ML20077J674
| ML20077J674 | |
| Person / Time | |
|---|---|
| Site: | 05000157 |
| Issue date: | 08/31/1983 |
| From: | Bernard H Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| NUREG-0984, NUREG-984, NUDOCS 8308170190 | |
| Download: ML20077J674 (84) | |
Text
--
' Safety Evaluation Report related to the renewal of the operating license for the Cornell University TRIGA Research Reactor Docket No. 50-157 U.S. Nuclear Regulatory Commission
)
Office of Nuclear Reactor Regulation August 1983 p "%,,
o k
I
$ggekoScd OSSo$$h>7 E
NOTICE Availability of Reference Materials Cited in NRC Publications Most documents cited in N RC publications will be available from one of the following sources:
- 1. The NRC Public Document Room,1717 H Street, N.W.
Washington, DC 20555
- 2. The NRC/GPO Sales Program, U.S. Nuclear Regulatory Commission, Washingtor,, DC 20555
- 3. The National Technical Information Service, Springfield, VA 22161 Although the listing that follows represents the majority of documents cited in NRC publications, it is not intended to be exhaustive.
Referenced documents available for inspection and copying for a fee from the NRC Public Docu-ment Room include NRC correspondence and internal N RC memoranda; NRC Office of Inspection and Enforcement bulletins, circulars, information notices, inspection and investigation notices; Licensee Event Reports; vendor reports and correspondence; Commission papers;and applicant and licensee documents and correspondence.
The following documents in the NUREG series are available for purchase from the NRC/GPO Sales Program: formal NRC staff and contractor reports, NRC-sponsored conference proceedings, and N RC booklets and brochures. Also available are Regulatory Guides, N RC regulations in the Code of Federal Regulations, and Nuclear Regulatory Commission issuances.
Documents available from the National Technical Information Service include NUREG series reports and technical reports prepared by other federal agencies and reports prepared by the Atomic Energy Commission, forerunner agency to the Nuclear Regulatory Commission.
Documents available from public and special technical libraries include all open literature items, such as books, journal and periodical articles, and transactions. Federal Register notices, federal and state legislation, and congressional reports can usually be obtained from these libraries.
Documents such as theses, dissertations, foreign reports and translations,and non-NRC conference proceedings are available for purchase from the organization sponsoring the publication cited.
Single copies of NRC draft reports are available free upon written request to the Division of Tech-nical Information and Document Control, U.S. Nuclear Regulatory Commission, Washington, DC 20555.
Copies of industry codes and standards used in a substantive manner in the NRC regulatcry process are maintained at the NRC Library, 7920 Norfolk Avenue, Bethesda, Maryland, and are available there for reference use by the public. Codes and standards are usually copyrighted and may be purchased from the originating organization or, if they are American National Standards, from the American National Standards Institute,1430 Broadway, New York, NY 10018.
GPO Printed copy price: _b_4.50
NUREG-0984 Safety Evaluation Report related to the renewal of the operating license for the Cornell University TRIGA Research Reactor Docket No. 50-157 U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation August 1983 j a"'%,,,
b$
ABSTRACT This Safety Evaluation Report for the application filed by the Cornell University for a renewal of Operating License R-80 to continue to operate a research reactor has been prepared by the Office of Nuclear Reactor Regulation of the U.S. Nuclear Regulatory Commission.
The facility is owned and operated by Cornell University and is located on the Cornell campus in Ithaca, New York.
The staff concludes that the TRIGA reactor facility can continue to be operated by Cornell without endangering the health and safety of the public.
i l
I
(
Ctrnell University SER iii
TABLE OF CONTENTS Pa2!!
ABSTRACT........................................................
iii 1
INTRODUCTION...............................................
1-1 1.1 Summary and Conclusions of Principal Safety Considerations......................................
1-2 1.2 Reactor Description...................................
1-3 1.3 Reactor Location......................................
1-3 1.4 Shared Facilities.....................................
1-4 1.5 Comparison With Similar Facilities.....................
1-4
- 1. 6 Nuclear Waste Policy Act of 1982.......................
1-4 2
SITE CHARACTERISTICS.......................................
2-1 2.1 Geography.............................................
2-1 2.2 Demography............................................
2-1 2.3 Nearby Industrial, Transportation, and Military Facilities..........................................
2-1 2.4 Meteorology...........................................
2-1 2.4.1 Severe Wind Considerations.....................
2-2 2.4.2 Precipitation and Flooding.....................
2-2 2.4.3 Co n cl u s i o n.....................................
2-2 2.5 Geology and Hydrology.................................
2-2 2.6 Seismology............................................
2-3 2.7 Conclusion............................................
2-3 3
DESIGN OF STRUCTURES, SYSTEMS, AND COMPONENTS..............
3-1 3.1 Wind Damage...........................................
3-1 3.2 Water Damage..........................................
3-1 3.3 Seismic-Induced Reactor Damage........................
3-1 3.4 Mechanical Systems and Components.....................
3-1 3.5 Conclusion............................................
3-1 4
REACTOR....................................................
4-1 4.1 Reactor Core..........................................
4-1 4.1.1 Fue l El eme nt s..................................
4-l' 4.1.2 Control Rods...................................
4-2 4.2 Reactor Pool..........................................
4-2 4.3 Support Structure.....................................
4-2 4.4 Reactor Instrumentation...............................
4-3 i
Cornell University SER v
s
1 i
TABLE OF CONTENTS (Continued)
P_ag 4.5 Biological Shield.....................................
4-3 4.6 Dynamic Design Evaluation.............................
4-3 4.6.1 Excess Reactivity and Shutdown Margin..........
4-4 4.6.2 Normal Operating Conditions....................
4-4 4.6.3 Conclusion.....................................
4-5 4.7 Functional Design of Reactivity Control System........
4-5 4.7.1 Standard Control Rods..........................
4-6 4.7.2 Transient Control Rod.........................
4-6 4.7.3 Scram-Logic Circuitry and Interlocks...........
4-7 4.7.4 Conclusions....................................
4-7 4.8 Operational Procedures................................
4-8 4.9 Conclusion............................................
4-8 5
REACTOR COOLANT AND ASSOCIATED SYSTEMS.....................
5-1 5.1 Primary Cooling System................................
5-1 5.2 Heat Exchanger nd Secondary Cooling Water System.....
5-1 5.3 Primary Water Purification and Makeup System..........
5-1 5.4 Nitrogen-16 Diffuser..................................
5-2 5.5 Conclusion............................................
5-2 6
ENGINEERED SAFETY FEATURES.................................
6-1 6.1 Reactor Building Heating and Ventilation System.......
6-1 6.2 Contamination Control Features........................
6-1 6.2.1 Normal Operating Conditions....................
6-1 6.2.2 Accident Conditions............................
6-2 6.3 Conclusion............................................
6-2 7
FACILITY CONTROL AND INSTRUMENTATION SYSTEMS...............
7-1 7.1 Control Console.......................................
7-1 7.2 Control Rods..........................................
7-1 7.3 Control-Rod-Drive Assemblies..........................
7-1 7.4 Pulse-Mode Switch.....................................
7-2 7.5 Instrumentation System................................
7-2 7.6 Scrams................................................
7-2 7.7 Conclusion............................................
7-3 8
ELECTRICAL POWER SYSTEM....................................
S-1 8.1 Main Power............................................
8-1 Cornell University SER vi
i TABLE OF CONTENTS (Continued)
Pag 8.2 Emergency Power.......................................
8-1 8.3 Conclusions..........................................
8-1 9
AUXILIARY SYSTEMS..........................................
9-1 9.1 Fire Protecion System.................................
9-1 9.2 Communications System.................................
9-1 9.3 Compressed Air System.................................
9-1 9.4 Equipment and Fuel-Handling Systems...................
9-1 9.5 Conclusion............................................
9-1 10 EXPERIMENTAL PROGRAMS......................................
10-1 10.1 Experimental Facilities...............................
10-1 10.1.1 Pool Irradiations..............................
10-1 10.1.2 Pneumatic Transfer Systems.....................
10-1 10.1.3 Beam Ports.....................................
10-1 10.1.4 Thermal Column.................................
10-2 10.2 Experimental Review...................................
10-2
- 10. 3 Experiment Reactivity Limitations.....................
10-2 10.4 Conclusion............................................
10-3 11 RADI0 ACTIVE WASTE MANAGEMENT...............................
11-1 11.1 ALARA Commitment......................................
11-1 11.2 Waste Generation and Handling Procedures..............
11-1 11.2.1 Solid Waste....................................
11-1 11.2.2 Liquid Waste...................................
11-1 11.2.3 Airborne Waste.................................
11-2 11.3 Conclusions...........................................
11-2 12 RADIATION PROTECTION PROGRAM...............................
12-1 12.1 A LARA Commi tme n t......................................
12-1 12.2 Health Physics Program................................
12-1 12.2.1 Health Physics Staffing........................
12-1 12.2.2 Procedures.....................................
12-1 12.2.3 Instrumentation................................
12-2 12.2.4 Training.......................................
12-2 12.3 Radiation Sources.....................................
12-2 12.3.1 Reactor........................................
12-2 12.3.2 Extraneous Sources.............................
12-2 Cornell University SER vii
1 l
TABLE OF CONTENTS (Continued)
PaSe 12.4 Routine Monitoring....................................
12-3 12.4.1 Fixed Position Monitors........................
12-3 12.4.2 Experimental Support...........................
12-3 12.5 Occupational Radiation Exposures......................
12-3 12.5.1 Personnel Monitoring Program...................
12-3 12.5.2 Personnel Exposures............................
12-3 12.6 Effluent Monitoring...................................
12-3 12.6.1 Airborne Effluents.............................
12-3
- 12. 6. 2 Li qui d E f fl ue nt................................
12-4
- 12. 7 Envi ronmental Monitori ng..............................
12-4 12.8 Potential Dose Assessments............................
12-4 12.9 Conclusion............................................
12-4 13 CONDUCT OF OPERATIONS......................................
13-1 13.1 Overall Organization....
13-1 13.2 Training..............................................
13-2 13.3 Emergency Planning....................................
13-2 13.4 Physical Security Plan................................
13-2 13.5 Conclusion............................................
13-3 14 ACCIDENT ANALYSIS..........................................
14-1 14.1 Rapid Insertion of Radioactivity......................
14-1 14.1.1 Scenarios......................................
14-2 14.1.2 Assessment.....................................
14-2 14.1.3 Conclusion.....................................
14-3 i
14.2 Loss of Coolant.......................................
14-3 14.2.1 Scenario.......................................
14-3 14.2.2 Assessment.....................................
14-4 14.2.3 Conclusion.....................................
14-4 14.3 Metal-Water Reactions.................................
14-4 14.4 Misplaced Experiments.................................
14-5 1
14.4.1 Assessment.....................................
14-5 l
14.4.2 Conclusion.....................................
14-5 l
l i
l Cornell University SER viii I
L
.=.
i TABLE OF CONTENTS (Continued)
P_ age 1
1 14.5 Mechanical Rearrangement of the Fuel..................
14-6 a
1 14.5.1 Assessment.....................................
14-6 j
14.5.2 Conclusion.....................................
14-6 14.6 Effects of Fuel Aging.................................
14-6 1
14.6.1-Assessment.....................................
14-6 14.6.2 Conclusion.....................................
14-7 l
14.7 Fuel-Handling Accident................................
14-7 i
14.7.1 Scenario.......................................
14-7 i
14.7.2 Assessment.....................................
14-8 14.7.3 Conclusion.....................................
14-9 2
14.8 Conclusion............................................
14-10 15 TECHNICAL SPECIFICATIONS...................................
15-1 16 FINANCIAL QUALIFICATIONS...................................
16-1 1
17 OTHER LICENSE CONSIDERATIONS...............................
17-1 17.1 Prior Reactor Utilization.............................
17-1 1
17.2 Multiple or Sequential Failures of Safety Components..
17-2 17.3 Conclusion............................................
17-3 i
18 CONCLUSIONS................................................
18-1 i
19 REFERENCES.................................................
19-1 I
J e
I l
i j
Cornell University SER ix l
E
LIST OF FIGURES Page 1.1 Plan View of Ward Laboratory..............................
1-5 1.2 Cross-Sectional View of Ward Laboratory...................
1-6
- 1. 3 - Map o f I th a c a A rea........................................
1-7 2.1 Cornell University Campus.................................
2-4 3
2.2 Ithaca Urban-Area Study...................................
2-5 4.1 Sectional View of CU TRIGA Reactor Tank...................
4-9 5.1 Pool Water Cooling System.................................
5-3 7.1 Block Diagram of Instrumentation and Scram System.........
7-4
- 7. 2 Schematic of Transient-Rod Drive Assembly.................
7-5 13.1 Organizational Structure for Radiation Protection.........
13-4 LIST OF TABLES 1.1 TRIGA and TRIGA-Fueled Reactors Licensed in the United States as of September 1982...............................
1-8.
7.1 Safety System Channels....................................
7-6 12.1 Number of Individuals in Exposure Interval................
12-5 14.1 Doses Resulting From Postulated Fuel-Handling Accident....
14-11 1
r i
Cornell University SER x
_1 INTRODUCTION _
The Cornell University (CU) (licensee / applicant) submitted a timely application to the U.S. Nuclear Regulatory Commission (NRC) (staff) for renewal of the Class 104 Operating License R-80 for its modified TRIGA research reactor by letter (with supporting documentation) dated May 27, 1980, as amended by letter and submittal dated September 15, 1980, for renewal of the operating license for 20 years and an increase in operating power level from the existing 250 kWt to 500 kWt plus a maximum pulse insertion of 3.00$.
Cornell University has held an operating license for the TRIGA reactor since 1967.
CU currently is permitted to operate the reactor within the conditions authorized in past amendments in accordance with Title 10 of the Code of Federal Regulations (10 CFR), Paragraph 2.109, until NRC action on the renewal request is completed.
The staff technical safety review with respect to issuing a renewal operating license to CU has been based on the information contained in the renewal application and supporting supplements, plus responses to requests for addi-tional information.
The renewal application includes:
a Physical Security Plan; proposed Technical Specifications; Environmental Report Data; Safety Analysis Report and supplements; Financial Qualifications as supplemented through April 20, 1981; Reactor Operator Requalification Program as. supple-mented through November 4, 1981; and an Emergency Plan dated October 29, 1982.
This material is available for review at the Commission's Public Document Room at 1717 H Street N.W., Washington, D.C.
The_ renewal application contains the information regarding the original design of the facility and includes information about modifications _to the facility made since initial licensing.
The Physical Security Plan is protected from public disclosure under 10 CFR 2.790(d)(1) and 10 CFR 9.5(a)(4).
The purpose of this Safety Evaluation Report (SER) is to summarize the results of the safety review of the CU TRIGA reactor and to delineate the. scope of the technical details considered in evaluating the radiological safety aspects of continued operation.
This SER will serve as the basis for. renewal of the license for operation of the CU facility at steady-state th'ermal power levels up to and including 500 kWt, and pulsed operation that permits step reactivity insertions up to 3.00$ (2.25% ak/k). The facility.was reviewed against the requirements of 10 CFR 20, 30, 50, 51, 55, 70, and 73, applicable Regulatory Guides (Division 2, Research and Test Reactors); and appropriate accepted industry standards (American National Standards Institute /American Nuclear Society (ANSI /ANS 15 series)).
Because there are no. specific accident-related regulations for research reactors, the staff has at times compared calculated dose values with related standards in 10 CFR 20, the standards for protection against radiation, both for employees and the public.
This Safety Evaluation Report was prepared by Harold Bernard, Project Manager, Division of Licensing, Office of Nuclear Reactor Regulation, Nuclear Regula-f tory Commission.
Major contributors to the technical review include the i
Cornell University SER 1-1 J
project manager and J. Hyder, D. Whittaker, and C. Thomas of Los Alamos National Laboratory (LANL) under contract to the NRC.
The CU reactor has been in operation since January 1962 using standard TRIGA fuel.
In the more than 20 years that the CU TRIGA reactor has operated, the average annual use in the experimental programs has been about 40 to 50 MW hours per year.
In terms of radiation exposure of reactor components or production of radioactive material, this amount of operational use corresponds to about 50 8-hour days per year at maximum authorized steady-state power.
TRIGA reactors--utilizing essentially the same kind of fuel, similar control rods and drive systems, and safety circuitry as at CU--have been constructed and operated in many countries of the world.
Among approximately 58 such reactors in operation, some since 1958, there have been no reported events that caused significant radiation risk to the public health and safety.
Some other TRIGA reactors have two to three times the power levels of the reactor at CU (500 kW) and have annual MW hours of operation at least a factor of 10 greater than the CU reactor, primarily because of different types of research programs.
1.1 Summary and Conclusions of Principal Safety Considerations The staff evaluation considered the information submitted by CU, past operating history recorded in annual reports submitted to the Commission by the applicant, reports by the Commission's Office of Inspection and Enforce-ment, and onsite observations.
In addition, as part of the licensing review, the staff obtained laboratory studies and analyses of several accidents postulated for the TRIGA-type reactor.
The principal matters reviewed for the CU reactor and the conclusions reached were the following:
(1) The design, testing, and performance of the reactor structure and systems and components important to safety during normal operation are inherently safe, and safe operation can reasonably be expected to continue.
(2) The expected consequences of a broad spectrum of postulated credible accidents have been considered, emphasizing those likely to cause loss of integrity of fuel-element cladding.
The staff performed conservative analyses of serious credible accidents and determined that the calculated potential radiation doses outside of the reactor room are not likely to exceed 10 CFR 20 doses in unrestricted areas.
(3) The applicant's mani.gement organization, conduct of training and research activities, and secLrity measures are adequate to ensure safe operation of the facility and protection of special nuclear material.
(4) The systems provided for control of radiological effluents can be operated to ensure that releases of radioactive wastes from the facility are within the limits of the Commission's regulations and are as low as reasonably achievable (ALARA).
1
\\
l Cornell University SER 1-2 l
(5) The applicant's Technical Specifications, which provide operating limits controlling operation of the facility, are such that there is a high degree of assurance that the facility will be operated safely and reliably.
(6) The financial data and information provided by the applicant are such that the staff has determined that the applicant has sufficient revenues to cover operating costs and to ensure protection of the public from radiation exposures when operations are terminated.
(7) The applicant's program for providing for the physical protection of the facility and its special nuclear material complies with the applicable requirements in 10 CFR 73.
(8) The applicant's procedures for training its reactor operators and the plan for operator requalification are adequate; they give reasonable assurance that the reactor facility will be operated competently.
(9) The applicant has submitted an emergency plan dated October 28, 1982, using guidance that was current at the time of license renewal applica-tion.
The plan is under review by the staff and will be reported sepa-rately following completion of the review.
1.2 Reactor Description The CU TRIGA is a heterogeneous pool-type reactor using 20% enriched 23sU.
The core is cooled by natural convection of light water, moderated by ZrH and x
water, and reflected by water and graphite.
The core is located in a 21,800 gal partially below ground pool which is, in turn, cooled and purified by external cooling and purification systems.
Reactor experimental facilities include incore irradiation positions, a thermal column, and numerous beam tubes.
TheexistingreactorsystemoperateswithstandardTRIGAfuelinthestead{U 23 state or pulsed modes.
Standard TRIGA fuel contains U-ZrHx. enriched in to less than 20%.
The maximum continuous steady-state power level is proposed to be 500 kWt and the maximum pulsed power level is with step reactivity insertions up to 3.00$.
The reactor has been operated safely for more than 20 years in its present configuration.
The safety of the system stems from the large, prompt, nega-tive temperature coefficient that is inherent in a water-moderated, U-ZrH x
fueled reactor.
1.3 Reactor Location The CU campus is in the City of Ithaca, New York, which lies on the southern end of Cayuga Lake, one of the Finger Lakes (see Figure 1.3).
The reactor is located in the Ward Laboratory building in the southern part of the southwest section of the Cornell University campus as shown in Figures 1.1 and 1.2.
The elevation of the ground floor of the laboratory is approximately 775 ft, which is 60 to 65 ft above the level of Cascadilla Creek.
There has been no new construction in the immediate vicinity of Ward Laboratory and-there are no plans for future construction, according to the applicant.
Cornell University SER 1-3
1.4 Shared Facilities Ward Laboratory, in addition to housing the CU TRIGA reactor, also contains a zero power reactor (License R-89) critical facility, a gamma irradiation cell licensed by New York State, and an accelerator facility.
The accelerator facility comprises three rooms below ground level at the east end of the reactor laboratory and a fourth room for service equipment on the south side.
Access to the control room, accelerator room, and target "oom area is through the machine shop in the basement of the reactor laboratory.
Services to the accelerator room from the equipment room enter through a sealed concrete labyrinth.
1.5 Comparison With Similar Facilities The Cornell University TRIGA reactor is similar to all TRIGA reactors using standard or FLIP-type fuel.
Of the 58 TRIGA reactors operating throughout the world, 27 are in the United States (24 licensed by NRC, 3 by the Department of Energy.
The instruments and controls are typical of NRC-licensed TRIGA research reactors; see Table 1.1).
1.6 Nuclear Waste Policy Act of 1982 Section 302(b)(1)(B) of the Nuclear Waste Policy Act of 1982 provides that the NRC may require, as a precondition to issuing or renewing an operating license for a research or test reactor, that the applicant shall have entered into an agreement with the Department of Energry (DOE) for the disposal of high-level radioactive waste and spent nuclear fuel.
DOE (R. L. Morgan) has determined that universities and other government agencies operating nonpower reactors have entered into contracts with DOE that provide that DOE retain title to fuel and is obligated to tade the spent fuel and/or high-level waste for storage or reprocessing.
Because the Cornell University reactor is a research reactor, it is in con-formance with the Waste Policy Act of 1982.
i Cornell University SER 1-4
l il
/
u
-~r d P i
7 q
... ~.. _
W-
%~
%u e..
r-I
==.=; - -
pdE.
h
) tads >
'4 '
s.e
,p~
a r--,.A a
'. oss 7_p
' ' * * +
- s
/p w w.r' 4_% Jm7" 4r
. v.
m= a.e.
i m
,p./
g - Q,, p (f a
.h.
,a
. i,-= " n cu er A
't 9 i
e-
~
- gg.,-
d
),--A(.
.,,q
._ut i.
m s
.. '*a.1 g
zpg k 6
- /rn, 1.n,, c g4x..
ue.a y i -
<r 1%h, h
l-
!Q
..r. C..
1 "*,rtz" r ar
- _h_
v c-
@ '. Ei L.._ T.,,S, g:NW 3.,-
7 v t:
c li _
i,..
.[
.aw
- u. --
c i
q fdd-
-t -
rc NErc}w lg g
r p'fi ' f y
[I 3T-.-
. l e &F
~, '
- 4 dp 3
-?.
, 9.;. ' r.9 6
t'
'o ti;tyn'_wr,_
p -
I
-, y
-m,
- tajg4
, - V"@;: j{ 4,,,
l N.
l
+tB
.gg-
_i
. _. _ 4 ?
- a. r.
>-i
' ate g w 1,a e f,.
. _N.,, g s*e 4
1,[.,vy-. e #n" Q". "" t.
egg ; = _ _,,~p* * %s..-.
u
.- A L i L
K2
- % ' _ _ _. _ - ~ - - _ _. L.
m--w.- W n/h,a e#-- ;
,. j g 1., r
- ggr
.y
-~~y
/y Lr mp.
.,.ms upy-rh. -,-<.
--a mw 7p rs.-g y
a
~r r*,
c ~~;- r *h
- h-- ; j.-!Q
- n. -
1 1;* ~h rp g'
,,a
.i
~~ s
(
It =~s F= 9pe._g,2 m c
5 g e g
,~j tif @
I" fa
. ~S.* 4i Figure 1.1 Plan view of Ward Laboratory l
1 l
Cornell University SER 1-5 l
- tifra.'.. ".,.
.r-
- _. _ _ _.. m.a.. - _.
, j. __.,> -. -..
,.,g.,
' e-y su_n.c tT
- c. -
s m
~~1 b
,R _ I'I tlII I (_ f
_u a
j,_._l. m! -
,I
= = -
o I ut2 t T ",,'.a
- ==h...
w m
A h, /,.i g o 2 -c_e
' ~ ~ ~
l!FM
- ,r
.-=rm -
I e
Latteeet l
J a
i....--
f -= '=;,
L
+
...,..n
..,,,, e c,.
ser -;
. Q,., p _.,
e g_ m 3
g-p r
&m
~
--- m =
a u
- ~..
Lj
.ji"'53
- v. -
- E
.........3.,
L.
ar n
,s..
- ".,...-._.__..i.....~i, a
rm=.
n.
.y
,,,,. r-
= ~ = -
~~ ~ ~
w w-Q 3 e c T
- o r'
I
- 1 l
- . :1n~
e...,~
'c
,=I X X-~1 I_ n I
- I__,
.I.
I
~ ~ ':!-3
....__.._--~y=
- ll
---~
--+
--y
,a 7"**.*.* D l I
"u 4.s 3
I M*,. N 'l. Y. Y
~
gu a~~*.
14.-...-
W r.r.e.- >.
w w...
,s
+,t_w_.
i
,D(Y:a_'*'..
A%,
5 /****'*'
r n/xw i'i" l l. l-ra
.n=3
, parts, m a+r e
o,--
- - r - ---...n m.
..==
,..,n -
r i
'55' l y==iWF i
aml :
s t
.uxez.
- . - r,.-c.~s:
g p, e-y a
1 3,
7,..
w-
-., = -
1_,,,,..
d
- /#
a we*
h 3 e c y o o a 3 - S Figure 1.2 Cross-sectional view of Ward Laboratory Cornell University SER 1-6 1
\\
\\
.u M_
V
'l l.:.:_:~;~g N
^
/
.i =
=*
~
i e
ceuou me-s
~
N=::.
- k*
\\
N!
N (hq g
\\s
\\,
- t o.
- ,l L /,
j
- \\
//
- .)
% ? g /~
w 4 [\\
U[ _-(<a m.Y (4~--*. ' g
'T,
\\
[
upps!r kb ;f ^
\\
g
~
ees si 8
't
)
Er. :.N -m 1
i rJ"N ] Ml1.\\I /....
/
s_
E
.2d E:.:...
p -~
Q
~
Reproduced by permission of the Office of University Publications, Cornell University i
l 5
Figure 1.3 Map of Ithaca area Cornell University SER l-7
. ~ __
i Table 1.1 TRIGA AND TRIGA-fueled reactors licensed in the United States as of September 1982 l
Authorized Licensee Type
- power, KWth AFRRI TRIGA MK F 1000 1
Aerotest Operations, Inc.
TRIGA MK I 250 Cornell University TRIGA 100 3
Dow Chemical Company TRIGA MK I 100 General Atomic TRIGA MK I 250 General Atomic TRIGA MK F 1500 Kansas State University TRIGA MK II 250 Michigan State University TRIGA MK I 250 4
Northrop Corporation TRIGA MK F 1000 Oregon State University TRIGA FLIP 1000 Pennsylvania State University TRIGA MK III 1000 Reed College TRIGA MK I 250 Texas A & M TRIGA FLIP 1000 U.S. Geological Survey TRIGA MK I 1000 University of Arizona TRIGA MK I 100 University of California / Berkeley TRIGA MK III 1000 University of California /Irvine TRIGA MK I 250 University of Illinois ADVANCED TRIGA 1500 i
University of Maryland TRIGA 250 University of Texas TRIGA MK I 250 University of Utah TRIGA 100 University of Wisconsin TRIGA FLIP 1000 Veterans Administration TRIGA 18 Washington State University TRIGA FLIP 1000
- The two basic models, Mark I and Mark II, are distinguished by whether they are located above or below ground.
The Mark III model has a movable core and pulsing capability.
1 i
r I
Cornell University SER 1-8
2 SITE CHARACTERISTICS 2.1 Geography Cornell University is located in the heart of the city of Ithaca, which lies in the Finger Lake's region of New York State.
The city of Ithaca is at the southern tip of Cayuga Lake in a very hilly area.
The elevation of the campus around Ward Laboratory is 760 to 770 ft.
Cascadilla Creek, the receiving stream for much of the campus runoff, is at an elevation of
- 710 ft, a drop of
- 60 to 65 ft from Ward Laboratory first-floor elevation and
- 100 ft away.
Cascadilla Creek also forms the southern boundary of the CU campus, separating the university from the adjacent residential area.
2.2 Demography Figure 2.1 shows the approximate boundary of the university campus.
Inside that boundary are classroom and laboratory buildings, service buildings, dormitories, and other usual university buildings.
The remaining part of the map (outside the campus boundary) is a residential area except for a commercial area along College Avenue and Eddy Street, along Dryden Road between Eddy Street and Lindan Avenue, and along Stewart Avenue.
The campus is separated from the residential and commercial areas to its south by the gorge of Casca-dilla Creek.
The population distribution of the Ithaca area is shown in Figure 2.2.
The total population in the area is about 25,000.
In the immediate vicinity, the nearest residence is about 300 ft away on the opposite side of Cascadilla Gorge.
2.3 Nearby Industrial, Transportation, and Military Facilities Ithaca is serviced by several commercial airlines using Tompkins County Airport, which is located approximately 2 air miles from the campus.
There are no railroads, heavy industry, or military installations in the vicinity of the campus.
l l
The staff concludes that there is virtually zero probability of risk of I
accidents to the reactor from activities associated with military, heavy industry, or transportation operations.
2.4 Meteorology l
The temperature at Ithaca averages between 25*F in winter to 68 F in summer.
Meteorological information is obtained from the Tompkins County Airport.
1 Prevailing winds during fall and winter are predominantly from the southeast.
l There is also a consistent strong wind from the northwest, which is more predominant in the spring and summer months.
Cornell University SER 2-1
When regional cyclonic meteorological disturbances are absent, local topograph-ical and thermal conditions produce a " night wind" described from Cornell University Extension Bulletin No. 764 as follows:
The so-called night wind of the Cayuga Lake Valley blows during the summer months at times when the absence of cyclonic disturbances gives full play to local influences.
Commonly it sets in a few hours after sunset as a light breeze from the south, gradually increases in strength until a velocity of about eight miles an hour is reached, and then continues steadily throughout the night.
This currerit has its origin on the hillsides at the southern end of the lake, and it flows northward down the watercourses converging into the main depression.
As it moves northward over the smooth surface of the lake, it is augmented by the numerous cool currents which join the main stream through the watercourses that debouch upon the valley from either side.
2.4.1 Severe Wind Considerations The weather in this region is usually controlled by the extra-tropical cyclones that frequently pass over the area.
These storms can cause occasional high winds and are usually from the southeast.
The peak 5-min wind in the period 1909 through 1943 was 70 mph in January 1939. While winds in the 45 to 70 mph range have been experienced in every month of the year, such velocities are not unusual for an area in the temperate latitudes.
The area is not in the usual path of tropical storms; only seven hurricanes have crossed central New York since 1800.
Hurricane Hazel (October 1954) passed through the area causing some damage to trees and roofs.
2.4.2 Precipitation and Flooding The maximum 24-hour precipitation of 7.90 in, was recorded in July 1935 and was associated with thunderstorm activity, and the highest 24-hour precipi-tation associated with a tropical storm occurred in 1972 with Hurricane Agnes (Letter from B. E. Dethier to Dr. K. B. Cady (CU), 1980).
On these two occa-sions local flooding was experienced in the Ithaca area.
However, flooding of the reactor building did not occur as it is virtually impossible because of the topography of the site.
2.4.3 Conclusion The staff concludes that there are no unique meteorological conditions that could produce or cause a significant risk to the safe operation of the CU TRIGA reactor.
2.5 Geology and Hydrology The unconsolidated mantle at the site of the CU reactor consists primarily of thin-bedded lacustrine silt and fine sand of moderate permeability and subor-dinate local pockets of highly permeable deltaic gravel.
The' underlying
. bedrock is the Enfield and Ithaca Formations, both of which consist of shale with numerous interbedded flaggy silt-stone units.
Because of the relative impermeability of shale,. flow through these formations is controlled by the Cornell University SER 2-2 l
well-developed conjugate joint system.
Ultimately the ground water would emerge to join surface flow in Cascadilla Gorge, mostly above the Central Avenue Bridge.
2.6 Seismology South central New York is considered to be moderately stable seismically.
The probability of a significant shock originating in the vicinity of Ithaca is low, but earthquakes of light-to-moderate intensity are fairly frequent in northeastern New York and the St. Lawrence Valley.
They are perceptible in the Ithaca region on an average of perhaps once a decade.
The nearest major earthquake had its epicenter near Attica (W'oming County) in western New York y
in 1929.
This earthquake was widely felt across New York and New England but structural damage was restricted to a radius of
- 10 mi around Attica.
2.7 Conclusion The staff concludes that there are no hydrological, geological, or seismo-logical conditions that pose unacceptable risks to the CU reactor facility or the contiguous public, i
{
l Cornell University SER 2-3 l
)
s
\\
\\
\\
,, s
~
l
\\.
q ca l
~
g'}
\\\\
'.i 5
/.\\
51 5
[}) t k @%,._[
/
C.8
/
=
-f:; l l,'~~
I
@)
g, -
--j i
s 0 ((fh l
,, / g -
w x
0i
.E;c 6-x
- w. o. c
_e
=
rs
~s 1
w
- gS#'
I "t
'i I
f y
\\e 4 I
I>
8 g#r
- L5 g
/N g
A
%G,.[ g \\de f
je 11 11 g e
1 I !.s e
f
.(
, Y /=
L 1
/ a;,
i
^
\\
I!
.b i-1
'y a
' T,, e '" p__t'"l i
i
_N
?
/
l' %!
i)j i
__sj,ACN 3
i k
oI' 1
Yik~ ~ *I
{I 3
Q h}
li
' TEsti M'=il i
2 01 o, d l = Bl 58
.R E
2 il s,
N ]1 i l<i,
glp 3
i
,e
,a,.. l s
a C!
l
>l a-lj 191-
',/,
d q
o a
3 as N
I gi c i
a f
-j I *-
. f.. !
A 111 b l
-y y,I.:sy1 / 4 9 1, s C } h a ) s L:
3
/ I t c; p
c y
w si ca 9 #
a,I c
ll p"$
ci)e ill 9
]I f
lCL
}
[!! c3 o$
i y, '
's
. o====
i r
I a,
'B 1
d!
cE i
g j
Cornell University SER 2-4
j k
p ^-7qs.;.,,/
y
?._
J.
V t
A x
e,
, \\,
h p.
y
)
p Y
~
- 7..,.
\\
x
[A..
-l /
~
-s
. - - - - a
\\
i.
['[%
/, - x
-y
' Q i
\\
/. i
,- s
.~ ~ '
A A-
.'a y, g.
'.... ;..4,..,,. j x.st R,~,s
.,e-
'.,,.o..
4
. *....., e.
,\\
/
, ig.;
' 4,,..g ;
-1....r 1
,s s
- *4 v.Jr..,*
.... x. f.,.* * *..,.../../ /
\\, ',
s f
-t s,
s,'
i r
\\
k-
)4 <
.q.:. v j
I 3
D e
. y,., y. e s. -\\.
)
i v
s,.
i.
l
....;,-N r.
'o.
s s
spa.. -
~
.,.*- ;k. ] c..
3 A
(
.ys.
s y
,-%..~ -
-(
.s i.,p. -...
-lQ 'y
' ~,..m...._
l.
.v e
-r
+e E
%7 low f t-
./,7,
%A <;-
- -' ". o
's'y
~
fu f t b
3 i !.
lj
.,._h.
I,
,7
- f. ' r
'$Nft 4
t
,i g
..- i
,8
U
.g
/
h5
'I ? M1M f t
.., 3W f t
!l
'T
',h [
~
~-
c :. N - E D ---
Y w.,h~ h '@. N.:p.,.p,cle.
f x--;-a
- e. y g?
3e
..p -y,..
a, '.
s,
.- x ks_ -gi.y.....xm.44, 2,My..., f.. ~,,;. -
j; s
u
-\\
., J. ../
3
-d
%. ';\\ m.. t,a.:
-w a
i o
.,: 7
. R.'..
v....
k r
~
1
. o..l
- u. 's
~. t-
- fI..::.
.::, z. -. [,\\.'b n
i
@(
j .
l-f-
.. s
- s/; j.
[,
.' 1,.
/,
r 1
~ y,r.
s.
~
ITu ACA untiN-AMA STHY i
, ' }.c p A'..:
nnuun muimin w m s
1 m
g/sy 4
N G,
.i
.... =
's s'
~-
-l
..i.,.,.
s.fi.
s,w s
<;d s
h 1
s, e ee. t u s.
)
5 8
$ PROPOSED SITE A(
N
}
q
))
M,/ j' d 'M Reproduced by the permission of the Office of University Publications, Cornell University Figure 2.2 Ithaca urban-area study Cornell University SER 2-5
3 DESIGN OF STRUCTURES, SYSTEMS, AND COMPONENTS 3.1 Wind Damage Except for perhaps superficial damage from the 1937 hurricane, wind has not caused any damage to any of the CU buildings, including Ward Laboratory.
3.2 Water Damage As stated in Section 2.1, Ward Laboratory is located ~ 60 ft above Cascadilla Creek.
In addition, drainage away from' the laboratory is excellent; accord-ingly, there is no risk of flooding at the reactor site.
3.3 Seismic-Induced Reactor Damage As stated in Section 2.6 of this report, the CU campus is located in a moderately stable seismological region.
In the unlikely case of severe seismic damage to the reactor structure and loss of coolant, there would be no nelting of the core fuel and, therefore, no dispersion of fission products (as dis-cussed further in Section 14.2).
The staff, therefore, concludes that the radiological consequences of seismic damage to the reactor facility and to the reactor core is infinitesimal.
3.4 Mechanical Systems and Components The mechanical systems of importance to safety are the neutron-absorbing control rods suspended from the superstructure, which also supports the reac-tor core.
The motors, gear boxes, electromagnets, switches, and wiring are above the level.of the water and readily accessible for testing and maintenance.
An extensive preventive maintenance program has been in operation for many years for the CU TRIGA reactor facility to conform and comply with the perfor-mance requirements of the Technical Specifications.
The effectiveness of this preventive maintenance program is attested to by the small number and types of malfunctions of equipment over the years of operation.
These malfunctions have almost exclusively been one of a kind (that is, no repeats) and/or of components that were fail safe or self-annunciating.
Therefore, the staff concludes that there appears to be no significant deter-ioration of equipment with time or with operation. Thus, there is reasonable as'surance that continued operation for the requested period of renewal will not increase the risks to the public.
3.5 Conclusion The CU reactor facility was designed and built to withstand all credible and probable wind and water damage contingencies associated with the site.
A seismic event has a small likelihood of occurring and the consequences of such occurrence would be minimal and would pose no radiological hazard to the public.
Cornell University SER 3-1
=.
4 4
1 4 REACTOR j
The CU reactor is a standard TRIGA heterogeneous pool-type reactor incorporating i
solid U-ZrH stainless-steel clad fuel elements.
The reactor core is immersed x
in a large open tank of light water that acts as both a moderator and coolant.
I Reactor control is achieved by insertion and withdrawal of neutron-absorbing control rods.
Pulses are initiated by withdrawal of a fast-moving transient rod.
I The CU TRIGA reactor first achieved criticality in January 1962 and began pulse operation in September 1964.
It has been used in a broad range of teaching, research, and service programs in basic and-applied areas of science and engineering.
In his license renewal application, the applicant requested an increase in the authorized, maximum, steady-state power level from 250 to 500 kWt and in the authorized, maximum, pulse mode reactivity insertion from 2.00$ to 3.00$.
The new limits are needed to permit more effective use of the reactor for several experimental programs as well as to make'some new ones possible.
4.1 Reactor Core t
The reactor core consists of a concentric cylindrical arrangement of approxi-mately 84 cylindrical U-ZrH fuel moderator elements and 4 control rods.
The x
elements are held in position by upper and lower aluminum grid plates.
The active (or fueled) region of the reactor core forms a right cylinder approxi-mately 16.5 in, in diameter and 15 in, high and contains 3.0 kg of _2ssU.
Water coolant occupies appro>imately 1/3 of the core volume.
)
A graphite radial reflector nominally 12 in. thick and 22 in. high-surrounds i
the core region.. Top and bottom axial reflection is provided by + 3.5-in.-long graphite plugs-incorporated in the individual fuel elements.
4.1.1 Fuel Elements The CU TRIGA reactor was fueled initially with aluminum-clad fuel-elements.
In 1970 authorization was granted to use stainless-steel-clad fuel elements in addition to the original aluminum clad.ones.
The stainless-steel clad fuel.
elements incorporate-zirconium. hydride with a significantly higher hydrogen-to-zirconium atom ratio, which allows' higher temperature operation.
The reactor currently is loaded with stainless-steel-clad fuel elements only.
In the renewal application and proposed Technical Specifications, the licensee states-that aluminum-clad fuel elements will no longer be used in the CU'TRIGA_ reactor.
-The fuel.part consists of a cylindrical rod of U-ZrH containing s 8.5. weight-x percent uranium ~enrichedzto'slightly less than 20%.
The nominal' weight of l
2ssU in-each fuel element is s 38 g. f The hydrogen-to-zirconium atom ratio of.
the fuel-moderator material is 1.7.
The fuel section of each element lis
.epproximately 15 in. long and 1.43 in. in diameter.
Graphite end: plugs, 3.44 in. long, located above.'and below the fuel-region,-serve as axial neutron
(
\\
l i
Cornell, University SER-4-1 l
1 1
l reflectors.
The fueled section and graphite end plugs are contained in a 0.020-in.-thick Type 304 stainless-steel-walled can.
The can is sealed by welds with stainless-steel fittings at the top and bottom.
Each element is
- 28.4 in long and weighs
- 3.4 kg.
At least one fuel position contains a special instrumented fuel element into which thermocouples were fitted during fabrication.
These elements are iden-tical to standard fuel elements in all other respects.
The thermocouples monitor the fuel element temperatures and provide a scram signal in case of overheating.
4.1.2 Control Rods Power in the CU TRIGA reactor is regulated by four control rods--two safety-shim rods, one regulating rod, and one safety-transient rod.
The neutron poison material in the shim, safety, and regulating rods is powdered boron carbide contained in sealed aluminum tubes.
The transient rod uses 8 C-4 impregnated graphite as the neutron poison.
The control rods are approximately 20 in. long with a vertical travel of s 15 in.
The outside diameter of the shim and transient rods is 1.25 in., and the diameter of the regulating rod is 0.875 in.
The control rods are screwed into the control-rod-drive assembly extension tube and operate in a perforated aluminum guide tube.
4.2 Reactor Pool The reactor core is positioned in the reactor tank under approximately 20 ft of light demineralized water.
The water serves as a radiation shield, neutron moderator, and reactor coolant.
The reactor tank is constructed of ilmenite concrete in the core region and ordinary concrete elsewhere.
The interior walls are coated with phenoline plastic.
A cross-section view of the tank showing the location of the reactor and the disposition of the normal and high density concrete is shown in Figure 4.1.
The approximate tank dimensions are 11 ft by 12 ft by 25 ft deep, with a resulting cavity of
- 22,000 gal of water.
The natural thermal convection of the water coolant adequately disperses the heat generated in the core during normal reactor operations in both the steady-state and pulse modes.
The pool water is pumped through an external heat exchanger system where it is cooled by chilled water supplied by the university.
Four storage wells located within the pool provide storage for irradiated fuel elements.
The 10.5-ft-deep wells are lined with 10-in.-inside-diameter alumi-num pipe and contain storage racks that can accommodate up to 19 fuel elements in each well.
These racks have been designed to be criticality safe when immersed in water.
4.3 Support Structure The fuel elements are positioned at the top and bottom by two 0.75-in.-thick aluminum grid plates.
The bottom grid plate supports the weight of the fuel elements.
Both grid plates are supported by the radial graphite reflector Cornell University SER 4-2
assembly, which in turn is supported by an aluminum frame structure at the bottam of the reactor pool.
4.4 Rcactar Instrumentation The operation of the CU TRIGA reactor is monitored by instrumentation channels that measure fuel element temperature and neutron density.
Thermocouples in at least one instrumented fuel assembly provide continuous information on fuel material temperature during both steady-state and pulse operation.
This signal is displayed in the control room and used to initiate a reactor scram if preset temperature limits are exceeded.
Four neutron-sensitive channels (one fission detector, two compensated ion chambers, and one uncompensated ion chamber), which measure reactor power, have a range from a source level of 1 MW to-full power.
During steady-state operations, two of those channels provide sc. ram signals to the safety circuits when preset power levels are exceeded.
The reactor period also may be measured during the steady-state mode.
In the pulse mode, the uncompensated ion chamber is connected to an integrating circuit that monitors and displays the integrated power and also provides a scram signal when preset levels are exceeded.
4.5 Biological Shield The thick concrete walls of the pool in which the reactor is immersed serve as an effective radiation shield.
Vertical shielding is provided by approximately 20 ft of pool water above the core and 2 ft of water plus 3 ft of ordinary concrete below.
The core is shielded horizontally by 1 ft of graphite, 1.5 ft of water, and a minimum of 6 ft of high-density concrete around approximately three quarters of its circumference.
The remaining quarter is surrounded by 6 to 10 ft of water and 2.5 ft of ordinary concrete.
The horizontal shield geometry can be seen in Figure 4.1.
All concrete immediately adjacent to the core and surrounding the ports and thermal column is of the high-density type.
Special shielding is provided for the experimental irradiation facilities.
When not in use, the beam ports are filled with stepped plugs consisting of boral, lead, concrete, and steel sections and closed off with a lead-lined door.
For the 4-in. and 6-in, ports, additional wooden plugs and lead-filled shutters are used inside the ports and supplemental 6-in.-thick iron shadow i
shields are imbedded in the concrete outside the ports (see Figure 4.1).
The thermal column and exponential experiment cavity above the beam tube are each closed off by removable 4-ft-thick stepped plugs of high-density concrete.
Further shielding for the cavity is furnished by lead plates and a boral sheath.
4.6 Dynamic Design Evaluation The safe operation of the CU TRIGA reactor is accomplished by the use and manipulation of control rods in response to observed changes in measured reactor parameters such as power, flux, and temperature provided by the instru-mentation channels.
In addition, interlocks prevent inadvertent -eactivity i
i additions and a scram system initiates rapid, automatic shutdown when safety i
settings are exceeded.
f Cornell University SER 4-3
Further stability and safety during both steady-state and transient conditions are incorporated into TRIGA reactors by virtue of the large, prompt, negative temperature coefficient inherent in the U-ZrH fuel-m derator material.
The x
negative temperature coefficient is primarily a result of the spectrum harden-ing properties of ZrH at elevated temperatures, which incraase neutron leakage x
from the fuel-bearing material into the water moderator where they are prefer-entially absorbed.
The reactivity decrease is a prompt effect because the fuel and ZrH are mixed intimately, and, thus, the ZrH temperature rises x
x essentially simultaneously with the reactor power.
An additional contribution to the prompt negative temperature coefficient is the Doppler broadening of 2380 resonances at high temperatures that increases nonproductive neutron capture in these resonances (Simnad et al., 1976; General Atomic Company (GA) 4314, 1980; GA-0471, 1958).
This inherent property of U-ZrH fuel has been the basis for designing TRIGA x
reactors with a pulse capability as a normal mode of operation.
The large, prompt, negative temperature coefficient rapidly and automatically compensates for step insertions of excess reactivity.
In the pulse mode it will terminate the resulting excursion without depending on electronic or mechanical safety systems or operator action.
In the steady-state mode it serves as a backup safety feature, mitigating the effects of accidental reactivity insertions (Simnad et al., 1976; GA-4314, 1980; GA-0471, 1958).
4.6.1 Excess Reactivity and Shutdown Margin The proposed Technical Specifications for the CU TRIGA reactor limit the excess reactivity to 4.00$ in the cold, xenon-free condition.
The Technical Specifications also require a minimum shutdown margin of 0.50$ with the highest worth control rod fully withdrawn.
The excess reactivity of the current CU TRIGA reactor core is 2.12$.
The control rod worths are 2.20$ for the shim rod, 1.99$ for the safety rod, 1.88$
for the transient rod, and 0.58$ for the regulating rod, yielding a total rod worth of 6.65$.
Under these conditions, the shutdown margin with the highest worth rod fully withdrawn is 2.33$ (6.65$, -2.20$, -2.12$).
Therefore, the current loading possesses an adequate shutdown margin.
4.6.2 Normal Operating Conditions In his license renewal application, the applicant has requested an increase in the authorized maximum steady-state power from 250 to 500 kWt and in the authorized maximum pulse reactivity insertion from 2.00$ to 3.00$.
The Technical Specifications prescribe a safety limit of 1,000*C as the maximum fuel temperature for stainless-steel-clad fuel elements.
This limit is estab-lished to provide confidence that the fuel elements will maintain their inte-grity and that no cladding damage will occur.
The safety limit for the high-hydride (ZrH,7) stainless-steel fuel elements is based on preventing excessive 2
stress buildup in the cladding because of hydrogen pressure from the disasso-ciation of zirconium hydride.
This limit is based on theoretical studies and a large mass of experimental evidence obtained during high performance reactor tests on these fuels (Simnad et al., 1976; GA-4314, 1980) and is consistent with those used at other TRIGA reactors.
Cornell University SER 4-4
To prevent exceeding the safety limit, the CU Technical Specifications estab-lish a limiting safety system setting (LSSS) for the scram setting on the instrumented fuel element temperature channel.
The proposed prescribed LSSS for the CU TRIGA reactor is 600 C for fuel elements in the B (hottest) ring as determined by the thermocoupled fuel element.
Lower settings are established for outer rings on the basis of observed' temperature distributions.
These values were selected to ensure that conditions would not arise that would allow the fuel element temperature to approach the safety limit.
The LSSS is set sufficiently below the safety limit to account for uncertainty in tempera-ture channel calibration as well as the differences between measured tempera-tures at the thermocouple location and actual local peak fuel temperatures.
On the basis of operating experience and supplemental calculations, the appli-cant has predicted maximum fuel element centerline temperatures of 292 C in the B ring at a steady-state power of 500 kW.
On the basis of experience at similar reactors, the applicant has estimated that a 3.00$ pulse would produce maximum measured fuel temperatures of 400 C in B-ring elements.
4.6.3 Conclusion The staff concludes that the inhere _nt large, prompt, negative temperature coefficient of reactivity of U-ZrH fuel m derator provides a basis for safe x
operation of the CU TRIGA reactor in the steady-state mode and is the essential characteristic supporting the capability of operation of the reactor in a pulse mode.
Furthermore, the Technical Specifications require that the core excess reac-tivity and experiment reactivity worths be limited so that the reactor can always be brought to a subtritical condition even if the highest worth control rod were totally removed from the reactor.
Current safety limits at the CU TRIGA reactor are based on theoretical and experimental investigations and are consistent with those used at other similar reactors.
Adherence to these limits provide confidence that fuel element integrity will be maintained.
CU's prediction of operating conditions at the new requested power level of 500 kW and pulse reactivity insertion of 3.00$ show the maximum fuel element i
temperatures to remain well below the current safety limits.
In addition, the applicant has submitted an operational plan that includes a step-wise approach to the new limits during which any deviations between observed and predicted conditions can be investigated and appropriate action taken.
TRIGA reactors similar to the CU reactor have demonstrated safe and reliable operation at steady-state power levels up to 1.5 MW and pulse reactivity insertions up to 5.00$ (Simnad et al., 1976; GA-4314, 1980).
Based on the above considerations, the staff concludes that under normal operating condi-tions there is reasonable assurance that the CU TRIGA reactor can be operated safely at 500 kW and 3.00$ pulses as limited by the requirements of Technical Specifications.
4.7 Functional Design of Reactivity Control System The power level in the CU TRIGA reactor is controlled by three standard control rods (one safety, one shim, and one regulating rod) and one transient rod, all of which contain boron as the neutron poison.
Rod movement is accomplished Cornell University SER 4-5
using rack-and pinion electromechanical drives for each standard control rod and a pneumatic electromechanical drive for the transient rod.
Each control rod drive system is energized from the control console through its own inde-pendent electrical cables and circuits, which tends to minimize the probability of multiple malfunctions of the drives.
Upon receipt of a scram signal, all four control rods will fall by gravity into the core, thereby shutting down the reactor.
4.7.1 Standard Control Rods The control rod drive assembly for a standard control rod consists of a non-synchronous, single phase electric motor coupled to a rack-and pinion drive system.
A draw tube connected to the rack supports an electromagnet that, in turn, engages an iron armature attached to the upper end of a long connecting rod.
The control rod proper is attached to the lower end of the connecting rod.
During normal operation the electromagnet is energized and the motorized system will insert or withdraw the control rod at a rate of approximately 11.4 in. per minute.
If power to the electromagnet is interrupted for any reason, the connecting rod is released and the control rod falls by gravity into the core rapidly shutting the reactor down (scramming).
Limit switches mounted on the drive assembly indicate on the control console the up (fully withdrawn) and down (fully inserted) positions of the magnet, the down position of the rod, and whether the magnet is in contact with the rod.
In addition, a helipot connected to the pinion generates position indi-cations for the shim and regulating rods that are displayed on the control console.
4.7.2 Transient Control Rod The transient control rod is operated by a pneumatic drive system that consists of a single-acting pneumatic cylinder whose piston is connected to the transient rod by a connecting rod.
For pulse operation, compressed air is admitted to the bottom of the cylinder through a solenoid valve, driving the piston upward in the cylinder and driving its connected transient rod out of the core.
At the end of its stroke the piston strikes the anvil of a shock absorber and decelerates at a controlled rate.
Adjustment of the anvil's position in relation to the piston head controls the piston's stroke length and, hence, the amount of reactivity inserted during a pulse.
The adjustment is performed manually at the rod drive housing where the position of the cylinder is dis-played also.
A limit switch is used to indicate at the control console the full-down position of the transient rod.
When the solenoid valve is deener-gized, air is vented from the cylinder, causing the transient rod to drop by gravity into the core.
In the steady-state mode an interlock prevents application of air to the tran-sient rod unless the shim and safety rods are in the full-in position.. Thus, in preparation for high power operation the transient rod is always withdrawn first, thereby becoming, in effect, an additional safety rod.
A key-operated switch has been installed to provide additional administrative control over pulsing operation.
Turning the rotary mode switch to the pulse mode will scram the reactor unless the key-operated switch is turned on.
Only authorized senior reactor operators are supplied access to the pulse mode switch key.
Cornell University SER 4-6
4.7.3 Scram-Logic Circuitry and Interlocks The scram-logic circuitry and interlocks ensure that several reactor core and operational conditions must be satisfied for reactor operations to occur or continue.
The scram-logic circuitry incorporates a set of open-on-failure-logic relay switches in series.
Any scram signal or component failure in the scram logic will result in loss of voltage to the electromagnets on the stan-dard ccntrol rods and solenoid on the transient rod causing these rods to scram and shut down the reactor.
A voltage sensor is installed in the scram relay power supply circuit to scram the reactor if the scram relay power supply voltage drops below a point that will not allow proper operation of the scram relays.
The Technical Specifications for the CU TRIGA reactor require the operability of one fuel element temperature scram in both the pulse and steady-state modes and two reactor power level scrams in the steady-state mode.
A manual scram also is required to allow ~the operator to shut down the reactor if an unsafe or abnormal condition occurs.
In addition to the scrams required by the Techncal Specifications, a short period scram in the steady-state mode and an integrated power scram in the pulse mode also are available.
On initiation of a scram, the control rods must insert 90% of their reactivity worth in less than 1 sec.
Appropriate surveillance checks, tests, and calibra-tions are required to verify continued operability and satisfactory performance of the scram functions.
Several safety interlocks are incorporated into the control rod circuitry to prevent inadvertent reactivity insertions.
During steadv-state operation, the interlocks prevent the simultaneous withdrawal of two or more standard control rods and the withdrawal of the trar.sient rod if the safety or shim rods are not fully inserted.
In the pulse m3de only the transient rod can be moved.
In addition, control rod withdrawal n, not allowed unless an adequate source signal is available to allow proper startup of the reactor.
The CU TRIGA reactor is equipped with safety and control systems typical of most nonpower reactors.
The control rods, rod drives, scram circuitry, and interlocks have performed reliably and satisfactorily in this reactor for many years, and similar equipment has shown satisfactory performance in many other TRIGA reactors over a long period of time.
The control system allows for an orderly approach to criticality and for safe shutdown during both normal and emergency conditions.
There is sufficient redundancy of control rods to ensure safety shutdown even if the most reactive rod fails to insert upon receiving a scram signal.
Interlocks are provided to preclude inadvertent rod movement that might lead to hazardous conditions; independent scram sensors and circuits, incorporated to automatically shut the reactor down, mitigate consequences of single malfunctions.
Several manual scram buttons allow operators to initiate a scram from strategic locations, including the control console and experimental areas.
4.7.4 Conclusions Because of the redundancy of shutdown mechanisms, the interlocks provided, and the inherent safety associated with the large, prompt, negative temperature Cornell University SER 4-7
l coefficient of reactivity inherent in the U-ZrH fuel m derator (discussed in x
Section 4.6) that automatically terminate or limit reactor transients, reactor transient accidents can be safely mitigated.
Therefore, the staff concludes that the reactivity control systems of the CU reactor are designed and function adequately to ensure safe operation and safe shutdown of the reactor under all operating and transient conditions.
4.8 Operational Procedures The applicant has implemented administrative controls that require review, audit, and written procedures for all safety-related activities.
A Ward Laboratory Safety Committee reviews all aspects of the CU TRIGA reactor opera-tion to ensure that the reactor facility is operated and used within the terms of the facility license consistent with safety of the public as well as of the operating personnel.
Responsibilities of this committee include review and approval of operating procedures, new experiments, and proposed changes to the facility or its Technical Specifications.
At least once a year the committee inspects the facility, reviewing safety measures and auditing operations.
Written procedures (reviewed and approved by the Ward Laboratory Safety Com-mittee) have been established for safety-related activities including reactor startup, operation, and shutdown; preventive or corrective maintenance; and periodic inspection, testing, and calibration of reactor equipment and instru-mentation.
The CU reactor is operated by trained NRC-licensed personnel in accordance with the above mentioned procedures.
In support of his request to increase the maximum authorized steady-state power level to 500 kW and maximum authorized pulse reactivity insertion to 3.00$, the applicant has submitted a plan for preoperational tests and initial operation at 500 kW and 3.00$ pulses. This plan includes adjustments and calibrations required for operation at the higher limits and a step-wise approach to the new limits during which important parameters, such as fuel element temperatures, control rod positions, and resulting radiation levels, are measured and recorded.
The staff has reviewed this plan and concludes that it provides for a safe and orderly approach to the proposed new limits and that any unexpected safety-related problems will be observed before a hazardous condition develops.
In addi' tion, there are currently 11 TRIGA reactors operating at 1 MW or greater with no safety-related problems.
4.9 Conclusion The staff review of the CU TRIGA reactor has included its specific design and installation, its control and safety instrumentation, and its specific preoper-ational and operating procedures.
On the basis of the review of the CU TRIGA reactor and experience with other similar facilities, the staff concludes that there is reasonable assurance that the CU reactor is capable of safe operation at 500 kW steady state and a maximum pulse insertion of 3.00$, as limited by its Technical Specifications, for the period of the license renewal.
Cornell University SER 4-8
I s
bdl 4.-
. =.
n Q
2 u
- s.
W A*
r ruou Hoto
?
'. E a enanimaz 2
/
Y.
tD s
oao,- r anawa c*oucears concerra 4
/
m i
4
-C,,
- i; a g*;.*f, *, :
4 g
k'.;:,s;-
',:p.y/*4.:*3.'**
titit* slow
' ~
- 'I*'"
Y i;
,.,,,,, y,a,,
.q 3.
,/ f
,Y-5,3. ;
- 8
~'e s
.. 4
}.
m,,
,i
/
N
., ; ': " W *-
x n.,
courua
^..
/
i
- i
,/
gra, &.,r-ar-roa no.-
,. -'x avenut a.
4 unna
'/
R'.RM
^
_.g.
mN n
g 1A
~
,A A s * *~~ r,
f l
~~
V weiaert l$ '. : 21.
'.::.,*'.Q
?
^
i esnecron r
m ronin a,oe Pluc_
45.
v.
hiH h
M *~'
f1(.h ff e.
, :.c. a G
y enemale.eja.
a : b:
..[*.. ". M ' '
I1..~'
- i?.*.sg)^4
'\\
A T,.
- h..-<,.:s).:: ( l- := y;.;._";ll[;
.Q l }
j ;
F a....
5 y
.,}l" flb= {,.*:'{j i.'.
hll'
\\
atw<w, cam.
a;-
t
- : ~. *..
- c. sua w.wirto SLIDING LEAD LM&
m,g pg,,,,,,
STorace wcu L,-
. b.
Figure 4.1 Sectional view of CU TRIGA reactor tank
5 REACTOR COOLANT AND ASSOCIATED SYSTEMS The coolant system for the CU TRIGA reactor consists primarily of a nonpres-surized pool, a heat exchanger system, a demineralizer system, and associated pipes, valves, pumps, and gauges.
Approximately 21,800 gal of light, deminer-alized water is maintained in the reactor pool and is pumped through a heat exchanger and demineralizer system.
5.1 Primary Cooling System The primary cooling system is composed of the reactor tank, a single-stage centrifugal pump, a 500-kW heat exchanger, a flowmeter, and associated valves and piping.
The primary coolant is demineralized water.
To ensure safe and reliable operations, the primary cooling system is provided with instrumenta-tion that gives the reactor operations personnel timely information concerning pool water level, pool water temperature, and the pressure difference between the secondary cooling water and the primary reactor pool water in the heat exchanger.
Coolant leakage from the reactor tank can be readily detected by the abnormal decrease in the level of water, followed by pump cavitation.
(During the time period of operation there has been no history of tank leakage.)
A schematic of the primary cooling system is given in Figure 5.1.
5.2 Heat Exchanger and Secondary Cooling Water System The secondary pool water cooling system, shown in Figure 5.1, uses campus-supplied chilled water in the secondary loop.
To guard against outleakage, the pool water is maintained at a lower pressure than the chilled water.
The chilled water is normally at 100 psi, and the shutoff head of the pool water pump is 32 psi.
However, because of the possi-bilities of reduced chilled water pressure, a pressure switch is installed in the chilled water line that shuts off the pool water pump if the chilled water pressure falls below 50 psi.
A conductivity probe is installed in the pool water line immediately downstream l
of the heat exchanger to detect inleakage of chilled water.
The presence of l
16N can be detected by a NaI crystal placed near the suction line at the top of the pool.
5.3 Primary Water Purification and Makeup System The demineralizer system is composed of a surface skimmer, pump, filter, flow 1
meter, demineralizer manifold, instrumentation, and associated piping.
This system is used to remove particulates and chemical impurities, to maintain the optical clarity of the water, and to provide makeup water to the primary cooling system.
Cornell University SER 5-1
t 5.4 Nitrogen-16 Diffuser The existing reactor does not have an 16N diffuser because of its low power.
It is anticipated that at the proposed power level of 500 kW a diffuser will still not be required.
However, if one is required the applicant indicates that the plumbing is available to immediately install the necessary piping and deflector.
5.5 Conclusion The staff has examined the design and installation of the coolant system of the CU reactor.
The design and operational characteristics of the system are simple and closely resemble those of cooling systems at other reactor facili-ties.
These factors, coupled with the tcmperature-monitoring capabilities of the instrumentation system as described in Section 7, lead the staff to con-clude that the coolant system at the CU TRIGA reactor facility is adequate for safe reactor operations.
1 4
Cornell University SER 5-2 '
n l
O W - MODULATING CONTROL VALVE y
l - BUTTERI'LT VALVE C - CONoucTaviTy PRost c
F - ANNUBAR IW ELEMENT 3.
P - PRESSURE GUACE REACTOR
'h h
BAY WALL T - THERHOMETER j
]
8 i
p - PRESSURE SWITCl!
,p 3
- TEMPERATURE CONTROLLER m
1 m
VI -
2 82*i k l
N p
QO Q O*
QT Q" V5 s
,V71 Al I
I
.i sse W' INTER 8
122 CPM I
SUPPLY l
HEAT g
y9 g;,.,, CHIM EXCHANGER 8
.. WATER l
f p
W I
anaN Vil V12 I
V8 3" STEEL PIPE 69*
V6 e
L M
s'N v
i 3
l V
I y mos' i
e 3Vw
...a
(.......
I 8
'8
& 6 i'
00 i
i i
i TRICA P00f. WATER C00 Lit!C SYSTEM Motet temperatures are for winter operation, summer temperatures are 78 cooler.
Figure 5.1 Pool water cooling system i
6 ENGINEERED SAFETY FEATURES The engineered safety features associated with the CU TRIGA reactor.
These include the reactor building heating and ventilation system and the contamina-tion control features.
i 6.1 Reactor Building Heating and Ventilation System The reactor bay is air conditioned for comfort and temperature control.
This system maintains an upper limit for humidity in the bay.
Heating is provided by hot-water reheating coils located in the duct system of the air conditioner.
Hot water radiators of the enclosed, finned-tube type also are used.
Approxi-mately 15% fresh air makeup is provided by the ventilation system and exhaust ducts with manually operated valves to maintain atmospheric pressure.
Valves located in the intake and exhaust ducts can be manually activated to isolate the reactor room (see Section 6.2.2 for details).
Heating for the upper floors of the office and laboratory wing of the building is provided by hot water radiators.
The basement of the wing is provided with forced-air ventilation combined with hot water heating units.
There is no air conditioning for this portion of the building, except for an air-conditioning unit located in the counting room.
An exhaust fan is provided in the classroom.
6.2 Contamination Control Features The building incorporates a number of features especially designed to facili-tate control of any release of radioactive material resulting-from either normal or accident conditions.
6.2.1 Normal Operating Conditions Normal activities in the building include the use of low-level radioactive-materials in experiments (for example, as tracers) and the production of radioisotopes.
The experiments are carried out under standard laboratory procedures for handling radioactive material.
Hoods with absolute filters are provided in the chemistry laboratory for use in experimental operations in which airborne radioactivity might be released and covered drums in the chemistry laboratory and elsewhere are used for collecting any resulting active liquid wastes.
Separate covered drums for collection of swipes, contaminated glassware, and other solid wastes also are furnished.
Disposal of the drums follows govern-ment and university regulations.
l l
A special room is provided for handling samples irradiated in the reactor.
Hoods and drums, like those described above are provided in this room.
After irradiation, samples in this room can be checked for activity level, surface contamination, vial breakage, and other parameters.
i Cornell University SER 6-1 l
The intense neutron flux in the TRIGA reactor core will produce radioactivity in the water and in the small amounts of air dissolved in the water that is irradiated by the core (specifically,18N in the water, 41Ar in the air, and miscellaneous activities in any corrosion products or foreign matter in the water).
Control of the 16N does not require special measures because its short half-life (7.1 sec) guarantees its decay to negligible levels before personnel can come in contact with it.
Control of the 41Ar, which has a 109-min half-life, is achieved by a venting system that continuously removes irradiated air from the reactor, dilutes it with large amounts of unirradiated air from the reactor bay, and discharges it from the building through a roof vent.
Control of activated water impurities is achieved by two measures:
(1) the water admitted to the core first is demineralized to keep the concen-tration of activatable impurities very low, and (2) the same purification system is used to remove those activated impurities.
When the demineralizing resins are regenerated, the effluent is monitored; if the level is above permissible values for immediate disposal, the effluent is pumped to a waste storage tank to decay while awaiting ultimate disposal.
6.2.2 Accident Conditions Accidents releasing radioactive material may range from minor spilling or breaking of isotope containers to a possible reactor accident.
To provide control of water-borne activity from spills, all drains in the reactor bay, including floor drains, empty into a sump in the reacter equip-ment room, where the activity can be monitored.
The level of activity will determine the disposal method in the same manner as described earlier for demineralizer effluent.
When the presence of airborne activity from container breakage is indicated by air monitors or other warning, the reactor bay ventilation system may be sealed off by closing valves in both intake and exhaust ducts by using any one of three manually operated switches.
These switches, which are located outside the re-actor bay at each of the doors connecting the bay with the rest of the building, also actuate a radiation alarm horn in the bay.
Closing off the ventilation system effectively isolates the entire reactor bay because of the following de-sign features:
(1) The bay is of reinforced concrete construction painted internally with a vapor barrier seal.
(2) There are three personnel doors and one freight door, all of which are gasketed, and there are no windows.
(3) All penetrations of the walls of the bay by pipes, conduit, and so on are caulked or otherwise sealed.
6.3 Conclusion The engineered safety features for the CU TRIGA reactor are designed primarily to prevent or mitigate the consequences of accidental radioactive spills and/or gaseous releases.
The ventilation system and reactor area isolation system, when operated in conjunction with strict procedures, appear to be capable of 1
i Cornell University SER 6-2
coping with any credible scenario under normal and accident conditions.
There-fore, the staff concludes that the engineered safety features are adequate to ensure safe operations for the period of the license.
l l
l l
i Cornell University SER 6-3
i 7 FACILITY CONTROL AND INSTRUMENTATION SYSTEMS The control and instrumentation systems at the CU TRIGA reactor provide auto-matic and manual scram capability in case of reactor malfunction.
A schematic of the CU reactor instrumentation and scram system is shown in Figure 7.1.
7.1 Control Console The reactor console contains the control, indicating, and recording instrumen-
[
tation required for operation of the reactor.
This ir:strumentation is located, as indicated, on either side of the dual pen recorder and operator's panel.
l On the operator's panel are (1) rod position indicators to show the position of the shim and regulating rods to within 0.2%; (2) control-rod buttons for the shim, safety, and regulating rods to control the position of each rod and the indicator lights to indicate the up and down positions of each rod and rod-magnet control; (3) a control button and indicator light for the safety-transient rod; (4) monitor alarm lights; and (5) additional pilot lights to indicate power on, cooling system on, and inadequate source strength.
Other annunciator lights on the console indicate the source of a scram signal.
A scram is initiated by (1) excessive power level on either of the power-level channels, (2) too short period, (3) ion-chamber power-supply failure, (4) oper-ating power failure, or (5) manual scram.
Following a scram or the dropping of any red (except the safety-transient rod) and after the rod reaches the full-in position, the drive mechanism autocatically follows the rod down to re-establish contact.
Table 7.1 shows the safety system channels and their functions.
7.2 Control Rods Four boron carbide control rods (one safety-transient, one safety, one shim, and one regulating) operate in perforated aluminum guide tubes.
The upper end of the control rod screws into the control-rod-drive-assembly extension tube.
The control rods are approximately 20 in. long.
The outside diameter of the regulating rod is 0.875 in., and the shim and safety-transient rods each measure 1.25 in. in diameter.
The vertical travel of the control rods is approxirately 15 in.
7.3 Control-Rod-Drive Assemblies The control-rod-drive assemblies for the shim, safety, and regulating rods are mounted on the center channel assembly and consist of a motor and reduction gear driving a rack and pinion.
A helipot connected to the pinion generates the position indication.
Each control rod has an extension tube that extends to a dashpot below the surface of the water. The dashpot and control rod assembly are connected to the rack through an electromagnet and armature.
In the event of a power failure or scram signal, all of the control rod magnets are deenergized and the rods fall into the core.
The rod-drive motor is nonsynchronous, single phase, and instantly reversible and will insert or Cornell University SER 7-1
L.
!~
withdraw the control rod at a rate of approximately 11.4 in. per minute.
Electrical dynamic and static braking on the motor are used for fast stops.
A i
control rod cannot be moved manually without major changes in the mechanical system.
Such changes are subject to strict administrative control.
Limit switches mounted on the drive assembly indicate the up and down positions of the magnet, the down position of the rod, and whether the magnet is in contact with the rod.
The transient-rod drive used for pulse operation is pneumatic.
It consists of a piston within a cylinder.
The transient rod is fastened to the piston, and the stroke of the piston is determined by the position of the cylinder.
In the deenergized position, the transient rod is complet.ely within the reactor core and is held there by gravity.
Application of pressurized air to the cylinder causes a rapid withdrawal of the rod for a distance determined by the l
prior position of the cylinder. With this scheme, any length of rod may be i
withdrawn (from zero to the full-rod length of approximately 15 in.).
The transient rod drive is used to expel the transient rod from the core in less than 0.1 sec by means of pneumatic pressure.
The drive assembly is illustrated schematically in Figure 7.2.
To prevent control rod withdrawal, this unit is fitted with a pressure switch in the air line.
This switch also energizes a warning light on the control console if the air pressure drops j
below 27 psig.
l j
7.4 Pulse-Mode Switch A key-operated switch is installed on the TRIGA console for the purpose of greater administrative control over pulse-mode operation. Without this switch in the on position, the reactor will scram when the operate-mode switch is turned to pulse mode.
Only authorized senior reactor operators have access to the pulse-mode switch key.
7.5 Instrumentation System The instrumentation system is composed of standard nuclear and process instru-mentation to provide operations personnel with timely and accurate information.
Included in the console are a startup, a log-N and period, and two power level channels; a log-N and power. level recorder; an crcimatic power control system; and a water radioactivity, a water conductivity, water temperature monitor,-
and radiation moniturs.
7.6 Scrams l
The CU TRIGA reactor is provided with several scram-logic circuits that are interlocked with the control system as discussed in Section 4.7.3.
The scram i
settings are indicated in Table 7.1.
Included also in the scram and. interlock system are three manual.ccram switches that are accessible to experimenters in the experimental area (one each on the east, south, and west external shield walls of the CU reactor). aAs an addi-tional safety feature, a voltage-sensitive device is installed in the scram relay power supply circuit for the purpose of scramming the reactor when.the-I Cornell University SER 7
scram relay power voltage drops below a point that will not permit proper operation of the scram relays.
7.7 Conclusion The control and instrumentation system at the CU reactor is very similar to those at other TRIGA facilities in the U.S.
The system is designed to provide timely information to the operator concerning both the nuclear and nonnuclear processes taking place during operations.
In addition, the control system allows for an orderly approach to criticality and for safe shutdown during both normal and emergency conditions.
There is sufficient redundancy of control rods to ensure safety shutdown even if the most reactive rod fails to insert upon receiving a scram signal.
Interlocks are provided to preclude inadvertent rod movement that might lead to hazardous conditions; independent scram sensors and circuits, incorporated to automatically shut the reactor down, mitigate consequences of single malfunctions.
Several manual scram buttons allow operators to initiate a scram-from strategic locations, includ-ing the control console and experimental areas.
Following examination and analysis of the system, the staff concludes that the control and instrumenta-tion system at the CU TRIGA reactor is adequate for safe reactor operation, under all operating, transient, and unplanned conditions.
(
Cornell University SER 7-3
t
(")
O 88'0" 3
START-UP CHANNEL PRE-LOG AMPLIFIER g LOG COUNT-R ATE COUNTER AMPLIFIER COUNT RATE
- SOUR CE METER g
a lNTERLDCK a
C I
I LOG n AND PERIOD COMPENSATE D LOGE PERIOD l~T LOG n AND G Af4NEL CHl.MBER AM PLIFIER HCUf7 j
RECORDER w
l-n.
r+
3' M
PERIOD METER m
i POWER-LEVEL COMPENSATED POWER-LEVEL i 8 CH AN NEL NO. y CHAMBER MONITOR T
I j
T r UNEAR POWER MAGNET g TO ROD R CRDER SCRAM
.- MAGNET j
i I
1 I
i POWER-LEVEL UNCOMPENSATED S
POWER-LEVEL l
I f W POWER METER CHANNEL NO 2 CH AM BER D
YONITOR g
g i
3 T
I l
I PEAK POWER l
3 i
AND b
3 p.-..-.-.-."..-.-.-...-..-..-.-.-.-------..
FUEL ELEMENT l
i l,
N(THERMOCOUPLESP POWER DEMANDI j
i
, _ _ _ _ - _ -., i
--t SERVO-g REGULATING.
I t.- _ONTROL_ --- i []7] 7 AMPUFIER.s_,* Roo I
/
8 C
1 i---.-.
+-
SCRAM f
a f TEMPERATURE METER l TEMP. PROBE l
+
Se (WATER AND FUEL ELEMENT)
WATER-MO NITOR COUNT-R ATE lGElGER COUNTER COUNT-RATE METER CHANNEL CIRCUlf
=
i WATER-CONDUCTIVITY lCOND. PROBE' MONITOR
- STEADY-STATE OPERATION (S)
-- TRANSIENT OPERATION (T)
-- AUTOM ATIC POWER CONTROL (OPTIONAL)
Figure 7.1 Block diagram of instrumentation and scram system
SHOCK ABSORBER VENT HOLES EXTERNALLY THREADED CYLINDER PISTON SUPPORT rm rvi I
i i
i 1
1 I
I il L_
IJ i
/
J g_
M BEARINGS WORM
=
/
s NW WORM GEAR a
8 s
[
=
HOUSING
~
o e_r PISTON ROD
=
VENT SUPPLY AIR SUPPLY HOSE SOLENOID VALVE "b
BOTTOM LIMIT c
CONTROL ROD Figure 7.2 Schematic of transient-rod drive assembly Cornell University SER 7-5
Table 7.1 Safety System Channels Minimum Safety System Channel Number Operating Mode or Interlock Operable Function Required Fuel element temperature 1
- Both modes Reactor power-level 2
Scram Steady-state mode Manual button 1
Scram Both modes Startup count rate 1
Prevent control rod Reactor interlock withdrawal when startup neutron count rate is less than 1/sec Standard control rod 1
Prevent withdrawal Steady-state position interlock of the transient rod mode when either the safety or shim con-trol rods are not fully inserted
- The drop time of a standard control rod from the fully withdrawn position to 90% of full reactivity insertion is less than 1 sec.
i
(
. Cornell University SER 7-6 l
l L
i 8 ELECTRICAL POWER SYSTEM The electrical power system at the CU TRIGA reactor facility is a standard and well-accepted electrical supply system designed and constructed to specifi-cations similar to those at other research reactor facilities.
8.1 Main Power Electrical power for building lighting and equipment power is 120/208 V, three phase, four wire, 60 cps.
The total estimated power requirement for the facility is 300 kVA.
The main power control panel is located in the electrical utility room, with subpanels locat d as required in other areas.
Fluorescent lighting is used throughout the bu lding except in the counting room, where incandescent lighting is used, and in the high-ceilinged part of the reactor bay, where mercury vapor lights are installed.
In addition, mercury vapor lights have been provided in the gamma irradiation cell.
8.2 Emergency Power Because the CU TRIGA reactor will scram in case of a power interruption and the decay heat generated in the core after scram is minimal, no emergency power is supplied except battery-operated emergency lighting for personnel movement during an emergency.
8.3 Conclusions The above factors, and the fact that the reactor will scram in the event of power failure and that ambient pool cooling can remove decay heat, leads the staff to conclude that the electrical power system is adequate for safe opera-tion of the CU TRIGA reactor for the time period of the license.
l I
Cornell University SER 8-1 2
9 AUXILIARY SYSTEMS 9.1 Fire Protection System Fire protection is provided by external fire hydrants and by portable extin-guishers located throughout the facility.
An automatic internal fire alarm system with heat-sensing elements in nearly every room is tied into the main university system.
A builtin CO gas-extinguishing system is part of the 2
gamma cell and can be operated from the cold work area.
9.2 Communications System Several communication systems are installed in the facility.
A public address system is provided, and there are microphones in the-TRIGA control room and in the supervisor's office.
An internal dial telephone system with stations suitably located throughout the building is provided.
Standard commercial telephone service connects into the main university system.
9.3 Compressed Air System The compresseo air system consists of a compressor located in the mechanical utility room and a series of pipes, regulators, and valves to provide com-pressed air to the reactor facility.
9.4 Equipment and Fuel-Handling Systems The reactor section of the facility is serviced by a pendant-operated 5-ton crane. Other specialized equipment is provided to service the reactor cores and the irradiation facility.
This equipment includes handling casks for radioactive samples, a handling and shipping cask for spent TRIGA fuel ele-ments, beam port plugs, and handling equipment for fuel elements for both reactors.
The equipment and fuel-handling systems are designed to enable personnel to service the reactors with a minimum of exposure-to radiation and other hazards.
9.5 Conclusion The auxiliary systems in use at the CU reactor facility are designed and used to assist in safe reactor operations, and to mitigate the consequences of accidents.
The handling and shipping casks for radioactive materials meet all existing regulatory requirements.
Design, workmanship, component quality, and mainte-nance levels are very high in all the auxiliary systems.
All the above factors, coupled with the administrative controls used in opera-tions at the CU reactor facility, lead'the staff.to conclude that the auxiliary systems are adequate to support the CU TRIGA reactor in a safe and. reliable manner.
l Cornell' University SER 9-1
v 1
10 EXPERIMENTAL PROGRAMS The CU TRIGA reactor is used on the average of 10 to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> per week, primar-ily for teaching and nuclear research purposes.
In addition to inpool irradia-tion capabilities, the experimental facilities include a pneumatic transfer system, several beam ports, a through port, and a thermal column.
10.1 Experimental Facilities 10.1.1 Pool Irradiations The open pool of the reactor permits the irradiation of experiments submerged in the vicinity of the core.
The decision to perform experiments in the reactor pool (as opposed to using the pneumatic transfetr system or a beam tube) is dictated by specimen size and the desired type and intensity of radiation fields.
Specialized positions include a wet central thimble in the center of the reactor core and three dry irradiation tubes within the graphite reflector.
The irradiaiton tubes are dry to preclude the necessity of sealing capsules against water leakage and thereby reducing the possibility of pressure buildup as a result of radiolysis or other chemical reactions.
The central thimble
~;
and the irradiation tubes are provided with offsets to prevent streaming of radiation from the reactor core.
To allow for placement of large experiments close to the reactor core, the reflector is provided with a flat face exposed to the bulk pool side of the reactor.
Thus, approximately 5 in. of graphite separate the large experiment from the core region of the reactor.
The actual placement of experiments or samples in the core region or the reactor pool is limited by the Technical Specifications.
10.1.2 Pneumatic Transfer System 1
(
The pneumatic transfer system allows small sealed samples to be transported
[
rapidly between the reactor and the radiochemistry laboratory.
The irradiation i
terminus is in the outer ring of fuel-element positions.
The exhaust air is filtered before it is discharged to the atmosphere through the building exhaust stack.
10.1.3 Beam Ports Eight horizontal penetrations through the shield wall at core level permit beam experiments and the insertion of samples for irradiation.
These include two 3-in. thermal beam ports, four 6-in. radial beam ports, and a 4-in. through-port. Two of the radial beam ports are perpendicular to the longitudinal axis of the reactor structure, and the remaining two are at a 45* angle to the axis.
The two diagonal radial beam ports extend as far inward as the outer face of the reflector can; inside the can, the graphite in line with the port is cut out, forming a 6-in.-diam cavity extending to the inner face of the Cornell University SER 10-1 a)
can, so that a beam consisting primarily of core spectrum neutrons is avail-able for experiments. The two perpendicular radial beam ports pierce the outer reflector can and the graphite reflector and terminate at the inner can, permitting insertion of samples inside the reflector.
The 4-in. through port penetrates both the reflector can and the graphite reflector; its axis is parallel to the perpendicular beam holes on a line passing the reactor core on the side opposite the thermal column 11.7 in, from the core centerline.
Two-3-in. ports are provided for the extraction of neutrons from the vertical sides of the thermal column.
These ports are located at a 45 angle to the longitudinal axis of the reactor structure.
Shielding is provided for each beam port to reduce the radiation outside the shield to acceptable levels when the port is not in use.
For the 4-in. through-port and the four 6-in. ports, the shielding consists of an inner concrete plug, an outer wooden plug, e lead shutter, and a steel door; only a concrete plug and a steel door are pravided for the 3-in. ports.
In all cases the inner concrete plug is stepped to prevent radiation streaming.
Shielding around the 4-in. and 6-in. ports is supplemented by 6-in.-thick iron shadow shields embedded in the concrete at the step in the penetrations. When the beam ports are in use, external shield walls, beam stops, or beam catchers are installed to control radiation levels in the experimental areas.
10.1.4 Thermal Column A graphite thermal column approximately 4 ft by 4 ft by 8 ft extends from the outer face of the reflector assembly into the concrete shield structure.
Horizontal access and shielding for the thermal column are provided by a track-mounted heavy concrete door.
There are two nesting coaxial removable plugs, square in cross section, in the center of the thermal-column door.
Inside the thermal column near the inner end are a 32-in. by 32-in. by 36-in.
hohlraum and a 24-in. by 24-in. by 28-in. removable graphite block.
Above the hohlraum there is a 3-ft 7-in.-square by 4-ft-deep cavity for experiments requiring vertical access to the thermal column. This cavity is shielded from the thermal column by about 6 in. of lead and is closed off from above by a removable 4-ft-thick stepped ilmenite concrete plug.
10.2 Experimental Review Before any new experiment can be conducted using the reactor or experimental facilities, it is reviewed by members of the Ward Laboratory Safety Committee.
This review and approval process for experiments allows personnel specifically trained in reactor operations to consider and recommend alternative operational conditions--such as different core positions, power levels, and irradiation times--that will minimize personnel exposure and/or the potential release of radioactive materials to the environment.
10.3 Experiment Reactivity Limitations The Technical Specifications include a 2.00$ limit of reactivity for each experiment and a maximum of 3.00$ reactivity insertion for all experiments in
(
the reactor.
In addition there is a limitation of 0.90$ of maximum negative reactivity of all experiments in the reactor.
Cornell University SER 10-2
_ = _ _ _ -. _ -.-
~.
9
)
In Section 14.1, the staff evaluated the consequences of inadvertent reactivity insertions of values higher than the above Technical Specification limitations.
The staff concluded that the reactivity limitations in the Technical Specifica-tions produce fuel and cladding temperature increases well below the critical limits.
i 10.4 Conclusion The staff concludes that the design of the experimental facilities combined with the detailed review and administrative procedures applied to all research activities is adequate to ensure that experiments (1) are unlikely to fail, (2) are unlikely to release significant radioactivity to the environment directly, and (3) are unlikely to cause damage to the reactor systems or its fuel.
Therefore, the staff considers that reasonable provisions have been made so that the experimental programs and facilities do not pose a significant risk j
of damage to the reactor or of uncontrolled release of radioactive materials.
L i
i it I
t r
i 1
L Cornell University SER 10-3
. ~
F i
I i
11 RADI0 ACTIVE WASTE MANAGEMENT 4
The CU zero power reactor and the CU TRIGA reactor are housed in the same building, and any radioactive wastes generated by their use are combined.
Therefore, this chapter addresses the management of radioactive waste resulting from the operations of both reactors.
The major radioactive waste generated by reactor operations is activated gases, principally 41Ar produced by neutron irradiation of air.
A small volume of radioactive solid water, primarily resins, is generated by reactor i
operations, and some additional solid waste is produced by the associated research programs.
No radioactive liquid wastes are generated directly by 4
i normal reactor operations.
However, liquid radioactive waste is produced by i
the regenerction of the resin bed in the water demineralizer system.
Addf-tional small amounts of radioactive liquid waste are developed as a result of several of the reactor-based research activities.
j 11.1 ALARA Commitment l
The Ward Laboratory Safety Committee instructs all personnel to develop pro-cedures to maintain the generation and possible release of radioactive waste materials to a level as low as reasonably achievable (ALARA).
Examples of adherence to the ALARA commitment include (1) the installation of j
a check valve in the CU TRIGA reactor pneumatic transfer line to minimize unnecessary 41Ar releases and (2) the change in the zero power reactor cell ventilation control system to require manual restart after reactor operations.
11.2 Waste Generation and Handling Procedures 4
11.2.1 Solid Waste Solid waste generated as a result of reactor operations consists _primarily of j
ion exchange resins and filters, potentially contaminated paper and gloves, and occasional small, activated components.
Some of the reactor-based research j
results in the generation of solid low-level radioactive waste in the form'of contaminated paper, gloves, and glassware.
This solid waste generation typi-cally has contained a. few millicuries of radionuclides per year.
4 Solid waste is collected by the CU health physics staff, combined with other-university generated waste, and held temporarily before being packaged and shipped to an approved disposal site in accordance with applicable regulations.
11.2.2 Liquid Waste i
Normal operations of these reactors produce no radioactive liquid waste.
However,- some of the. research~ activities are capable-of generating limited volumes of such waste, and the sinks in the two laboratories are connected to portable, disposable drums.
Cornell University SER-11-1
4 All drains in the reactor bay lead to the sump in the reactor equipment room.
Sump contents can be pumped to a 2,000 gal waste storage tank for decay if significant radioactivity is detected.
The largest volume of potentially contaminated water is produced by the regen-eration of the demineralizer. This periodically generated effluent also can be discharged to the 2,000 gal waste storage tank.
Before any releases to the sanitary sewer system, representative samples are collected from the sump or from the storage tank and analyzed by standard techniques.
If the concen-trations of radioactive materials in the water are less than the guideline values of 10 CFR 20.303, the contents are discharged directly to the sewer.
If the activities exceed the values of 10 CFR 20.303, additional dilution water may be added to reach acceptable concentrations.
11.2.3 Airborne Waste-The potential airborne waste is composed of gaseous 41Ar and neutron-activated dust particulates.
No fission products escape from the fuel cladding during normal operations.
Radioactive airborne emissions are produced ' principally by the neutron irradiation of air and airborne particulate materials in the reactor thermal column and beam ports. This air is swept constantly from the beam room and discharged to the environment through the Ward Laboratory exhaust stack where it is diluted with air from the ventilation syste'n at the rate of 3
2,200 ft per minute.
A stack monitoring system measures the gaseous concentrations in the effluent.
During normal operations, no measurable radioactive particulates are released in the air effluents from this stack.
Cornell has measured the release of 41Ar over the years with gas-sampling instruments calibrated with known quanti-ties of 41Ar.
During the years since the ' reactor was first licensed, CU has reported an average annual release of less than 0.2 Ci of 41Ar.
Both the applicant's and the staff's evaluations show that this amount of release would lead to exposures in unrestricted areas that are small fractions of the limits specified in 10 CFR 20.
In accordance with the ALARA principles, Cornell has committed itself in its Technical Specifications to several conditions that will minimize 4 tar' production and subsequent release.
I 11.3 Conclusions The staff concludes that the waste management activities of tnis reactor facility have been conducted and are expected to continue to be conducted in a manner consistent with 10 CFR 20 and with the ALARA principles.
Among other guidance, the staff review has'followe'd the methods of ANSI /ANS 15.11, " Radio-logical Control at Research Reactor Facilities," 1977.
Because 41Ar is the only potentially significant radionuclide released by the
~
reactor to the environment during normal operations.
The staff has reviewed the history, current practice, and future expectations of the quantity and handling of 41Ar at the-CU facility and concludes that the doses in unrestricted areas as a result of actual releases of 41Ar have been and are expected to remain a small percentage of the limits specified in 10 CFR 20 when averaged over a year.
Furthermore, the staff's conservative computations of the dose beyond the limits of Ward Laboratory give reasonable assurance that potential 3
doses to the public as a result of 41Ar would not be significant even if there were a major change.in the operating schedule of the CU reactors.
Cornell University _SER-
,11-2 I
i
12 RADIATION PROTECTION PROGRAM The CU zero power reactor and the CU TRIGA reactor are housed in the same build-ing, and the radiation protection programs are a combined project.
Therefore, this chapter addresses the radiation protection policies and practices as they affect the operations of both reactors.
Cornell University has a structured radiation safety program with a health physics staff equipped with radiation detection instrumentation to determine, control, and document occupational radiation exposures at its reactor facility.
In addition, the reactor facility has monitors to identify both liquid and air-borne effluents at the points of release to comply with applicable guidelines.
12.1 ALARA Commitment As stated in Section 11.1, the Ward Laboratory Safety Committee has formally established the policy that operations are to be conducted in a manner to keep all radiation exposures ALARA.
All proposed experiments and procedures at the reactor are reviewed for ways to minimize the potential exposures of personnel.
All unanticipated or unusual reactor-related exposures will be investigated by both the health physics and the operations staff to develop methods to prevent recurrences.
12.2 Health Physics Program 12.2.1 Health Physics Staffing The normal health physics staff at CU consists of two professionals and several technicians. This staff provides radiation safety support to the entire university complex, including an accelerator and many radioisotope laboratories.
The routine health physics-type activites at the reactors are performed by the operations staff.
The formal health physics staff is available for consulta-l tion and the head of the Environmental Health Department is a member of the Ward Laboratory Safety Committee.
The staff believes that the radiation safety support is adequate for the research efforts within this reactor facility.
12.2.2 Procedures Detailed written procedures have been prepared that address the radiation safety support that is expected to be provided to the routine operations of the university's research reactor facility.
These procedures identify the interactions between the operational and experimental personnel.
They also specify numerous administrative limits or action points as well as appropriate responses and corrective action if these limits or action points are reached or exceeded.
Copies of these procedures are readily available to the opera-
/
tional and research staffs and to the administrative personnel.
Cornell University SER 12-1
12.2.3 Instrumentation The University has acquired a variety of detecting and measuring instruments for monitoring potentially hazardous ionizing radiation.
The instrument cali-bration procedures and techniques ensure that any credible type of radiation and any significant intensities will be detected promptly and measured correctly.
12.2.4 Training All reactor-related personnel are given an indoctrination in radiation safety before they assume their work responsibilities.
Additional radiation safety instructions are provided to those who will be working directly with radiation or radioactive materials. The training program is designed to identify the particular hazards of each specific type of work to be undertaken and methods to mitigate their consequences.
Retraining in radiation safety is provided as well.
As an example, all reactor operators are given an examination on health physics practices and procedures at least every 2 years.
The level of retrain-ing given is determined by the examination results.
12.3 Radiation Sources 12.3.1 Reactor Sources of radiation directly related to reactor operations include radiation from the reactor core, ion exchange columns, filters in the water cleanup systems, and radioactive gases (primarily 41Ar).
The fission products are contained in the fuel's stainless steel cladding.
Radiation exposures from the reactor cores are reduced to acceptable levels by water and concrete shielding.
The ion exchange resins and filters are changed routine'y before high levels of radioactive materials have accumulated, thereby limiting personnel exposure.
Personnel exposure to the radiation from chemically inert 41Ar is limited by dilution and prompt removal of this gas from the reactor room and experimental areas and its discharge to the atmosphere, where it diffuses further before reaching occupied areas.
12.3.2 Extraneous Sources Sources of radiaton that may be considered as incidental to the normal reactor operation but associated with reactor use include radioactive isotopes produced for research, activated components of experiments, and activated samples or specimens.
Personnel exposure to radiation from intentionally produced radioactive material as well as from the required manipulation of activated experimental components is controlled by rigidly developed and reviewed operating procedures that use the normal protective measures of time, distance, and shielding.
N Cornell University SER 12-2
12.4 Routine Monitoring 12.4.1 Fixed Position Monitors The zero power reactor has a fixed position radiation monitor in the reactor i
cell and another in the zero power reactor control room.
The CU TRIGA reactor facility has two fixed position radiation monitors in the experimental area, a radiation monitor on the bridge above the reactor, and a radioactive gas monitor located near the top of the reactor.
All monitors have adjustable
. alarm set points and read out in the control room.
12.4.2 Experimental Support The health physics staff participates in experiment planning by reviewing all proposed procedures for methods of minimizing personnel exposures and limiting the generation of radioactive waste. Approved procedures specify the type and degree of radiation safety support required by each activity.
I 12.5 Occupational Radiation Exposures j
12.5.1 Personnel Monitoring Program The CU personnel monitoring program is described in its Radiation Safety Instructions.
To summarize the program, personnel exposures are measured by the use of film badges assigned to individuals who might be exposed to radia-j tion.
In addition, non-self-reading pocket chambers are used, and instrument dose rate and time measurements are used to administratively keep occupational l
exposures below the applicable limits in 10 CFR 20.
All visitors are provided with non-self-reading pocket chambers for monitoring purposes.
12.5.2 Personnel Exposures The annual exposure history for the last 5 years for CU is given in Table 12.1.
12.6 Effluent Monitoring 12.6.1 Airborne Effluents As discussed in Section 11, airborne affluents from the reactor facility consist principally of activated gaset.
l The stack gas monitoring system measures the radioactive gases discharged from the entire reactor complex; the only identifiable radioactive gas is 41Ar.
I The system consists of a Geiger-Mueller tube positioned inside of the stack.
The instrumentation readout consists of a meter and a strip-chart recorder in the control room.
The detector count rate is proportional to the amount of radioactive gases in the stack and hence to the concentration in the air stream.
High concentrations and detector failure activate alarms in tho control room.
This gaseous monitoring system has been calibrated by releasing a small, known quantity of 41Ar into a closed loop " mock-up" of the stack.
It now is checked annually with a known external gamma source.
Cornell University SER 12-3
As stated in Section 5.4, 18N is not a problem at 250 kW power.
The applicant has stated that a diffuser will be installed if 18N is found to be inordinately high at the proposed power level of 500 kW.
The plumbing in the core has been arranged to enable installation of a diffuser without any modification to the existing piping.
12.6.2 Liquid Effluents As described in detail in Section 11.2.2 the reactor generates very limited radioactive liquid waste during routine operations.
Exposure of personnel or the public in unrestricted areas as a result of liquid effluents have been too low to measure dry radioactivity for normal operations.
4 12.7 Environmental Monitoring The program of sampling environmental air and water previously conducted by the Sanitary Engineering Department has been discontinued.
This program was terminated after s 15 years of operations because of the lack of statistically valid positive findings as a result of the radioactivity readings being undis-tinguishable from normal background values.
12.8 Potential Dose Assessments Natural background radiation levels in the Ithaca area result in an exposure of about 100 mrems per year to each individual residing there.
At least an additional 8% (approximately 8 mrems per year) will be received by those living in a brick or masonry structure.
Any medical diagnosis by X-ray exami-nation will add to these natural background radiations, increasing the total accumulative annual exposure.
Conservative calculations by the staff based on the amount of 41Ar released during normal operations from the reactor facility stack predict a maximum i
annual dose of less than 1 mrem in the unrestricted areas.
12.9 Conclusion The staff's review of the radiation protection program at CU indicated that (1) the program is properly staffed and equipped (2) the reactor health physics staff has adequate authority and lines of communication (3) the procedures are integrated correctly into the research plans (4) surveys verify that operations and procedures achieve ALARA principles (5) effluent monitoring prograrrs con-ducted by university personnel are adequate to promptly identify significant releases of radioactivity to predict maximum exposures to individuals in the unrestricted area On the basis of the above, the staff concludes that the CU radiation protec-tion program is acceptable, receives appropriate support from the university administration, and procedures will continue to protect the health and safety of the public during routine reactor operations.
9 I
Cornell University SER 12-4
i i
Table 12.1 Number of individuals in exposure interval Number of individuals in each range Whole-body exposure range (rem) 1977 1978 1979 1980 1981 No measurable exposure 25 18 19 21 25 Measurable exposure
{
> 0.1 0
2 0
1 1
< 0.1 0
0 0
0 0
Number of individuals monitored 25 20 19 22 26 1
l f
l Cornell University SER 5 l
13 CONDUCT OF OPERATIONS 13.1 Overall Organization Radiation safety at the Cornell University Ward Laboratory is subject to regulations of the NRC, the New York State Department of Public Health, and the Cornell University Radiation Safety Committee.
Governmental and. university regulations, and the terms of the AEC-issued operating license form the frame-work within which administrative procedures, safety rules, operating procedures, and emergency procedures have been drawn up, and enforced, by the CU reactor personnel. An organization chart showing the lines of responsibility and communication is given in Figure 13.1.
The individual with overall responsibility for radiation safety at the labora-tory is the Director of the Laboratory.
The CU Radiation Safety Committee has authority over all radiation safety matters at the university.
It is independent of the Ward Laboratory organiza-tion and both regulates and inspects radiation activities over the entire campus, including the Ward Laboratory.
The individuals designated on the organizational structure chart (Figure 13.1) as " Responsible Person on Duty" are the persons who have the direct responsi-bility for close-contact procedures and for immediate supervision of operations in progress.
These people are relied on to ensure that safety rules are actu-ally enforced, operating and emergency procedures are actually followed, and, in general, that hazardous or potentially hazardous activities are prevented.
Each area of responsibility b, rules and procedures that were established initially by the Ward Laboratory Director and consequently approved by the CU Radiation Safety Committee.
In summary, the system for achieving radiation safety at Ward Laboratory con-sists of safety rules, operating procedures, and emergency procedures which implement the applicable governmental and university regulations; administra-tive procedures to assign responsibilities and to establish and modify rules and procedures; and a chain of individual responsibilities.
The ultimate reli-ance is placed on personnel acting responsibly within a framework of rules and procedures.
Responsible persons on duty (Figure 13.1) are appointed by the Director of Ward Laboratory with the approval of the Department Director and of the CU Radiation Safety Committee.
Because of the wide variety of activities that take place in the laboratory and that may involve potential hazards to people or property, activities'are grouped into the following areas of responsibility.
1 (1) TRIGA - operation of the CU TRIGA reactor including isotope production
~
but not miniature lattice or exponential experiments Cornell University SER 13-1
(2) ZPR - operation of the zero power reactor lattice and exponential experi-ments and activities in the zero power reactor laboratory (room 111)
(3) Gamma cell - activities using the gamma irradiation cell, the set-up room, and the cold work area (4) Dynamitron - activities using the accelerator aid facilities (5) Radioactive sources and materials - activities involving use of all portable radioactive sources (especially the strong neutron sources in the subcritical reactor and the graphite sigma pile) and of radioactive materials used in tracer and similar experiments For each area of responsibility there is a separate person who is responsible for keeping the Director informed of the state of affairs and efficacy of safety measures in that area.
13.2 Training Most of the training of reactor operators is accomplished by inhouse personnel.
The applicant's Operator Requalification Program has been reviewed, and the staff concludes that it meets applicable regulations (10 CFR 50.34(b)(8)).
13.3 Emergency Planning 10 CFR 50.54(q) and (r) require that a licensee authorized to possess and/or operate a research reactor shall follow and maintain in effect an emergency plan that meets the requirements of Appendix E to 10 CFR 50.
In 1979 the guidance available to licensees was contained in RG 2.6 (1978 For Comment Issue) and in ANS 15.16 (1978 Draft).
In 1980, new regulations were promul-gated, and licensees were advised that revised guidance would be forthcoming.
Thus, revised ANS 15.16 (November 29, 1981 Draft) and RG 2.6 (March.1982 For Comment) were issued.
On May 6, 1982, an amendment to 10 CFR 50.54 was pub-lished in the Federal Register (47 8 19512, May 6, 1982) recommending these guides to licensees and establishing new submittal dated for emergency plans from all research reactor licensees.
The deadline for submittal from a licensee in a class 2 MW was November 3, 1982.
The applicant made a timely transmittal of an emergency plan, thereby complying with existing applicable regulations.
13.4 Physical Security Plan Cornell University has established and maintained a program designed to protect the reactor and its fuel and to ensure its security. The NRC staff has reviewed the plan and concludes that the plan meets the requirements of 10 CFR 73.67 for special nuclear materials of low strategic significance.
CU's licensed authoriza-tion for reactor fuel falls within that category.
Both the Physical Security Plan and the staff's evaluation are withheld from public disclosure under 10 CFR 2.790(d).
s Cornell University SER 13-2
13.5 Conclusion Based on the above discussions, the staff concludes that the licensee has sufficient experience, management structure, and procedures to provide reason-able assurance that the reactor will be managed in a way that will cause no significant risk to the health and safety of the public.
(
l l
Y I
Cornell University SER 13-3
.=. _ _..
n I
E Board of Trustees o
t.$2 I
g President of 7
the University 1
i l
I E
Vice President for Provost m
Facilities and Business E
Operations i
Life Safety Services University Radiation
- Dean, Director, Safety Committee College of Engineering
- Director, Nuclear Science and g
Engineering Program Ward Laboratory Safety Committee Director.of
- Director,
~ - - - - ~ ~ ~ - - - - - - - - - ~
~~-
Ward Laboratory Radiation Safety I
IIcalth Physics Reactor Staff Supervisor i
I I
Responsible Person L
1 on Duty y
Analogous lines for other Areas R'eac to r f Responsibility Operator line of responsibility l
line of communication User Figure 13.1 Organizational structure for radiation protection
.x
14 ACCIDENT ANALYSIS As part of its evaluation of several pending license renewals for nonpower reactors, the staff asked Battelle Pacific Northwest Laboratory to analyze generic reactor accidents for uranium-zirconium-hydride fueled reactors (NUREG/CR-2387) and asked the Los Alamos National Laboratory to evaluate the applicant's submitted documentation and analysis of potential site-specific events.
These analyses included the various types of possible accidents and the potential consequences to the public.
The following potential accidents or effects were considered to be sufficiently credible for evaluation and analysis.
(1) rapid insertion of reactivity (nuclear excursion)
(2) loss of coolant (3) metal-water reactions (4) misplaced experiments (5) mechanical rearrangement of the fuel (6) effects of fuel aging (7) fuel-handling accident Of these potential accidents, the loss of cladding integrity of an irradiated fuel element in air in the reactor room, is the only one that poses a potential impact to the environment outside the CU TRIGA reactor building.
For purposes of classification, the staff will call this the " fuel-handling accident" which has been designated as the maximum hypothetical accident (MHA).
An MHA is defined as an accident for which the risk to the public health and safety is greater than that from any event that can be postulated mechanistically.
Thus, the staff assumes that the accident occurs but does not attempt to describe or evaluate all of the mechanical details of the accident or the probability of its occurrence.
Only the consequences are considered.
14.1 Rapid Insertion of Reactivity As discussed in Section 4.7, theoretical calculations have predicted and experimental measurements have confirmed that U-ZrH fuel exhibits a strong, x
prompt, negative temperature coefficient of reactivity.
This temperature coefficient not only terminates a pulse or nuclear excursion but also causes a loss of reactivity as the steady-state temperature of the fuel is raised.
These results have been verified at many operating TRIGA reactors using U-ZrH*
fuel.
Although it may be possible theoretically to rapidly add sufficient excess reactivity under accident conditions to create an excursion that would not be terminated before fuel damage occurred, the limits imposed by the design and Technical Specifications of the CU TRIGA reactor make such an event not credible.
f.
The staff has postulated and evaluated two potential accidents in which a large amount of reactivity is inserted into the CU TRIGA reactor.
Cornell University SER 14-1
14.1.1 Scenarios The two reactivity accidents postulated by the staff are (1) insertion of the maximum transient rod reactivity worth and (2) initiation of a pulse while operating at high steady-state power levels.
In some reactor, configurations, full withdrawal of the transient rod could result in a reactivity insertion greater than the authorized maximum pulse insertion of 3.00$.
Administrative controls are applied to the adjustment of the transient rod stroke to ensure that the maximum allowed pulse reactivity is not exceeded.
The first postulated reactivity accident assumes that the transient rod is inadvertently fully withdrawn.
In the current CU reactor core, full withdrawal of the transient rod is worth
- 1.90$, which is actually less than the maximum allowable pulse.
However, because the reactivity dif-ference between full and operational withdrawal of the transient rod can change when the reactor core is reconfigured for 3.00$ pulses, the staff conservatively assumed a reactivity difference of 1.00$.
Therefore, the first postulated accident is the initiation of a 4.00$ pulse in the CU TRIGA reactor.
The total excess reactivity allowed in the proposed Technical Specifications also is 4.00$.
The second postulated accident is initiation of a pulse while the reactor is operating at a significant power level, as high as 500 kWt.
Such an accident could occur if the operator were to first fully withdraw all of the standard control rods and, after an equilibriuia power is reached, eject the transient rod from the hot reactor.
This scenario would require (1) the reactor oper-ator to violate approved operating procedures as well as the Technical Speci-fications and (2) the failure of an installed interlock designed to prevent accidental pulsing from significant power levels.
If the CU TRIGA reactor core were loaded to the maximum allowable excess reactivity of 4.00$ and the transient rod were set for the maximum allowable pulse insertion of 3.00$, the maximum equilibrium power before the pulse would be
- 180 kW (corresponding to a compensated reactivity of 1.00$).
At that power level the expected peak and average fuel temperatures are approximately 170 C and 90*7 respectively.
At power levels above
- 180 kW the reactivity available for a pulse would be less than 3.00$, decreasing to
- 1.80$ at the maximum authorized steady-state power of 500 kW.
14.1.2 Assessment The potential significant consequences of the reactivity insertion accidents considered by the staff are melting of the fuel or cladding material and failure of the cladding as a result of high internal gas pressures and/or phase changes in the fuel matrix. The primary cause of cladding failure at elevated temperatures in stainless-steel-clad elements would be excessive stress buildup in the cladding caused by hydrogen pressure from dissociation of the zirconium hydride.
Calculations performed by GA and confirmed in many reactor pulses indicate that cladding integrity is maintained at peak fuel l
temperatures as high as 1,175*C (GA-6874, 1967; Simnad et al., 1976; GA-4314, 1980).
w Cornell University SER 14-2
The staff has reviewed the literature for large reactivity insertions into reactor cores similar to the CU TRIGA reactor.
On the basis of this review, the staff has estimated the expected peak fuel temperature for the postulated reactivity insertion accidents.
General Atomics has performed many experiments with reactivity insertions as high as 5.00$ in an 85-element TRIGA core. GA measured, among other parameters, the temperature of fuel in the hottest core position, and also esamined fuel elements afterward (GA-6874, 1967; Simnad et al., 1976).
There was no indication of undue stress in the cladding and no indication of either cladding or fuel melting.
The measured maximum tempera-ture for the 5.00$ pulse was
- 750 C, and the estimated peak transient tempera-ture at any localized point in the fuel was 1,175 C.
Because the radial temperature distribution in a fuel element immediately following a pulse is similar to the radial power distribution, the peak transient temperature immediately after the pulse is located at the periphery of the hottest fuel element.
This peak temperature falls rapidly, within seconds, as the heat flows toward the cladding and fuel center.
GA also observed that for a 5.00$
pulse the maximum measured pressure rise within an instrumented fuel element was far below the expected equilibrium value at the peak temperature (GA-9064, 1970; GA-6874, 1967; Simnad et al., 1976).
A conservative analysis of the second postulated accident, a pulse from full power, yielded maximum fuel temperatures well below the allowable limits in stainless-steel-clad elements.
14.1.3 Conclusion From the above considerations, the staff concludes that there is no credible nuclear excursion in the CU TRIGA reactor that could lead to fuel melting or cladding failure resulting from high temperature or high internal gas pressure.
Therefore, there is reasonable assurance that fission product radioactivity will not be released from the fuel to the environment as a result of an acci-dental reactor pulse or excess reactivity transient.
14.2 Loss of Coolant A potential accident that would result in increases in temperatures of the fuel and cladding is the loss of coolant shortly after the reactor has been operating.
Because the water is required for adequate neutr:n moderation, its removal would terminate any significant neutron chain reactian.
However, the residual radioactivity from fission product decay would continue to deposit heat energy in the fuel.
14.2.1 Scenario It is assumed that the reactor has been operating at the licensed power of 500 kW long enough to achieve fission product equilibrium (a conservative assumption based on expected usage) and is shut down at the initiation of a gross cooling-water leak.
It is further assumed that heat is removed by convective water cooling until the top of the core becomes uncovered, after which heat removal is provided only by air convection.
E Several investigations have evaluated such scenarios under various assumptions (GA-6596, 1970; GA-9064, 1970; Oregon State University, 1968; Texas A & M, 1979).
In the CU TRIGA reactor, the core will remain completely immersed in Cornell University SER 14-3
water as long as the water level is at least 5 ft above the tank bottom.
That would require about 4,000 gal of water in the tank and allow the leakage of 18,000 gal before the top of the core becomes uncovered.
If it is assumed that a gross constant leak of 1,000 gal per minute occurs, the core would remain covered for at least 18 min.
Under these conditions the peak tempera-ture reached in the fuel would be less than 500*C and would not be reached for at least I hour after the core is uncovered.
A pulse performed immediately before water loss would not contribute signifi-cantly to the fission product decay heat. The heat generated during the pulse would be removed during the 18-min interval before the core becomes uncovered.
14.2.2 Assessment The Technical Specifications require the reactor to be shut down if the water level falls below 18.5 ft from the top of the core.
In addition, neutron and radiation monitors would alert the operating staff to a low water condition.
Even if the coolant loss was preceeded by an extended reactor run at the maximum authorized power level of 500 kW followed by a 3.00$ pulse, the resul-tant maximum fuel and/or cladding temperatures would not cause fuel damage or fission product release.
In addition, the time scale for the entire event would allow for remedial action such as supplying water from drinking water supplies through existing piping or fine hoses.
14.2.3 Conclusion From the above evaluation, the staff concludes that a loss-of-coolant accident would not melt the fuel or cladding and would not cause the release of fission products from the fuel element.
14.3 Metal-Water Reactions Chemical reactions, especially oxidation, may occur if sufficiently hot metal is brought into contact with water.
This has been an area of concern and study in designing reactors since the early 1950s, and ther.e is an extensive body of literature on the subject (Mertin, 1959; Baker and Just, 1962; Buttrey et al., 1965; Baker and Liimatakinen, 1973).
No scenarios could be realistically or mechanically developed that would lead to potential spontaneous metal-water reactors as explained in the following discussion.
From the laboratory tests, it is concluded that the metal (reactor fuel) would have to be heated to very high temperatures (for example, above the melting point) and/or be fragmented into small hot particles and injected into water to support a rapid (explosive) chemical reaction.
Either of these conditions implies a prior catastrophic event of some sort, which presumably would have to originate with a nuclear excursion or loss of coolant.
In Sections 14.1 and 14.2, these events were shown to be not credible in a 0.5 MW U-ZrH* fueled reactor like the one authorized for operation at Cornell University.
Additionally, some of the studies (Baker and Liimatakinen, 1973) include f
metal-air and metal-steam chemical reactions.
Violent (explosive) reactions j
do not appear to be possible in air or steam at atmospheric pressure, even
)
Cornell University SER 14-4
I though rapid reactions may oocur at sufficiently high temperatures with spe-cially prepared samples and conditions. As for the possible metal-water reaction, a prior cataclysmic event would be necessary even to approach those conditions, and the discussions in Sections 14.1 and 14.2 show that such an event is not credible.
In addition to the investigations referenced above, GA has experimentally plunged heated samples of unclad zirconium hydride into water to examine possible conditions for initiating and sustaining a metal-water reaction I
(Lindgren and Simnad, 1979).
At temperatures up to about 1,200 C, there was no chemical reaction of the metal except for the formation of a relatively inert oxide film.
Furthermore, most of the hydrogen may have been driven off in the hottest unclad test samples, so the metal surface in contact with the water could have been mostly zirconium.
On the basis of the above considerations, the staff concludes that there is reasonable assurance that violent metal-water, metal-air, or metal-steam reactions will not occur in a reactor of the TRIGA type operating at 1 MW or below with the maximum available excess reactivity as authorized at the CU TRIGA reactor facility.
14.4 Misplaced Experiments This type of potential accident is one in which an experimental sample or device is inadvertently located in an experimental facility where the irradia-tion conditions could exceed the design specifications.
In that case, the sample might become overheated or develop pressures that could cause a failure of the experiment container.
14.4.1 Assessment As discussed in Section 10.3, the Technical Specifications limit the reactivity of each experiment and the total reactivity of experiments in the core. The staff has reviewed the impact of large reactivity insertions in Section 14.1 l
and has concluded that reactivity insertion of the Technical Specification limits for experimental reactivities will not produce a hazard.
In addition, Sections 10 and 13 indicate that all new experiments at the CU TRIGA reactor facility are reviewed before insertion, and that all experiments in the region of the core are separated from the fuel cladding by at least one barrier, such as pneumatic transfer and irradiation tubes, beam ports, central thimble, or reflector assembly.
14.4.2 Conclusion The staff concludes that the experimental facilities and the procedures for experiment review at CU are adequate to provide reasonable assurance that failure of experiments is not likely, and even if failure occurred, breaching of the reactor fuel cladding will not occur.
Furthermore, if an experiment should fail and release radioactivity within an experimental facility, there is reasonable assurance that the amount of radioactivity released to the envi-ronment would not be more than that from the accident discussed in Section 14.7.
Cornell University SER 14-5
14.5 Mechanical Rearrangement of the Fuel This type of potential accident would involve the failure of some reactor sys-tem, such as the support structure, or could involve an externally originated event that disperses the fuel and in so doing breaches the cladding of one or more fuel elements.
14.5.1 Assessment The staff could not develop an operational scenario for such accidents that would develop consequences greater than those considered in Section 14.7, which discusses a scenario that assumes the failure of the cladding of an element after extended reactor operation and evaluates possible doses result-ing from various hypothetical scenarios for release of the inventory of radioactivity.
14.5.2 Conclusion Accordingly the staff concludes that no mechanical rearrangement that is credible would lead to an accident with more severe consequences than those accidents considered in Sections 14.1 or 14.7.
14.6 Effects of Fuel Aging The staff has included the process of fuel aging in this section so that all credible effects are addressed.
However, fuel aging should be considered normal with use of the reactor and is expected to occur gradually (see Section 17 for more detail). The reactions external to the cladding that might occur also are addressed in Section 17;- the possibility of internal reactions is discussed in this section.
14.6.1 Assessment There is some evidence that U-ZrH fuel tends to fragment with use, probably x
because of the stresses caused by high temperature gradients and the high rate-of heating during pulsing (GA-9064,1970; GA-4314,1980).
Some of the possible consequences of fragmentation are (1) a decrease in thermal conductivity across cracks, leading to higher central fuel temperatures during' steady-state operation (temperature distributions during pulsing would not be affected significantly by changes in conductivity because a pulse is completed before significant heat redistribution by conduction occurs) and (2) more fission products would be released into the cracks in the fuel.
With regard to the first item above, hot cell examination of thermally stressed hydride fuel bodies.has shown relatively widely spaced radial cracks that would cause minimal interference with. radial heat flow (GA-9064, 1970; GA-4314, 1980).
However, after pulsing, TRIGA reactors have exhibited an increase.in both steady-state fuel temperatures and power reactivity coefficients.
At'
_ power levels of 500 kW, temperatures have increased by
- 20C', and power reactivity coefficients have increased by
- 20% (GA-5400, 1965; Armed Forces Radiology Research Institute (AFRRI), 1960).
GA'has attributed these changes y
to an increased gap between'the fuel material and cladding (caused by rapid 1
1 fuel expansion during pulse heating) that reduces the heat transfer coefficient.
Experience has shown that the observed changes-occur mostly during the first Cornell University'SER' 14-6
several pulses and have essentially saturated after 100 pulses.
Therefore, the CU TRIGA reactor should not experience any further changes in the fuel-cladding gap caused by 2.00$ pulses.
However, the initiation of 3.00$ pulses may produce a slightly higher fuel temperature than initially expected.
Because these effects are small and have been observed in many TRIGA reactors operated at pulses up to 5.00$ and power levels as high as 1.5 MW, they are not considered to pose any hazards during continued operation of the CU reactor.
Two mechanisms for fission product release from TRIGA fuel have been identi-fied by GA (GA-8597, 1968; Foushee and Peters 1971; GA-4314, 1980).
The first mechanism is fission fragment recoil into connected gaps within the fuel cladding.
This effect predominates up to about 400 C and is independent of fuel temperature.
GA has postulated that in a closed system such as exists in a TRIGA fuel element, fragmentation of the fuel material within the cladding will not cause an increase in the fission product release fraction (GA-8597, 1968).
The reason for this is that the total free volume available for fission products remains constant within the confines of the cladding.
Under these conditions, the formation of a new gap or widening of an existing gap must cause a corresponding narrowing of an existing gap at some other location.
Such a narrowing allows more fission fragments to traverse the gap and become embedded in the fuel or cladding material on the other side.
In a closed system the average gap size, and therefore the fission product release rate, remains approximately constant independent of the degree to which fuel material is broken up.
Above s 400 C the controlling mechanism for fission product release is diffusion, and the amount released is dependent on fuel temperature and fuel surface-to-volume ratio.
However, release fractions used for safety evaluation are based on a conservative calculation that assumed a degree of fuel fragmentation greater than expected in actual operation.
14.6.2 Conclusion The staff concludes that the two likely effects of aging of the U-ZrH I"'I x
moderator would not have a significant effect on the operating temperature of the fuel or on the assumed release of gaseous fissicn products from the clad-ding.
Therefore, the staff also concludes that there is reasonable assurance that fuel aging will not significantly increase the likelihood of fuel-cladding failure or the calculated consequences of an accidental release in the event of loss of cladding integrity.
14.7 Fuel-Handling Accident This potential accident includes various incidents involving one or more of the irradiated fuel elements, with the reactor shut down, in which the fuel cladding might be breached or ruptured.
14.7.1 Scenario To be general, the staff let the scenario include the time scale from immedi-p ately aftar a long run at full-licensed power to any later time, such as, for example, with moving stored irradiated fuel from a rack in the pool into the reactor room. Also to remain general, the staff did not try to develop a detailed scenario but simply assumed that the cladding of one fuel element Cornell University SER 14-7
s 1
l certainly fails and that the volatile fission products accumulated in the free volume between the fuel and the cladding are released abruptly.
t t
14.7.2 Assessment 4
Several series of experiments at GA have given data on the species and frac-tions of fission products released from U-ZrH under various conditions (GA-4314, x
l 1980; GA-8597, 1968; Foushee and Peters, 1971). The noble gases were the principal species found to be released. When the fuel specimens were irradi-ated at temperatures below 350*C, the fraction released could be summarized as a constant equal to 1.5 x 10 5, independent of the temperature.
At tempera-tures greater than 350*C, the species released remained the same, but the fraction released increased significantly with increasing temperature.
GA has proposed a theory describing the release mechanisms in the two tempera-ture regimes that appears to be valid, although the data do not agree in detail (GA-8597, 1968; Foushee and Peters,'1971).
It seems reasonable to accept the interpretation of the low-temperature results, which imply that the fraction released for a typical TRIGA fuel element will be a constant, independent of operating history or details of operating' temperatures, and will apply to fuel
'whose temperature is not raised above
- 400*C.
This means that the 1.5 x 10 5 release fraction could be reasonab1'y applied to TRIGA reactors operating up to at least 800 kW steady state.
4 The theory in the fuel temperature regime above
- 400*C is not as well estab-l lished.
The proposed theory of release of the fission products incorporates a diffusion process that is a function of temperature and time.
Therefore, in principle, details of the operating history and temperature distributions in fuel elements would be required to obtain actual values for release fractions at the higher temperatures.
In situations where a fuel cladding failure was assumed, the staff used the GA results to estimate fission product release i
fractions.
The staff considers these results to be conservative in that they represent a theoretical maximum release that is greater than corresponding experimental observations.
For the fuel-handling accident, the staff estimated a fission product release i
fraction 8 x 10 5 of the inventory of both noble gases and halogens.
Based on an extrapolation of a GA analysis, this fraction is a conservative estimate of the release following a 3.00$ pulse, with a maximum local temperature of 7
j 650*C, performed after a prolonged steady-state operation.
Because the GA analysis assumed infinite operating time, it is likely that this approach will give a conservatively high release value.
Also the activity released is weighted toward the shorter half-lived nuclides and will decrease rapidly after the pulse.
Because the noble gases do not condense or combine chemically, it is' correct 2
to assume that any of these gases released from the cla'dding will diffuse in the air until the decay.
On the other hand, the iodines are chemically active and are not volatile below
- 180*C.
Therefore, some of the radioiodines will be trapped by materials with which they come in contact, such as water and structures.
In fact, evidence indicates that most of the iodines either will y
not become or will.not remain airborne-under many accident scenarios that are applicable to nonpower reactors (NUREG-0771).
However, to be certain that the fuel-cladding-failure scenario discussed below led to upper-limit dose estimates Cornell University-SER 14-8
> - ~
..,y_
v
__m. _ _ _
for all events, the staff and the applicant assumed that 100% of the iodines in the gap become airborne if water is absent from the pool.
This assumption will lead to computed doses that may be at least a factor of 100 too high in some scenarios, for example, those in wnich the pool water or water vapor are present.
In the application, the applicant analyzed a cladding failure in air and calcu-lated the resultant doses in both restricted and unrestricted areas.
The analy-sis assumed that a cladding failure occurs in a B-ring fuel element following an extended run at the authorized maximum power.
All the noble gases and halo-gens in the fuel cladding gap are postulated to have been released from the fuel element and form a uniform distribution in the reactor room.
It is further assumed that the ventilation system is shut down at the time of the accident and the air in the reactor room containing the fission products subse-quently escapes from the building at a uniform leak rate.
The applicant calcu-lated the wholebody (immersion) dose and thyroid dose by iodine inhalation to an individual in the reactor room as well as in an unrestricted area 100 m downwind of the release point.
The latter calculation assumed Pasquell Type F atmospheric conditions, which produce the highest calculated exposures, and a 2 m/sec wind speed.
Because the winds do not normally blow in the direction of the closest dwelling and because the assumed atmospheric conditions do not usu-ally occur, these assumptions yield a conservative estimate of the unrestricted exposures.
The staff has reviewed the applicant's assumption and calculations and, except for the assumed fission product release fraction, finds them reasonable and acceptable.
Whereas the applicant assumed that the fraction of noble gases released from a fuel element with a cladding failure would be 1.5 x 10 5, representative of the release from the low temperature (<400*C) regime, the staff believes a more appropriately conservative estimate would be a value of 8 x 10 5, which would be representative of the release following a 3.00$
pulse.
Therefore, the staff has increased the doses calculated by the appli-cant by a factor of 5.3.
The resulting increased value of the doses are presented in Tale 14.1.
Because there is no credible way in which the postulated accident could occur without operating personnel being alerted immediately, orderly evacuation of the reactor bay would be accomplished within minutes.
Therefore, the doses shown in Table 14.1 represent a conservative upper limit.
14.7.3 Conclusion l
In accordance with the discussions and analyses above, the staff concludes that if one f uel rod from the CU TRIGA reactor were to release 100% of the l
noble gases and halogen fission products accumulated in the fuel cladding gap, l
radiation doses to both occupational personnel and to the public in unre-stricted areas would be only a small fraction of the limits stipulated in 10 CFR 20.
These assumptions correspond to a very conservative scenario.
i Furthermore, from the results the staff obtained, even if all of the fuel rods l
failed simultaneously, the expected doses in unrestricted areas beyond 100 m l
would be less than 1 mrem and thus would be well within 10 CFR 20 limits.
Y l
In addition, the accident scenario assumed the added conservatism that the l
building isolation system did not function.
Therefore, the staff concludes l
l Cornell University SER 14-9 l
that even in the event of a single or multiple fuel-cladding failure in the CU reactor, there would be no significant risk to the health and safety of the public.
14.8 Conclusion The staff has reviewed the several postulated credible and hypothetical tran-sients and accidents for the CU TRIGA reactor.
On the basis of this review, the postulated accident with the greatest potential effect on the environment is the fuel-handling accident involving the loss of cladding integrity of one irradiated fuel element in the reactor room.
Although conservative assump-tions were used throughout the analysis, exposures to the public in unrestricted areas were a small fraction of the exposure limits given in the guidelines of 10 CFR 20.
Moreover, the analysis of this accident has shown that even if several fuel rods failed at once, the expected dose equivalents in unrestricted areas would still be below 10 CFR 20 limits.
Therefore, the staff concludes that the design of the facility and the Technical Specifications provide reasonable assurance that the CU TRIGA reactor can be operated with no signifi-cant risk to the public's health and safety.
1 Cornell University SER 14-10 m
T Table 14.1 Doses resulting from postulated fuel-handling accident Whole-body (p + y)
Thyroid Exposure levels immersion dose
- dose
- 1-hour occupational dose in the reactor bay 70 mrem 4.5 rem 6-hour public dose 100 m downwind 11.1 prem 1.1 mrem 10 CFR 20.105 annual aver:ge 0.5 rem in any 1-hour period 2.0 mrem in 7-day period 100.0 mrem
- The whole-body and thyroid doses are small fractions of the annual l
average values in 10 CFR 20.
1 1
i I
if i
Cornell University SER
.14-11 l
15 TECHNICAL SPECIFICATIONS The applicant's Technical Specifications evaluated in this licensing action define certain features, characteristics, and conditions governing the con-tinued operation of this facility.
These Technical Specifications are explicitly included in the renewal license as Appendix A.
Formats and contents acceptable to the NRC have been used in the development of these Technical Specifications, and the staff has reviewed them using the Draft Standard ANS 15.1 (September 1981) as a guide.
Based on its review, the staff concludes that normal plant operation within the limits of the Technical Specifications will not result in offsite radi-stion exposures in excess of 10 CFR 20 limits.
Furthermore, the limiting conditions for operation, surveillance requirements, and engineered safety features will limit the likelihood of malfunctions and mitigate the conse-quences to the public of offnormal or accident events.
i Cornell University SER 15-1
r-16 FINANCIAL QUALIFICATIONS The CU TRIGA reactor is owned and operated by Cornell University in support of its role in education and research.
Therefore, the staff concludes that funds will be made available, as necessary, to support continued operations and eventually to shut down the facility and maintain it in a condition that would constitute no risk to the public.
The applicant's financial status was reviewed and found to be acceptable in accordance with the requirements of 10 CFR 50.33(f).
9 I
Cornell University SER-16-1 i....
17 OTHER LICENSE CONSIDERATIONS 17.1 Prior Reactor Utilization Previous sections of this SER concluded that normal operation of the reactor causes insignificant risk of radiation exposure to the public and that only an offnormal or accident event could cause some significant exposure.
Even a maximum hypothetical accident would not lead to a dose to the most exposed individual greater than applicable guidelines or regulations (10 CFR 20).
In this section, the staff reviews the impact of prior operation of the facil-i ity on the risk of radiation exposure to the public.
The two parameters involved are the likelihood of an accident and the consequences if an accident occurred.
Because the staff has concluded that the reactor was initially designed and constructed to be inherently safe, with additional engineered safety features, the staff also must consider whether operation will cause significant degrada-tion in these features.
Furthermore, because loss of integrity of fuel clad-ding is the design-basis accident, the staff must consider mechanisms which could increase the likelihood of failure.
Possible mechanisms are (1) radia-tion degradation of cladding strength, (2) high internal pressure caused by high temperature leading to exceeding the elastic limits of the cladding, (3) corrosion, or erosion of the cladding leading to thinning or other weakening, (4) mechanical damage as a result of handling or experimental use, and (5) degradation of safety components or systems.
The staff's conclusions regarding these parameters, in the order in which they were identified above, are as follows:
(1) Some of the standard TRIGA fuel in the CU TRIGA core has been in use since 1963 and has been subjected to a maximum of 15% burnup of 2ssy, Some TRIGA fuel at more extensively used reactors has been in use for at least four times as much.burnup, with no observable degradation of clad-ding as a result of radiation.
It is unlikely that the CU:TRIGA reactor program will change during the renewal period and alter this contiusion.
(2) The possibility of approaching such pre'ssures would occur if the entire fuel element including the cladding were to be heated to more than 930*C, (GA-4314, 1980; Simnad, et al., 1976).
Although some points in the fuel may approach this temperature for a few seconds following a 3.00$ (2.25%,
Ak/k) pulse, only a simultaneous and instantaneous total loss of coolant could cause the cladding temperature to exceed a few hundred degrees.
Because the staff considers that there is no credible scenario involving all of these assumptions, the staff concludes that there is no realistic event that would cause the elastic limit of the cladding to be exceeded.
(3) Water flow through the core is obtained by natural thermal convection,.so.
the staff concludes that erosion effects as a result of high flow velocity will be. negligible.
High primary water purity is maintained by continuous l
Cornell University SER 17-1
passage through the filter and demineralizer system that maintains pool water quality at a conductivity of 5 pmho-em 1 With conductivity limited by the Technical Specifications to below 5 pmho-cm 1, corrosion of the stainless-steel cladding is expected to be negligible, even over a total 40 year period.
(4) The fuel is handled as infrequently as possible, consistent with periodic surveillance. Any indications of possible damage or degradation are investigated immediately.
The only experiments that are placed near the core are isolated from the fuel cladding by a water gap and at least one metal barrier, such as the pneumatic tubes or the core experiment tube.
Therefore, the staff concludes that loss of integrity of cladding through damage does not constitute a significant risk to the public.
(5) University personnel perform regular preventive and corrective mainte-nance and replace components as necessary.
Nevertheless, there have been some malfunctions of equipment.
However, the staff review indicates that most of these malfunctims have been random one-of-a-kind incidents, typical of even good quality electromechanical instrumentation.
There is no indications of significant degradation of the instrumentation, and the staff further concludes that the preventive maintenance program would lead to adequate identification replacement before significant degra-dation occurred.
Therefore, the staff concludes that there has been no apparent significant degradation of safety equipment and, because there is strong evidence that any future degradation will lead to prompt remed-ial action at the CU TRIGA reactor facility, there is reasonable assurance that there will be no significant increase in the likelihood of occurrence of a reactor accident as a result of component malfunction.
The second aspect of risk to the public involves the consequences of an accident.
Because the CU TRIGA reactor has not and is not expected to operate on the maximum continuous available schedule, the inventory of radioactive fission products will be far below the values that were postulated in the evaluation of the maximum hypothetical accident analyzed in Section 14.7.
Therefore, the staff concludes (1) that the risk of radiation exposure to the public has been acceptable and will continue to be well within all applicable regulations and guidelines during the history of the reactor, and (2) that there is reasonable assurance that there will be no increase in that risk in pny discernible way during this renewal period.
17.2 Multiple or Sequential Failures of Safety Components Of the many accident scenarios hypothesized for the CU TRIGA reactor, none produce consequences more severe than the hypothetical design-basis accidents reviewed and evaluated in Section 14.
The only multiple mode failure of more severe consequences would be failure of the cladding of more than one fuel element.
This scenario was determined to result in a small fraction of the values in 10 CFR 20 (see Section 14.7).
However, as only one fuel elemant is moved at a time, the staff could not develop a credible scenario that would lead to the failure of more than one g
fuel in any one accident.
Additionally, the reactor contains redundant safety-related measuring channels and control rods.
Failure of all but one control rod and all but one safety channel would not prevent reactor shutdown to a Cornell University SFR 17-2
f safe condition. The staff review has revealed no mechanism by which failure er malfunction of one of these safety-related components could lead to a nonsafe failure of a second component.
17.3 Conclusion For the reasons stated above, the staff concludes that none of the various conditions or mechanisms (as stated in Section 17.1) will cause any significant risk to the health and welfare of the public.
f Cornell University SER 17-3
T 18 CONCLUSIONS Based on its evaluation of the application as set forth above, the staff has determined that (1) The application for renewal of Operating License R-80 for its research reactor filed by the Cornell University, dated May 27, 1980, as amended, complies with the requirements of the Atomic Energy Act of_1954, as amended (the Act), and the Commission's regulations set forth in 10 CFR, Chapter 1.
(2) The facility will operate in conformity with the application as amended, the provisions of the Act, and the rules and regulations of the Commission.
(3) There is reasonable assurance (a) that the activities authorized by the operating license can be conducted without endangering the health and safety of the public; and (b) that such activities will be conducted in compliance with the regulations of the Commission set forth in 10 CFR, Chapter 1.
(4) The applicant is technically and financially qualified to engage in the activities authorized by the license in accordance with the regulations of the Commission set forth in 10 CFR, Chapter 1.
(5) The renewal of this license will not be inimical to the common defense and security or to the health and safety of the public.
l 7
Cornell University SER 18-1
19 REFERENCES Armed Forces Radiology Research Institute (AFRRI), Final Safeguards Report for the AFRRI TRIGA Reactor (Docket 50-170) Appendix A, Nov. 1960.
Baker, L. Jr., and L. C. Just, " Studies of Metal-Water Reactions at High Temperatures, III, Experimental and Theoretical Studies of the Zirconium Water Reaction," ANL-6548, Argonne National. Laboratory,1962.
- Baker, L., Jr., and R. C. Liimatakinen, " Chemical Reactions in Reactor Materials and Engineering," The Technology of Nuclear Reactor Safety, Vol. 2, Thompson and Beckerly, eds., The MIT Press, Cambridge, Mass.,
pp. 419-513, 1973.
Code of Federal Regulations, Title 10, " Energy," U.S. Government Printing Office, Washington, D.C.
Cornell University Extension Bulletin No. 764, "The Climate of New York,"
July 1964.
Foushee, F. C., and R. H. Peters, " Summary of TRIGA Fuel Fission Product Release Experiments," Gulf-EES-A10801, San Diego, California, Sept. 1971.
General Atomics Company, GA-0471, " Technical Foundations of TRIGA," Aug. 1958.
--, GA-4314, M. T. Simnad, "The U-ZrHz Alloy:
Its Properties and Use in TRIGA Fuel," E-117-833, Feb. 1980.
--, GA-5400, "Thermionic Research TRIGA Reactor Description and Analysis,"
Rev. C, Nov. 1, 1965, transmitted by letter dated Feb. 28, 1966 (Docket No. 50-227).
--, GA-6874, C. O. Coffer et al., " Stability of U-ZrH TRIGA Fuel Subjected to Large Reactivity Insertion," Jan. 1966, transmitted by letter dated July 25, 1967 (Docket No. 50-163).
--, GA-6596, J. F. Shoptaugh, Jr., "$1mulated Loss-of-Coolant Accident for TRIGA Reactors, transmitted by letter dated Sept. 22, 1970 (Docket No. 50-227).
--, GA-8597, F. C. Foushee, " Release of Rare Gas Fission Products from U-ZrH Fuel Material", March 1968.
--, GA-9064, G. B. West, Safety Analysis Report for the Torrey Pines TRIGA Mark III Reactor," Jan. 5, 1970, transmitted by. letter dated Jan. 29, 1970 (Docket No. 50-227).
)
Letter, Sept. 15, 1980, from B. E. Dethier to K. B. Cady (CU),
Subject:
Application Supplement.
Cornell University SER' 19-1
i Lindgren, J. R., and M. T. Simnad, " Low-Enriched TRIGA Fuel Water-Quench Safety Tests," Transactions of the American Nuclear Society 33(276),
1979.
Merten, U., R. S. Stone, and W. P. Wallace, " Uranium-Zirconium Hydride Fuel Elements," in Nuclear Fuel Elements, H. H. Housner and J. F. Schuman, eds., Reinhold Publishing Co., 1959.
Oregon State University, SAR for the Oregon State University TRIGA Research Reactor (Docket 50-243), Aug. 1968.
Simnad, M.
l., F. C. Foushee, and G. c. West, " Fuel Elements for Pulsed TRIGA Research Reactors," Nuclear Technology, 28:31-56, 1976.
Texas A & M SAR for the Nuclear Science Center Reactor, Texas A & M University (Docket 50-128), June 1979.
U.S. Nuclear Regulatory Commission, NUREG-0771, " Regulatory Impact of Nuclear Reactor Accident Source Term A.ssumptions," for comment (June 1981).
--, NUREG/CR-2387, S. C. Hawley, et al., " Generic Credible Accident Analysis for TRIGA Fueled Reactors," Batelle Pacific Northwest Laboratories, 1982.
--, RG 2.6, " Emergency Planning for Research Reactors," For Comment Issue, 1978 and March 1982.
Industry Codes and Standards American National Standards Institute /American Nuclear Society (ANSI /ANS),
15.11, " Radiological Control at Research Reactor Facilities," 1977.
American Nuclear Society (ANS) 15.1, " Standard for the Development of Technical Specification for Research Reactors," Sept. 1981.
--, ANS 15.16, " Standard for Emergency Planning for Research Reactor,"
Draft 1978 and Draft 2, Nov. 1981.
v Cornell University SER 19-2
U.S. NUCLEA2 REIULATORY COWIdlSSloN BIBLIOGRAPHIC DATA SHEET NUREG-0984
- 4. TlTLE AN D SU8 TITLE (Add Volume No., et aptconewJ
- 2. (Leave 6/e94/
Safety Evaluation Report Related to the Renewal of the Operating License for the Cornell University TRIGA
- 3. RECIPIENT'S ACCESSION NO.
Research Reactor
- 7. AUTHOR (S)
- 5. DATE REPORT COMPLETED MoNTs l YE AR H. Bernard August 1983
- 9. PERFORMING ORGANIZATION N AME AND MAILING ADDRESS (Indu* Iw Codel DATE REPORT ISSUED Division of Licensing oo~rs lvEAR Office.of Nuclear Reactor Regulation August 1983 U.S. Nuclear Regulatdry Commission
- e. aem u.,,*/
Washington, D. C.
20555
- 8. (Leave Nashi
- 12. SPONSORfNC ORGANIZATION NAME AND MAILING ADDRESS (Include le Codel
- 10. PROJECT / TASK / WORK UNIT NO.
Same as 9, above.
- 11. FIN NO.
- 13. TYPE OF REPORT PE RIOD COVE RE D (inclusive dams /
Safety Evaluation Report
- 15. SUPPLEMENTARY NOTES
- 14. (Leave oIa>&1
- 16. ABSTR ACT (200 words or less/
This Safety Evaluation Report for the application filed by Cornell University for renewal of operating license number R-80 to continue to operate the TRIGA research reactor has been prepared by the Office of Nuclear Reactor Regulation of the U. S.
Nuclear Regulatory Commission. The facility is owned and operated by Cornell University and is located on the university campus in Ithac, New York. The staff concludes that the TRIGA reactor facility can continue to be operated by Cornell University without endangering the health and safety of the public.
- 17. KEY WORD:n AND DOCUMENT AN ALYSIS 17a. DESCRIPTORS Non Power Reactor Cornell University TRIGA License Renewal
)
17b IDENTIFIERS /OPEN ENDE D TERMS l
18 AVAILABILITY STATEMENT 19 SE OR TY A
(TA,s reporrt
- 21. NO. 0F P AGES.
Unlimited 2gEc" iWeg tra,s awl s
- 22. P RICE C5 j
Nec,ocu us m n.
1 UNITE 3 STATES
,,3, gg, gu, NUCLEAR REGULATORY COMMISSION Postact a se ss emo
.a$.."o' c WASHINGTON. D.C. 20666
.... ut OFFICIAL BUSINESS PENALTY FOR PRIVATE USE. s300 1 1Ali p
CF TIDC POLICY g W-Soy PUB 99 7 g POR nupEG O
kASHINGTON DC 20555 2mrr*
CZE m
2Ea
<l-4*
c.
D-2h m>
2O' Z
2m O.
HO 2.
i
,