fuel tem 3erature at the core 30sition representative of the MEPD in that loating shall not excecc 650'C.

Hahts l

a. Thermal and hydraulic calculations and operational experience indicate that a compact TRIGA reactor core can be safelv operated 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 S AR.

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 Ilypothetical Accident (MilA), See Safety Analysis Report, section IX.

Pmpned Amendment No.32 C/23N8)

10 3.1.2 lintClivity Limitation Applicability 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 is operated within the limits analyzed in the Safet exceeded. y Analysis Report and to ensure that the safety limit will not be Specification The maximum exuc wmty above cold, clean, critical plus samarium poison of the core cor.dsgr hon with experiments and experimental facilities in place shall be 4.9% Ak/k (~$7.00).

Ihtiis Limiting the excess reactivity of the core to 4.9% Ak/k (~$7.00) prevents the fuel temperature in the core from exceeding Il50'C under any assumed accident condition as described in the Safety Analysis lleport,Section IX ,

3.1.3 Shutdown Margin Applicability This specification applies to the reactivity condition of the reactor and the reactivity worth of control rods, ex applies to all modes of operation. periments, and experimental facilities. It 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, Snecification l

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

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

b. The highest reactivity worth control rod fully withdrawn.

Ihtsis A shutdown margin of 0.175% Ak/k (~$0.25) ensures that the reactor can be made suberitical from any operating condition even if the highest worth ._

control rod should remain in the fully withdrawn position. The shutdown margin requirement may be more restrictive than Specification 3.1.2.

Pmposed Amendment No. 32 (2/23/98)

11 3.1.4 Pulse Mode Operation Applicability These specifications apply to the energy generated in the reactor as a result of a pulse insenion of reactivity.

Objecth e The objective is to ensure that the safety limit will not be exceeded during l l pulse mode operation.

Specifications

a. The stepped reactivity insenion for pulse operation shall not exceed 1 2.45% Ak/k (~$3.50) and the maximum worth of the poison section of I the transient rod shall be limited to 2.45% Ak/k (~$3.50).

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

Ihtic

a. Experiments and analyses described in the Safety Analysis Report,Section IX C., show that the peak pulse temperatures can be predicted for new 12 wt% fuel placed in any core position. These experiments and analyses show that the maximum allowed pulse reactivity of 2.45% Ak/k

(~$3.50), prevents the maximum fuel temperature from reaching the safety limit (1150*C) for any core configuration that meets the requirements of 3.1.5.

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

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

3.1.5 Core Configuration Limitation Applicability These specifications apply to all core configurations except as noted. l Objective The objective is to ensure that the safety limit (ll50*C) will not be exceeded due to power peaking effects in the various core configurations.

Propned Amendment No. 32 (2/23/98) s

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, 12 Snecifications

a. The critical core shall be an assembly of either 8.5 wt% U.Zrli stainless steel clad or a mixture of 8.5 wt% and 12 wt% U.Zrl! stainless steel clad TRIGA fuel moderator 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 the core excess, or administratively), the maximum NP may be increased. The maximum NP may be increased above 2.2 as long as the calculated maximum fuel tem scrature does not cxceed the safety limit with that maximum allowed Ise worth and NP. In addition, the Reactivity Accident in the S fety Analpis Report shall be evaluated to ensure that the safety limit is not exceeded with the new conditions (See Safety Analysis Repon,Section IX.).

Bases

a. The safety analysis is based on an assembly of either 8.5 wt% U.Zrli stainless steel clad or a mixture of 8.5 wt% and 12 wt% U.Zrli stainless steci 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 03eration places an upper bound on the elemental heat pnx!uction anc the source term of the PSBR to that used in the analysis of a Loss of Coolant Accident (LOCA) and Maximum liypothetical Accident (MilA) respectively. See Safety Analysis Repon,Section IX.

c. The maximum NP for a given core loading determines the peak pulse temperature with the maximum albwed pulse wonh. If the maxlmum 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 acce documented by an administratively approved procedure.pted met 3.1.6 TRIGA Fuel Elements Applicability These specifications apply to the mechanical condition of the fuel. l Obiective The objective is to ensure that the reactor is not operated with damaged fuel l that might allow release of fission products.

Pmpowd Amendment No. 32 (2/23/98)

13 S.pccifictllons l The rer, tor shall not be operated with damaged fuel except to detect and identify the fuel element for removal. A TRIGA fuel 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 clongation,its length exceeds its orig lnal length by 0.125 inch,

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

Bases

a. The limit of transverse bend has been shown to result in no difficulty in disassembling the core. Analysis of the removal of heat from touching ,

fuel elements shows that 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 Centigrade below the safety limit (ll50*C), and (2) during a pulse, the cladding temperatures remain well below their stress limit. The clongation limit has been s ycined to ensure that the cladding material will not be subjected to stralns that could cause a loss of fuel integrity and to ensure adequate coolant flow 3.2 Reactor Control and Reactor Safety System l 3.2.1 Reactor Control Rods Applicability This specification applies to the reactor control rods.

Obiective The objective is to ensure that sufficient control rods are operrble to l maintain the reactor suberitical.

Specification There shall be a minimum of three cperable control rods in the reactor core, lhtill The shutdown margin and excess reactivity specifications rec uire that the reru tor can be made suberitical with the mest reactive contro rod withdrawn. This specification helps ensure it. l Pmpowd Amendment No.32 (2/23M8)

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14 3.2.2 Manual Control and Automatic Control l Applicability This specification applies to the maximum reactivity insertion rate associated with movement of a standard control rod out of the core.

Objective The oljective is to ensure that adequate control of the reactor can bc l mainttuned during manual and 1,2, or 3 rod automatic control.

Specification The rate of teactivity insertion associated with movement of either the regulating, shim, or safety control rod shall be not greater than 0.63% Ak/k

(~$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 reactivity of those rods shall be not greater than 0.63% Ak/k (~$0.90) per second when averaged over full travel.

IhtS15 The ramp accident analysis (refer to Safety Analysis Report, 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 Control SystCm Applicability This specification applies to the information which must be available to the reactor operator during reactor operation.

Objective The objective is to require that sufficient information is available to the operator to ensure safe operation of the reactor. l Specification The reactor shall not be operated unless the measuring channels listed in Table 1 are operable. (Note that MN, AU, and SW are abbreviations for manual control mode, automatic control mode, and square wave mode, respectively).

Propowd Amendment No.32 (2/23/98)

15 Table 1 Measuring Channels Min. No. Effective Mode Measuring Channel Operable MN. AU & SW Eglg Fuel Element Temperature I X X Wide Range Instrument Linear Power i X Log Power i X Reactor Period /SUR 1 X Power Range Instrument Linear Power 1 X Pulse Peak Power i X lhlsh Fuel temperature displayed at the control console gives continuous infonnation on this parameter which has a specified safety limit The power level monitors ensure that the reactor power level is adequately monitored 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 assurance that the reactor l is operated at all times within the limits allowed by these Technical

Specifications.

! 3.2.4 Reactor Safety System and Reactor Interlocks l

Applicability This specification applies to the reactor safety system channels, the reactor interlocks, and the watchdog circuit.

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

Bases

a. A temperature SCRAM and two power level SCRAMS ensure the 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).

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i Proposed Amendment No. 32 (F23/98)

16 l

Table 2a i

hiinitr.um Reactor Safety l

System Channels Number liffective hiode Chrad Dunbh EuEtica hiN. AU l'ula MY 1 uel Temperature i SCRAhi s 650'C' X X X liigh Power 2 SCRAh! s 110'I of maximum X X reactor operational power not to excred 1.1 Mw Detector Power Supply 1 SCRAh! on failure of supply X X voltage SCR Aht ilar on i hianual SCRAh! X X X Console Picset Timer i Transient Rod SCRAh! 15 X seconds orless after pulse Watchdog Circuit 1 SCRAhi on sofIware or self- X X X check failure

• The limit of 650*C shall be reduced as required by specification 2.2.

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Table 2b hiinimum Reactor Interlocks Number liffective blode Channel Operable Punction hiN. AU Eula SY Source level i Prevent rod withdrawal X without a ricutron induced.

signal on the los power channel Pulse hiode inhibit 1 Prevent pulsing from levels X above i kw Transient Rod 1 Prevent applications of air X unless cylinder is fully inserted Shim, Safety, and 1 Prevent movement of any X Regulating Rod rod except the transient rod Simultaneous Rod 1 Prevent simultaneous manut.1 X X Withdrawal withdrawal of Iwo rods Proposed Amendment No. 32 01:3/98) l

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b. The maximum reactor operational power may be administratively limited to less than 1 Mw depending on section 3.1.5.b of this Technical Specification. The high power SCR AMs shall be set to no more than ,

1 l0% of the administratively limited maximum reactor operational 1 power if it is less than 1 Mw.

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

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

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

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

g. The interlock to prevent startup of the reactor without a neutron induced signal ensures that sufficient neutrons are available for proper startup in all allowable modes of operation,

h. The interlock to prevent the initiation of a pu!se above 1 kw is to ensure that fuel temperature is approximately poo. 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 rod unless the cylinder is fully inserted is to prevent pulsing the reactor in the manual control or automatic control mode.

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

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

3.2.5 core Loading and Unloading _ Operation Applicability This specification applies to the source level interlock. l Obiective The objective of this specification is to a!!ow bypass of the source level interlock during operations with a suberitical core.

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

Propned Amendmem No.32 (2/23/98)

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18 BAih During core loading and unloading, the reactor is suberitical. Thus, momentarily defeating the source level interlock is a safe operation.

Should the core become ir advertently supercritical, the ace dental insertion of reactivity wi'4 not allow fuel temperature to exceed the ll50*C safety limit bect.use no single TRIGA fuel element is worth more than 1% Ak/k ( 41.43)in the mo;t reactive core position.

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

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

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

Bash This specification ensures that the reactor will be aromptly shut down when a SCRAM signal is initiated. Experience and ana ysis, Safety Analysis Repon.Section IX, have indicated that for the range of transients anticipated for a TRIGA reactor, the specified SCRAM time is adequate to ensure the safety of 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 limit.

3.3 Coolant System 3.3.1 Coolant Level Limits Anolicability This specification applies to operation of the reactor with respect to a required depth of water above the top of the bottom grid plate.

Objective The objective is to ensure that water is , resent to provide adequate I- personnel shielding and core cooling w len the reactor is operated, and during a LOCA.

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

Pmposed Amendment No. 32 (2/23/98)

- u

19 Ilush When the water is more than approximately 18 ft, above the top of the bottom grid plate, the water provides sufficient shleiding to protect personnel during operation at 1 Mw, and core cooling is achleved with natural circulation of the water through the core. Should the water level drop below approximately while operating at 1 Mw, a low pool 18.25 ft. above the top (see Technicalof the botto level alarm Specifications 3.3.2) will alert the operator who is required by adminidratively a aproved procedure to shutdown the reactor. Once this alarm o; curs it wi I take longer than 1300 seconds before the core is completely uncovered because of a break in the 6" pipe connected to the bottom of the pool Tests and calculations show that, during a LOCA,680 seconds is suflelent decay time after shutdown (see Safety Analysis Report,Section IX ) to prevent the fuel temperature from reaching 950*C. To prevent cladding rupture, the fuel and the cladding tem perature must not exceed 950 *C (it is assumed that the fuel and the cladding are the same temperature during air cooling).

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

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

Specification l A pool level alarm shall be activated and corrective action taken when the i pool level drops 26 cm from a level where the pool is full, lhish The alarm occurs when the water level is approximately 18.25 ft. above the top of the bottom grid plate. The point at which the poolis fullis approximately 19.1 ft. above the top of the bottom grid plate. The reactor staff shall take action to keep the core covered with water according to existing procedures. The alarm is also transmitted to the Police Services annunciato panel which is monitored 24 hrs. a day. The alarm provides a signal that occurs at all times (see Safety Analysis Report, Section Vil).

Thus, the alarm provides time to initiate corrective action before the radiation from the core poses a serious hazard.

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

Objective The objective is to ensure that fibsion products from ilcaking fuel element l are detected to provide opportunity to take protective action.

Proposed Amendment No.32 (2/23/98)

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

IlA515 This unit will be sensitive to airborne radioactive particulate matter containing Hssion products and Ossion gases and will alert personnel in time to take protective action.

3.3.4 Pool Water Supolv for Leak Protection Apolicability This specification applies to pool water supplies for the reactor pool for leak protection.

Objective The objective is to ensure that a supply of water is available to replenish reactor pool water in the event of pool water leakage.

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

Dasis Provisions for both of these supplies are in 31 ace and will supply more than the specified flow rate. This flow rate will x more than sufficient to handle leak rates that have occurred in the past or any anticipated leak that might occur in the future.

3.3.5 Coolant Conductivity Limits l Applicability This specification applies to the conductivity of the water in the pool.

Objectives l The objectives are: l

a. To prevent activated contaminants from 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 microsiemens/cm (5 micrombos/cm).

Pmposed Anwndment No. 32 (2/23/98)

21 lhtsis Experience indicates that 5 microsiemens/em is an acceptable level of water l contaminants in an aluminum / stainless steel system such as that at the PSlW. Based on experience, 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/cm.

3.3.6 Coolant Temnerature Limig Apnlicability This specification applies to the pool water temperature.

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

Specification l An alarm shall annunciat; and corrective action shall be taken if during operatiun the bulk pool water temperature reaches 10(TF (37.8'C).

Basis This s ycification 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.

Objective The objective is to ensure that no large air passages exist to the reactor bay during l reactor operation.

Snecifications The reactor bay truck door shall be closed and the reactor bay personnel doors shall not be blocked open and left unattended if either of the following conditions are true,

a. The reactor is not secured, or

b. Irradiated fuel or a fueled experiment with significant fission product inventory is being moved outside containers, systems or storage areas.

Pmposed Amendment No,32 (2/23/98)

22 Dmb This specification helps to ensure that the air pressure in the reactor bay is lower l than the remainder of the building 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 Enginected Safety Features Facility ExhausLSygem and Emergenev Exhaust i System Applicability This specification applies to the operation of the facility exhaust system and the emergency exhaust system.

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

Snecification

u. If the reactor is not secured, at least one facility exhaust fan shall be operating and, 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 /> during maintenance or repair, the emergency exhaust system shall be operable,

b. Ifirradiated fuel or a fueled experiment with significant fission product inventory is being moved outside containers, systems or storage areas, at least one facility exhaust fan shall be operating and the emergency exhaust system shall be operable.

13ases l P. iring normal operation, the concentration of airborne radioactivity in unrestricted areas is below effluent release limits as described in the Safety Analysis Report,Section IX. In the event of a substantial release of airborne radioactivity, an air radiation monitor and/or un area radiation monitor will sound a building evacuation alarm which will automatically cause the facility exhaust system to close and the exhausted air to be passed through the emergency exhaust system filters before release. This reduces the radiation within the building. The filters will remove

= 90% all of the particulate fission products that escape to the atmosphere.

The emergency exhaust system activities only during an evacuation where upon all personnel are required to evacuate the building (section 3.6.2), if there is an evacuation while the emergency exhaust system is out of service for maintenance or repair, percannel evacuation is not prevented.

Personnel dose to the public will be equivalent or less whether or not the emergency exhaust system functions in the unlikely event an accident occurs during the maintenance or repair.

Propmed Amendment No. 32 (2/23/98)

23 3.6 Radiation hionitoring System 3.6.1 ((adiation hionitoring Information Applicability This specification applies to the radiation monitoring information which must be available to the reactor operatos during reactor operation, l

Objective 1 The objective is to ensure that sufficient radiation monitoring information is available to the operator to ensure personr.r i radiation safety during reactor operation.

Specification J

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

Table 3 Radiation hionitoring Channels Radiation hionitorine ~

Channels Function Number Area Radiation hionitor hionitor radiation levels 1 in the reactor bay.

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

Heamhole Laboratory hionitor radiation in the I hionitor Beamhole Laboratory required only when the laboratory is in use.

llairs

a. The radiation monitors provide information to operating personnel of l any impending or existing danger from radiation so that there will be sulficient time to evacuate the facility and to take the necessary steps to control the spread of radioactivity to the surroundings,

b. The area radiation monitorin 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 Aoplicability This specification applies to the evacuation alarm.

Pmposed Amendment No. 32 (2/23/98)

24 DWcc1hc The objective is to ensure that all personnel are alerted to evacuate the l PSBR building when a potential radiation hazard exists within this building.

Specification The reactor shall not be operated unless the evacuation alarm is operable and audible to personnel within the PSUR bunding when activated by the radiation monitoring channels in Table 3 or a manual switch.

11A111 The evacuation alarm produces a loud pulsating sound throughout the PSBR building when there is any impending or existin dan ;cr from radiation.

The sound notifies all personnel within the PSHg3 dingbul to evacuate the building as prescribed by the PSBR emergency procedure.

3.6.3 Argon-41 Discharge Limit Applicability This specification applies to the concentration of Argon-41 that may be discharged from the PSBR.

Objective 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.

Specification All Argon-41 concentrations produced by the operation of the reactor shall l be below the limits imposed by 10 CFR Pan 20 when averaged over a year.

Ilasis 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 i 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 EnvironmentalImpact Appraisal, December 12,1996),

3.6.4 As low As Reasonably Achievable (ALARA)

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

Oh!rc1hc The objective is to maintain all exposures to radiation and release of radioactive effluents to the environs ALARA.

Proposed Anwndnwn No.32 (2/23/98)

_ , _ . ~ _ . .

25 SPIdf1 in An ALARA program shall be in efrect, llash llaving an ALARA program will ensure that occupational e xposures to radiation and the release of radioactive effluents to the environs will be ALARA. Itaving such a formal program will keep the staff cognizant of the i importance to minimize radiation exposures and efnuent releases.

3.7 Limitations of Experiments-1 Apnlicabilltv l

These specifications apply to experiments installed in the reactor and its '

experimental facilities.

Obiective The objective is to prevent damage to the reactor and to minimize release of l

radioactive materia.s in the event of an experiment failure, Specifications The reactor shall not be operated unless the following conditions governirig experiments exist:

a. The reactivity of a movable experiment and/or movable ponions of a secured experiment plus the maximum allowed pulse reactivity s ball be less than 2.45%

Ak/k (~$3.50). Ilowever, the reactivity of a movable experiment and/or movabla ponions of a secured experiment shall have a reactivity worth less than 1.4% Ak/t:

(~$2,00). When a movable experiment is used, the maximum allowed pulse shall l be reduced below the allowed pulse reactivity insenion of 2,45% Ak/k (~$3.50) to ensure that the sum is less 2.45% Ak/k (~$3.50).

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

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

c. When the keff of the core is less than I with all control rods at their upper limit and no ex criments 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 i

conjunction with the reactor provided its use does not constitute an unreviewed

(

i safety question. The failure mechanisms that shall be analyzed include, but are not hmited to corrosion, overheating, impact from projectiles, chemical, and mechanical explosions.

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

Pronned Amendment No 32 (2/23/98)

26 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 performed to determine the consequences and the need for corrective action.

The results of the ins petion and any corrective action taken shall be reviewed by the Director or e designated alternate and determined to be satisfactory before operation of the reactor is resumed.  ; )

c. Experiment materials, except fuel materials, which could off gas, sublime, volatilize, or produce acrosals under (1) norma! operatin ; conditions of the cxperiment and reactor,(2) credible accident conditions n the reactor, or (3) possib!c 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.

I When calculating activity limits, the following assumptions shall be used:

1) If an ex 3eriment fails and releases radioactive gases or aerosols to the reactor lay or atmosphere,100% of the gases cr acrosols escape.

2) If the effluent from an ex 3erimental facility exhausts thmugh 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 effluant from an experimentul facility exhausts through a filter installation designed for greater than 99% efficienc particles, at kast 10% of these vapors can escape, y for 0.3 micron i 4) For materials whose boiling point is above 130'F and where vapors formed by boiling this material can escape only through an undisturbed column of water above the core, at least 10% of these vapors can escape.

f. Each fueled experiment shall be controlled such that the total inventory of iodine isotopes 131 through 135 in the experiment is no greater than 1.5 curies.

In addidon, any fueled experiment which would generate an inventory of more than 5 millieuries (mci) ef I 131 through I 135 shall be reviewed to ensure thht in the case of an accident, the total release of io line will not exceed that postulated for the MilA (se Safety Analysis Repon,Section IX ).

Bases

a. This specification limits the sura of the reactivities of a maximum aVawed pulse and a movable experiment to the specified maximum reactivity of the umsiere rod. This limits the effects of a pulse simaltaneous with the failure of the rnovable experiment to the effects analyzed for a 2.45% Ak/k (-$3.50) pulse. In addition, the maximum power attainable with the ramp insertion of 1.4% Ak/k

(-$2.00)is less than 500 kw starting frorn critical,

b. The maximum worth of all experiments is limited to 2.45% Ak/k (~$3.50) so that their inadver1cnt sudden removal from 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 worth 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 inte;ted by a ramp increase, the maximum power attainable is less than 1 Mw.

Pmined Amendment No 32 (2/23/98)

27

c. Since the initial core is suberitical, adding and then inadvertently removing all negative reactivity experiments leaves the core in its initial suberitical condition.

d. The design basis accident is the MilA (See Safety Analysis Re, ort,Section IX).

A chemical explosion (such as detonated TNT) or a mechanica , explosion (such as a steam exp osion 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 refcase of fission products or loss of reactor shutdown capability. Accelerated I corrosion of the fuel cladding due to material released by a failed experiment could also lead to release of fission products.

if an experiment failure occurs a special investigation is required to ensure that all effects from the failure are known before operation proceeds.

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

- to the atmosphere outside the facility boundary.

l

f. The 5 mci limitation on 1 131 through 1 135 ensures that in the event of failure l of a fueled experiment, the exposure dose at the exclusion area boundary will be less than that postulated for the MilA (See Safety Analysis Report,Section IX) l cven if the iodine is released in the air.

4.0 SURVEILI ANCE REQUIREMENTS 4.1 Reactor Parameters 4.1.1 Reactor Power Calibration Applicability This specification applies to the surveillance of the reactor power calibration.

Oblective The objective is to verify the performance and operability of the power measuring channel.

Soccification l

A thermal power channel calibration shall be made on the linear power level monitoring channel annually, not to exceed 15 months, llash The thermal power level channel calibration will ensure that the reactor is l

operated at the authorized power levels.

4.1.2 Reactor Excess Reactivity Anolicability This specilihation applies to surveillance of core excess reactivityi Progned Amendment No,32 (2/23/95)

28 Objective The objective is to ensure that the reactor excess reactivity does not execed the Technical Specifications and the limit analyzed in Safety Analysis Repon,Section IX.F.

Specification The excess reactivity of the core shall be measured annually, not to exceed

, 15 months, and following core or control rod changes equal to or greater .

than 0.7% Ak/k (~$1.00).

lhtsia Excess reactivity measurements on this schedule ensure that no unexpected I 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 Apolicability This specification applies to the surveillance requirements for the TRIGA fuel elements.

Objective The objective is to verify the continuing integrity of the fuel element cladding.

Spreification Fuel elements and control rods with fuel followers shali be inspected visually for dama ;e or deterioration and measured for length and bend in accordance with tie following:

a, Before being placed in the core for the first time or before retum 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.

Basis The frequency of inspection and measurement schedule is based on the ,

parameters most like ,y to affect the fuel cladding of a pulsing reactor operated at moderate aulsing levels and utilizing fuel el-ments whose characteristics are we i known.

I'ropned Amendment No. 32 (2/23d8)

29 4,2 Reactor Control and Safety System 4.2.1 Reactivity Wor $

Applicability This specification applies to the reactivity wonh of the control rods.

Obiective l l

The objective is to ensure that the control ods are capable of malataining l the reactor suberitical.

Snecification The reactivity worth of each control rod and the shutdown margin for the core loading in use shall be determined annually, not to exceed 15 months, or following core or control rod changes equal to or greater than 0.7% Ak/k

(-$ 1,00),

lush 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 Insenion Rate Applicability This specification applies to control rod movement speed, Obiective The objective is to ensure that the reactivity addition rate specification is not l violated and that the control rod drives are functioning.

Enecification The rod drive speed both up and down and the time from SCRAAi 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.

Ihsh This specification ensures that the reactor will be aromptly shut down when l a SCRAM signalis initiated. Experience and una ysis have indicated that

, for the range of transients anticipated for a TRIGA 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 Amendrnem No. 32 (2/23/98)

30 4.2.3 Reactor Safety System l

Applicability l The specifications apply to the surveillance requirements for rneasurements, channel tests, and channel checks of the reactor safety systems and watchdog circuit.

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

Specifications

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

b. A channel test of the detector power supply SCRAM functions for both l

the wide rance and the power range and the watchdog circuit shall bc l performed annually, not to exceed 15 months,

c. Channel checks for operability shall be performed daily on fuel element temperatua wide range linear power, wide range log power, wide range reactor periCUR, and power raupe linear power when th:: reactor is to be operated, or prior to each operation that extends moic than one day,

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

e. The pulse peak power channel shall be compared to the fuel temperature each ti,me the reactor is pulsed, to ensure propei peak power channel l operation.

Bases System co.mponents 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 system works, based on past expenence as recorded in the operation log book. An annual channel test of the watchdog circuit is sufficient to ensure 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.

Propeyd Amendnwnt No. 32 (2/23/98)

31 l

c. Comparison of the peak pulse power to the fuel temperature for each pulse will ensure the detection of possible channel drift or deterioration  ;

of its operational characteristics.

4.2.4 Reactor Interlocks Applicability These specifications apply to the surveillance requirements for the reactor

• control system interlocks.

Objective The objective is to ensure performance and operability of the reactor control l system interlocks.

Specifications i

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 is greater than the setpoint when the source is removed from the core.

not to exceed 71/2

b. A channel on thetest shall modebe performed semi annually,h prevents puls 4 months, pulse inhibit interlock whic power levels higher than one kilowatt.

c. A channel check shall be performed semi annually, not to exceed 71/2 months, on the transient rod interlock which revents ap to the transient rod unless the cylinderis fu!! inserted. plication of

d. A channel check shall be performe<1 semi annually, not to exceed 71/2 months, on the rod drive interlock which prevents movement of any rod 3

except the transient rod in pulse mode.

, c. A channel check shall be performed semi annually, not to exceed 71/2 months, on the rod drive interlock which prevents simultaneous manual withdrawal of more than one rod.

Eu l The channel test and checks will vetify operation of the reactor interlock system. Expericcce at the pSBR indicates that the prescribed frequency is adequate to ensure operability. l After extended operation, the photo neutron source strength may be high enough that removing the source may not drop the neutron signal below the e

setpoint of the source 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 Overoower SCRAM Apylicability This specification applies to the high power and fuel temperature SCRAM channels.

Proposed A.nendment No. M (2/D/94

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

Soccification The high power and fuel temperature SCRAM's shall be tested annually, not

.a exceed 15 months.

13 asis Experience with the PSilR for more than a decade, as recorded in the operatien log books, indicates that this interval is adequate to ensure operability.

4.2.6 Transient Rod Test Applicability These specifications apply to surveillance of the transient rod mechanism. l Objective The objective is to ensure that the transient rod drive mechanism is l maintained in an operable condition.

Specifications l

a. On each day that pulse mode operation of the reactor is planned, a functional perfonnance 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 annually, not to exceed 15 months,

b. The reactor shall be pulr. d annually, not to exceed 15 months, to compare fuel temperature measurements and peak power levels with those of previous ?ulses of the same reactivity value or the reactor shall not be pulsed unti such comparative pulse measurements are performed.

Dash 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 Coolant System 4.3.1 Fire llose insocetion Applicability This specifier.tlon applies to the dedicated fire hoses used to supply water to l the pool in an err.crgency.

Proposed Anendment No. 32 (2/23/98)

33 Objective E

The objective is to ensure that these hoses are operable. l Specification The two (2) dedicated fire hoses that 3rovide supply water to the poolin an h emergency shall be visually inspectec for damage and wear annually, not to exceed 15 months.

Basis This frequency is adequate to ensure that significant degradation has not l occurred since the previcus inspection.

E a,-

J.3.2 601 Water Temnerature Applicabihty i

~

This specification applies to pool water temperature.

Objective The objective is to limit pool water temperau.re, k Snecification The pool temperature alarm shall be calibrated annually, not to exceed 15 months.

hs Experience has shown this instrument to be drift-free and that this interval is adequate to ensure operability. l 4.3.3 hol WatetConductivity Annlicability This specification applies to surveillance of pool water conductivity.

Objective The objective is to ensure tha; pool water mineral content is maintained at l an acceptable level.

Specification Pool wate; 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 interval.; provides acceptable surveillance of limits that ensure that fuel clad corrosion and neutron l activation of dissolved materials will not occur.

Proixned Amendment No.32 (2/23/98) h . . _ _ _ _ . .

34 4.3.4 Pool Water Izvel Alarm Applicability This specification applies to the surveillance requirements 'o the pool . ,vei alann.

Objective The objective is to verify the operability of the pool water level alarm. l Specification The pool water level alarm shall be channel checked monthly, not to exceed 6 weeks, to ensure its operability, Dillis Experience, as exhibited by past periodic checks, has shown that monthly checks of the pool water level alarm ensures operability of the system l during the month.

4.4 Confinemeji Anplicability This specification applies to reactor bay doors.

Objective The objective is to ensure that reactor bay doors are kept closed .: ner Specification J 3.4.

Snecification Doors to the reactor bay shall be locked or under supervision by an authorized keyholder.

Basis A keyholder is authorized by the Director or his designee. l 4.5 Facility Exhaust _ System and Emergency Exhaust System Annlicability These specifications apply to the facility v.haust system and emergency exhaust system.

Obiective The objective is to ensure the proper operation of the facility exhaust system and l emergency exhaust system in controllir.g releaset of radioactive material to the uncontrolled environment.

Proposed AmendMw do. 32 (2/23/98)

35 Snecifications l

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

b. It shall be verified monthly, not to exceed 6 weeks, whenever operation is scheduled, that the facility exhaust system is secured when the emergency exhaust system activates during an evacuatior, larm (See Technical Specificationn 3.6.2 and 5.5).

Basis Experience, based on periodic checks performed over 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. This pro' ides l reasonable assurance on the control of the release of radioactive material.

4.6 Radiation Monitoring System and Effluents 4.6.1 Radiation Monitoring System and Evacuation Alarm Apolicability This specification applies to surveillance requirements for the area radiation monitor, the Beamhole Laboratory radiation monitor, the air radiation monitor, and the evacuation alami.

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. They shall be verified to be operable by a channel check daily when the reactor is to be operated, and shall be calibrated annually, not to exceed 15 months.

Basis Experience has shown this frec uency of verification of the radiation monitor set points and operability and tae evacuation ahrm operability is adequate to correct for any variation in the system due to a change of operating characteristics. Annual channel calibration ensures that units are within the l specifications defined by procedures.

4.6.2 Argon-41 Anolicability This specification applies to surveillance of the Argon-41 produced during reactor operation.

Proposed Amendment No. 32 (2/23/98) l 1

36 Objective To ensure that the production of Argon-41 does not e. meed the limits l specified by 10 CFR Part 20.

Sprification The production of Argon 41 shall be meatured and/or calculated for each j

new experiment or experimental facility that is estimated to produce a dose greater than i mrem at the exclusion boundary.

Basis One (!) mrem dose per experi.nent or experimental facility represents 1% of the maximum 10 Cl R 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 annuallimit.

4.6.3 ALARA Applicabihty 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.

i Obiective

{

1 The objective is to provide surveillance of all operations that could lead to occupational exposures to radiation or the release of radioactive effluents to i the environs.

Snecification As part of the review of all operations, consideration shall be given to alternative operational modes that might reduce staff exposures, release of radioactive materials to the environment, or both.

Basis Experience has shawn 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 radiaSm exposures. Similarly, overall reactor scheduling achieves significant reductions in staff exposures. Consequently, ALARA taust be a part of both overall reactor scheduling and the detailed experiment planning.

4.7 Experiments Aonlicability This specification applies to surveillance requirements for experiments.

RhlNiirc The objective is to ensure that the conditions and restrictions of Specification 3.7 l are met.

Proposed Amendment No. 32 (2/23/98)

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

, experiment.

Basis Authorized reviewers are appointed by the facility director. l 5.0 DESIGN FEATURES 5.1 Reactor Fuel Specifications The individual unirradiated TRIGA fuel elements shall have the following characteristics:

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

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

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

5.2 Reactor Core Specifications

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

b. The reflector, excluding experiments and experimental facilities, shall be water, or D 20, or graphite, or any combination of the three moderator materials, 5.3 Control Rods Specifications l

a. The shim, safety, and regulating control rods shall have SCRAM capability and contain borated graphite, B4C powder, or boron and its compounds in solid form as 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 control rod shall have SCRAM capability and contain borated graphite, B4C 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.

Proposed Amendment No. 32 (2/23/98)

38 5.4 Fuel Storage Specifications

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

b. Irradiated fuel elements shall be stored in an array which shall permit sufficient natural convection cooling by water such that th: fuel element temperature shall not reach the safety limit as defined in Section 2.1 of the Technical Specifications.

5.5 Rcactor Bay and Exhaust Systems Specifications

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

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

b. The reactor bay shall be equipped with two exhaust systems. Under normal l operating conditions, the facility exhaust system exhausts unfiltered reactor bay l

air to the environment releasing it at a point at least 24 feet above ground level.

Upon initiation of a buu ling evacuation alarm, the previously mentioned system is automatically s: cured and an emergency exhaust system automatically starts.

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 Snecification l

The reactor core shall be cooled by natural convective water flow.

6.0 ADMINISTRATIVE CONTROLS 6.1 Organization 6.1.1 Structure The University Vice President for Research Dean of the Graduate School (level 1) has the responsibility for tne 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 Head of the Nuclear Engineering Department and the Dean of the College of Engineering. Administrative and fiscal responsibility is within the offices of the Department Head and the Dean.

The minimum qualifications for the position of Director of the PSBR are an advanced degree in science or engineering, and 2 years experience in reactor operation. Five years of experience directing reactor operations may be substituted for an advanced degree.

The Manager of Radiation Protection reports through the Director of Environmental Health and Safety, the assistant Vice President for Safety and Environmental Services, and to the Senior Vice President for Finance and Business /freasurer. The qualifications for the Manager of Radiation Proposed Amendment No. 32 (2/23/98)

39 r 3 r 3 Senior Vice President for Finance Vice President for Research and llusinessffreasurer

( ) (Dean of the Uraduate School >

r 3 Assistant Vice President for Safety and Environmental Services r 3 y of Engineering j r 3 Director of Environmental IIcalth and Safety Nuclear Engineering Department Head r r 3 Manager of Penn State Reactor Radiatk' protectior Safeguards Committee t J t  ;

i r 3 l

l Director Penn State lireazeale Reactor - ---l (Level 2)

< J r 3 Manager of Onerations and Training (Level 3) ,

r 3 Operating Staff Senior Reactor Operators and

Reactor Operators (Level 4) t J ORGANIZATION CHART Proposed Amendment No. 32 C/23/98)

40 Protection position are the equivalent of a graduate degree in radiation l protection,3 to 5 years experience with a broad byproduct material license, and certification by The American Board of IIcalth Physics or eligibility for certification.

6.1.2 Resnonsibility 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 specificatiens.

In all instances, res 3onsibilities of one level may be asstmed by designated alternates or by hig ler levels, conditional upon appropr ate qualifications.

6.1.3 Staffing

a. The minimum staffing when the reactor is not secured shall be:

1) A licensed operator present in the control room, in accordance with applicable regulations.

2) A second person present at the facility able to carry out prescribed written instructions.

3) If a senior reactor operator is not present at the iacility, one shall be available by tel a e and able to be at the facility within 30 minutes.

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

1) Management personnel.

2) Radiation safety personnel.

L

3) Other operations personnel,

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

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

Pmpned Amendment No 32 (2/23/93) l

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-15.4-1988, Sections 4 6. l 6.2 Review and Audit 6.2.1 Safeguards Committee Comnosition A Penn State Reactor Safeguards Committee (PSRSC) shall exist to provide an independent review and audit of the safety aspects of reactor facihty operations. The committee shall hav, a minimum of 5 members and shall collectively represent a broad spectrum of expertise in reactor technology and other science and engineering fields. The committee shall have at least one member with health physics expertise. The committee shall be appointed by and re 3 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 charter, including provisions for:

a. Meeting frequency - not less than once per calendar year not to exceed 15 months, b, Quorums - at least one-half of the voting membership shall be present 4 (the Director wno is ex-officio shall not vote) and no more than one-half of the voting members present shall be members of the reactor staff,

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

d. Minutes of the meetings - shall be recorded, disseminated, reviewed, and approved 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 upon reactivity or upon the release of radioactivity.

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

c. Violations of technical specifications, license, or charter. Violations of internal procedures or instructions having safety significance.

Proposed Amendment No. 32 (2/23/98)

42

d. Operating abnormalities having safcty significance. l

c. Special reports listed in 6.6.2. l

f. Audit reports. l 6.2.4 Audil The audit function shall be performed annually, not to exceed 15 months, preferably by a non-member of the reactor staff. The audit function shall tu performed by a person not directly involved with the function being audited.

The audit function shall include selective (but comprehensive) examinations of operating records, logs, and other documents, Discussions with operating xtsonnel and observation of operations should also be used as appropriate.

Jeficiencies uncovered that affect reactor safety shall promptly be reorted to the Nuclear Engineering Department Head and the Dean of the Co lege of Engineering The following items shall be audited:

a. Facility operations for conformance to Technical Specifications, license, and procedures (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

+ actor facility equipment, systems, structures, or methods of operations

.at affect reactor safety (at least once per calendar year with the interval not to exceed 15 months),

u. The reactor facility emergency plan and implementing procedures (at least once every other calendar year with the interval not to exceed 30 months).

6.3 Onerating Procedures Written procedures shall be reviewed and ap ? roved prior to the initiation of activities covered by them in accordance witi Section 6.2.3. Written procedures shall be adequate to ensure the safe operation of the reactor, but shall not preclude the use of independent judgment and action should the situation require such.

Operating procedures shall be in effect and shall be followed for at least the following items:

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

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

c. Routine maintenance of major components of systems that could hsve an  ;

effect on reactor safety. 1

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

c. Radiation, evacuation, and alarm checks.

f. Release ofIrradiated Samples. l

g. Evacuation. l Proposed Amendment No. 32 (2/23/98)

43 h, Fire or Explosion,

i. Gaseous Release, J .- Medical Emergencies.

k. Civil Disorder.

1. Bomb Threat.

m. Threat of Theft of Special Nuclear Material,

n. Theft of Special Nuclear Material,

o. Industrial Sabotage,

p. Experiment Evaluation and Authorization.

l q. Reactor Operation Using a Beam Port.

l r. D 20 Handling.

L l s. Health Physics Orientation Requirements.

t. Hot Cell Entry Procedure,

u. Implementation of emergency and security plans.

v. Radiation instrument calibration L w. Loss of pool water.

6.4 Review and Approval of Exoeriments

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

b. Substantive changes to experiments previously reviewed by the PSRSC shall be made only after review and approval in writing by level 2 management or designated attemate.

6.5 Required Action 6.5.1 Action to be Taken in the Event the Safety Limit is Exceeded in the event the safety limit (ll50*C)is exceeded:

a. The reactor shall be shut down and reactor operation shall not be resumed until authorized by the U.S. Nuclear Regulatory Commission.

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

. c. An immediate report of the occurrence shall be made to the Chairman, PSRSC and reports shall be made to the USNRC in accordance with Specification 6.6. l Proposed Amendment No. 32 (2/23/98)

44

d. a report shall be prepared w hich shall include an analysis of ne causes and extent of possible resultant damage, efficacy of corrective action, and recommendations for measures to preven: or rtduce the probability of recurrence. This report shall be submitted to 'he PSRSC for review.

6.5.2 Action to be Taken in the Event of a Reportable Occurrents

.in the event of a reportable occurrence,(1.1.34) the following action shall be l taken:

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

b. The Director or a designated alternate shall be notified and corrective action taken with respect to the operations involved,

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

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

c. A report shall be made to the PSRSC which shall include an analysis of the cause of the occurrence, efficacy of corrective action, and recommendations for measures to prevent or reduce the probability of recurrence. This report shall be reviewed by the PSRSC at their next meeting.

f. A reaort shall be made to the Document Control Desk, USNRC Was ,ington, DC 20555.

6.6 Reports 6.6.1 Operating Reports An annual report shall be submitted within 6 months - ..e end of The Pennsylvania State University fiscal year to the Document Control Desk, j USNRC, Washington, DC 20555, including at least the following items:

1 [

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

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

c. Tabulation of major preventive and corrective maintenance operations having safety significance.

Proposed Amendment No. 32 (2/23/98)

45

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

c. A summary cf the nature and amount of radioactive effluents released or discharged to environs beyond the effective control of the owner-operator as determined at or before the point of such release or discharge. The summary shall include to the extent practicable an estimate of individual radionuclides present in the effluent. If the estimated average release after dilution or diffusion is less than 25 percent of the concentration allowed or recommended, only a statement to this effect need be presented,

f. A siimmarized result of environmental surveys performed outside the facility.

6.6.2 Special Reports i

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

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, Washington, DC 20555, that describes the circumstances of the event within 14 days of any of the follawing:

1) Violation of safety limits (Fee 6.5.1)

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

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 1-2 personnel.

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

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

6.7.1 Records to be Rctained for at least Five Years

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

b. Checks and calibrations procedure file.

Proposed Amer.dment No. 32 i2/23/98)

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 experiments,

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

g. Preventive and corrective maintenance records of associated reactor equipment.

h. Facility radintion and contamination surveys.

3 i. Fuel inventories and transfers.

j. Surveillance activities as required by the Technical Specifications,

k. Records of PSRSC reviews and audits.

6.7.2 Records to be Retained for at least One 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 environs,

d. Drawings of the reactor facility including changes.

Propmed Amendment No. 32 (2/23/98)

_ _ . . . . . . . .