ML20199H120

Draft SAR for Mcclellan Nuclear Radiation Ctr. Page 2-5 in Chapter 2 of Incoming Submittal Not Included

ML20199H120
Person / Time
Site:	University of California-Davis
Issue date:	07/05/1996
From:	AIR FORCE, DEPT. OF
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References
NUDOCS 9711260024
Download: ML20199H120 (450)

Text

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-DOCUMENT-CONTROL DESK: '

Attached is.;the Safety Analysis Report (CE)- submitted by

.. McClellan Air Force Base. in- support of . their -license' application,:

Docket-Number 50-607.

Please distribute:to the Public! Document P.oom and-Facility File.

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CHAPTER 1 INTRODUCTION AND GENERAL DESCRIPTION OF PLANT 4

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TABLE OF CONTENTS

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1.0 INTRODUCTION

AND GENERAL DESCRIPTION OF PLANT . . . . . . . . . . 1-1 1- 1 1.1 I ntroduction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .j% . . .

1-1 1.1.1 Purpose of Facility . . . . . . . . . . . . . . . . . . y .. w......

........ 1-1 1.2 1.1.2 Location General of Facility Plant Description . . .. .. ... .. .. .4 . . ....3m

.he. . . .

- ....... 1-2 1.2.1 Building . . . . . . . . . . . . . . . 4/$ . . . 7dh . . . . 3 1.2.2 Reactor . . . . . . . . . . . . . . . ..... . . .N h. ,. - - - - -

• 10 1.3 Relation of MNRC to Other TRIGA 1.4 Safety Summary . . . . . ..w. . . . . . ..........i .gors

. . . . . . . : 2.Y 1-13 1-12 1-13 1.4.1 Nuclear .

1.4.2 Building . . . . . . . . . . . . . . . < W""h , . . . . . . . . . . . . 1 -17 V

LIST n l-1 Typice] Design Parameters . .. . . . . .

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1.2 Mc car ter . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-4 1.3 MNRC "M"- Maik ..............................1-5 1.4 MN3Gj ~

n r............................... 1-6 1.5 MNRC ElevalloiE A-A ................................. 1-7 1.6 NRC ElevatiNSintidB-B . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-8 1.7 MNRC Elevatio$ysydons . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-9

1. M N RC ReactorsY. fy

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t l.0 'lNTRODUCTION AND GENERAL DESCRIPTION OF PLANT 1.1 Introduction' This Safety Analysis Report supports an application to the United States 4 Air F (USAF) by McClellar. Air Force Base (AFB) for the utilization of a steady-state 200gWTRIGA fueled reactor with a pulsing capability with a maximum reactivi p ins (* tToaMS_1.75. The reactor is owned by the USAF and is operated by McC FB eutr?n rad' graphy and irradiation services for both military and ' ilitary '. ___A is k i as the McClellan Nuclear Radiation Center (MNRC). %M I W This dor ament addresses only the safety issues associa e operation of th %NRC reactor. This document reflects the as-built condition op and includes experience with the operation and performance of the reactor syps, eys, and personnel exposure histories related to operation cf the MNRC at 1 M ity insertion acci-dents are analyzed in Chapter 13. The industri ' sues in handling of radiographic parts and irradiation experime in the . C Operational Safety Hazards Analysis and support documents i 1.1.1 Purpose of Facility The MNRC provides a broad r radi hic adiation services to the military and non-military sector. The facill ovides radi hy bays and consequently four beanui l

of neutro radiograp "ses, in~ 'the radiography bays, the MNRC reactor l .

core a expeds~ ilities etely accessible for the irradiation of materi adiatiMggnelude, silicon doping, isotope production, both medical and industr . ron aw w.salysis (e.g., geological samples). Although al! four radiography b able ofia lography film techniques, Bays 1,2, and 3 are equipped with _

brmallyl lectronic imagir.g devices. All bays contain the equipmc ~Iredl arts r inspection as well as the radiography equipment. To meet ty use requi reactor system and associated experiment facilities are des to operate thre ' y per day.

l 1 Location of F ity '

in the MNRC Building (38' 39' 51.2"N,121* 23' 31.5"W), on

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or is government-owned site of 2600 acres located approximately 8 miles nu. _sm cramento, California.

McClellan AFB is adequately suited for the location of the MNRC reactor. This is substantiated by this document and by the fact that over 50 TRIGA reactors are in operation world wide, including 28 in the United States. Many of these reactors are located on i

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4J university campuses and in hospitals with surrounding high populated areas.

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Rev. A. 7/5/96 1-2 1.2 General Plant Descrintion 1.2.1 Building The MNRC is a three-level 18,000 ft 2rectangular shaped building that'locor . a TRIGA reactor, as shown in Figs.1.1 through 1.7. This facility prpvides spf~e7 ing, and environmental control for the radiography and irradiation gscyices Work , uate room has been provided to handle the experiments and componen n the facilit[ manner.

The ground-level elements of the MNRC are construcle(pf reinf6tced coner

$ and:cotprete unit masonry with minor elements of exposed s'. eel. TMiiEgii6r walls of the uhiir$rtions feature factory-colored metal panels, concrete, and con asonry walls.

The exterior walls of the radiography bays are made rein _ e and vary in

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thickness from 2 to 3 feet. The interior walls and,thevoofs of hy bra are constructed of 2-ft thick reinforced concr:te. N7 Y j The reactor room is above the radiogra bays. , Kytructed of standard-filled reinforced concrete block and it has a t ical metaideck roof.

The reactor is located in a cylin 'umin8 walled with the core positioned approximately 2 ft below grade 'e', t mk lom i3 61/2 ft below grade), Fig.1.2. The reactor tanksis surrounded onolithi .

nforced concrete. Below ground level, the coner proxi th'ck. ~ ~ ground level, the concrete varies in thickness from ap% 4 10 " !ftwith the smaller dimension at the tank top. The tank is supported 17gte pad'appronrmgly 9-1/2 ft thick.

The basic pyrpiier

= Q~ Ngjy massive cogerete structures is to provide biological shielding for personnelporkin; in jir$ dad the MNRC. However, due to the massiveness of these structys, they ptovidep3 cell 66f protection for the reactor core against natural phenomena.

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'lls; facility exhaust systems are designed to maintain the reactor room and radiography bays at redsfure with respect to surrounding areas to prevent the spread of radio-ahtly acave; contamination p?These systems also maintain concentrations negative of radioact reabc@ room andith'e radiography bays to levels that are below the 10 CFR Part 20 limits for rIsMANWThe reactor and radiography control rooms each have their own air handling sysiihi6" There is a system of interlocks and warning devices to prevent personnel from inadvertent exposure to high radiation levels. Interlocks prevent personnel from entering the radiography bays whenever the beam tube shutters are open and the reactor is operating. This system also prevents the beam tube shutters from being opened when the reactor is operating and personnel are in the radiography bays when the bay doors are open. There are " Reactor On" lights throughout the facility that indicates the reactor operating status. Beam tube shutter

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Rev. A. 7/5/96 1-10 positions are monitored in the reactor and radiography control rocms. Audible and visual alarms are sounded in the radiography bays when the shutters are opening. Manual and auto-matic reactor shutdown devices are locmd in the reactor room, and each radiography bay, so immediate reactor shutdown can be init ated by anyone occupying these areas should it become necessary.

M 4

p The MNRC contains the electrical, water, and sewer utilitie(requiredIIkopfe ion. In addition, the facility has both fire detection and suppressiobystemMineE:dhtsystems, radiation monitoring systems, security systems, parts pfido"ning equijnkSt[lihd,iation i and[

radiography equipment. F VMLgf um w#

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'MMA The MNRC reactor is a 2000 kW, natural-convection-cdol6LTRIGA teactor with a graphite

, reflector designed to accept the source ends of the fooPneutrilh%miaphy beam tubes which terminate in four separate neutron radiography baystgbe reactorQs located near the bottom of a water-filled aluminum tank 7 ft in diameter _atidabout:24-1/2 ft d6ep;(See Fig.1.8). Direct visual and mechanical access to the core and1EeilEinidilI$mponents are available from the top of the tank for inspection, maintenance,.aiEd' fuel hdNiiriid1T$MMter provides adequate shielding for personnel standing at thegp of thela'nk. Bih3nir'ol rod drives are mounted diameteyrof the tank. The reactor is above monitored and thecontrolled tank by ona computer-based a bridge structu,rhnningg' instrumentation and co color graphics display and autod$ tic logg%Af vitabli$ formation. Both manual and automatic control options are availab$o Ee operatofS5$p;6 IN nno 4 The rea6tdr~co{he,ic is kh lo'Nhk c5isdiImitBegeactor control room and manages all control rod movement 6abcNditleg for }$iftilitgs as interlocks and choice of particular operating modes.

It processes a:IdMisp pulse characterjitidy$ysjnformlii66~dnontrol rod positions, a:Lifother sys'tN[ parameters. The reactor console performs many other power eactor usage and storage of historical operating data for replay functionsfiUEh at a later' time. T(ks! as m'Dditd-ih[I/

8 j[$V Fuel;for ae MNRC reactor is standard TRIGA reactor fuel having 8.5%,12.5% or 20% by weight of uranium enriched to less than 20% U-235. TRIGA reactor fuel is characterized by inherent safety, high) fission product retention, and the demonstrated ability to withstand water qussiddaggithicfidverse reaction from temperatures to 1150*C, The inherent safety of TRKiA"reissodas been demonstrated by extensive experience acquired from similar TRIGA sysiinis?UirTiughout the world. This safety arises from the large prompt negative temperature coefficient that is characteristic of uranium-zirconium hydride fuel-moderator elements used in TRIGA systems. As the fuel temperature increases, this coefficient immediately compensates for reactivity insertions. This results in a mechanism whereby reactor power excursions are limited / terminated quickly and safely.

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Rev. A. 7/5/96 1-12 lleat produced by the reactor core is removed by the primary and secondary cooling systems.

The primary system circulates tank water through a water-to-water heat exchange. The secondary water system gains heat in the heat exchanger and rejects it by use of a cooling tower. A purification system circulates a small amount of tank water through a filter and resin tanks to maintain purity and optical clarity, All of these systems contain the necepary instruments and controls for operations and monitoring performance.

4 dt

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@j?pdA 1.3 Relation of MNRC to Other TRIGA Reactors SS)

The design of the MNRC fuel is similar to those of ap;d i ately I56 rgz tates. Thirty-nine ofinesennetys currently operating world-wide with 28 in the United {i@gefumber of these EAfj were constructed in the late 1950's and 1960's. Since been in operation for many years, considerable operatidAsi~iAfihnation is avsilable and their r

characteristics are well documented.

4[ %Vsss $

eration at f0ddh d seven at even There are seventeen TRIGA reactors higher steady-state power, g$$ approvedTM forsop$sh f?%$$lik /

Four of the 1000 kW TRIGA reactors are$!censedhhsirfthe United States (T are eleven 1G00 kW TRIGA'S licensedfn the U,Sh whicVn$5haracteristics similar to the MNRC reactor. These four reactors *$i:locatedit Penn Sdte (1966), the U.S. Geological Survey Center - Denver (1969),,0%n Staf niversg Corvallis (1%7), and the University of Texas - Austin (ljg0). Therege five (5JTRIGA 3 operating at powers above 1000 kW. The two reactodopefating at 2MFanthe Korean and Thailand TRIGA reactors.

Mk hk kb$f7 Table Iddiins Ae! principhl;designparameters for the 2000 kW MNRC reactor and the Thailand 2%Rdihrnr. Itihdiiktlihnoted that these parameters may vary slightly depending on the use andMj63iing. *Q@iEP gum gp

.dWQD$ Y The functi6isl chara$eristikbf the MNRC Reactors' Instrumentation and Control (I&C)

System $r'e the same as' TAN $ proximately 50 TRIGA reactors operating in the United StateOnd throughout tlis,Inid. However, the standard instrument and control system has l

beefilijireplaced by one witit a' computer-based design incorporating the use of a GA developed, analo[g-type neutrognonitoring channel. The channels are com prow'de redundahtdafety channels. In addition, the NM-1000 channel provides wide-range log am-w powerperpand multirange linear power. The control system logic is contained in a separatet6ntrol system computer with a color graphics display which is the interface between operator and the reactor.

Both the control rod and pulse rod drives are slightly different than those used on the earlier standard TRIGA systems. The MNRC control rod drives, with the exception of the motor are i essentially the same as the drives used on other TRIGA system. The MNRC drives use a l

l-13 Rev. A. 7/5/%

f(3-) . stepping-type motor rather than the non-synchronous, single-phase motors used on earlier drives. The design and operation of the stepping motor type drive has been fully developed ,

and is used on the Sandia National Laboratory ' '.lGA Reactor System.

The adjustable fast transient rod drive used on the MNRC is a modifie rs' f the standard fast transient rod drive. The modified design consists of a gombinatjgT standard rack-and pinion control rod drive and the standard fast transie trol reI! is design has been thoroughly developed, tested, and operationally p ' n.

The only other significant difference between the MN ct eind others i . ggspnector has been modified to accept the source-end of the bee is modifiestionTFiiiinor significance rud discussed in more detail in Chapter 1

,1.4 safety Summary 1.4.1 Nuclear The analyses presented in this report de trate ctor has been destined and constructed and can be operated, escri crein t undue risk to the heahh and ,

safety of MNRC employees and th ral c.

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The approach taken in this doc t to d trat safety of the MNRC reactor is to:

Shag t the mentation and control systems are of (a)

T glifjgign,% tor fuel baHEMBht_ operating experience of systems with the same or similar IitJ9 " 'ed for operation by U.S. Government agencies.

des @ ave %

an ~ident conditions of the MNRC reactor are no greater (b) Shoy ose ( , _ _ _. reactors using the same fuel systems, and therefore present undue risk t6@jgqtth and safety of the public.

NRC reactor fuepcontrol-rod drives, control rods, and experimental systems are to many othep, systems used throughout the United States. These items have well-ed oper ' ~ experience and no new significant reactor-design activity was required, ity has been specially designed to accommodate the reactor. The tank is e m a massive reinforced-concrete block, which is, in turn, sun:ounded by the reinforced concrete walls and roofs of the radiography bays. The core is approximately 4-1/2 ft below ground level. The reactor shielding configuration is similar to other TRIGA reactors, where radiation levels exterior to the shield have been measured at less than 1 mr/hr. The

( i reactor bulk shielding and the radiography bays walls and reinforew roofs provide biological Lf 1

Rev. A. 7/5/96 1-14 TABLE 1-1 TYPICAL PRINCIPAL DESIGN PARAMETERS Parameter MNRC ,

Thall $ tid d X% WW Maximum steady-state power 2000 level ,/';ts [/EYlAJ2000 RWd FuelType /f[~ gh6[I$4, [

Fuel Moderator material U-ZrH m., [k gf7 %2iilMd[F

. . ...__...m Uranium Enrichment Up 'o 20% U m gQsl),:4 Up to 20KV5 ym m Uranium Content 8.5 to 20 wt % jfg (A 8.5 wt %

~. ,n c4,

. Shape Cylindrical 't y ?N Cylindrical fs e,  %.- ; ;

Length of Fuel 38 cm (15Jn)pveral!h, 'ggy~38 cm (15 in) overall

.,-.m-.x ,. - n Diameter of Fuel 3.63 cm:(1:43 in);ODyg.h jfy3.63 cm (1.43 in) OD a n +m:n :q Claddirg Material 0.051, cm (0.020)nj 3M SSfE h 0.051 cm (0.020 in) 3M SS ug gyyp 100ya hg 100 Number of Fuel Elements Excess Reactivity [k.Ed% Ak[kfcold,eje$hd 6.3 % Ak/k (cold, clean)

Number of Control Rods ,,[ h}( (i3)kk[ 5 l . Am#m qu s . 3. e =-

Regulatmggggggg g gg 1 I vtuwm- g1# %

Safety-Transiertygd may4 2 3

= , e,

l Shim ,d7 (gh, Ty' 3 1 n- m Total Reactivify Worth of R&s.,: . s% 8.7% Ak/F 10.12% Ak/k l Reactor' Cooling sh

~

? Natural Convection of Pool Water Natural Convection of Pool Water Aa c.

(a) #5hroximate value. O CM

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\ shielding to keep perso nel exposures as low as reasonably achievable, and protects the reactor from natural phenomena. The reactor room air handling system maintains the reactor room at a negative pressure with respect to surrounding areas to control and prevent the spread of air-borne radioactive materials. The air from the reactor room passes through HEP)tand t activated charcoal filters prior to being discharged to the atmosphere, fgthe,efeht of a release of radioactive material within the reactor room, the reactor room airA'shd{t(system automatically isolates the room preventing the release ofMity to fliefstMisphere. The room air can then be recirculated within the reactor room an ugh the 1 Pggegarcoal fil y

~

to remove particulates, 11 ~

The reactor operates at a nominal steady-state power . The average ensity is approximately 20 kW/ element, whereas the same fu E0$cessfully operated at other facilities at much higher power densities. The reactor giffdM%d with a pulsing

, capability. This is the same type of pulsing operatiodiat ha4 hssfully demonstrated with many other TRIGA type reactfiggpie total 4 fuelloading of the 7re provides excess reactivity similar to other-- --Wtors th = routinely pulsed with step insertions up to $3.00.

~

l X*MP' e larg2gibtnpt negative temperature The inherent safety of the reactor lies , maril '

coefficient of reactivity characteri tha T A fuel erator mater.al. Thus, even A ity ade an e reactor power rises on a short

(

4 when large sudden insertions of period, the prompt negative r ity fec M - p~ by an increase in temperature causes the power e,xcursion to bejirebated bef[ Iemperature approaches its safety limit.

The pror$4tiigdown ansilitiitkharacterfg heactors fueled with TRIGA fuel have been demonsfrateiidE[ ring traNsM$onducted at GA in La Jolla, California as well as other facilities. Mis 7 Z ure-type containment usually required for powerstra urban areas in'Ii@slMia$rgithouF reactors. Chapslidhiity3fiscu'ssfis characteristic in detail, sY M M itia AbnoptdIconditions o accidents discussed in this report include:

[

E Y

( Maximum Hyppetical Accident (MHA)

Reactivity insertion less of cooldif

..LosspfhiIremoval system pla'dding failure 1

l; (f) rcraft crashes (g) Pyrotechnic detonation In the first three postulated accidents (using actual measured rod worth 6. fuel and c. adding f

(

- temperatures remain at levels below those sufficient to produce cladW.

release of fission products would occur.

4ilure, and thus, no

Rev. A. 7/5/96 1-16 The limiting fault condition (i.e., the Maximum Hypothetical Accident (MHA), which assumes falh n of fuel clad and an air release of fission products from one fuel element, will result in radiation doses to operations and base personnel and the general public for both thyroid and whole body that is orders of magnitude below those allowed by ANS 15.7 (see Section 2.1.2 for boundary definitions). Chapter 13 contains a detailed discussiortof this accident scenario. A Y p%AW facil[hhaging the The calculations of the probability reactor has been analyzed, it has been found that the ppoba asiaccidgnt of than 104/ year and is, therefore, considered incredible.sfhe aircraft 35pa t an airp is lessf i t analygris summarized in Chapter 13. The complete probabilistiekessmrifor an airEMtil Yij$6 McClellan AFB is containM in Appendix C. EM?f NE!5M)tfikat k? %

The amount of explosive material allowed in the radiolodSNI$yskt any given time will be

, limited to prevent damage to the reactor (see ChapterII0 anIlli $$h.

m MiEE%

Radiation exposures to personnel working in,thpC)from t>o [and airborne radiation during normal operation have beetatiilyfedMIn*additiongactual radiation levels were measured during one megawatt jopg tTon andfa$$%isholatted to two megawatt operation. This analysis and measurements showfthat thehgust exposures occur when personnel are working in the radiogr^akf5y baysdiien the idctor is operatii.g (beam tube bulk shutters closed). Under these codditi0ns, perssdnnel willN subjected to a maximum radiation field of less than 10 mr/hr. Mal 5/fg a condrEative asi$Inption on the time personnel will be in this field iffdm radiographs Ided, the (6tiljToilE[@ill be 100 mr/wk. Chapter 11 contains the persoMposure aralfsngn actual $$ti28, exposures should be lower since electronic imaging %eiskes are used'iiiBUs]h2 and 3. When using these devices, the amount of time

~

perronnelMIMystbese 6$$yilhe, lower than assumed in the analysis. All personnel entering the areas %BFbe.closelFiii6sitortd, exposures kept as low as possible, and in no case will they be EiAEdeed the'$~tFR Part 20 guidelines.

A-Muh8imM#%

The effects of Ar-41 arid! Nil (28ncentrations during normal operation of the reactor have also o@istions personnel and the general public. These isotopes result in beenjiyaluated e%Iures of only a fesy for both[Jnr/yr to operating personnel. Their release through the MNRC stack, results in a maximum down wind concentration below the 10 CFR ggguidelines ffunrestricted areas, see Chapter 11 and Appendix A for analysis.

WEkM&BV W51EffddtsMYlingle fuel element clad failure Arr$sMI noeration have been evaluated for'b6tif 6jifrations personnel and the general public. The results show exposures below the 10 CFR Part 20 limits, see Chapter 11 and Appendix kfor analysis.

Radiation-monitoring equipment has been installed at ey locations te r.ionitor radiation levels and to sound alarms if preset values are exceeded. Also, a system of reactor scrams, interlocks and administrative controls have been provided to prevent operating personnel from

l-17 Rev. A. 7/5/%

^

(

\

entering, high radiation areas, namely the radiography bays. Included in the reacter scram

- chains nre a number of rincords in the radiography bays. These rip cords allow personnel in the tr.diography bays to terminate reactor operations if radiation levels become abnormally hi',n. A 1.4.2 Building g pgg/

The MNRC reactor is housed in a building specifically e d for ion. It includes the many systems needed to support this type operati, The , te consists of one building which houses the reactor, rad h s, and sup m _.jithe MNRC is a three-story facility. The Oc:rior walls ar from reinforMit36Ecrete '

and block to a height of 24 ft, and the remaining super vered with corrugated he radiography bays are steel. The roof is a weather. sealed steel deck. The in

. constructed of reinforced standard concrete ranging 2t t The roof of hese r.reas is constructed of 2-ft thick reir. forced concr_ge react nstructed of standard reinforced concrete block with a bui The structural design of the MNRC coni s to ti ral Design Criieria (AFM 88-15), the Uniform Building Code, AISC ifica to the ACI Code. The ce Unifo uilding Code 2*one 3 criteria. The F MNRC design seismic load is in '

provide protection from naturci

( massive concrete walls and roof tro efac e reac e . r can withstand reactivity-insertion phenomenon. This, coupled wL and loss-of lant accidegy, out rel products, and the low exposures associate, e desigggH speident, Mates that the structure is adequate for housing : react Fire detectio N ission e been installed throughout the facility. In addition, nd radiography control consoles have been equipped the !astrumgtaggggg gAhe rea ,

with fire asppression'W ELA Th or room and e jn room crancs have been designed and constructed in a ance with OSHAp9 CFR Part 1910.184, Overhead and Monorail Cranes. All parts en designed fgtesultant static loads based on rated capacity with a factor of safety of five based the ultimate strength of the material used. The fuel transfer cask lifting -

ed using the ANSI /ASME code as guidelines. The design analysis shows an six when the entire load of the cask is on one lifting lug. In addition, all of ansfer hoisting equipment will be load tested, maintained and operated in accordance with ANSI /ASME during all fuel handling operations. This design, fabrication and testing approach coupled with the low exposures associated with fuel element clad failures shows that this system is adequate for its intended use (see Section 9.1),

g b

7/02/96 s

(

CHAPTER 2 l

i 1

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l I

! SITE CHARACTERISTICS I

i J

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Rev. A 7/5/%

( ^T ABI E OF CONTENTS

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2-1 2.0 SITE CH A RACTERISTICS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

2-1 2.1 Geography and Demography . . . . . . . . . . . . . . . . . . . . . .g' .

s.. ..

2.1.1 Site location and Description . . . . . . . . . . . f c 2-1 2.1.2 Exclusion Area Authority and Control. . . . . ^+r 4. . . . . . . 2-5 2.1.3 Population Distribution . . . . . . . $. . . .e ...... 2-9 2.2 Nearby Industrial, Transportation, and itary Fac' ... 2 2.2. l lndustry . . . . . . . . . . . . . . . , ..... . . . . , . 2

........s 2-12 2.2.2 Transportation . . . . . . . . . . .g...

2.2.3 IL atry Facilities . . . . . . . . .

........... .. 2-13 2.2.4 Evaluation of Potential Acciden ...............-2-15 2.3 Meteorology . . . . . . . . . . . . . . . . . ............ 2-15 2.3.1 Regional Climatology . . . . . ..... .......... 2-15 2.3.2 Iecal Meteorology . . . . ...... ........ 2-16 2.3.2.1 Temper .... .......... 2-16 2.3.2.2 Precip n. .. ............. 2-16 2.3.2.3 Hu paty . . . .............. 2-16 2.3.2.4 V ds and ility . . . . . . . . . . . . . . . . 2- 16

. /"N 2.3.2.5 cre er . . . ................... 2-19

! 2,4 Hydrologic Engij .g.. ..... .................... 2-19 2.4.1 Hydrolog scrip .. ..................... 2-19 4 2 4.2 F1 i

..... . . . . . . . . . . . . . . . . . . . . . . . 2-2 0 Ah .4.3 Release Effluents in Surface Waters . . . . . 2-23 logy, Bay, and Geotechnical Engineering . . . . . . . . . . . . 2-23

T OF TABI Es 2- Summny of Pu $$N and Military Airports in SACOG Region . . . . . . . . . . 2-14 2 Normal and Exttieme Temperatures . . . . . . . . . , , . . . . . . . . . . . . . . . . . 2-18 i W I

!m .

-i-

)

i l

4 I

Rev. A 7/5/96 UST OF FIGURES 2-2 2.1 California Map . . . . . . . . . . . . ..............................

McClellan Air Force Base - General 1.ocation Map . . . . . . . . . . . . . .... 2-3 2.2 2.3 Map of McClellan Air Force Base ................. .. .....26

........ 2-7 2.4 MNRC Plot Plan . ................. .

2-8 2.5 MNRC Axonometric View . . . . . . . . . . . . . .d. .IA . .......

2.6 Historical and Projected Population Growth in t[Vi . . . . ~

.. MUdillh . gff10

[

- 2.7 of McClellan AFB Flight Operations Regional Airport System - 1983 . . . . . . . . .q

. . . . . . . T* T W2-11

.. 2-17 2.8 Annual Wind Rose for McClellan AFB . . . . fg. ..........

2.9 McClellan Air Force Base Intermittent Drainage . . . . . . . . . . . . . . 2-21 2.10 Sacramento Area Rivers . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 -2 2 2.11 McClellan Air Force Base - 100 Year Floodp n... .......... 2-24

. . . . . . . . 2-25 2.12 Earthquake Epicenter Map of California (Rastial) . . . . .

2.13 Sacramento Area Significant Faults . .[WRK3,,, .... ........... 2-27 n.- aw _

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REFERENCES 2.1 C. B. Hurt, " Natural Regions of the United States and Canada," W. H. Freeman and Co.,1967. Ab 2.2 "McClellan AFB Compatible Land Use Report," June 1983.

U.S. Soil Conservation Service, " Soil Survey o ramento ~ gton, 2.3 D.C.,1954. i/

2.4 U.S. Department of Commerce, " Statistical A _ ;ihe United Statelflik9%" 104th f Ed.  % j U.S. Geological Survey, Sacramento, Califor 1 e, Land Use and Land 2.5 Cover and Associated Maps, Washington 1979. -p National Oceanic and Atmospheric ims lima bf the States," 2 Vols, 2.6 Second Edition, Gale Research , Detro 1  ;

on, January 1966.

O 2.7 Climatography of the Nati eacto sting S  ;

Meteorology and Atomi rgy, TI ;90.

2.8 2.9 . J. "P n Atm ispersion Program for Evaluating Design ntal dioactive Materials from Nuclear Power Stations,"

NU 8 Pa .

' west Laboratory, November 1982, 2.10 U. y, "%nal Atlas of the United States of America,"

ngton, _ p. 66, Major Recorded Earthquakes.

2.1 J. H. Bennett, " Is Fault Systems and the Auburn Dam," Calif Geology, August 1978.

ousson R pazada, at el., " Annual Technical Report - Fiscal Year 1981-1981 Isoseismic Maps and Summaries of Reported Effects for Pre-190G Calif.

ces," September 1981.

2.13 Tousson R. Toppazada, " Annual Technical Report - Fiscal Year 1981-1982, Areas Damaged by California Earthquakes."

f~

(

'- -lii-

2-1 Rev. A 7/5/96 2.0 SITE CHARACTERISTICS 0

This chapter provides information on the site characteristics of the McClellan AFB and vicinity as they relate to the safety considerations for operation of the MNRC reactor.

The conclusion reached in this chapter and throughout this document thatihe selected site is -

well suited for the MMRC facility when considering the reliiively b5dik;Tcidizating characteristics of the reactor including the Maximum Hy%etical A "" " ). This i[

consistent with the conclusions reached for the other 50lTRIGA r rs o ~

at the world,28 in the United States. Many of them arebted odIniversity _

hospitals, and other highly populated areas. EN[

NN MM 2.1 Geograohy and Demograohy

/s[05*

7 w!A 2.1.1 Site Ixcation and Description ._a;W

=s t SacrafriEnto, California.

McClellan AFB is located a few miles no ElitshiiiiI(range and the Sierra Sacrumento Nevada, about 90 milesandnortheast the base lie in the CentrgIfalley of SalIFrancisio s @$ Cali7$$iATs$e F adjacent lands are located in the G ^ halleyg$'d ivision{fie Pi:,cific Border Physiographic Province (Ref. 2.1). The area is f "ted onge alluvialhlains of the Sacramento River and its tributaries (Ref. 2.2). The land ' relative ,r g in elevation from 50-75 ft (15-23 m) above meartsea level. So,il" _ ' of abo ) consists of sandy loam (Ref. 2.3). The le but I has low permeability. The soils have surface s J 3noderat . ~~

se a slight erosion hazard.

mMer::.R M_ '" ~ ding idish m

%Wayt McClellan AFBn Grq

"- 2600 acresd.%.overnment owned land and is located approximately g

eight miles nM'Yidwqtov.N$$famento, California. (See Figure 2.2). The base and the city ar(IESacrariisini8toonty, California and the base is located northwest of the intersecildn of Watt AM$5%eville Road, and I-80 and is between the communities of NorfHighlands-FoothilllT;iims, Arden-Arcade, Rio Linda-Elverth, and North Sacramento.

m h bf hE;[hl5?W$fe$y NI eQ a$

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, -g McCLELLAN AFB 9 SACRAMENTO g STOCKTON e OAKLAND SAN FRANCISCO ,

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1 LOS ANGELES SAN DIEGO l

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MCCLELLAN AIR FORCE BASE - GENERAL LOCATION MAP FIG. 2.2

l Rev. A 7/5/96 2-4

(~)

( / McClellan AFB has one active. runway,10,60(, ft long and 200 ft wide, made of concrete.

The south end has a 1,100 ft asphalt overrun, while the north end has a 1,000 ft asphalt overrun. The runway is capable of handling any aircraft in the Air Force inventory. The taxi-

- way system consists of 383,276 square yards of pavec surface. Aircraft aprons total ^ 18.9 acres.

The Air Force maintains a 1,000 ft safety zone on each sidesof the er!!ne,3,000 x tyizone mer of each 3,000 ft clear zones at the ends of the runway, a 200 ft safe [of ap taxiway, and 125 ft minimum safety zone from the outsg uired bet n s, cargo pad [d g;grgolge are located at the base, with a 1,250 ft safety distance inhabited structures. b5--~

%W Navigational aids include high intensity runway lights, w Qjty approach lighting, Visual idalhding System (SSILS),

Approach Slope Indicator (VASI) lights, Solid-State I L Area Surveillance Radar (ASR), VHF Omni range a acticd n Station (VORTAC), and UHF transmitters and receivers.

During the past 16 years. from 1970 to I r at McClellan AFB have varied from a low of 57,180 to a h f 104 16 year period, there were a total of 1,272,700 operations, which an annd avera ' 544. The following table summarizes these operations.

V w

4 :gj MEJAL Al ERATIONS8 pg =.n y7 197TAh1 68,858 1988 83,333 h 1979 a=L 99,9s-- g' 1980 76,467 1989 85,826 82,985 1990 78,811 1972 g8,125%Q 1981 87,713 59,055

-11 - 1982 1991 19734.: _'ig l g~

1983 79,251 1992 52,138 1974" 1984 76,381 1993 58,593 i

[/1975 1976 ,i,

,180 1985 1986 72,160 90,175 1994 1995 50,717 43,516 1977 1978 58,822 1987 80,861 f

= 1,850,833 verage = 74,033 NOTE: Imt 5 Year Average = 52,804 m-

' Source: McClellan Control Tower p

Rev. A 7/5/% 24 I

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29 Rev. A 7/5/96 p) t 2.1.3 Population Distribution The McClellan AFB is situated approximately 8 miles (13 km) north-by-northeast of downtown Sacramento, California. Metropolitan Sacramento has a population of'about 1,093,000 (1992 census), an increase of about 26% since 1970, Ref. 24, Tholhiajor population center lies south by southwest of the base. App (16 km) of the base, Fig. 2.7 (Ref. 2.15). y p ggd p g ds; McClellan AFB is surrounded by communities. To . east anfn6rtheast is to the northwest, Rio Linda; to the west is the city of ~ '"to; and to the so den-Arcade. The highest density developments are directly""" in North Highlands; to the southwest, in the Del Paso Heights area of the city o d to the south in

.Secramento County.

e . _ _ . .'_s Existing land uses around the base are gener North filohlands: Mostly single family identia s bout six units per acre, with retail and other business uns een ed alon att Elkhorn Boulevard. There are also some commercial and lig ustrial s alon orn Boulevard. Scattered in the

')

(V residential areas within about a J the . , are 1 family housing is scattered in ti ca. In ial de ementary schools. Some multi-pment is centered along Orange Grove Avenue. A gh ==ieg p

n g p5 Rio I.in--- -

ingl idential uses with few retail or business uses. There are four eleme s, a school, and a high school located hi this community.

There is also a grt I two miles to the west of the McClellan runway.

North Snatamento: le family residential uses, with lower densitbs near Rio Linda and higVdensities in tights. There are some commercial and business uses along Ma ille Boulevard a d Avenue. There are nine elementary schools, a junior high

, a high school, ' a hospital located in this area.

A, ly urbanized area with single family and multiple family residential

~'d mmercial, and business uses centered on arterial streets: Marconi Avenue, nue, Howe Avenue, Auburn Boulevard, Fulton Avenue, Watt Avenue, and Ar y. There are 13 elementary schools, three high schools, a hospital, major shopping centers, and a community park located in the area.

em No significant population variations due to transient population or transient land use occur in (k the area surrounding the base. Although there are some recreational areas within 10 miles (16 km) of the base, none attract large numbers of people and most are used by local residents.

1 I

Rev, A 7/5/96 2 10 r.

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F ) MAP C-2

/((REA 19501 1960 1920 1910 1990 1995

/F A 23,360 33,180 30,700 29,130 34,000 29,000 59,667 59,678 24,620 25,910 25,630 yl%{/S M

B 1,0,V60 gg&7,840 23,850 37,440 46,250 65,000 74,843 INP (With C) 12,180 14,620 15,590 18,000 19,417 ArdasMtiliP 41,960 93,830 108,670 116,600 146,000 213,605 Sacramento 277,140 502,780 634,370 783,380 996,900 1,093,000 Total (1992)

HISTORICAL AND PROJECTED POPULATION GROWTH IN THE VICINITY OF MCCLELLAN AFB FIG. 2,6

2 11 Rev. A 7/5/%

l

1. RPORT SYMBOLS '

../

p'  ; . PRIVATE USE AIRPORT 3 MluTARY USE NR BASE .

PUBUC USE AIRPORT O PUBUC OWNERSHIP ,pg , , ,

t e O PRIVATE OWNERSHIP RUNWAY SURFACE N E :.u = PAVED

.i

- UNPAVED -

AIRSPACE INFORMAT10N \ . y) -

,3

-~ CONTROL ZONE EXTENDS .""

I.#

UPWARD FROM SURFACE /150 /

' INDICATES NRPORTS Wi1H NR CONTROL TOWER. .

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g g U REGIONAL AIRPORT SYSTEM - 1983 FIG.2.7

Rev. A 7/5/96 2 12 2.2 Nearby Industrial. Transoortation. and Military Facilitics 2.2.1 Industry There are no major industrial facilities in the Sacramento area that need be of conjern from the MNRC safety standpoint. The areas' economy is primarily based on icultuteand govern-ment with much smaller contributions by such things as: mining, Miliig of durable goods, lumber and wood products, and metal fabricationdip clos '" . !tofinery is located at Martinez, California, approximately 85 miles to the southsfist. Ngh 7 f

• 2.2.2 Transportation

%y 9 m[v((

• liighway Transportation Y

. The Sacramento area is at the cross roads of t pys: the transcontinental 180, and border to border,h5>J 80 goeY rancisco to the west, and to Reno to the east. Business 80 passii$rdish the dow n area and connects with I 80 in west Sacramento, and inNRGEdento y att Avenue. As mentioned earlier, McClellan's th[eimain MId%1Nih6iion Watt Avenue about a mile north of the 180/ Watt Avfdue interdctior.MV s hf /f f Interstate 5 passes throughdodntown#near the Sacramento River; traveling north, it leads to Oregon and Wa Egton; sMI-5 leadTto les Angeles and San Diego.

A Ak hdS8Y downtopiW'to points east; Rancho Cordova, Folsom, El U.SEHighway .%Iinnis ' '

Dotsd6Hil.a, nd South 12ke Tahoe.

ls PlacEiF.&i $es.,

~ sg h4 State lilith499,generallyparallels 1-5 to southern California, joining I 5 south of Bakersflil@%yah %7 4

/p %gfrg+u-,.

• A

/ iroorts }qydut.yf 79

[4f f There are 71 airisrts within the Sacramento Area Council of Governments, SACOG, gt the Region onjvhich records are kept. Of those,16 are public use,53 are private, and

';t two (includirig'McClellan AFB) are military. There are other private landing strips that

$r f@kM[a'e;usedliMnfrequently that no dnifof interest for the public use and military airports in the SACOG region. The record "lochtion of these airports with respect to McClellan AFB is shown in Figure 2.7.

O

2-13 Rev. A 7/5/96 (O)

• %ter Transnortatlan Sacramento has the largest river system in California. A ship channel between Rio Vista and Sacramento was dredged by the Army Corps of Engineers w there was an existing lake area. It is the Port of Sacramento, operated bype S ento Yolo Port district, and lies 79 tautical miles from the Pacific Oceagd roximately 5 miles from McClellan AFB. /ib V -i e nsive; go s ha Since its opening in 1963, the port has Avelop .

facilitics, largely focusing on rice, wheat, and ; ch tommodit .

• Rail Trantnortation b

Sacramento is served by two major rail carrie nd t s. Southern Pacific connects Sacramento with 21 western, , een thern states.

n Com ovide feeder, con-Sacramento Northern and Central Cali nector and intertie services to the r  !.

McClellan's warehouse and maintena stora ,tric ed by a spur of the Southern Pacific.

( r 2.2.3 Military Facilities .

, l There ar ilitary the vie acramento: McClellan AFB and Beale AFB. FB ussed in detail earlier in the chapter, y y P ale AFB is I ba proximately 13 miles east of Marysville and 75 miles from McCle Lt 12, x 300 ft runway was completed in 1959. At present, Beale AFa' ploys y 4,800 military and civilian personnel.

4

~

Th ery different ma ' rations are housed at Beale AFB. Rese are an air refueling n, a reconnaissans wing, and missile warning squadron. These operations use four nt types of airgdit, each with varying speeds and airspace requirements. The aircraft used are;4KC-135 Refueling Tankers, U-2 liigh Altitude Photographic, T-38 gl liigh Speed - liigh Altitude Photographic.

Th tres at Beale AFB are not available for use by the general aviation or air carrier operators. There were an estimated 85,000 aircraft operations at Beale AFB in 1981.

t Q

I i

Rev. A 7/5/96 2 14 j TABLE 21 ,

SUhiMARY OF PUBLIC USE AND hilLITARY AIRPORTS IN SACOG REGION Ownership Use Acreage Runways Aircraft Operations Airport

,, y, Military 86,000 12,000' x 300'? N/A 85,000 lleale AFB USAF a a. ri Private Public 25 N113, 3,800, Hrownsville 2,240' x 40';b, #

47 h Airport e

,o .

. .. c +

Franklin Field Public Public 496 /3,105' 7.,,60' y ~"o "j 4 70,000 23,030

_ . . .s - . . " x 60 ' -

40,000 Lincoln Public Public 695 '6,000% J00' 220 hiunicipal #SU) A

• s .~ - ( : . ...%

Public 2,875 ,. 300'd g4 (Civ) 18,300

'hfather Airport Public

.,# k II,300*xl V J 41 (htil) 7,300 pm..- m. s .v hiilitary . A,7076 #

10,600! x .200' 60 88,000 hicClellan AFB USAF o ,y ,n . av Publicy 80 2,600'Y30' 63 20,000 Natornas Field Private

,a o r Publid /T/6 3ih00' x 75' 51 15,500 Rancho hiurleta Private w,

4,500 Riego Fligh' Public !Public iy' h 10,,4 /2,380' x 35' 13 Strip A .A 1 / W#f

, 1 U$

Rio 1.indaf,. -M Privit'en a Public 4 2,620' x 30' 160 34,000

,e ,1 y_ us

'40 5,503' x 150' 115,000 Sacramento %,. iPublic N ;,Pubilo s 463 3,482' x 150' j) ,fu. A' i Executive t.i V Ni' 3,834' x 100' Sacram qth PufllicY , ' Public 3,400 8,600' x 150' 4 175,000 hietrop611 tan R _

3 o ,

18,000 Sunset Skyranch Private Public 108 2,780' x 150' 38 vn ie Public 170 3,040' x 75' 72 352,000 Sutter County Public 3,185' x 50' 37,300 LIniYersdNm[ 'bniversity Public 95 67 Alf;6tC #

Woodland-Watts Private Public 100 3,770' x 60' 90 63,000 Yolo County Public Public 400 6,000' x 100' 70 60,000 Yuba County Public Public 933 6,000' x 150' 100 63,000

2 15 Rev. A 7/5/96 m

2.2.4 Evaluation of Potential Accidents There are no nearby industrial, transportation, or military facilities with the potential of causing a credible accident that would result in a release of radioactive material # frbm MNRC that would excced the general public exposure limits of 10 CFR Part The basic MNRC design and structure provide significanj j tio gtor. As described in Chapter 1, the reactor core is below grade surround  : ithicbl f of reinforced concrete from six to nine ft $1ck. Also, above e str ity co e and rei te that surround the reactor tank are constructed of rein bleck.

The accident, from sources outside the MNRC that is r discussion is one

. involving an aircraft since the facility is located near airs . ssibility of an aircraft impact involving the MNRC reactor has been ev see C occurri

%s

,es than 10-8 Appendix C per year.

it has been determined that the probability of Therefore, this type of accident is conside cr y 2.3 Meteorology q

'V 2.3.1 Regional Climatology V

J

&g Sacrament situated in 'r ia's Cer between the Sierra Nevada and Coastal Range. is ch by hot ! '(July mean maximum temperature 105'F) and col mum temperature 28'F) (Ref. 2.6) As in most of California, of th 'crage precipitation, about 17 in. (40 cm), falls in the winter month pre thds in the area are from the south to south by.

southeast. 9 The e rn most mou form a barrier that protects much of California from the ex ly cold air from at Basin in the winter. There are occasions when cold air from a ensive high press area spreads westward and southward over California. Even in these the warming ompression as the air flows down the slopes of the mountains into the prevents s cold damage. The ranges of mountains to the west i

v

Rev. A 7/5/96 2-16 offer some protection to the interior from the strong flow of air off the Pacific Ocean.

Between the two mountain chains and over much of the desert area the temperature regime is intermediate between the maritime and the continental models, liot sumraers are the rule while winters are moderate to cold.

A 2.3.2 lual Meteorology /N ,#

red

,Cw$f A A The summary of meteorological conditions for McClellayA('B are 6 sed 68the records obtained by officials of the National Oceanic and Atm ric Administratbn/JJ S ClimatedithSiniifs?y e .

Department of Commerce and published in Volume 11 specific data, for the most part, is from the weather st; ' n at thgacramentdt Airport. MY TM

%w;W;Y h 2.3.2.1 Temperatures // Q; AS$

4r Nop nMh :wy, The normal and extreme temperatures for the Sacramento area af(showdjn Table 2 2. The normal temperatures are climatological standardpislii((1Q31 1960)fThe normal daily minimt.m temperature of 37.2*F occurs inJsThEaryMWe normal , daffy maximum temperature,93.4*F, occurs in July. EMrne tenMt@have~yinged from a low of 23'F in January of 1963 to ll5'F in Jun f1961. 9 pay 2.3.2.2 Precipitation [J[

The normal recipitation folth crame ds[$issMi$.29 in./yr with the highest amounts, approxim 3,2 in, ooE@p the moidi,6f;6ecember and January. The maximum monthlfiaisfall,512.M i}nfidllNDecember 1955. The maximum rainfall oser a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period of tl75.5h.in., ociu'r"redYOctober 1962.

\f.ym.A.%

.t %M@

2.3.2.3 liv ( %yr.

The hurNdities in the hEnie$ area range from a low of 28% in July to a high of 91% in Decciiii>er and January.}V[

sb$ j 213.2.4 Winds ~and Stability

[M1 B' ika$hal,windke for the Sacramento area is shown in Figure 2.8. The data to prepare this Eollected for the periods 1969-70 and 1973-80. As can be seen, the prevailing wiridiossM' whidfiWilie area are from the south to south-by southeast.

O

2 17 Rev. A 7/5/%

(~T U

N NNE fJNW

' NE NW 21 %

18 N ENT WNW 12

/ -

No c>-

X E

(h</ ,

6 .

. 9 I

ESE WS 15 18 2I'4 N SE SW -

~ /

S5 SSE PERIOD OF RECORD $ VIND SPEED IN KNOTS ay - .3%

sw- vo, n-so 0 ANNUAL WIND ROSE FOR MCCLELLAN AFD FIG. 2.8

l l

Rev. A 7/5/96 2 18 TABl.h 2 2 NORMAL AND EXTREME TEMPERATURES Temperature Normal Extreme Daily Daily Record Record Maximum Monthly High Year Lowest Year Month Maximum

.r e . v s\ 9 ' /,

(a) (b) (b) (b) 93- M' N@a Q)_%}l Ry y, ...

J 53.2 37.2 45.2 674 0 flh6 . ,:

23 1963 49.2 ,76 9; 1964tg .- 28 1962 F 58.6 39.8

, n. 'y 42.0 53.4 4 O$88h 1966T; 28 1966 M 64.8

. , , , n m. .a ,x 71.4 45.3 58.4 L 915 " V s 1968+ 34 1967+

A u ,, . m .w M 78.2 49.7 p64.0 f ,e " 10l'v- ?d1967 37 1964

+~ .

y Il5dy 54.4 4 70.57 1961 43 1966 J 86.5 j^; ", ',. "

/"4 .a *C/

k'

s. l* V5. + 3 J' 4_r,57:41 75:4' . F113 1961 50 1960 J / M3.4 wm'3 %

+ ~a  ;( 56 1966 A ,+Y Ji91;9, 74.1 107 1968+ 49

i. W, . . g S %s. . . m88 ~.2%... 55.0btf 1A1.6 104 1969+ 43 1965

e. _ r e u D . G77.66 - R.49.4 s*O 63.5 99 1963 38 1969+

N8 y

64.2 i e

d4'l'$>

1 52.9 87 1960 26 1961 sD 54.6  ! 38.1 46.4 72 1967 24 1965 r i a

.,g .Nf e ., .e - l Jan.

,. lg_ A/ June 23 1963 3YRi ~ ' ~3733 47.2 60.4 115 1961 ma .x 0

I i

2 19 Rev. A 7/5/96

[ \

2.3.2.5 Severe Weather

' Tornadoes have been reported in California, but they are infrequent. They are generally not severe and mest cases cause only minor damage to trees or light buildings." gp h j 2.4 llydrolocic Fncineerinc 4 Mh'y'y 4% #

Dr$%.

2.4.1 Ilyrologic Description fi g V&.

r The base and adjacent lands are located in the Great y subd[ :on of tiiieWluvial plains of th aeramento Physiographic Province (Ref. 2.1). They are situated River and its tributaries (Ref. 2.2). The land is relativ E$sdnp ing in elevation from 50-75 ft above mean sea level, Soll cover of about 4 ft consistit loam (Ref. 2.3). The

. surface soll is moderately permeable but the subsoil N# low " . The soils have moderate water-holding capacity and pose a sli n hazar il W@

McClellan AFB is underlain by a thick ( n blidated sediments ada, posits are termed the deposited by streams draining the Sierr Victor Formation which is approxi Clellan AFB. The Victor 50 to JO ft th O

Formation is composed of 'the he neo fting s ms that drained the Sierra Nevada sa grave channel like structures that grade C/ in Pleistocene time. These strea*

lay I anne provides little correlation of materials laterally and vertically into silt or low sloping alluvial fan deposits."

from area to: area. This isdaracteristic stor Fo Isa series of alluvial deposits, termed the Laguna or Fair Underlf ese alin ts are composed of *a heterogeneous assemblage of Oaks FormW[%'ejbd witif bf gravel deposited on westward-sloping floodplains beds of silt, cl7p @shcams.*Sode of the sands are clean and by meanderingfd the grave /afe extrei%s9%and poorly sorted. Sediments of the Laguna are variable; for in one area ilhfbtmati6n consists of compact silt, clay with lenses of poorly sorted examg" gr 7, sand, and silt, a1M%ithers it contains sand with only a few interbe W

lying the Victor Laguna, and Fair Oaks Formations is a volcanic unit termed the Formationj. the vicinity of McClellan AFB, this formation is composed of itslts derived from reworking of andesitic tuff breccias which issued from

[he nt Sierra Nevada. Typically, these are referred to as ' black sa dri .

'The black sands genera!!y are fairly soft and well sorted. They are formed as fluviel deposits, having been derived from andesitic detritus washed down the slopes of the Sierra Nevada. Beds of black sand are commonly about 2 meters thick, although beds up to 6 meters or more have been reported. Where exposed in road cuts, these beds exhibit

,q i crossbedding, indicating a steam laid mode of origin. Associated with the black sands are (d lenticular beds of streans gravel containing andesitic cobbles and boulders up to a meter or more in diameter. Also associated with the sands are beds of brown to blue clay and silt." In

2-20 Rev. A 7/5/%

addition to these sedimentary units, volcanic mudflow units have apparently also been encountered.

I The Mehtren Formation is the major aquifer of the Sacramento area. The thickness of the ft.

Mehtren formation in the vicinity of the base is unknown, but probably exceed ,

2.4.2 Floods A cp s y

lon of a The natural surface drainage around McClellan AFB ha(tieen altered ^

series of storm drains. The North Sacramento, Del Paglielghts,g ' a, and f to Elverta areas drain storm water runoff to the west thr " (ArcadCreek, h liebMIand swales (se '2.9).

Linda Creek, Dry Creek, and a series of shallow natu l' as they once did, these Rather than emptying onto the flat farmland of the Nat creeks and ditches are intercepted by the East Natomasif Canal and carried via

. level ranges from

. Bannon Slough to the Sacramento River. In the arcatiEfevat exte ed, double channel about 90 ft in the northeast to 50 ft in the southw serving the study

, drainage-Dry Creek is the mosi important component of:

coll from a large watershed wea. Dry Creek begins to the east in Place (C ""

in the Roseville vicinity.

i Two rivers, the Sacramento and A n, fl . troug'n) e Sacramento area, Figure 2.10.

The American River flows appr ly iniles soup of McClellan AFB. There are two rox! 20 upstream. The major dam which Good control dams on this rive Lake is an es structur . ilownstream of Folsom Dam is Nimbus forms Fols poncre and forto_s atomias. The Sacramento River flows Dam. T approxilt , h miles Clellan AFB This river handles the runoff from areas north of Sa

%x. m.,

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= DRAINAGE COURSES O

MCCLELLAN AIR FORCE BASE INTERMITTENT DRAINAGE COURSES FIG. 2.9 i- ___ _ _ _ _ . . __ _ _ -

.... s 2,s. -22 g

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FIG. 2.10

2-23 Rev. A 7/5/% l t

I O)

(

Neither of these rivers presents a flocx! hazard to the MNRC facility. The nearest 100 yr floodplain is about 3.400 ft (1,037 m) from the site of the MNRC, see Figure 2.11, 2.4.3 Accidental Release of Liquid Effluents in Surface Waters &

I The probability of an accidental release of radioactive liqu ffluen MNRC in 1 I

surface waters is extremely low. Two (2) MNRC syste y con tive liquid:

The reactor primary and the water purification systems f the c r these j systems; reactor tank, pumps, heat exchangers, filters sin tanks located within the MNRC reactor and equipment too n minat from this equipment will be wiped up and disposed of ed in Chapter 1: nly other areas where contaminated water may be encounte radiography bays and the men's washroom. The radiography bays have a dral s to a sump in Bay 1.

Any water collected in the sump is pumped into an a e gr orage ts.tk. The also dr storage tank.

decontamination shower located in the men's w the i 1 waste. Any water There are no floor drains in the men's was entering the tank, even if other than the r a d for radioactive materials if radioactive materials are discussed in Chapter 11.

it wi]

2.5 Geology. Selsmology. and hnic ngin m _

The Sacramento area is located i elsmi 3 Uniform Building Code, in general, seismic acti is not as the area e coastal areas (Refs. 2.10, 2.11, 2,12 and 2.1 on a q istorica , the maximum intensity earthquake in  :

Sacram ical I en about Vil on the Modified Mercalli scale (Refs. 2.12 and 2.13), w I f earthquakes centered about 20 mi (32 km) west of Sacramento w ted I f 6.0 to 6.5 on the Richter scale. Earthquakes of the intensit ter collapse of weak chimneys, moderate damage to maso , fall m high buildings, and fall of some nonstructural, unrei ed brick wal 2 and 2.13). However, earthquakes of higher intensity co tave occurred pri e coverage of the historical record, and higher intensity uakes are possib ' n the future. Figure 2.12 is a historical summary of the seismic in the area.

[

V

Rev. A 7/5/96 2 24

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g MCCLELLAN AIR FORCE BASE - 100 YEAR FLOODPLAIN FIG. 2.11

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Rev. A 7/5/96 2 26 California contains innumerable earthquake faults. Some of these faults are shown in Fig. '

2.13, including the known faults around Sacramento (Ref. 2.14), it is quite probable that other surface and subsurface faults also exist; however, this can only be positively determined l by ac' equate explorations. The fact that no surface faults appear on the map in the Sacramento I or San Joaquin Valleys may only indicate that sediments laid down during late ge61ogic time cover the fault scars. On the other hand, rock or the firmer sediments:gallyfo6nd in the hill and mountain areas retain the evidence of faults over long thne periodWf uth

.y , 4 W A. -

l As shown in the figure, surface faulting has been identipe,d"m the fault Beg %

zone some 25 miles cast of Sacramento and in the Rumsey yels area giff W A ,

r

car Sacram d of subsurface faults have been found during exploratiotejf,orasurface by the Division of Oil and Gas of the California DepaItsede ggConservatio faulting is reported near Freeport and Clarksburg just th$ f Sacramento; in the Todhunter Lake area a few miles north and east of DagW lo Vista area, to identify nee of subsurface

.a few areas near Sacramento. Data are not available To indic'a faulting nearer to or within the City of Sacrament gf v

Geologic investigations to date, known to ' AiIth'o' lsqt discodred evidence indicating movement on subsurface faults in the Sipiainento , ey;ngepeiknt than Eocene time, about 40 million years ago. Eocene rocks exte'nd genpfly frogju#rface of the ground to 0.5 to 0.75 kilometer depth. One fault imth6 Folsograrca, recently mapped by the California

$d%Ien intehreted as; Iving moved during the Quaternary Division Period.

of Mines and Geology.)f fj g 4 A

One cenclusios) based th gvklence og/e,N is thaf,##

Icept for the possibly more recent movement on tie $has been no near surface fault displacement in, or within the faulfiniltt&om arfi[ksir Mdpcramen the past 40 million years. The focal depth of California close proxirnity%digithblow earthquakes (tl 7bf thethT earth to the start of the rupture in the rock that provides thpen$rgygtlIe,q,uake Tahi;es from a few kilometers to 15 to 20 kilometers, and therefore cartliquakes of a' smaller magnitude could have originated here during the past 40 million' pears, but the fayiti$fght not have extended into or through this layer of post-Eocci5d rocks. rjV 8/ 7 A sscond conclusion is' bat faulting did extend to the surface, but the evidence for this surface h[lle va g 3; g y hg either haspi yet been found or is undiscernible in the sediments which fill the

%AWW wwe O

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2 27 Rev. A 7/5/%  ;

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V SACRAMENTO AREA SIGNIFICANT FAULTS Y FIG. 2.13

Rev. A 7/5/96 2 28 California's approximately 200 year recorded history is short, indeed, compared with the estimated 4.5 billion year age of the earth, it is a certainty that the Sacramento area has experienced violent earthquake motion during a part of this geologic time. From recorded information readily available for the past 200 years, however, it apper s that Sacramento has not experienced violent earthquake motion of a nature compared with that experie'pced by several othei areas within California. > [

.A was Probably the greatest amount of earthquake shaking experienced in Sacramento during the '

recent past occurred on April 21,1892. This earthquakt"'oduced edng'Ikdircage to to

/ Qd ,Mshf some 25 miles west of Sacramento.

.g 4 9[

As noted above, the April 21.1892 earthquake, alongWitliths[' quake two days'Nrli$

probably produced the most vigorous carthquake shakiriglIn1Santamento during recorded history. There is some evidence that the epicenters of tliesOINidwere in the area between

. Winters and Vacaville. Both of lese towns, as wellI.DaviUDixotPfand Woodland experienced significant damage to many structurescAhhough tIIe36dbfor the fault responsible for the 1892 earthquakes is not kno# spi alifornia Iihdpin of Mines and Geology and the U.S. Geological Survey hafh YecciitiM5did (Mayd972) that the Green W$sliVmovements just to the Valley fault, west of Fair 6 eld, #is showir$ctive fd1tp$@?

south of Interstate 80 highway.

A lineament on the east flank of  ; Imniga[ Hills has hen mapped recently by the U.S.

xkessiop6l'a fault that has moved recently.

Geological A Survey.

Mh It may be t[ Turfacc[dKY e

In recenttiiriObere wat'about$ .000 dal@aNS the Sacramento Filtration from th6 DMegalpy eartlI , cast of Fallon, Nevada, December 16,1954 - a Richter magnitude 762'earthip'ake. 'Ih . out 185 miles northeast of Sacramento and clearly indicates that tNidng period ea%A?ivaves resulting from distant carthquakes can have definite effectsMiitsMicr_es or'thejIcontents. Damage also occurred to the digestion tanks at the SacriIiEnto S'e"wagh Tr?catment Plant and to a clarifier tank at the Campbell Soup Compady.

%k g'Y Therp appears to be a str,ong northwesterly structural

• grain" to California geology.

Earthquakes having epjcenters towards the west have not affected Sacramento in the past to the istne%xtent as thog(centered cast and south of Sacramento. The 1892 Winters earthquake shelidho be an%cption to the general statement. To explain further, the April 18, 1906, Saif4ailisticifshock of Richter magnitude 8.25 with its epicenter about 80 miles west of SacIsnYestfwas probably felt in Sacramento with about the same intensity as the Owens Valley quake of March 12,1872, which has been estimated to be between 8.0 and 8.25 Richter magnitude and was about 230 miles southeast of Sacramento. Also, the Boca Reservoir earthquake of Richter magnitude 6.0 on August 12,1966,95 miles northeast of Sacramento was strongly felt in the Sacramento area as well as the above mentioned Dixic Valley earthquake 185 miles northeast of Sacramento.

2-29 Rev. A 7/5/%

O V The University of California Seismographic Station Reports that since 1932 there have been approximately 700 earthquakes of Richter magnitude 4 and greater in the area bounded l

between longitudes 118'W and 124'W and between latitudes 36.5'N and 40.5'N, In general, this area is from Eastgate, located in west central Nevada, to the Pacific Ocean from south of Fresno to Redding. Also within this area there were approximatel akes of magnitude 5 and some 15 carthquakes of magnitude 6 duri this As noted above, the distance of the clo.',est fault to Mc AFB siting requirements of ANSI 15.7, Section 3.2, which states pr

- closer than 400 neters from the surface location of a fault." ,,

1 1

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CHAPTER 3 ,

i I

i 4

i

- DESIGN OF STRUCTURES, COMPONENTS, i

EQUIPMENT AND SYSTEMS 4

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Rev. A. 7/5/96 m h (Q TABI E OF CONTENTS 3.0 DESIGN OF STRUCTURES, COMPONENTS, EQUIPMENT UD SYSTEMS .................................. $ . . . 31 2.1 Conformance with NRC General Design Criteria .....;.. ~....

. 3-1 3.1.1 Introduction . . . . . . . . . . . . . . ... .....f . . . . . . . 3- 1 3.1.2 Overall Requirements (Criteria 1 -y . . ~ .....i.. ....... 3-1 3.1.3 Protect ion by Multiple Fission Drodu iers

.... .. 7 (Criteria 10-19) ......... .. .....

3.1.4 Protection and Reactivity Control S s

........... .. 3 11 (Criteria 20-29) ............

3.1.5 Fluid Systems (Criteria 30-46) . . . . .............. 3-14 3.1.6 Reactor Containment (Criteria 50-5 ............ 3-16 3.1.7 Fuel Radioactivity Control (Criteri -64 ......... 3-18 3.2 Classincation of Structures, Compo Syste ......... 3-20 3.2.1 Seismic Classincation . . . .... . . . . . . . . . . . 3 20 3.2.2 Systems-Quality Group .. ............. 3-20 3.3 Wind and Tornado Conside ns . . . .............. 3-20 3.4 Flood Protection . . . . . . ..... ..... ................320 3.5 Missile Protection . . .... ..... ................... 3-21

- 3.6 Protection Against D Ef Associ with the Postulated Rupture of Pipi ...... ... ..................... 3-21 3.7 ismic Desig .f ....................... 3 21 n of __ @i . . . . .

Structur ......................... 3-21 3-21

. ompone nts . . . . . . . . . . . . . . . . . . . . . . . . .

plicalrol RSJ ' ' L ,7__ ............................... 3-21

3. 1 pport ..............................3-23

~

,1 Guld s ............................. 3 23

.9.4 ? .

...................................3-23

' 9.5 Fuel ssemblies . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-23 3.9.6 Bea he'A sse mi>11es . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-23

[ LIST OF TABI FS 3- icability of Compliance with General Design Criteria . . . . . . . . . . . . . , . 3-2

. f3 -i-(J) l

Rev. A. 7/5/96 I IST OF FIGURES 3.1 Rack-and Pinion Control Rod-Drive Mechanism . . . . . . . . . . . . . . . . . . . . 3-22

. . . . 3-24 3.2 Adjustable Fast Transient Rod Drive . . . . . . . . . . . . . . . . . . . . . .

3-25 3.3 TRIG A Reactor . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

I n-Tank Beam Tube . . . . . . . . . . . . . . . . . . . . g. . . . . . . . . . . . . 3-26 3.4 1

51

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-ii-

Rev. A. 7/5/96 REFERENCES is

)

3-1 W,11. Strohmayer and M. G, Stamatelatos, "Probabilistic Assessment of the Airplane Crash Risk for the McClellan Air Force Base TRIGA Reactor." ggP h

, i ,+ -  ;

4 $7sM.

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31 Rev. A. 7/5/96 l cs k

3.0 DESIGN OF STRUCTURES, COMPONENTS, EQUIPMENT AND SYSTEMS 3.1 Conformance with NRC General Design Cr!1ctia s

A 3.1.1 Introduction

!hk3df+f This chapter discusses the " General Design Criteria for tjb PcMPlhiis(as set forth in 10CFR50, Appendix A, as they apply formulated for the purpose of establishing minimum r to the MNRC.

irements,fofthe peiidi %MIGeneral criteria to be utilized for water-cooled nuclear power # ' s. Fuftfier, they UNM@I[desig new design and construction of plants similar in desig been previously issued. Since the"many MNRC is acannot of the systems research' be __

logically categorized according to power-plant applica , _e, the discussions here are

, oriented with regard to the individual criterion, rathe6than t cation of areas of noncompliance and corrective actions. Mjhk The r.ominal MNRC steady-state power 1 W ._

The maximum peak power for a $1.7 lse is M icted $,$00 MW, with a half-width C's of a few msec. Thus, the fission- et iny Iry is or 's of magnitude less than those of

() the conventional power reactors (lch .

i ral Des Criteria were primarily prepared.

A conservative upper limit of e yrel r att . tre year of operation would be about 700 MW-days,. These corugiari ens illus . 'MNRC may be placed in a category of much lo rous review for compliance with the General"_ ' Criteria.Q and tr '" " rdingly

& gyp. - A ~d h.

ex The accidents Jn Cha cident Analysis, conservatively demonstrate that instrumented 4tiGNIEyi@i@hns andi$i11 ding confinement are not necessary to ensure that radiologicdNoses wiNKhifoec,d aliowable limits. Table 3-1-1 presents a synopsis of the cor Ens regarding a riof the General Design Criteria to the MNRC.

3. ' Overall Requireynts (Criteria 1-5)

Af n 1- Qualit/ Standards and Records mf sies, systems, and components important to safety were designed, fabricated, co , and/or tested to design specifications (MAN-NDI-86-03) and associated standards, n.

/ \

L)

Rev. A. 7/5/96 3-2 i

TABLE 3-1 APPLICABILITY OF COMPLI ANCE WITH GENERAL DESIGN CRITERIA Compliance Compliance Conditional, Conditional Criterion Numbcr and Utle Not Noi. compliance- Compliance Repired /[

t m -

1. Quality Standards and Records X f! Q,47 e m

2. Design Basis for Protections Against Natural Phenomena X /'

s?

.w

's fy %g

, ' m ru. , 1 Xfy, x, fp tgEL9{L ,9

3. Fire Proicction

4. Environmental and Missile Design Basis ik /[ N(.h Y/

5. Sharing of Structures. Systems and Componenta Xi^;~' FA YO, I ____

y ;19;% c. .. A ffhU?%

Comitiance compilanid A Conditional Conditional

- Criterion Number and Title Compliance a@ NoFMI INoncomphance

,,C 'gj h, Requirbd M 7 l'RO1T.cTION BY MULTdlEfis510NYRODtiCT BARRII3is

. .a , , . . , . .w -

10. Reactor Design [7 /k XYi ?M/ u s. ,

y

}f ' " '

11. Reactor inherent Protection pfs / X fi:

M Xg-

12. Suppressions of Reactoc Power vy f Oscillatiog/11 tr ?

-r

13. Instrumeatation and Control e gP

[/e

.=r r4 4

AN -f U .. [.i;3k ew A .

I4. Reactor Cdolant Pressure 11ounoery m g ';P X e

.n

,4 :::m s

w,$~ w y M %a %glms 2

15. Reactor Coolant System Design i tvi !4 X g:3 y e- Q  % . ; r. . A 3 q.:p q? mT %

16. Containmets Desige H g X YUMN T5P5 YYN;$k.h X

17. Electrical Power,3ystems(Q p T yc k %i:@)

Wf

18. Ins;<ction and Te'sntig cf Doctrkal Power Systems X fy wi 1g

19. C0:5o1 Room NC ~ :U# X lll (i%y

$$ ?f Comphance Compliance conditional Conditional A Criteriott Number and Title Not Noncompliance Compliance 6A 1

@($N'x -/

7 Required lh5Kyyn 7 PROlTCTION AND REACTIVil'Y CONTROL SYSTDis w .--. -

20. Protectio +n Systems l' unctions X

21. Protection System Reliability and Testability X

22. Protection System IMerendence X O

33 Rev. A. 7/5/96 m

\'t Compliance Compliance Conditional Conditional Criterion Number and Title Not Noncompliance Compliance Required PROTLCTION AND REACTIVIIT CON 11tOL SYSTD15 (CONT.) f;,

y ,e

23. Prmection System l'ailure Modes X pih Alf

.e +a y

24. Separation o Protection and Control System X6 A;Wi?.pe L

~ . , . , . . . ,

23. Pracction System Requirernenta for Reactivity Control Malfuncdons /[NF, 16 3h,il'3.,. [t?

. qp .q

26. Reac6vity Corsrol System Redundancy and Capability kX /_gy.

p,?2 ;n:A Q3 gh

27. Combined Reacivity Control Systems Capability hX, j f \fMNdk w

28. Reactiv.ity ta. nuts gAgypj: rrd ,

29. Protection Against Anticipated Operational Occurrences rf.3XU'is wu. __MA sy v Wg $:qQ i mar.--

.,4mm WW?%

Criterion Number and Title 8 "" [I . ,

CompliakoR ' Conditional Y

Conditional Compliance

' N Noncompliance e am wa

[arLtJID SYgMS +Q;1937 h 30. Quality of Reactor Coolant Pressu 4 Doundwy [

s,

(( X 31, fracture Prevention of Reactor Coolant te Bounda - ,. git / X

32. Inspectiongf Reactor Coolant igte BNndary y; $ gd)Y

/%

33. ReactyCoolant, Makeup g;xpQgg jf$yhh UN'.-4 gp @y X C;WiMA GY: ' L -

34. ResidualHeatRes% unal $QM; X V.cw n)vi.hr 3 --h. '- x:m.

- ,g

35. Emergency cose Coalitigg g32.,.g X

?iltCr$h'g %yMY X

36. Inspection of13 amtgency_

J Case Cooling Systeen jQ"%Wo (Nih X

37. TestitiefEmergency Cnc$ ling System Ay' \tXhi&.Q X

38. C6niainment lleat RemovalWEPM

$Y ff 39lf Inspection of Containment Heat Removal System X AM #$ X 40.? Testing of Containment Hest Removal System qA m

.e.e wn 41ECtetainment Atmosphere Cleanup X ems 2. o WWw d % s ?gr 421. kapselic7o ef Opatainment Atmosphere Cleanup System X

%dMliM;FLa#'

43NTestisIgh Containment Atmosphere Cleanup System X d Cooling Water X

45. Inspection of Cooling Water System X D \ 46. Testing of Cooling Water System X fr

)

L

Rev. A. 7/5/96 3-4 Compliance Compliance Conditional Conditional Criterion Number and Title -

Not Noncompliance Compliance Required REACTOR CONTAINMFST

50. Containment Design Basis X gh a

[QN gp' St. Trscture P.evention of Containment Pressure Boundary X s @s,~b ;ih

52. Capability for Containment 12akage Rate Testing gp;g

+

w

/fg eng

53. Provisions for Containment Testing and Inspection gy X$vM; a Us h.h

--, .~ m

54. System Penetrating Containment M- AfA q) MA [k, y W.;r s a r

55. Reactor Coohnt Pressure Boualary Penetrating Containment h/o;;h4. 4;;t;jg X

56. Primary Containment Isolation Y$(hh A. X 57 Closed Systems isolation Valves 4[M NA X W4MA

.AN \. . h criterion Number and Title gW y ICSCMi @OAE 3 Compli M conditional Noncompliance conditional Compliance v ,==

gE ?M s+, Not' fi a 3 *1uIred

~ ..a w-~

FUF14ND RADIOACITVT1Y COpOD'g rn a ej

60. Control of Releases of Radioactive Materialsendse Environament X#3

61. Fuel Storage and llandling and Radioaca h /

4 Fyf ontrolfM [ON ,

62 Preventionb( Cnticality in Fuci stocage end llandlire ibpy;Si y#.. - X l MJNs AASMSk Monitueng hel~and Wuste' Storage) fQ k'D M TU X 63.

64.

R1Mia2r Monitorbia'?Mity ReleaseaMl9fl%k%1%+]Q X

'a.E-M M Pp. TtMHI5?iteya-

'%" V6, h TWW o rM %E TAA ig+tO?Y

)EN Np $4g Y gi? %Q:fd?4 v v;= w #

jy y .g-sy

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G:

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1.f9.i ?k ?lh?:Y

%j%[q:a m s_ve-O

3-5 Rev. A. 7/5/%

gO All design and construction work was monitored by McClellan AFB engineers to assure that the specifications incorporated appropriate standards, and the design and construction were in accordance with these specifications. Modifications have been made in accordance with existing standards and requirements.

LW BY Criterion 2 Desien Bases for Protection Acninst Natural PhenomenaWA8 qMME Hurricanes, tsunamis, and seiches do not occur in the mento 7 in the ar could be caused by run-off from local rainstorm activi r by a catMTro jf F . m Dam. liowever, the MNRC is situated some 3,400 ft (the406-yr floodpgg w

Only a small number of tornadoes, one or two per year, reported in California.

Based on the small probability of occurrences, postulag , the intermittent type of

. reactor operation and low fission-product inventory, Merite .,does have been established for the MNRC structure. However ings at to withstand the area wind loads.

The Sacramento area is classified es bei Seis ned in the Uniform Building Code; the MNRC structures e bee ig structed in accordance with this code, with an importance fact 1.5a AFM 15, Chapter 13. Seismic activity V in the region has registered as hi ich .0-6.5 i istorical time which indicates an fer t upper limit on the most likel ic eve tion 2.5). Since the MNRC is designed t e Uniform , Code an importance factor of 1.5, there is ample 'sm in tg r the expected event. The MNRC structures may su age" ic event of the highest possible yield, but, as previously noted, lo ope fission product inventory minimize the consequences vent, nt radiological doses would be within the ranges evaluated in iden ysis.

Criter 3: Fire Prot ,

T ctor room and ri}petor control room structurcs, built of steel, concrete,

, are highly fire _ istant. However, material inventories inside the rooms could include flammab ~ erials (paper, wood, etc.), and these coupled with potential ignition

_ t fires be considered.

Se es reduce both the likelihood and the consequences of a fire. First, periodic fire-safety inspections are made by Fire Safety engineers. Second, periodic in-house inspections are made for the explicit purpose of reducing nonessential combustible material inventory.

g Third, fire detection and suppression systems are installed in the facility. If these systems are g

activated, or a fire alarm is tripped, the McClellan AFB Fire Department is automatically L alerted and will respond to the MNRC within a few minutes. Fourth, a closed circuit

i Rev. A. 7/5/96 3-6 i

television camera in the reactor room with r. monitor in the control room permits the reactor operator to continuously observe the reactor rcam, so that immediate action can be taken to minimize the effects of a fire; established emergency procedures will be put into effect in the event of a fire. Fifth, the large volume of water in the reactor tank would protect the core from any conceivable fire. Sixth, the reactor is fail safe and will shutdown shoul&a fire damage the instrumentation or control system. Seventh, it has been s n irtdfapter 8 that cause an charging of the uninterruptable power system will not produce " gase "

explosion and fire.

Criterion 4! Environmental and Missile Design Bases 5 jy %h "W W x.,=

a The construction of the facility precludes catastrophic , _ _ of the reactor ere is no source in the reactor room for generating large, sustiinBdysitive pressure differentials which would breach the reactor room integrity.

[9erig%

%w:w The amount of explosive materials allowed in the, radiographyn b$pecshiisi limited to preclude damage to the reactor should they deiosinZPlancs coveriditfientrance to beam kthat thelIe will not fail from a tubes pressure pulsehave beengenerated explosive sized,fromjImaxin[m*s as discussed in ChapnWID'5iidn%gantities o Further, each experiment containing ejfosivesglfi be argsixf'to show that detonation will not produce pressure or fragments that will damage the reactor. The reactor core is protected from missiles by being below gro$iidTevel ad[surroundM by a large block of reinforced concrete. Dynamic effects of sfconditidas whip [$g pipes are not a problem because teds. The didsgpi6ns are anchored and do not penetrate the there are nopigh tank walbdidithey pressureg%eivably1Fth"e reactor. The probability of an ev couldTsi conditio'ri$li NJromb" [ cts of.ofmissiles, is very smht3Pr8bability the Airplane Crash aircraft, Risk for etc.,

the McClellan causing Air a reactor incid Force Base TRIGER$ictor" s e probability of an aircraft accident impacting the facility is %0si;TsjaherefolQi6nsidered incredible.

,p/

NWJh ve m Criterwf5: Sharine of Str6Etures. Systems. and Comoonents nr qxw Ei ical power constithes the only system shared by the MNRC. Sharing is based on the fd't[ihat c the MNRC cl$ltric power is supplied from a distribution point within the adjacent NDFJacility. Lospofpower results in the shutdown of the reactor since all control circuits are and mi f$dfilife)kh5$

A(Q " r failure atpower istherequired any point in MNRC networkfor safe will not shutdown detrimentally affect the or to maint reactorP' 9

1

37 Rev. A. 7/5/%

A kv )

3.1.3 Protection by Multiple Fission - Product Barriers (Criteria 10-19)

Criterion 10: Reactor Design The safety limit placed on the :emperature of the reactor fuel for MNRG operations is 1100'C

/k I l

when the clad is less than 500'C and 930*C when the clad tempera ' "al to the fuel 4A S temperature.

v iM+p:h I

Accident analyses presented in Chapter 13 show that u r credib ccideg.\ hsens,',

be e safety limit on the temperature of the reactor would be no fission product release that would exceedi10 Part 20 fuel willk(!PR:,xs, ided. C allowableNdhitfon levels. -Zk

%g Criterion 11: Renctor Inl erent Protection Ishg a

C Because of the fuel material (U-ZrH) and co e is a sfghnt prompt negative temperature reactivity coefficient. Routin ati6n is performed with the transient, shim, and regulating rods part y withc . - __ _ in Chapters 4 and 13, the most rapid possible reactivity insertio ; tes are equate _

5ensated for by the negative temperature reactivity coefficient 1 %/* 1 x 104

• C) .

. L) -

r Powekeillatanm Criterion 12 Suporession of R _

, ik $D N E Due to ime core aliR5Ffower levels, the reactor is inherently stable to spac 1enon 0 , , , .

_k Criterion 13; f?"Mationk,i n -~kh.1 M

.e N N A, Y The instytntation hdj66statisystem for the MNRC TRIGA reactor is a computer-based syste 'ncorporating th4tii$i bas eutron monitorir{gic%jojlGA-developed, nnel and a NPP-1000 analog type neutron multifunction, monitoring channel NM-1000

( to Chapter 7 forfylther detail) The NM-1000 system provides a safety channel nt power with scram), a wide-iange log percent power channel (below source level to er), peric41Hdication, and a multirange linear power channel (source level to full 000 system provides a second safety channel for redundancy (percent

). In the pulse mode of operation, the Data Acquisition Computer (DAC) m ain change in the NPP-1000 safety channel to provide NV and NVT indication along with a peak pulse power scram. The NM-1000 is bypassed once a pulse has been initiated.

The control system logic is contained in a separate Control System Computer (CSC) with a p}

('- color graphics display. While information from the NM-1000, NPP-1000, and fuel temperature channels is processed and displayed by the CSC, each is direct wired to its own

Rev. A. 7/5/96 3-8 i output display, and the safety channel connects directly to the protective system scram circuit.

That is, signals to the scram circuits are not processed by the Data Acquisition computer or the control computer. The nuclear information goes directly from the detectors to either the NM-1000 or NPP-1000 where it is processed. The processed signals connect directly to the scram circuit switches. Fuel temperature information goes directly to " action pack mod 61es' for amplification and then to the scram circuit switches.  ; r The NM-1000 digital neutron monitor channels were devel6ded for '

power indus 5M and generatingare fully plant. Their designqualified is based on a special noise ultra fast pulse amplifier. The NPP-1000 safety _ nel for use Gd esignedpion in designed the glo demandi criteria as the NM-1000 channels. 2 ,

p .a The hkt ings as interlocks and choice CSC manages of particular operating all control modes. rod movements, It also processeir an accounti,

@mation d dF"==

on control rod positions, power level, fuel and water temperatur_ _ utse ch- ^=bs. The CSC also performs many other functions, such as mon

• sage,Y8, oring historical operating data for replay at a later time. A1fiiyiuter ntr y'stem has many advantages over an analog system: spe . accuracj/7N_ bility for self-calibration, improved diagnostics, graphic displ nd the jogging formation.

The MNRC reactor can be opera four, s: m , automatic, square wave, and pulse. The operations are con from . reactorconsole mode control and the rod control panels. The manual'ho utomatie 1 de" steady-state reactor conditions; the square-waSMui pulse mIR$$the con @5ftfiplied by their names and require the use of the puls$gy) w k %ghwas The manual andeutomatic reacto@betrolmodes are used for reactor operation from source level to 100gji[piner$Tgtwo "nixTes are used for manual reactor startup, change in power level, andl steady-stateKoperation.

1. yMs9 Int ks prevent the nN$knt of the rods in the up direction under the following conditions:

m Miy 19 n

jlg Scrams not reset.

p2&1 Source levplyelow minimum count.

NM[hoyKsWitches depressed at the same time.

MjyMnitch in the PULSE position.

5. "Mdie switch in the AUTOMATIC position [servocontrolled rod (s) only).

6. Mode switch in the SQUARE WAVE position.

Automatic power control can be obtained by switching from manual operation to automatic operation on the mode control panel. All the instrumentation, safety, and interlock circuitry for steady operation applies to this mode. However, the servocontrolled rod (s) is (are)

39 Rev. A. 7/5/%

controlled automatically to a power level and period signal. The reactor power level is .

compared with the demand level set by the operator, on the mode control panel, and used to bring the reactor power to the demand level on a fixed preset period. The purpose of this feature is to maintain automatically the preset power level d. iring long erm po runs.

The square wave mode allows the reactor power to be qui y raise 4 Tddied power level.

In a square-wave operation, the reactor is first brought to, ity%"e$$

manual mode, leaving the transient rod partially in the The des @tTotle.kW in the the reactor operator using the power demand selector I ted on gtiode' of the steady-state instrumentation is in operation. TIMansie Tod is ejec re by means of the transient rod FIRE pushbutton locat control panel the power level reaches the demand level, it is maintained atic mode, p rN

, Reactor control in the pulsing mode consists of manuarTy esta ~ 'ticality at a flux level below one kW in the steady-state mode. This is ished b . **of the control rods, leaving the transient rod either fully or parti e pul e selector switch located on the mode control panel is then E ECTOR switch automatically causes the DAC to make m cha safety channel to monitor and record peak flux (NV), cattgy relea (N ..

o provide a peak pulse power O scram. The pulse is initiated by a mg th IRE p tton. Once a pulse has been initiated and it is detected by th , the -1000 ty scram is bypassed. Pulsing can be initiated f om either the criti 'or sube rea state.

CriterionMibetor C--- m ure P " Y ~ ~

Ngg The ctorrea%%% ooling' grate at low pressure and temperature. The vessel is open to the at ,,

is construct,e4 '

nd<ndthe.,p$,py the mary coolant incans for pressurizing system components are aluminum the or system. T ges . The ,. gnents outside the reactor tank have a low probability of stainlessa$akage seriogl or yprs of gwhich would propagate. Further, the design of the system is a that even thoug p or component ruptures, only a small amount of water would be ed from the 3 ft) (see Chapter 5). Rupture of the reactor tank is virtually Lr i sible, since it ippported on the bottom and side by reinforced concrete.

M or Coolant System Design ThM' tank is an open system and the maximum pressure in the primary system is that due to the static head (about 23-1/2 ft).' The primary cooling system, the secondary cooling system, and the purification system are pressurized by small capacity pumps. The secondary system water pressure is maintained slightly higher than the primary system. This feature (a)

'd prevents any radioactive primary water from entering the secondary system, and the environment, should a leak develop in the heat exchanger. There are no instrumentation

l I

l Rev. A. 7/5/96 3-10 systems that derive signals from any portion of the reactor coolant systems to initiate either control or protection actions. Piping and valves in the primary and purification systems are stainless steel or aluminum and of such size to provide adequate operating margins. The secondary system components are carbon steel. Chapter 5 describes the cooling system r in detail. g;r Criterion 16: Contninment Design .4 /

w A en wm The structure surrounding the reactor constitutes a confinem; pent buildingyg&mdhan providi absolute containment. Because of the low fission-prod $6iinventorfifeakagifr6idthe structure can be tolerated. N( 8 W$@

hT39rf NW Criterion 17: Electric Power Systems k An uninterruptible power supply (UPS) ical po provides ggeactorelecfr#v" console, DAC, and translator rack during normal reactor operationsgge UPS pfiltered and regulated power source to the computers and .,coqiponentibf;the reactor control systems. If there was a loss of electrical UPS$viltsupplydlectrical power to all components for fifteen minutes. Beca e react #IiMlt intural convection, and there is no requirement to provide forced _

g flo [r theIpjiMEm(i'of heat, there is suf time for the reactor operator to sht gn the : ctor and/q6nfirm the reactor is shutdown.

The UPS also provides an additiod Tour hg[ of powefto the stack continuous air monitor (CAM) and all remote area radi n moniMRA$[#

A Criterion.mfilnsocction'*nM A ino of d_-- SPo:siV

~

wer Systems

~

%$0!

The prima'ry.Yidk)hdistribut TON '

pplying commercial power to MNRC is maintained by McClellan'$%tMtrical crews. Routine inspections of the systems are performed. i; Y The hjlkC can toleraEaki8ss of electric power with no adverse effects on the safety of diejdiEIlity. There are kilNtrical power (distribution) systems desipated as necessary to pr,0vhe power to the hjtlRC during either normal or abnormal conditions (see Chapter 13).

M disbn 19: Contthi Room M

M & w=m W IKKeviGiiity an accident and operations instructions require shutdown of the reactor, contlilii6&or even partial occupancy of the control room is not a requirement since the reactor has been shut down and experiments in progress terminated. The control room is equipped ti1* mp rate exhaust system and can tolerate to some extent those accidents which do not

mit ti a L 'ach of the control room structure. Exposure levels from radiation sources mu%; is .1 an accident would be significantly reduced in magnitude (due to the location of the canam : com with respect to the reactor room). Consequently, control room radiation

3-11 Rev. A. 7/5/96

( \

s /

levels may not be higher than the allowable tolerance levels. Nevertheless, the MNRC Emergency Plan describes actiom for mitigating accident situations which require control room evacuation.

-4 A fP 3.1.4 Protection and Reactivity Contral Systems (Criteria 20-29) iib [L gn-Criterion 20: Protection System Functions gd(h C h h, Y 4 toI been desigh

[

The MNRC Reactor Protection System has to re to in that fuel design limits are not exceeded by anticipated pationafl5ccurrenceY conditions. The automatic actions are initiated by two LNsAprtrumentation c ' and two fuel temperature channels. The Reactor Protective'5yMautomatically scrams the control rods when trip settings are exceeded, see Chapp no other automatic

, actions required by MNRC systems to keep fuel tem;Wfature .' being exceeded. The Reactor Protective System satisnes the intent of IN23-197 -

of redundancy, diversity, power-loss fail-safe protection, isol ?ageillancegy

. na :t%. JJ Criterion 21: Protection System Relinh and T T. C;&d?

( The MNRC Reactor Protection Syg desi to be f any sub-channel loss that causes the channel to lose its ability ltorperfodet its intepd ftmetion results in initiation of shutdown action. Protective ac .' is mangd thtp#ghifseveral independent scram inputs arranged i series such thaig by anygts (.urrent to the scram magnets resulting in shut e reac ,

ancy owds is provided and in addition, a loss of any wer will result in a scram. Scram action is, therefore, channel circulCR is. A1 ntation is provided with testing capability. The on a one-ou$ '

. dent of the IEEE-323-1974 standard, Reactor Protech n satis Criteriond Protec h%. k T h Indeoendence WWNP Th otective system sENI!(the intent of IEEE-323-1974 " Criteria for Protective Systems clear Power Genijlating Stations." Protective functions are initiated through two i ndent nuclear add two independent fuel temperature channels, and there is a diversity of odes. Fur '*more, the protective system is fail-safe upon loss of power.

tective System and the magnet power supply are, for the most part, physically a lly isolated from the remainder of the control system. The cables between the control room and reactor room are enclosed in conduit. There is a separate conduit for each safety channel and one for the magnet power supply.

/3 b

Rev. A. 7/5/96 3-12 l Criterion 23: Protecdon System Failure Modes The reactor protective system is designed and constructed to fail safe in event of a failure of a  !

safety channel. Failure of a safety channel will result in removal of power to the control rod and transient rod magnets, allowing the control rods to fall into the core. Simup[peously, loss of a safety channel causes the transient rod's solenoid valve to de-energize thufremoving any gas pressure that may be on the pneumatic cylinder. This causes thejiideN# rod to fall into the core. The reactor protective system contains no controthetions2W6infore, loss of a protective uncontrolled reactivity function will not necessarily Id affect the operatj Criterion 24: 5%)

Seoaration of Protection and Control Sy '9Mah wwm The MNRC has two nuclear instrumentation and two fucTtemperstge channels. One of the nuclear for both safetychannels utilizes (scram) and control a wel action as fission as; i chamber lf i and a GA'NM-1000(

wide power range. The second channel utilizora[6M[gqa s or mocham channel provides % power for safety (scram [A8118iinMiil as neutrdh monitoring capability for pulse operation. Fuel temperature ip[Easured[Qils placed within the special instrumented fuel elements. While information fjom theseMnels is processed and displayed indenendcint, has its own output displays, and by connects the directly control to thesystem computer, eagchannel[reuit, seeEriterion 13, se safety sysfefpIcramf,t n and Mol func(fins. The ability of this configuration to technique used to separate p^ rot meet the intant of protect' item requitedicTWif6Pieliability, redundancy, and indepen TRIG " ctors hEMiWeepted by the NRC, Separate co 'argused RtNdf$ty channel and control system cabling from the NM-1000 and the 10bblocateil@tM r,eactor room) to the control console.

A%+mmch WP 6%Q'2MM W Finally,,ths ^

control and;sa%.fety systems are fail safe and will scram the malf tition. No contr efy systems are required to maintain a safe shutdown condition.

Critation 25 Protectin!$ System Reauirements for Reactivity Control Malfunction g ~

~

+6 N %q py gMNRC Protectiini System is designed to assure that fuel tempetature limits are not e3ceIEMdsfocan[iEgle malfunction of the reactivity control system. However Chapter 13 shMdl$dhi(ental runout of all rods simultaneously from the core at their normal drive speYdWili*not result in exceeding fuel temperature limits.

Criterion 26: Renetivity Control System Redundancy and Cannbility The MNRC has six independent reactivity control rods: four shim rcds, one regulating rod and one transient rod Each of the rods has its own drive mechanism and control circuit and

- - - = . . - . -

3-13 Rev. A. 7/5/%

>p)-

" they are operated individually. The shim aneegulating rods and drives are similar. The regulating rod is used to control power either manually or by automatic control.

Upon receipt of a scram signal, all six rods are released from their drives and cIpped into the core. Insertion of five of the six rods ensures reactor shutdown. j; 9e A khM??

Criterkn 27 Combined Reactivity Control System Canabiliiv gig?g Emergency core cooling capability is not required for 1 RC,fI klj shown that the worst conditions rcsulting in instant los' lanido not catistii int temperatures to reach the safety limit.

• Total worth of the rods is more than adequate to main suberitical level, with

. the most reactive rod stuck out of the core. 'f M v=='

,,gg g

'M

[-

Qiterion 28 Reactivity Limits No conceivable malfunction of the react contr kciiul[ result in a reactivity tered ng a til Mm-yield pulse. As shown in accident worse than the conditions en , draw oolant will cause undue heating of p Chapter 13, neither continuous rodN r loss l

i Q the fuel. Identified accidents wi or otherwise disturb the core sow $lesult.

ignifi to add gvi > , e system.

movement of adjacent fuel elements f

A g Since tl . cool  :

peratts'ppheric pressure, control-rod ejection is not a credible" '

shi "

ulating rod, and the transient rod cannot drop out of the core becau in the fusii _

. sition are approximately one inch above the safety plate located travel out of the core in the downward position is j

ktom bnsient of}4@gd' system is specially designed for r therefore elj(din.2il ystem ejection might cause a reactivity accident in the sense insertiop a

- that ip not planned; tivity additions constitute the normal pulse mode, howsyer, and the maxi g rMpactivity change and the rate of addition are limited by the design c' rods, the rod driyy system, and the available excess reactivity such that the reactor is ted against the co6 sequences of this type of accident.

A f

' 292!d5ction Acainst Anticioated Ooerational Occurrences W ~

ThefritEtwo scram loops, using different input signals, to provide redundancy in scram

! capability. The protection and reactivity control system satisfy all existing design standards.

Periodic checks (i.e., startup, shutdown, and maintenance procedures) of all reactor protective system channels and reactivity control systems demonstrate that they perform their intended i function.

(O I

i l

l

1 1

Rev. A. 7/5/96 3-14 If there was a loss of electrical power the UPS will supply electrical power to all components for fifteen minujtes. Because the reactor is cooled by natural convection, and there is no requirement to provide forced cooling flow for the removal of heat, there is sufficient time for the reactor operator to shutdown the reactor and confirm the reactor is shutdown.

4 3.1.5 Fluid Systems (Criteria 30-46) -

p.

Criterion 30: Quality of Reactor Coolant Pressure Boun dd yfs

~d Q.

Kgim -

2 The reactor tank is open to the atmosphere and is subjecied w only lentWhhnt.

M m components containing primary coolant (.i.e., reactor tan ( prm3try coolant s e steel, using stan es for purification quality control. There system) are constructed is no requirement for leak detect $m of aluminum and stainlg' ith purification loop since no conceivable leak condition cali{dityttgtank water level to lower

. more than approximately three feet. W y@h Criterion 31: Fracture Prevention p3 of Reactor QD NNPrbsture BouN[R p#%MMQ Y g~

Since the coolant system is open to the ospher 3:gpfant pressure boundary exists, i jf }"g;J#

Criterion 33: Insnection of Reaci nolant ssure Bdn dary The reactor,. tank is surroutdxl thick crete block which prevents external forces frMng direct (dEddni N@ind precludes movement of the tank. The tank walltannot be, inspected;b5a,tted to tobservation through the water ny means other than visual from inside ~i thiitiinlihAll pipiijsfaXy.0 meters, etc., associated with the primary water system are loIMidYdpen spaces ~iGidVd'readily accessible for periodic inspections.

wv gpuwh4 u g gep% m y The MNRC'bperates wrilEt$ely low powers and temperatures. Because of the moderate j

flueneg ivels and low if5hdPre factors, no significant change in material properties is expecied, pf M,, yq C'Wrion r 33: ReactoriCoolant Makeun

$NA gb TMNRQwaty' purification system design includes a system for makeup of primary coolant wagginystem is manually operated and can be easily replenished should the 300 gal storagetm'k'be depleted (see Chapter 5).

Criterion 34: Residual Heat Removal Natural convection cooling is adequate to dissipate core afterheat. Many years of operations with TRIGA reactors have shown that natural convection will provide adequate flow for the

I I

3-15 Rev. A. 7/5/96

[m)

LJ removal of heat after several hours of maximum steady state operation. Further, calculations performed for loss of coolant (see Chapter 13) show that fuel temperatures will not reach the safety limit even under loss-of-coolant conditions.

Based on the above, there is no requirement to provide a residual heat,lemova( ability.

Criterion 35: Fmergency Core Cooling System ,

An emergency core-cooling system is not required for p followi omygcc w nm7

1. The system is not pressurized and de N not o v., a h coolant temp N

2. Natural convection cooling will adequately di @ $ core heat.

3. If all water is lost for any reason, air shown in Chapter 13) will satisfactorily dissipate heat and preve e safe s.

Criterion 36: Inspection of Emergency Coo

.- 1.# 1 This criterion is not applicable. /_

Criterion 37: Testino of Emer, .v Core, .line R ' m m

='

This cri _ ot app Q_ V Criterion NN .- hment !!EIS -eal i There are ncta _s, q , experiments, etc., with sufficient stored energy to require a bi$at removd f

Crit 41on 39: Insnectio a=f ntninment Heat Removal System EY riterion is not applicable.

~

25 7 f 41 ~T.fiig of Containment Heat Removal System Thts Tek6h is not applicable.

Criterion 41: Contninment Atmosnhere Cleanun I

-(m Postaccident activities are not contmgent upon maintaining the integrity of the building structure. Accident analyses Chapter 13 have shown that downwind doses would not exceed l

j

Rev. A. 7/5/96 3-16 10 CFR Part 20 or ANSI 15.7 guidelines in any credible accident. Routine operations result in in isotopes of concern being produced: Argon-41 in the reactor room and radiography bays and Nitrogen-16 from the irradiation of oxygen in the tank water. Analysis in Chapters 11 show corsentrations to be below ANSI 15.7 guideline for accident situations and below 10 CFR Part 20 guidelines for normal operation 3.

Criterion 42: Insnection of Containment Atmosnhere Cleanuo Syste N #

m This criterion is not applicable. [7  %

pf bew

• L#

j Criterion 43 Testing of Containnient Atmosphere Clamann Systems (Qnf Y

This criterion is not applicable. g Criterion 44: Cooling Water h A coolant system is utilized to cool reactor end g norm tion of the reactor.

The MNRC requires no auxiliary cooling s , in I _

gf rea,c)or tank water upon shutdown.

p e Ew2V%_a f<

Criterion 45: Insnection of Coolind ter Sv f

Cooling equipment used in norm 41 operat a

the r . or is located either in the reactor room, equipment room, opMd the bu , $dequate space provided to permit inspection 1isn' testing o nents, n of the bulk coolant and cooling tower system F3%on a d 'or to reactor operation. During this checkout, the performanIIoksystem ,

w ith emphasis on pump outlet pressures, pressure ow ratespjly differentials 456 $4 alisysg*hb;.

Criterion 4I esIiN

^

w.

ino W er System~

1. DjNjp MyR[ reactor cooling fy7t$Ns are routinely checked, tested, and maintained.

3[ Reactor Contal ent (Criterion 50-57)

% 1 Critarinh60:-Containment Desien Basis

~

• hutigd%,y

3 Un conditions of a loss of coolant, it is conceivable that the temperature at the reactor room could increase slightly due to heating of the air flowing through the core. However, since the building is not leaktight, it will not pressurize from the heating of the air.

l Further, there is no requirement from a radiological-exposure viewpoint for a containment structure; hence, only confinement capability is provided. In addition, there is no source of

3-17 Rev. A. 7/5/%

f, ~)

energy (from an accident) which would provide a significant driving force (AP) if no corrective action were taken.

Criterion 51: Fracture Prevention Boundarv />

The confinement structure (the reactor room) is a reinforce 4 filled c

-til k structure with a conventional built-up roof. The entire structure is ' ed tow 8Ohmal external envirotunental conditions and internal environmental co itions are @& regulated [

conditions. [ qg@y -

M isi^%'

les during normMrepitations.

The structure Postulated will not accident conditions beresult cannot subjected in significa to significant interrk* Min the p due to the non leaktightness of the structure, i g Criterica52LCapabil p

This criterion is not applicable. j

y Criterion 53

Provisions for Containnet Testi: nd I n g er -

The reactor room confinement , is ked on ;

aily basis prior to reactor operation and routinely during reactor ions. eckf Ives monitoring the pressure differentials:between the r_ ' tdoom and ~ Iding areas. The reactor room exhaust ,

recirculaggm is c i bnthly t$ 1 proper operation.

.Va Cr.iterion Sw = T @h.g.

-yvstems _-. Contamment gg vg#

Piping systemC^ hh b ilding walls have no effect on the safety 4f6perat@Qe pen ^ f ons throug t e reactor u-e, isolation, re systemflinot required'YSP jY C ' ion 55: Reactor Ylant Pressure Boundarv. Penetratine Containment tor room [ cither designed nor constructed as a containment structure but provides ement capability. As pointed out in the responses to previous criteria, irements for containment (or confinement) capabilities. The only systems that i pendmitF e reactor room are the ventilation system, primary coolant system, demineralizer system, helium pressurization system and air monitor lines for CAM and remote sampling.

Reactor room wall penetrations are packed with fill material. l

\v)

Rev. A. 7/5/96 3-18 Criterion 56: Primary Contninment holation Penetrations through the building walls have no effect on the safety of reactor operations; therefore, isolation systems r.re not required in the MNRC.

Criterion 57: Closed System Isolation Valves e f.b ility The MNRC reactor building# was designed to prov valves are not required. F

?[: AY R"u?bh 3.1.7 Fuel Radioactivity Control (Criteria 60-64) h /V YYN~

p j/

Criterion 60: Control of Relene of Radioactive Materf Environment Arvf;ym*w There is no readily available path for liquid waste to fie"dischsg6dfirbetly to the environment.

Liquids in the reactor room which could subsequentlyibe injecte6Kth$ environment may result from spills, washdown of the floor, etegibbe[ikulds are col 0EdEd in a storage tank within the MNRC, analyzed for radioactivitj?diffdijissid'of accortilfigly, gf Nh Y l

l f

Significant dilution of gaseous materialbeleasedpo the and 2lgsgdfre, at soil permeability coefficients are such that transn* - timejs of groundifulds are very long

1. f l/)

! (see Chapter 13).

if $$

Criterion 61s Fuel Storace4ndi andlinc5%86Gidivity Control 4% $ A$$V The mak'@$%ngelatNe$[@$i g he handling, and control o is shielding 8@l gare e theri store d in special racks (see Criterion 62) in the reactord tan %giated fq reactor tanjkg, pitslhgreactor room iges a thijiiinum shield (seeof Chapter thickness 9). amount at least 18 ft. This When of fu water also.provides s'cg.pungf any fission products should any escape from the fuel elemeptir! lead covers 6Mihielding for elements stored in the reactor room storage pits.

Cogg is not required ihTe$ low burnup and no large decay heat source is present in the My y eles$C tats arcfuel. Irradiateg transferteid'onc uel at a time so elements are handled they are in a criticality-safe either configuration under water or (see Ser9).95$h u g,8 S trradiated, MNRC fuel elements may be stored in a criticality-safe con [utation in the reactor room. These elements require no special handling arrangements or radiation shields.

For some experiments, special core loadings may be required. Fuel elements removed from l

the core can be placed in a criticality safe fuel storage rack attached to the tank wall.

l

3 19 Rev. A. 7/5/%

g .

Criterion 62: Prevention of Criticality in Fuel Storage and liandling Fuel-storage capability is provided by storage tacks mounted in the tank and fuel storage pits which are located in the reactor room floor. The sterage locations are criticalitygafe due to the geometry utilized and the limited quantity of fuel clemenu which c&u be sadfd (see Ch:p'er 9). A @!IM?

y.m

-Since only one fuel element can be handied at a time, I Mmg does riticali

" !l ,

problem. _ _.

Criterion 63: Monitoring Fuel and Waste Storage -

' ~

No residual heat removal or temperature measurmg tred for irradiated MNRC

. fuel elements. Fuel burnup is low, therefore, only a mimu t source is present.

The reactor room and the MNRC fuel s: ora level tored with both a RAM system and a CAM system.

Criterion 64: Monitoring Rndinar tivit eleme y O

V The radiation monitoring system monitor the reactor room, and ted at M

tsi consis f the RAM's and CAM's. RAM's reactor room for gamma activity.

There are four CAMS in hd ty.

The M t staclhi with a CAM which provides continuous readings of radiation nd iculates released from the facility.

- a E 'Witors ir exhausted from the reactor room for radioactive The iodine reactor.t"bntna)5"assues. andnoble gases. Actions initiated to reduce the rele radioad, betiligiity if the set @ are exceeded are discussed in Chapter 9 and is instrument Ch 11. A third C itors the radiography bays for Ar-41 and beta / gamma particu-la The sample line . this unit are manifolded and valved so that one bay may be red at a time. addition to providing routine surveillance of the bays, this unit will be help deter the source of activity should the stack monitor alarm. All three of I readouts and alarms as well as remote recdouts and alarms in the r oom.

The fourth CAM which is currently located in the Bay 1/2 staging area monitors the al- for Ar-41, beta / gamma particulates, and noble gases. It has local readouts and alarms. This unit p can be moved anywhere in the facility as deemed necessary.

- T D

Rev. A. 7/5/96 3-20 3.2 Classification of Structures. Comoonents. and Systems The MNRC reactor does not havs structures, components, or systems that are important to safety in the same context as nuclear power plants. For the MNRC, a loss of the coolant systems, failure of the protection system, or any other credible accident does not$' ave the es of9 ANSI 15.7.

potential for causing off-site exposure comparable to the guideline ex df A s requiring a Thus, th MNRC facility does not have any structures, c,ognents ave been Category I classification. }{owever, certain structures,gynponents. .

designed to withstand natural phenomena nat have a reasonable ili ren during the plant lifetime. These design considerations" iscussed la :he V subsections.

%=#

3.2.1 Seismic Classificatiois 4  ;

The MNRC site is in a UBC Zone 3 risk area (s _ er 2). g building, reactor foundation, shielding structure, reactor tank, ;#il;' . . ort structgyare designed in accor-dance with AFM-38-15, Chapter 13 and UB,$PZ8E" genpith an importance factor

--on be returned to operation of 1.5 Meeting these requirements sho;fd ensure without structural repair; following an enrthqu g

- kelyIg@during the plant lifetime.

Furthermore, failure of the react andI f the c lant in the event of a very large earthquake has been considered apter nd the ~ equences found acceptable from the standpoint of public safety.

• 8 A

3.2.2 Sy[p Quality gsificaifjisy

%eA .s into quality groups in accordance with the ClassificatR4hfMMNRC "

• Regulatory GtildEli25' quality- ification system is considered inappropriate, because these systemstkyjdSain functid$ 1 to ensure that the reactor can be maintained in a safe shutdownc'olidition andMsTdvent th'e release of significant quantities of radioactive material to the es'rIironment. WN&

/

3.Jg Wind and Tornado Considerations

!$$ )?

PWiMNRC reactor core is protected frora damage by high winds or tornedoes by virtue of its hMJocaiIid$ and the thick reinforced concrete structure surrounding the reactor tank.

The ynfre of the MNRC has been designed for area wind loads.

3.4 Flood Protection As discussed in Chapter 2, flooding is not expected at the MNRC site. However, even if flooding occurred, reactor safety would not be an issue since the core is located in a water pool.

__--___-__-___________-____=___________-_____________

3-21 Rev. A. 7/5/%

q 3.5 Missile Protection Missile protection is provided for the MNRC reactor by virtue of the building design and the below grade location of the core which is surrounded by a seven (7) ft thick rei,np'rced concrete block (see Chapter 1 for building design). Chapter 13 also sps that'in aircraft accident damaging the facility is not probable. 5 Mh/7

$ %5$b 3.6 Protection Against Dynamic Effects Associated the PostuisGefM re of PinindE y

There are no pipes in floor level. Furthermore, the lowest elevation in the PMj the o MNRC 3 radiograp hy~ 2lifacility ay 1 floor) con mp. If cap rst a predetermined water level is reached, the sump pumpf. gTpglatically start.

. 3.7 seismic Desien .

Seismic considerations applicable to the MNP ,, 1l& disc apter 2.

jy , n 3.8 Design of Category I Structures , h -

O The MNRC facility does not have teg structu ,

c A"i 3.9 Mechanical systems antdomoone ey

~

.%Ah 3.9.1 at Rod Dr' ,

N3Yh% . rm ntrol rods are mounted on the reactor bridge The control!" T fa assem structure. The stand drive mechanisms manufactured by GA. A drive mechanism gu sg rods is shown in Fig. 3.1. The mechanism consists of a steppirg otor a ear, a rack and pinion, an electromagnet and armature, a dash afssembly, and od extension shaft. Rod-position data are obtained from po ometers. Limit sy , s are provided to indicate the up and down positions of the t and the down position of the rod. The drive motor is of the ng type and is instantly reversible. ' Die nominal drive speeds for the shim and the

' g rods is 24;iE./ min, respectively. The stepping motor speeds are adjustable with a of 7.50 in./sec. Rod reactivity insertion accidents, Chapter 13, use this i

v

l Rev. A. 7/5/96 3 22 uoawtexc CDM SCOT WQUNTING P(A4 WOU@HG PLAtt it I I MM R .

/

m

=/ i

: /e nucr. cRv. t ucua 90704 CRM CDA#. 3

% m - ,wct ao 0= cm  ! iyC000L.-R00 WOLAN 3

i,

"' N WRD, y K Q G CPAN TUBE uCET Pull-400 HOU$NG

,uu ,00 ,=

m == .' G PCT Po m Co m ot R00 e

m Imi RDCOR $R:0Gt

.a TYPICAL RACK-AND-PINION CONTROL-ROD-DRIVE MECHANISM FIG. 3.1

3 23 Rev.- A. 7/5/96

'T

[^J

~

During a scram, the control rod, rod extension, and naagnet armature are detached from the electromagnet and drop by gravity. The dashpot assembly slows the rate of insertion near the bottom of the stroke to limit deceleration forces.

The transient rod drive mechanism is shown in Fig. 3.2. This is a an 4Qustabl[ ast transient TRIGA pneumatic pulse drive system. The operability andgeliability%L ; system has been proven over many years of use at Sandia National Laboradries. 4K Ae W $

3.9.2 Core-Support Structure k w The fuel elements and graphite assemblies are supporti core-support s own in Fig. 3.3. The MNRC grid plate has been designed ickness and hole pattern identical to those of other TRIGA reactors with hexag 't 3.9.3 Instrument Guide Tubes wamesge yr The nuclear instrument chamber guide tu M by the core support structure as shown in Fig. 3.3. There tree .y two are used during normal operations. -w O /

(d 3.9.4 Neutron Source _

The startu Jource is appra+b ly 4 Ci -Beryllium held in a triple encapsulated stainles ntainer gftely3i

=

1 in, in diameter. The capsule is held in a war the reactor core centerline, see Fig. 3.3. Chapter 4 contain 'ons _

gives a de' jtion M di$kcapsule and holder, a w .__

3.9.5 Fuej Md @] g ies Y

='

wwa.. ~

Five . storage racksi tlf holding 20 elements each are mounted in the reactor tank.

Ou tank storage for it , lete core is provided by five pits within the reactor room. Each p' . a storage capac' for 19 elements. The storage systems are described in Chapter 9.

r Beam-Tubo , ssemblics

<W Moriginate within the graphite reflector approximately 90' apart. The beam tu scribed in Chapter 9 and mounting arrangements are shown in Fig. 3.4.

p s

v

Rev. A. 7/5/96 3-24

~

M W1Rt C'Jol cturER SWCH WOU@N0 PWI C'CNG M7E WTOR. PortnoutTUt ,

uouGNG PWE

_m, pWW l@CX DRNEN CCM uoren ont acas

' % / * *""mg

]% g,.*AO. W2M RC CRf4 CEM-a L -com-ax um

-l h

~" E "

STRONG-BW g & GAD.

G-,

MCG Puti-ROD HOUS:NG

- == : CONNECCNC R00 Puu-A00% % -

MbMW MDON)

,xr / i=- -

PNEVWADC CMCOt

C

)O 3

2 MsTCN ANO M. TON R:NG

_a 5"?! a P a m 24j-s 7ccN= - ,mt O

TYPICAL ADJUSTABLE FAST TRANSIENT ROD DRIVE .

FIG. 3.2

3-25 Rev. A. 7/5/%

D

! \

\v/

C4mtRAL Co. FactLITY

,. a f

! .h UPPt.04CPLATE

.s E  ! -

m u s h.e : omslada' E r r . n.u.

g nms unsa.

F

[ ,

4 i

j

- / .

. r i

I cons pannt

-,. 7-l gl  : .

gp . . . ,un eu==

y

&- 3 s

O i

j

( . .

J ,

i 1 l'  :

,.o.."lC<'nsir"3"2 # '

, l

l. v

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aorma aaa na

p<my rr 'ua<a

.-, .oo. ,.

is  ! o a

! O J

""""^"

6 y O s e o 3:

s# amunI m v muut4

, lOl t L.m . s. , i . } Ama AE.' M2EL12.m2 F21.AE.2.A&.Af 52.2.2.24. M.dn.E r/ anactonvanso,ttu 1.; co ..n . .u..no ., , .o.cio. . u.=, .u. w . .n m . ni

/^N U TRIGA REACTOR FIG. 3.3 L

I.-

Rev. A 7/5/96 A

E

• f 9'

,/

ROJOYMtf .

BEAM T1JEE

] /

E '

u Ro -

x -

20' .J s s

(TYP. 3 PLACES)

I _

N , _ _q A RD40TE REMOVMLE F1MCE JOINT- -

_ n ;7

.\ .m M 4 3'-lj/#

amtcrpa _

MP _d y ,

1 r i REActDR TMK TYPICAL IN-TANK REACTOR CORE AND BEAM TUBE ASSEMBLY FIG. 3.4

7/03/96 CHAPTER 4 4

MNRC TRIGA REACTOR O .

OI i u __ _ . _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _

Rev. A 7/5/96 TABl E OF CONTENIS 4-1 4.0 MNRC TRIGA REACTOR . . . . . . . . . . . ...................... 4- 1 l 4.1 I ntrod uct ion . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ..

J 4.2 R eactor Core . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Ah . . . . 4-6 46 4.2.1 Reactor Fuel . . . . . . . . . . . . . . . . . . . . . . b. . . 8. . . . . l 4.2.1.1 Fuel Moderator Element . . ,s. . . . .M.%......46 4.2.1.2 Instrumented Fuel Moderats ElemedtOMOh, . . . . . . . 48 jill 4.2.1.3 Evaluation of Fuel Graphite Dummy Element . . 4j . . . . .fg. . .Y Ls ..I Eleme5[IIesign .jjd!j%g*. .

4.2.2

............%,..g....., 4-12 4.2.3 Control Rods ,

4.2.3.1 Control Function . . . .{2d.?TW, . . . . . . . . .

4.2.3.2 Evaluation of Control RM%k" ............ . . . . .4-14

........4 4.2.4 Reflector Assembly . . . . . . f. .d.WE. . . . . . . . . . 4 16

. 4.2.5 Neutron Source and 11 older . ....

cr . . . . . . , :

4.2.6 Grid Plates 4.2.6.1 . . . . Plate Top Grid . . . . ./ I$' .... Y.h.........4-16

. . . . . . . . . 4 16 4-18 4.2.6.2 Bottom GridSt e .'p, ,.

4.2.6.3 Safety Plaf. . . . . : W M N. #.............

4.3 Reactor Tank . . . . . . . ..... r..... N N .............. 4 18

............. 4-21 O 4.4 Biological Shield . . gb . . . .

( 4.5 Nuclear Design . ./g . . . .

4.5.1

.'....f(...................421 TRIGA fpeIs . . . {Q. . .,.d. .....................421

. . . . . . . . . . . . . . . . 4 21 4.5 tAGWFuels . . . . . . . . . . . . . . . , . . , 4 21 scriptio ,ek}dr6fum 1.caded Fuels . . . . . . . . . . . . 4-24 4 formancs.d

4. ials Properties . . . . . . . . . . . . . . . . . . . . . . . . 4 25 esigd ...............................427

.. e gn CinariaWeference Cores . . . . . . . . . . . . . . . . . . . . 4-28 4 ' r Cole $iameters . . . . . . . . . . ...............432 Y";$ l4i Reactor Fuel Tampernure . . . . . . . . . . . . . . . . . . . . 4 32

- @N.5.4.1.2 Ilydrogen Pressure in TRIG 4.5.4.1.3 Zril Fuel Temperature Limits . . . . . . . . . . . 4 56

) 4.5.4.1.4 Performance of liigh Uranit.m wt.% Fuels . . 4-58

[i.5.4.2 Prompt Negative Temperature CoeiYicient . . .

4.5.4.2.1 ZrH Model . . . . . . . . . . . . , , . . . . . . . 661 4.5.4.2.2 Prompt Negative Temperature Coefficient Calculations . . . . . . . . . . . . . . . . . . . . . . 4 62 4.5.4.2.3 Codes Used for Calculations . . . . . , . . . . . 4 62 4.5.4.3 Cross Section Generation . . . . . . . . . . . . . . . . . . . . 4 63

("

\

U .i.

1 l

l l

Rev. A 7/5/96 4.5.5 Reactor Physics Analysis - Reference Cores . . . . . . . . . . . . . . 4 66 4.5.5.1 Ilexagonal Z Calculation Medel . . . . . . . . . . . . . . . . 4-66 4.5.5.1.1 Validation of flexagonal Z Model Analysis . . 4 70 4.5.5.2 Referenee Core MlxJ - Mixed Fuel Loading . . . . . . . . 4 73 4.5.5.3 Reference Core 20E #,. ,., ,. . 4.80. . . . . . . . . .

4.5.5.4 4.5.5.5 Power PeakingReference Factors Analygis 4.5.5.6 Neutron Flux Analysis . ./gh. . . .'qf.e;( ,..... 4 84 Core... Considerations . . . .[

.hf[.......481 4.5.5.7 Fission Product Release lYaction

1. /

O[ .....

6 i 6 Thermal Thermal and Ilydraulle and IlydraulicDesignAnalysis . . . . . N. . . . . .h.

. 87 4.6.1 4.6.2 Steady State Results ............ . . . . . . .kk.b,

~^ ..[4 88 4.7 Operating Limits . . . . . . . . . . . . . . . fd .............. 4-92 4.7.1 Operating Parameters . . . . . . . ............ 4 92 Limiting Safety System Setting . . . . . . .......... 4-94

. 4.7.2

f. (

c ?>

' 17 f' ,/M,y d,

n , .,-

m[T:;z w1 *s[f/

s  ;.

.. @ /f$}%i Wn&b%

1l:j g j(: fh.*#

+Qy ,m.

%QYf%y ;p 4,

.% h b[ ,I p wta gnfo w

fT

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gy bf Q

unl)ep$

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Rev. A 7/5/96

/m \

\v/ l IST OF FIGURES 4-3 4.1 Typical Six Control. Rod Core Arrangement . . . . . . . . . . . . . . . . . . . . . . . .

4-4 4.2 Reactor Assembly - Elevation . . . . . . . . . . . . . . . . . . . . . . . . . . . ., . . . .

Reactor and Reflector Assembly + Plan View . . . . . . . . . . . . . . . . . :0: . . . 45 4.3 4.4 Typical TRIGA Fuel Element Assembly . . . . . . . . . . . . . . A . . 8I . . .4.7 4.:s Typical TRIGA Instrumented Fuel Element . . . . . , . . . . . O$dT . . . . . . 4 9 4.6 8u 4.7 Graphite Dummy Reflector and Beam Tube Assembly . . . . . . .Element . . . . . . . . . . . . . .h . 3. . .Qh.%

4.8 h......d.#.Nh.....4

....d . M. 5.b... 17 4-19 4.9 Top Bottom Grid Plate Grid Plate . . . . . . . . . . . . . . . . . . . .h,N.

% . . 4-20.M.#.......%.1 4.10 S a fe ty Pla te . . . . . . . . . . . . . . . . . . . . . . OA 4.ii Reacior Tanu . . . . . . . . . . . . . . . . . . . . . #%. . . . . . . . . .. . . . . . .

4.12 Typical in Tank Reactor Core ani Beam Tube . hlib . . . . . . . . . . . . 4 23

. 4.13 Mixj Reference Core . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 -2 9 4.14 20E Reference Core ...... ...... 4t 4.15 Phase Dlagram of the Zirconiu'.n lly . .(mperamre

. . . ?/F. .........4

......... 4-33

. . . . . . . . . . 4 36 4.16 Equilibrium liydrogen Pressures ,

4.17 Strength of Type 304 Stainless as a Fj rature . . . . . . . . . 4-37 Strength and Applied Stress as uncti ..............4-38 4.18 (6f Tem _

. . . . . . . . . . . . . . . . . . 4 -4 0 O 4.19 Radial Power Distribution 14 Zril Eleme

( 4.20 4.21 Axial Power Distributio Subcooled Boiling Heat nsfer uel ent A ed for Thermal Analysis . . . 4-41 atera f./. . . . . . . . . . . . . . . . . . . . . 4-4 3 4.22 Clad < Temperature at M oint of ' d Fuel Element . . . . . . . . . . . . 4-44 4.23 Fp6lShdyTempdad Midpla: il Bonded U ZrH Element After Pulse 4-45 4.24 StifbNi&at FlulMMidplane of Well Bonded U Zrli Element After Pulse . . . 4-46 4.25 SuifacEM Flux DIElbi5t$ for Standard Non-Gapped Fuel Element After l ukk 590%$$p. . . . , . . . , , . . . . . . . . . . . . . . . . . . 4-48 i

4.26 Distribu'f$for Standard Non Gapped Fuel Element Surface'Jhat A!siWii ulse,'Q Mp%3y5; . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

4.27 $brface lleat Flinx:Distihkiution for Standard Non-Gapped Fuel Element A fter Pulse h,,,, y'25 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-50

4. Surface lleat Flu at Midpoint Versus Time for Standard Non Gapped Fuel Element After Pulse . . ................................4-51

< l Re fe re nce Cdre' M ixj . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-74 or e 2 0 E . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 -7 8 4 fle"ction of Central Experiment Facility with Al/C Plug . . . . . . . . . . . . 4-82 4.3 RC REl A PS Model . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-89 4.33 M NRC Power D!stribution . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-90 4.34 Calculated Coolant Temperature and Void Distribution . . . . . . . . . . . . . . . . 4 93 p

(v) .iii.

Rev. A 7/5/96 LIST OF TABI ES 41 Summary of Fuel Element Specifications . . . . . . . . . . . . . . . . . . . . . . . . . 4-10 42 Thermocouple Specifications . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-11 43 Parameters for TRIG A Fuels . . . . . . . . . . . . . . . . . . . . . . . . . , . fh . . . 4-24 44 Physical Properties of Delta Phase U ZrH ...........4..

. . . . . . 4-5 9 45 46 ORR in Pile Irradiation Cross Section Accuracy Test -Infinite Lattice of 20 Test of "B Cell Average MicroscopicN ll#Parameters h . . . . . . 4-64 Absorptio/pfoss

.I.%..... Se 4-7 Descriptions of Some important Element Cell Qes . . . /s[. . . iVl#'..

k 8 48  ? 4-69 49 Compositions Used in the llex Z MNRC Modh .N. /f. . . . . . .. . . . .

4-10 Symbols Used in Reactor Grid Loading Diagra L Comparison of Measured and Calculated Excess' of flex Z Modeled 4 11

............ 4-72 Cor es . . .of. Measured Comparison . . . . . . .and

. . Calculated

. . . . . . . Rod

. . . . . $. .rths Atodeled Cores . 4-73

. 4-12 Fuel Dummy Substitution Worth in MixJ ......... 4-75 4-13 4 14 Calculated Rod Worths in MixJ W'c ..... ...........

. . .d.734-77 4-76 4-15 4 16 Calculated Rod Worths in 20E . (E@. . . . , Fuel Dummy Substitu 4-17 Power Peaking Factors . . . . . h. . . .... ,, . . . mN,...............

. . . . . . . . . . .4-83

. . . 4-79 4 18 Core-Average Neutron Flu ..... ..... ................... 4-84 4-19 Ileat Transfer and liydraylic; Parameters For O tion at 2 MW With 101 Fuel Elements .8 M f 4 20 Limiting

- Safety n S,ypom httings . .

[.......................4-91

..................... 4-94 hy opn gs . e%;,wnA jms Y,  ;?& E0{ b$f

%q%g%;% %W, r v4 VgdVA s $a9%g %n V

41 P ' g~My ,; p ?

d y &p q d A

y fs W G>f ff'I i

w y kk g .f

)ry;~ 'x .tn l\"  % _ u;?Nl

?f

%7 iv-

Rev. A 7/5/%

O REFERENCES

(]

4.1 U. S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation, " Safety Evaluation Report on liigh Uranium Content, low Enriched Uranium Zirconium liydride Fuels for TRIGA Reactor", NUREG-1282, Aug.1987. pv UZR-14A, "TRIGA LEU Shrouded Fuel Cluster De gn for t etween 2.0 4.2 fik MW and 10 MW (Thermal)," GA Technologies.

M kg&hbk ff Y

4.3 UZR 23 "Special 4 Rod Unshrouded Cluster y T Technologies (Feb.1988). e g; 4.4 ANL Supplement to Final Safety Analysis Repo > Upgrade,1996 I

E. Baker, U.S. Air Force Safety Agency, Pe alC n,1996.

4.5 n in U fZirconium Alloys, "

4.6 Merten, U., et2L, " Thermal Migrati General Dynamics, General Atom f

hsu A-

,1962.

G. A. Report, " Low Enriched IGA F Wat Safety Test," GA-15413, 4.7 .

June 1979.

U 1 -

fariation#(licat Transfer, and Void Volume 4.8 Johnson, II. A., eLal, mperat Dev$pment in tholfrant ent At $K8diling of Water," SAN-1001, University of Calhipirnia, Berk %Iik. "7 4.9 Y h$ W,lI., llemit-)'

McAdampj$h N

QNbI 'w 'ulon,3rd Ed, McGraw-lilli Book Co., New York, 1954.' % li h

,dMsCh 'QW Y 4,10 Sjpiitow, E. MMtd AD. Cess, "The Effect of Subcooled Liquid on Film Boiling,"

td Hent Transfer, lidp)%;*ifj %~156(1962).

4 Speigler, P., cid, " Onset of Stable Film Boiling and the Foam Limit," Int. L lleat nnd Mass Trnnfrir, 6, 987 989 (1963).

O c%, bilydrodynamic Aspects of Boiling 11 eat Transfer," AEC Report AECV-

TIS, ORNL,1959.

4.13 Rohsenow, W., and 11. Chol,11ent. Mass and Momentum Transfer, Prentice liali, 1961, pp. 231-232,

(

-V-

Rev. A 7/5/>6 4.14 Ellion, M. E., " A Study of the Mechanism of Boiling lleat Transfer " Jet Pro.nulsion Laboratory Memo, No. 20 88, 1954.

4.15 Coffer, C. O., et al, " Characteristics of Large Reactivity insertions in a liigh Performance TRIGA U Zril Core," General Dynamics, General Atomic;I)) vision

, Report GA-6216,1965. b f/

A 4.16 Fenech,11., and W. Rosenow, " Thermal Conductakof M 80kN faces in Contact," USAEC NYO 2130,1959. jf {

4.17 Graff, W. J., " Thermal Conductance Across Mdal Jointsi MachineMEM" 15, 1960, pp.166172. Qg.F T5F Fenech,11., and J. J. Ilenry, "An Analysis of ajTif kbu tact Resistance," Trans.

4.18 Am. Nucl. Soc. 5, 476 (1962). 4 %3g

.. I k S.

e Bernath, L., " A Theory of local Boil and its NIlon to Existing 4.19 h'Iium Series, Storrs.

Data," lleat Transfer - Chemical ,

r; 'incerij; %

, A.11. " Quarterly Tec , leal Re ,SP t," April, May and June 4.20 a

4.21 Dee, J. B., et. al, " An. r Core e R ctpr," General Dynamics, General Atomics Division R(poriUACD- 6tnent 2,1966.

'k

~

j$[A d$N .

Disassociation Pressures of Modified SNAP Fuel," Report 4,22 Jifthisonlii E., "H" ^

lonal,19M.

NEALSR-9Ib5, Atom

'Myg 9 4.23 My7Jimnad> d G. L. Copeland, " Final Results from TRIGA LEU G. B,$$AQ'idiatioh Examination and Evaluation Following Long Term Fuefl iost Irr de' sting in the O'RFG$*-A18641, Nov.1986, V [}%f f

4 [4f N. L. Baldwin,JC C. Foushee, and J. S. Greenwood, " Fission Product Release from

%1 TRIGA LEU Re$ctor Fuels," GA A16287, Nov.1980.

f;;GL M 4I21$ET, Stab and G. B. West, " Post Irradiation Examination and Evaluation of

$*7RlG5ffEU Fuel Irradiated in the Oak Ridge Research Reactor," GA-A18599, Ma 1986.

4.26 Beyster, J. R., cLal, " Neutron Thermalization in Zirconium 11ydride," USAEC Report, General Dynamics, General Atomic Division Report GA-4581,1963.

-Vi-

Rev. A 7/5/%

4.27 Woods, A. D. B., cLaL, ' Energy Distribution of Neutrons Scattered from Graphite, Light and lleavy Water, Ice, Zirconium liydride, Lithium liydride, Sodium liydride, and Ammonium Chloride, by the Beryllium Detector hiethod," in Proc. Symp.

Inelastic Scattering of Neutrons in Solids and Liquids, Vienna, Austria, Oct,11 14, International Atomic Energy Agency,1960.

A 4 p

4.28 Adir, J., cLat, ' Users and Programmers Manual foy,the GGd'kkuffkroup Cro Section Code,' General Dynamics, General Atomics Divisick t(OA 7157,1967.

4,29 Lenihan, S. R., ' GAZE-2:

kV A One DimensionalgMultigrpup),t$i Ne ty hh Theory Code for the IBM 7090,* General Dynkinics, General Atomic'd rt GA-3152,1962. M- S7

@Yk@F h 4.30 Dorsey, J. P., and R. Frochlich, ' GAMBLE-5gf " .for the Solution of the Multigroup Neutron Diffusion Equations in Two Di .th Arbitrary Group Scattering, for the UNIVAC-1108 Com Gulf Ge .

ic Report GA 8188, 1%7.

4.31 Lathrop, D. K., "DTF IV, A F RAN. lying the Multigroup Transport Equation with Aniso ic Scaid ings, j Report LA-3373, los Alamos Scientific 12borat ew h co,19 4.32 Adler, F. T., G. W.,111L n,and . N 9tdhtim, "The Quantitative Evaluation of Conf. Peaceful Uses At. Energy (A/ CONF.

Resonance Integr (rfroc. 2n, 15#/1983), Ge nationaP Energy Agency,1958.

%AtMlh Jr., Gi&dh$Iral Atomic Inc., 'TilERMIDOR - A FORTRAN 11 Code 4.33 BroMiMA[

for Cal' ~ Nelk'1(Sainding Kernal for Bound 11ydrogen (A Modification of Robe Mblished%tE.

belkin,M.S., Dj &ng of Slow Neutrons by Water," Phys. Rev. 119, 741-746 4.34

[(1960). V W., cLaL, ' Neutron Thermalization by Chemically Bound 11ydrogen McReynolds, fn Proc. 2nd intern. Conf. Peaceful Uses At. Energy (A/ CON and Carbo.n/[A' 5/pgGeneva, International Atomic Energy Agency,1958.

kp

4. M@iitemore, W. L., ' Neutron Interactions in Zirconium liydride," USAEC Report GA-4490 (Rev,), General Dynamics, General Atomic Division,1964.

A

.vil-

Rev. A 7/5/96 4.37 Bell, J., 'SUhthilT: An IBht-7090 Program for the Computation of Crystalline Scattering Kernels,' USAEC Report, General Dynamics, General Atomic Division Report GA 2492,1962.

4.38 " Documentation for CCC-576/WlhtS D4 Code Package," RSIC Compuje) Code Collection, Oak Ridge National Laboratory,1990. /( g#

dtonal C ibon,1995.

4.39 W. L. Woodruff, Argonne National laboratory,yf N$

f t$biatL, 4,40 R. N. Blomquist, "VIh1," Proc. Int. Topl. Aftg.Qfdvances,l f ~$1. 28 5, - h1ay Computations, and Reactor Physics, PittsburglGy, Apri

p. 39.4 2-1 American Nuclear Society, ISBN: 6 89448 l61-4, 1991.

f 4.41 D. C. Wade, "hionte Carlo-Based Validation o. MC2-II/SDX Cell Ilomogenization Path," ANL-79 5, Argonne $ tionaL 1979.

4.42 R.D. lawrence, "The DIF3D Nodal tro Neu/D on for' and Three-Dimensional Diffusion Theory Calpdfallo'n? n nal Geometry," ANL-83-1 Argonne National Laboratory,1 37 [% r W. L. Whittemore, Gener mics tsonal munication,1995.

4.43 W. S. Yang, P.J. Finck ' d 11. "Recomtruction of Pin Burnup Characteristics 4.44 Ery, " Proc. OfIntl. Cortf. On the frotpRodal Calculat finbtlon, Hexa D Computatio'n, hiarseille, France, Vol. 2, P/pfcs g Reac 91}{j%!MMg2Mpgil DA 23 4.45 hi. T. simnadMThe U- Hihil'6y: Its Properties and Use in TRIGA Fuel", General Atomicd&d 7%3 3 Feb.T9f0.

,gP 4.46 4 .KE. Carlson, c QREIAP5th!OD3 Code bianual Volume 1: Code Structure, System g/ hiodels, and Methods. Sol %dr(): NUREGICR 5535, EGG-2596, EG&G Idaho, Inc.,

tf

.$ June 1990.

}. V. Del Be , General Atomics, Personal Communication, June 1995.

! We/%

4i48gl%fa SAR 1

-Vill-

Rev. A 7/5/%

4.49 G. B. West and R. H. Chesworth, " Update on World Wide Use of TRIGA LEU Fuel including inss of Flow Tests." Proc.1990 Int. Mtg. On Reduced Enrichmentfor Research and Test Reactors, Newport, Rhode Island, September 23 27,1990, p. 249, ANL/RERTR/TM 18, CONF-9009108 Argonne National Laboratory,1990.

4.50 R. H. Chesworth, J. Rarvi end W. L. Whittemore, "TRIGA r ctor Activities Around the World," Trans. Am. Nucl. , 41, 1.

U k s i

b .ix.

41 Rev. A 7/5/%

('

t 4.0 MNRC TRIGA REACTOR 1 Introduction n y

The MNRC reactor is a hexagonal grid, natural convection water coo TRI reactor  !

designed to operate at a nominal 2 MW steady state power to include square wave operationc Prior to 2 MW upgrade the reactor operated a:Wiiominai ' steady state power with pulse and square wave operations. The reactor util% speciaM " ' yphite radl[

reflector to accept the source ends of four neuson radifgraphy beam tubest beams,4 terminate in four separate neutron radiography bays. eac core is I eter and 7.47 m deep.

bottom of a water filled aluminum tank 2.13 m (7.0 The water provides adequate radiation shielding at the t ,

tank.

. Standard TRIGA fuel of three types, 8.5,12.5, and t %, .

Ms, an enrichment less than 20 % U 235 can be utilized in reactor core lg g. Initiatropepitp will be with a -

mixed core fuelloading (e.g.,8.5/20 and 2 c. MixJ refetsnee core section 4.5.5.2) to be followed later with operation of a ful D (se E core section 4.5.5.3).

Results of the analysis of reference coreJdi ings . . .

O wt % fuels are presented in this section to illustrate the safe operftlon of e fuel . . . _

s. Mixed core loadings p utilizing 12.5 wt % standard TRI ^ Melcan be opetited safely at 2 MW, however, no Q core loadings of this type are pr fplan .

TRIGA fuelds characterizcd ierent ssion product retention, and the demonstratiid$b{lity to aterq lth no adverse reaction from temperatures up to !!'00fC32012'F).' ent safety of this TRIGA reactor has been demonstrated by the extenside%nce acq'O .similar TRIGA systems throughout the world. This safety arises fionfMlar proI" 'e temperature coefficient that is characteristic of uranium-zircoaliSQ , fuel m , rator elements used in TRIGA systems. As the fuel temperatugincreases) cient immediately compensates for reactivity insertions. The i a mechanism whereby reactor power excursions are nega compensation ter ted quickly and dhly(See Section 4.5.4.2),

ding the stead te power rating of the MNRC from 1 MW to 2 MW involves primarily sfer and 1-hydraulic considerations but there are associated neutronics issues..

sideration is the need to increase the capacity of the cooling system so that temperature can be maintained. The irregular spacing of fuel elements that is ch of the circular pattem in the current reactor grid would lead to excessive local fuel and coolant temperatures if the power level were doubled. Consequently, the upgraded design'

~

. has an hexagonal grid pattern, which characteristically has a uniform ' element spacing (1.714" pitch).e This necessitates replacement of three major components: the top grid plate, the bottom

/^ grid plate and the safety plate. The grid change makes it possible to create a more accurate and

( computationally efficient neutronics model. At a given power level, the neutronics performance i

1 i

Rev. A 7/5/96 42 of the reactor is not signincantly alTected by the changes in element spacing, so doubling the power level essentially doubles the peak power. Power distributions were predicted by neutronics calculations and these values were input to thermal hydraulic calculations so that the fuel and coolant temperatures could be predicted. The main thrust of the reactor physics analysis, as far as safety is concerned, is to identify reactor grid loading patterns that have acceptable values of peak power (temperature), excess reactivity and shutdown [cactivity.

M%ty The reactor physics analysis takes into account the reason d bebil the f'o#es u"pgrade. The purpose of the upgrade is to raise the Oux level to make thkMNRC morEeffective in meeting present mission and allowing expansion ofits capabilitieQThe designlri'g'of grid ibading patterns included evaluation of ways to improve the nedten Dux (ma'gnitude' arid $sd or .

these applications. In addition, a large, central core irrhiition fa'cility providesT0exibli environment for in-core irradiations. The reactor physiWaislysi(includes evaluatior%f the aux that will be attained there, as well as reactivity and poweryakmg issues associated with this facility. [ ypQ bmA A typical mixed core arrangement of reactor fuel elbisnts, grapfiltb%me'nts, and control rods is shown in Fig. 4.1. The two operational corftasirigdd,up of approximately 100 fuel-moderator elements including fuel followedlcontrolj6drand approximately 10 graphite elements.

/f p f %)?f y The reactor core assembly is n,FigureN2.

showpj. O The r}eactor grid plates and fuel / loa contained within a core barrel appr6ximatelyl24 inches,lEdiameter by 40 inches in height.

The diameterreactor and 23 in, and reflector assembifl dating tintanihtial neutron radiography b high acd>n{ano Submergodiffthe reactorftank?the reflecthf ins mb'ly rests on a platform, which raises the lower edh6fjhlre0cetNiss$ddly.about Ift above the tank Door. Coolant water occupies core soldt MCooling of the reactor fuel elements is provided by about one-third %f the[ihe tank WaicMiit heat dissipated to the tank water is rem natural convelEtidn'of circulating primary spGm thed(ieactWphirIto tattwater) is removed byaa pri'mityTo secondary system secondary cooling tower. heat exchanger. Th M 7[f The NNRC reactor normpilf; operates in the steady-state mode, however, pulse and square wave" modes of operatioh are also provided. During steady state operation, the reactivity in ths'dactor core is contfolled by up to five standard control rods and drives and one pulse rod a'nd Elrive. An optioriaflow worth Type 304 stainless steel control rod can be substituted in the MlxTCore for a staridard control rod. The control rods have fuel followers and are sealed in

%ction_contains se a neutron absorber (boron carbide in solid form) and neutron absorber is a fuel follower section containing 8.5,12.5 or 20 wt % uranium enriched to less than 20% 2"U These control rods are attached to drive assemblies mounted on a bridge which spans the tank top. The drive assembly consist of a motor and reduction gear driving a rack and-pinion. The control rod together with i's segmented connecting rod is connected to the rack through an electromagnet and armature.

l

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- TYPICAL LOCATION OF PNEUMA TIC TRANSFER TUBE TYPICAL EXPERIMENTAL TUBE LOCATION (TYPICAL OF 4 PUCE 5) 5HIM ROO TRANSIENT ROO FUEL OR GRAPC'TE ELEMENT POSITIONS

" ~

neoUurlNo ROO O TYPICAL SIX CONTROL-ROD CORE ARRANGEMENT l

FIG. 4.1

Rev. A 7/5/96 44 O

CtNTRAL CORT F AClufY L'PPER CRID PLA tt O.  ?

grmmf\a T man .} El 3 3 i p 3

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v iO / RiaCroRtAux orreu ummmuurmxme_uwemmmmmawre sce :szerr-NOTt; CORT BARRtt,is SUPPORito 8Y THE REPktCTOR ASSEWOLY BASE SU*PORY(Stt F10 4181 O

REACTOR ASSEMBLY - ELEVATION FIG,4.2

45 Rev. A 7/5/M i O

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l < 3 4..

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- REACTOR AND REFLECTOR ASSEMBLY - PLAN VIEW FIG. 4.3

Rev. A 7/5/96 46 pulsing is accomplished by the use of a rod attached to an adjustable fast transient drive. The transient rod drive is also mounted on the reactor bridge and is a combination of the standard GA rack and-pinion control rod drive and the standard pneumatic fast transient rod drive. The rod is a 44.25 in, long by 1.25 in. diameter aluminum tube. The top portion of the rod contains solid boron carbide for neutron absorption and the bottom an air followefjsection. A complete description of both the pulse and steady state control rods and# rives 7 contained in Chapters 4 and 7. f. [f$h[

/D %kh The Instrumentation and Contcol System (ICS) for the TRIGA reactorp4pgtogujer-based design incorporating a GA developed, multifunction mfctoprocesso$ based'neuttert'moni channel and an analog type neutron monitoring channf18Thesep6 units prM), d i t safety channels (percent power with scram), wide rangelldp6wer (below sourfeleief*fo full power), period, and multirange linear power (source lehl;to full p,ower). The control system logic is contained in a separate Control System ComputeE(CSC) pith a color graphics display

. which is the interface between the operator and the tcEtor. betailidth,e centrol system logic can be found in Chapter 7. p W ud?-

gb hbs N 4.2 Reactor Core MFNMM ,h5$$hdW

/

M l This section describes, thsg and f where

, the safejy app plate, the fuel [jid platlate, ev modcretor assembly, the reflector elements, including instrumentedassembly,flNnents,neutron' source, the graphite dummy elements, the control rods and drives, theherimen lacilities$iInd the beam tubes. A detailed description of the control ro'd qstem can $5%ImdWChapter 7.

h?%F eel 9deh hh kW 4.2.1 Reacql$l%

M  %$$W 4.2.1.1 Fuel Mddein't'or,ElemeniiM9 awmp g3 s;

gm n:w%

The actigjart of chh'fuelslement, shown h Figure 4.4 is approximately 1.43 in, in IQrEactor fuel is a solid, homogeneous mixture of uranium-diametef andalloy!c zirconfum hydride 15 in. lon'g@inialning 8.5,12.5, or 20 % by weight of ura

$U./ The hydrogen-to-Mrconium atom ratio within the MNRC fuel varies from 1.6 to 1.7.

To facilitate hydridingva small hole is drilled through the center of the active fuel section and a ziMonium rod is.irise'rted in this hole after hydriding is complete.

8@Shd7 Pge inent is' clad with a 0.020 in, thick stainless steel can, and all closures are made by heharc@ welding, Two sections of graphite are inserted in the can, one the fuel, to serve as top and bottom reflectors for the core. Stainless steel end fixtures are ,

attached to both ends of the can, making the overall length of the fuel moderator element approximately 29.0 in. Table 41 gives a summary of fuel element specifications.

47 Rev. A 7/5/96 j

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STANDARD FUEL ELEMENT L

TYPICAL TRIGA FUEL ELEMENT ASSEMBLY

l. FIO 4.4 l l l

Rev. A 7/5/96 48 The lower end fixture supports the fuel moderator element on the bottom grid plate. The upper end fixture consists of a knob for attachment of the fuel handling tool and a triangular spacer, which permits cooling water to flow through the upper grid plate. The total weight of a fully loaded fuel element is about 3.18 kg (7 lb).

Ar pg 4.2.1.2 Instrumented Fuel Moderator Element r

jwYy[

it An instrumented fuel moderator element will have three ~

thsinocoul es"cinli6dded in th 2cated abou As shown in Figure 4.5, the,f,/jsensing tips of a.e f 0.3 in. Radially from the vertical centerline. [ '%

%B The thermocouple leadout wires pass through a seal in ?6%r end fixture. A'fesdE0t tube provides a watertight conduit carrying the leadout wiresitiov(tim, water surface in the reactor tank. Thermocouple specifications are listed in Tablefj2Nn othebrespects, the instrumented

. fuel moderator element is identical to the standard element.

ylb h

s g $yn$h 4.2.1.3 Evaluation of Fuel Element L Qg v Designy&sf r

/

General Atomics has acquired extensive,c,xierience/ n and operation of high ions of this Safety Analysis hydride, stainless steel clad fuel elements. As shown in Report, the stresses associated witgtdjfuel ardEladding, mperatures in all modes of operation, normal and abnormal (arewithin fuel temperature safety limits, it is concluded that the chedic. bility ZiN h fuel moderator material does not impose a sdfefftlimit onfidoI$perationMNcI[ ion 4.5.4.1.4 9M gh DimensionM@$hil ty of the NSMITdel element has been excellent operation. AilslNis of the heat'ridevid fromI elements that touch due to transverse bending shows that the 5:06tSilnot reshlj iffhot spots that damage the fuel.

f?' Q$Q Tests )dve been conducted onJRIGA fuel elements to determine the strength of the fuel eleme,dt clad when subjgiddjo internal pressure. At room temperature the clads ruptured at apt 3 2050 psig. This,cdrresponds to a hoop stress at rupture of about 72,000 psi which comp;ves favorably pth the minimum expected value for 3(M stainless steel, ffNk /

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l FIG. 4.5 j

,,.--A-,--,,-a-n-n .,.n , --,,,--~e , - . , , , - - - - , , - - , - - - - - ,, - > ~n--

Rev. A 7/5/96 4 10 TABLE 41

SUMMARY

OF FUEL ELEMENT SPECIFICATIONS.

Nominal Value

.. . .L Fucl Moderator Material

/

• r, ny ll/Zr ratio 1.6 to 1.7 (actual)h 4.. %, i , gq 4

.8.5l12.5, 1 0MW . ,,EA.

Uranium content Enrichment (U 235) /Y s 20%7Y[

!!attA 39 g/$lement V w- ~$%df 8.5 wt % Uranium 235 O_ , jf g .cr s *y p? ~ 'i S5'g/ element 12.5 wt % Urani*.m 235 n

O'%a, hc s99 ekiment 20 wt % Uranium 235 y m : ,r. A I;43 in! A Diameter g n ,'ax .a y Length .41 15 inE1 Graphite End ReDectors );bp';shEtu [ Lower 9 it ws e 20%b/

Porosity // / 20 %

ty // at Diameter e r J/ // 1 AJ in. 1.43 in. _

1xngth [/ M Idaries 3.47 in,

/gl sCladding ./U-N NIUi$c

... w ,-

t 3

(i V M aterial @ . 3 %g _ Type 3N SS

1. (% ea.; Wall m

+ thicknessM:PA

, c, 0.020 in.

1. Mih3h. Nb[ 22.10 in.

1,nn . . e.

p c9 End Fixtures amt Spacer Type 304 SS

/>/g Overall Eieddd..

q fb]j Outside' diameter 1.47 in.

NA ,.

Length 28.4 in. and 29.5 in.

QQ m.;: -

K , %'~ h C:AVelpht  ? lb

% y:%g:. v.9 -

9

4 11 Rev. A 7/5/96 N

I .

TABLE 4 2 TilERMOCOUPLE SPECIFICATIONS.

Type Chromel alumel ,, ,

h' Wire diameter 0.005 in. ,s /YNb !7 Resistance 24.08 ohEsIdoublelosi'llfdh

• 68'F[/'.

AV 'T6.,ti .

Junction Grouk gf7 9. [ig,,

Sheath material StainlNSk Sheath diameter 0.0%1sn.W y++$uh .. . - - Q

~~ Insulation Mg, Nhhk #

Lead out wire j[ %Ik$)y hY Material [ [oEf M iE (s Size d [0AWQf~

Color code ,[pf d Chraff- yellow (positive) n Afhf[ @ 45el - red (negative)

, ResistamiW4 0.59 ohms / double foot at s

5 - - %- . Yd. . _ 75'F kkh pw gs~,_ e,rs s.

v

-yh,sj f(flkf ,

fjRV

,f  :.:.

Mi- ..

/$di. //lf ggh n

X_)

Rev. A 7/5/96 4 12 1 4.2.2 Graphite Dummy Elements -

Graphite dummy elements, shown in Figure 4 6, rnay be used to fill grid positions not filled by the fuel moderator elements or other core components. They are of the same general dimensions and construction as the fuel-moderator elements, but are filled entirey,with  ;

graphite and are clad with aluminum. Graphite dummy elements are aftjntegppart of core )

loadings being utilized as neutron reflectors. 6 5f 4

}&p. 8 4.2.3 Control Rods j;f Mn*%g JGGi 4.2.3.1 Control Function [

wq w The reactiv:iy of the MNRC reactor is controlled by up't6fMi standard control. rods. The

~_

control and transient rod drives are mounted on a bridge;NQggthe reactor. tank. The

, drives are connected to the control and transient rods'through's connectilig ro']

d assembly.

3 Every core loadir.g includes four or five fuel kolrod control rods that have uppermoe oction is olid boron carbide a fuel section below the absorber section.

neutron absorber. Immediat:ly below t torberdEth$g$idM6'n consisting of U-Zrilu enrkhed in 235 to less than 19.7%. T[ weigh heentKurah' Ism in the fuel is either 8.5, 12.5, or 20, depending on the cor ing. bottomyction of the rod has an air filled Tre seal in Typg3N stainless steel tubes approximately vold. The fuel and absorber sec}

43 inches long by 1.35 laches i t ameter JV One contr61 A i ital.coctaining[k 5hlusstec

'8n absorber section pnd no fuel follower may nadiff Monly four fuel-folicwed control rods. The low reactivity be inclu'dsd%j%

worth of tidsind s

s ver' teamlylty control. A detailed description of the control rod system, contro1M ivest isifiMjded in Chapter 7.

,.y kgA Y 3 gg a[g ,.[ l f; N Aih o

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v' GRAPHITE DUMMY ELEMENT FIG. 4.6 l

Rev.A 7/5/96 4 14 4.2.3.2 Evaluation of Control Rod System The reactivity worth and speed of travel for the control rods are adequate to allow complete control of the reactor system during operation from a shutdown condition to full power. The scram times for the rods are quite adequate since the TRIGA system does not relyjon speed of control as being paramount to the safety of the reactor. The inherent shutdow'rt mechanism of the TRIGA prevents unsafe excursions and the control ion.

system is used$r shutdown of the reactor and to control the power level in sf$dy-state $' aMh

/*W 4 wvA The reactivity worth of the control system can be varied by the placement'of the c4ntrolg in the core. The control system has been configured th"$rovidefof the exceINEctislIyj needed for 2 MW operation 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> per day (includikE6ori6verride) and wlitN8pable of providing a shutdown reactivity greater than 30 centsNEYES Nith the most reactive control rod in its most reactive position and moveable experimenb?hY th$iflmost'reftive positSn. -4 Y Nk5h ~"

The* nominal speed of the control rods is about 2 Mapper min arid lthsMel Ts about 15 in, llowever, the drive system is capable of movingtoM 14a maxirkmurjipeed of 7 ?./2 in, per sec. If the rods travel at the maximum speed /tne reactor <cim be taken from low power to full power in less than three seconds. Also,iittee seconds. FIdt}W@511 r power can be accomplished in less than e

once. The system is fail-safe, that ilhiultiplofailures ar$'rkulred to get uncontrolled rod withdrawa% at the me.imum speNp pr 8 f t hb iWW Af 4.2.4 Reflector Assembig i M!)h dN b The reffectorS6wn in Figf4;7h@a ring shaped block of graphite that surrou radially. 'IWyMhhis ab55t43 ln.1 thick radially with an inside diameter of about 21-1/2 in, and a heigfit'offabout 23 in.Tliiffaphite is protected from water penetration by a leak-tight weld 6DilStiniEm.can. ' Vertical tubes attached to the outer diameter of the reflector assemblypriiit achudiedreprod'ucible positioning of fission and ion chambers used to monitor'Teactor operatiEiUN?

T y/ Ejf T1pfeflector was redep[gned to accommodate four tangentia This)ew design provides space for a removable in tank beam tube section referred to as the (eflecht insert.I) four reDector inserts begin the shaping of gracially widening, conical neutron)eams from[the reactor core to the plane of radiography. Each c

fthidho'gWgraphite surrounded by a leak tight aluminum can. The inserts fit into four perpendiciilar cutouts in the reDector assembly with each perpendicular cutout being tangential to the reactor core.

9

4 15 Rev. A 7/5/%

20' REMOVABLE BEAM TUBE TYP 4 PLACES l'

REFLECTOR k

.~ ~)- Y

J.

m

/ "

N ,

ic-1/ . .

L \

E l i 20' /' I'

. _ V' -

/ '

s n -

g-_s' .

g CORE REGION 20' (INCLINE ANGLE TYP)

REACTOR AND BEAM TUBE ASSEMBLY FIG.4.7 l t

i

l Rev. A 7/5/96 4-16 Since the reactor is located below grade level, the inserts are inclined to direct neutrons I upward towards the plane of radiography. Three of the inserts are inclined at a 20' angle and one at a 30' angle.

The reDector assembly rests on an aluminum platform at the bottom of the tankgPour lugs are provided for lifting the assembly.

h

,%_ xM

-,a 4.2.5 Neutron Source and lloider C?

S% QifM

, n#h h e Wm f/).

6 foretcactor'startup//

A 4 curie (-9 x 10 n/sec) americium beryllium neuttprt; source is tis The source material is triple encapsulated in welded stainless ste<f The capsile'isN approximately 1 '.... In diamster and approximately 3 iB%g;jThe neutron sotif6d .

in an aluminum cylindrical shaped source holder. Thenst'c 6 bider can be installed in any fuel location in the top grid plate. A shoulder at the uppFeSNo(the holder 30pports the -Q lOL cavity in the lower assembly on the upper grid plate. The neutron sourc61s coni $i@hM1pp of the cdre portit n of the source holder at the horizontal centerlines portions of the holder are screwed together andylindd.TSince theligend fixture of the source holder is similar to that of the fuel ele $er717tf06uIcepoldetrean be installed or tif6pp6;end fiiture has a small hole removed with the fuel handling tool. In,5!Iltion,$ be insertedi$ facilitate handling through which one end of a stainless steei wire m from the top of the tank. f[ 7 4.2.6 Grid Plates q A i,Y Q h%M/

8 4.2.6.1 4g T6p 5

gn, Grid Plate [g$%gg DF lNe il'an alurdinshi'to 21 inches in diameter and 1 1/4 inches thick The top grld [ini(iidcentral Mgioh) tlEat provides accurate lateral pos (3/4 inches th components Tisplatelisupported[f six one-half inch stainless steel rods that are attached to the bottorfgrid plitiath anodized to resist wear and corrosion, y[ M9/ n One htindred twenty one_(12,1) holes, on a 1.714 lach pitch are drilled into theiop grid plate in sit idxagonal rings arojind a central hole to locate the fuel moderator and graphite dummy elements, the control rods, guide tube, and the pneumatic transfer tube (see Fig. 4.8) The dentir lJole accommdd'ates the central plug. Small holes at various positions in the top grid err.cinserdon of foils into the co.c to obtain Dux data.

Wy n.: Oy hhy& e w A heingonal secthn can be removed from the center of the upper grid plate for the insertion of specimens up to ap:.roximately 4.4 inches in diameter into the region of highest aux; this requires prior relocation of the seven fuel elements from the center to the outer portion of the Core.

4 17 Rev. A 7/5/96 Etu0VABLE C(NTER GRID N ATE INttRT TYPICAL GRID PLAf E IN5tRT (4 PLACE $) N p' -\- - -

Y ih _ ]

r r

J S AR HEX GRID NUMBER SYSTEM (HEX RING A, B, C, D E. F,0) .

U 00 ,

o o 1

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Q B41 A41 8 44 O g/

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I  %. # 6M116 ALUMINUM i 25 in THICK 121 GRID LOCATIONS TYPICAL CONTRCL ROD LOCATIONS TOP GRID PLATE FIG. 4.8

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

v .

I i

l 1

Rev A 7/5/96 4 18 Four triangular shaped sections are cut out of the upper grid plate. When fuel elements are placed in these locations, their lateral support is provided by a special fixture. When the fuel elements and support are removed, there is room for inserting specimens up to 2 fra:hes in diameter.

The MNRC reactor is equipped with a TRIGA type pneumatic transfer,'systemi irradiation of small specimens. The in-core section of this system is typically lontee iri tfie outer portion of the reactor core. g% $k The differential area between the fitting at tiie top of ths h elfgolegri eleme6I ahd the top grid plate permits passage of cooling water thrgthe p late, g%5ggy 4.2,6.2 Bottom G:id Plate M[bu

% Cl NN id plate is an aluminum plate 1 1/4 in, th!ck, wh($$hA ch'sspports

' if the entire weight of

'The bottom gr the core'and provides accurate spacing between the fuel-moderai5614 ants (See' Fig. 4.9).

Six adapters are bolted to pads welded to a ring %5cid$ hip turn, > WeiMto the core barrel to support the bottom grid plate. qscg f/g9mmm}, Dhr Holes 1.25 inch in diameter in the bot idp,teareaj ith fuel element holes in the top grid plate. They are p receigthe adapp%ned countersug$)gmend gr of the fuel moderator elements and the adaptor-end of the pneurpatio transfer; tube, f/

Eight additional 1.505 in, dfam holes ibiii ith upper grid plate holes to provide passage of'fki-follower nt'rn!" rods. thy 63e bleiin the bottom grid plate not occupied by control fMfdliske[s are[$pyddhith removable fuel elemerit ada plate. ThEsE iers are alumiiinMIYubing 1.5 in. OD x 1.25 in. ID by 18 in. long. Slotted channels are m@achilid'In.the sihdi6tMtubing to provide for coolant flow.

is a fitting that$3ocodu$'odated billiole in the safety plate. The upper end of the cylinder is flush witigtiTupper'idfdof the bbttom grid plate when the adaptor is in place. This end of the adaft'o r is countersthdmila'r to that in the bo; tom grid plate for accepting the fuel elensit lower end fittinjif yith the adaptor in place, a position formerly occupEd by a codifol fod with a fuel follower will now accept a standard fuel element. -

!?:h N 4.2.623 Safety Plat 8 W%&Y Ksafety) lit 61s*provided to preclude the possibility of control rods falling out of the core (See Fig?4:10)T lt is a 1 inch thick machined aluminum plate that is suspended from the lower grid plate by 18.25 inch long stainless steel rods.

f

4 19 Rev. A 7/5/96 REMOVABLE CENTER GRID

• LATE IMERT

! GRID SUPPORT ROD q

LOCATIONS (6 PLACE S) . .- *).

23 . 3 j.

SAR HEX CRID NUMBER SYSTEM (MEX RINO3 A. D, C,0, E. F,0) o C42 C48 841 A41 844 O x

l T q

..: G 0

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606116 ALUMINUM 175.wlTHICK

' 171 GRID LOCATIONS TYPICAL CONTROL ROD LOCATIONS O BOTTOM GRID PLATE FIG. 4.9

Rev. A 7/5/96 4-20 l

l l

l 1

l 1

CONTROL ROD O~t FUEL ADAPTER LOCATIONS (8) i f

\

/ SUPPOPT RC'; LOCATIONS (8)

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l 1/2-13 TAPPED HOLE 2

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o/ - o IN. DIAMETER (f-I

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l 6061-T6 ALUMINUM 1-IN. THICK i

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SAFETY PLATE FIG. 4.10

4-21 Rev. A 7/5/96 O)

\

- 4.3 Reactor Tank The '..iNRC reactor core assembly is located near the bottom of a cylindrical ahiminum tank surrounded by a reinforced concrete structure (see Fig. 4.11). The reactor core ahd beam tube j assembly installation is shown in Fig. 4.12. The reactor tad i; a weld 6d alumih, sin vessel with 1/4 in, thick walls, a diameter of approximately 7 ft, an'. c depa of a%5i6ftely 24-1/2 ft.

The tank is all welded for water tightness. The integrity o((hj wel($ia,.yerified by radiographic testing, dye penetrant checking, and leak tesp. ; The o4t: .. . e tank is A coated for corrosion protection. 5 #

55Mk 8 maggf Four beam tubes clamp onto the reactor tank at 90' intentgpcing tangentia greflector assembly and core (See Figs. 4.7 and 4-12). The tank *silliciation of the beam tuMs consists of a 12-% in. diameter pipe welded to the tank wall. T%%gflanges ag. welded to the '

in tank end for water tightness. The beam tubes clarq[8nto eh ,

' the bulk shielding concrete that surrcunds the reactor ank. T$gi all a~nd-extend wbes aye positionedGro'igh at a 20' angle from horizontal and a fourth beam tubbi sitioned angle from horizontal as shown in Fig. 4.12. $.

4.4 Biological Shield hk Y(f

~ %y$E*r lV The reactor tank a

is cinfor surrounded by*g'in6n dard concrete bulk shield structure. Below ground level, the neretelisrapproxi .. ely 11 ft, thick. Above ground level, the concrete varies in thickness fro rox' ly 10 ft. to 3-1/4 ft, with the smaller dimension at th tank top.dh/dnk e

is su" ' y a concrete pad approximately 9-1/2 ft.

p m.9 thick. Aggig g g#g The mas hg v#%

sive %[Niictete bulk shIdkistructure provides radiation shielding for perso in and aroundWNNR . AlsEDENssiveness of the concrete bulk shield structure provides exceld damage to the reac@toncorodhg for tM[idctor core against natural

.- # D5ShD .

)

4. Nuclear Design [

~

4dif~ TRIGA Fuels /b' ihM A Wu , M s a brief description of TRIGA fuels followed by evaluations of neutron section ,

lconsiderapro @tions, materials properties, irradiation performance, fission pd fislikfimd limiting design basis (Ref. 4.1).

4.5.1._1 Description of TRIGA Fuels f7- The uranium-zirconium hydride fuel used in TRIGA reactors is fabricated by r.jdriding an

( ) alloy that is a solid solution of uranium '.n zirconium. The zirconium is selectively hydridcd v

Rev. A 7/5/96 4-22 0

_ (LCC4 vatti t(Vf ir N I .

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23 Rev. A 7/5/%-

E RwavatE -

_ BEAM IUBE e

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.t (h?. 3; PUCES) i -

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TYPICAL IN-TANK REACTOR CORE AND BEAM TUBE ASSEMBLY FIG. 4.12

Rev. A 7/5/96 4-24 and the uranium remains as small metallic inclusions in the zirconium hydride matrix. The size of the uranium particles increases from 1 to 5 pm with increasing uranium content from 8.5 to 45 wt%. Some important parameters for TRIGA fuels are provided in Table 4-3.

$Y TABLE 4 3. A PARAMETERS FOR TRIGA F.UELS. h&[7 Weight Percent Uranium a x 105 Core Uranium Type of . Uranium-235

- (g/ element) Enrichment (Ak/k'C) lifetime Volume Fuel

• Uranium Erbium (%) (Mwd) Percent

_% .e w. ov Original 8.5 0.0 39 20 % 7' 9.5 100.3 % yF2.6 FLIP 8.5 1.6 137 7Uhif;I [II, 0.5 3500 2.6 ,

~

== .. '

LEU' 20 0.5 99 45, g[10.5;bf T200 6.8 LEIT 30 0.9 162 glM R (8[$ *3db0 11.2 LEU 45 1.8 282 A s N N UE h SM 4000 19.5

$?

• FLIP = Fuel Life improvement 4 N%$%dv?

Program conductefat GA: // fr LEU = low-enriched uranium. .Sii) /[

10 N N

l/3y' g m1 At a Ri HLM The use of cibium burnabl6yois$n in conju'ndti6fijith the higher U-235 loadings permits longer coi$11fitimes thaiN6EINe obtainaN5%Ilh the original TRIGA fuel. It also permits maintaidthgylnde prompYdbkaliyhtemperature coefficient of reactivity, a, that is changed little from thhidf tMoriginafftiE6hio' ugh at least the 30- wt% LEU fuel. As shown in Table 4-3, the volumyI5Nc%i3v,%) ofihnitim' increases with the increasing uranium loading but remains a smallWalu@incicasing from 2.6 v% in the original fuel to 11.2 v% for the 30-w:%

LEU fueJ/and to 19.'5N% ioEthe 45 wt% fuel, y plij7 ._

4.5 h2 Performance ofMr$itim Loaded Fuels n

e Thyrimary intent of fGA Technologies Reactor Physics Qualification Program was to show that;neutronically tlie'20-20 and 30-20 TRIGA LEU fuels behave essentially the same as the currentif~apptcived TRIG A Fuel Life improvement Program (FLIP) fuel. The following was c6ddldddlf(REf.'4.1):

- The power peaking factors in the LEU and FLIP fuels are very comparable. Any variations are due mainly to differences in the contained U-235 (not the total uranium loading).

4 25 Rev. A 7/5/%

tm I \'

Qj

- The prompt negative temperature coefficient the reactivity worth, and the core lifetime of the TRIGA LEU and FLIP fuels are comparable primarily because of the adjustment of the erbium poison concentration. Also, the reactor kinetics parameters most important to power / burst behavior, prompt neutron lifetime, and effective' delayed neutron fraction are similar. .

h&aMf[

The data in Table 4-3 represent a single average estimatep[It, RefMA2 and 4.3 present p! cts of the effect of fuel temperature on the prompt fd co9f e'nt for 2,0%%slel g These [

sources show that, at low burnup, the feedback coeffic$ int is a strogfun?tElitA16mperaafrt, ranging from about -7 x 10 5 at 200'C to about -15 x M.at 7, 4.2, an integrated average over the 23 to 750*C rangeliMl0. 10-5/'C. Simil~djbN@(6r a the ranges 23 - 800* and 23 - 1000*C, the averages are -lF41E . _and -11.8 x 10 5/*C, y

respectively, from the curve in Ref. 4.3. All of these villbThidliinijas ate thaJl he prompt -.)

negative feedback characteristics are retained with theTrbium g am the 20 wt. % fuel 4.5.1.3 Materials Properties _

V The materials properties of TRIGA fuel hig pa nts were reviewed relative to those 8.5 wt% TRIGA fuels (Ref. Twith clusion E(diIo, ws:.

(V Measurements were made of the _

zirconium hydride fuels. The from l co ~ ctivity 9jf 8.5 , 30 , and 45-wt% uranium-meas ~ ents, in conjunction with density and specific heat' data, were us termin conductivity of these materials. The be ind .f uranium content within this range.

thermal Mn con'@@i'cti ' ity was:

wss The specifidtiest anium dride was calculated as a function of uranium content using , ific h _ nium and zirconium hydride and a linear interpolatiop 51 3a stra rward and acceptable apprmch, and the resulting values forhe,at capac n adequately factored into the analyses of kinetic behavior of

g tlie bifiloaded LEU oefficient of thernih expansion was measured for 45-wt% uranium fuel and compared hat for 8- to 11M% fuel. For a maximum power density TRIGA fuel element, the

_ m radial exg/Ihion would be about 0.6% for 45 wt% fuel as compared with 0.5% for S?f,uelqwhich is not a significai t change.

nng Y

Th oring of hydrcgen pressure d aring 1.ydriding in the fabrication of high uranium content fuels showed that the equ3ibrium hydrogen dissociation pressure of the fuel depends only on the hydrogen / zirconium (H/Zr) atio and the fuel temperature. It is independent of the

,m. uranium content.

-( )-

u l

L

l Rev. A 7/5/96 4-26 Thermal cycling tests were performed on 45-wt% uranium fuel over the temperature range of 500 to 725*C, which includes the orthorhombic-to-tetragona. phase transformation at 653*C.

Specimens were cycled 100 times out of pile and then 32 times in a neutron .,ux of 4 x 10'2 n/cm2 s. There were no significant changes in dimensions in the out-of-pile tests, and a small decrease in weight was measured. The in-pik cycling test showed a small decrease in both length and diameter, which may be related to a loss of hydrogen. The'dimensfolial stability of the high uranium content fuel is understandable considering the fine dji$rsi$ Eof the uranium in the zirconium hydride matrix. The dispersion of uranighilp partic{sjleifthyn 5 m in diameter e tidently precludes pK anisotropic grow transformation because of accommodation by the matrigwhich nfes up volume in the case of 45-wt% uranium fuel. kh ,ff y$NhF kiNM Uranium and zirconium form eutectics with iron, nickels?andinromium, the principal

~-

constituents of the four alloys (304 or 304 L stainless stniQl@8,00;antDiastelloy-x) that are licensed for use for fuel rod cladding according tithe Tech;nicil3pecifications. The uranhim eutectics have lower melting temperatures than those ofi.lic6nis'm, w1iich is tied up as a hydride in any case. The melting points ofJhd[eEtidtkwith ura'niiisiIre: iron,725'C; nickel,740*C; and chromium,859'C. AgtiEYrif I6 tent of[ fuel is increased, the potential for the formation of low-meltinseutecticsj,tiuin l s ei$aisiidyl ocalized fuel melting has c been observed in 45-wt% uranium fuellin contac3 Mi th Incogd600 thermocouple sheating at temperatures above 1050*C, The esteyt of potbntial euteft'ic melting due to fuel / cladding interaction should be less in the 20]End 30-wts uraniuffuels than in 45-wt% uranium fuel, 8.ht% uranNM fuel,Mfall cases, the extent of eutectic but more than in the original,Mie:relatively3@inkil melting would be limited by v% or lessT8r*the fuels $dO3New). Th ijErature at which cutectic fuel melting has been obkM6050*C)M1Mabove the lowest temperature at which cladding failure by hydrogen owr %Nire is prEdi&$ der conditions in which the cladding is at approximately the fuel temperdtdPeNITherefor6fEhishillfuel/ cladding melting does not constitute a more severe limit _forjiEMid$legrity tEiid50es hydrogen overpressure. h does, however, have

^

the potential't$ prodMfujet fuelting at temperatures about 80*C lower than the uranium meltigthoint. This mkinnisidould lead to somewhat higher releases of fission products from^the fuel rod in theitliiipefature range 1050 to 1130*C under some accident conditions (suOfas loss of coolantjyr during film boiling; however, these temperatures are above the sa'fe@ limit of 930 C4which applies if the fuel rods are not immersed in water.

Mi [

DiiriQtainedjirfadiation, hydrogen tends to migrate from the hot radial center of the fuel tolaib61cf ahnulus near the pellet periphery. Hydrogen / zirconium (H/Zr) ratios can vary by i10To'15% of their initial values. The increased H/Zr ratio near the outer radius of the (pel, coupled with high peak fuel temperatures that occur at the outer radius during a pulse, can cause excessive hydrogen pressures in the fuel matrix, which can weaken and deform the fuel matrix and cause excessive swelling and fuel element deformation. Experience suggests that pulse sizes or maximum fuel temperatures should be limited in higher burnup cores to account for the effects of hydrogen redistribution. This effect, however, is independent of uranium

4-27 Rev. A 7/5/%

\v content in the TRIGA fuel, and the evidence suggests that an equilibrium hydrogen distribution is established within a moderate time scale.

A 45 wt% uranium LEU fuel rod that was instrt.mented for measuring temperat.ufiand pressure was subjected to a series of 30 power pulses in a TRIGA reac(or to ridximum fss than 2 psi) temperatures in the inrange pressure pulses were measured ofa result the rod as 1050 to 1100'C. Only very modes of th,eVisising, data on negligible hydrogen release during the pulsing ofA5-wt% u,raj_ 8 temperatures up to ll50*C. All surveillance examina __ on rod diform ,g satisfactory. Tests have shown that the pulse respo ' uraniufzirconiu A fuel is independent of the uranium content of the fuel . .

inated by the of the zirconium hydride, along with the prompt temperature f reactivity, Q

-m

. As mentioned earlier pulse sizes or maximum fuel ted>eratu ~ limited in higher

~

burnup cores to account for the effects of hydrogendtributi ential problem is adequately addressed by imposing limits on :T,g$ munnggating te 'ures in standard TRIGA fuels. The effects of hydrogen m 5n wannes to viewed fission product releases if these restrictions are applied ; -  :

j?

4.5.2 Design 11ases  ;

(

\

The reactor design bases are es shed b maxinspm operational capability for the fuel elements and configuration 4es d in tlGi% YThe TRIGA reactor system has three major arcaiiMich are usii$Mnc the @ design bases:

NUEMS. M783h b m perature coefficient, g$I P We y The uJtihIate safety limYOf.%on fuel temperature, while the negative temperature '4 coefficient contributes tdyiebnherent safety of the TRIGA reactor. A limit on reactor power is to ensure operatin3% low the fuel temperature safety limit. A summary-of the con-ns of the analyseifthat supports these limits is presented below.

.Y The M rature is a limit in both steady-state and pulse mode operation. This limit stems from the out-gassing of hydrogen from U-ZrH fuel and the subsequent stress produced in the fuel element cladding material. The strength of the cladding as a function of temperature sets n the upper limit on the fuel temperature. Fuel temperature limits of 1100*C (with clad (v) <500'C) and 930*C (with clad > 500*C) for U-ZrH with a lii4r ratio less ihsn 1.70 have

l Rev. A 7/5/96 4-28 been set to preclude the loss of clad integrity (see 4.5.4.1.3). These temperature limits are less than the basic limits for TRIGA fuel of Il50'C and 950*C as stated in Reference 4.1.

Promot Negative Temperature Coefficient M

The basic parameter which provides the TRIGA system with a large safety factiir in steady-state operation and under transient conditions is the prompt negative,MdIIre coefficient which is rather constant with temperature, as described inSMLion 4f512$This coefficient is a function of the fuel composition and core geometry. Tidalue for[ Mile,temperatu

• coefficient in TRIGA fuels is 0.01 %/*C (1 x 104 Ak/k/f C). 47 WKWMgeD .

Il 57 Reactor Power hs;;wf gw Q @M ?

yw. W,. m .

A d V lli i i l ulsted that ~ ~

Fuel and dad temperature define the safety limit. ensures that the fuel and c ~

analysis indicates that operation at up to 1675 kW,with.a 75 elerMuti30NPand 2300 kW with an 101 element core (95'F inlet water temperatirijMiMnatural cEycdtbe flow will not allow film boiling, and therefore, high fuel and elsi (YperiitidNEwhich ccYld cause loss of clad integrity could not occur. [ $OMA$@[ NGNW' n k?~ f the TRIGA fuel, including the ts o funda The increase to 2 MW will not affoett[dpNature$ptental fety limits, aspand fission produ plMilent power and radicactive releases reactivity Section 4.5.4). The feedback effect of th coefficients,$lifower r

are addressed,in Section 4.5}5/TJ The the#itiAlIli)$

in$ Ease ondli c performance is discussed in Section M M MW 4.6- +A  %&g #

QD$9 eria%- Ref'fsincNSic$d>res 4.5.3 DesIdnilhg Dlh TnSM) uw Design decision.x.nnsgers made based'oithe results of a scoping study of reactor cores containing mixed 8.55n"d 20 wiMifuil[ sad all 20 wt.% fuel loadings (Ref. 4.4). The design decisions are reJ1(died in the refelfndlAlings described below.

g my '

N TWCreference loadings'y Tarc defined for use in the safety calculations. One case, Mix], is a tri.iEd-fuel loading witlI20/20 fuel in Hex Rings F and G, and a nearly equal amount of 8j5/20, fuel in HexcRings C through E (Fig. 4-13). The other case, 20E, is an all-20/20 fuel loadis[(except:854t.% fuel followed control rods) having as much fuel as possible in Hex RQiDihr6uili G and some fuel in Hex Ring C (Fig. 4-14). Both loadings contain a central irra'diittidiffacility, which occupies Hex Ring A and B.

It is assumed that the initial loading for 2 MW operation will be similar to MixJ and that, as the existing inventory of 8.5/20 fuel is burned up, there will be a smooth transition to a loading similar to 20E. In this case, the safety and performance characteristics are bracketed by those of the Mix] and 20E loadings.

l l

1 4 29 Rev. A 7/5/96

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5 t,v/ 3.

. . . .. ::. 3 3 B[- /. ')/ef i

TRANSIENT Roo , 20*20 FUEL ELEMENT

' ~~

- Low woRrn s.s coNTRolRoo 4% """

'?E on FUEL roLLowEo CONTROL Roo $I' GRAPHITE oUMMY ELEMENT FUEL FolloWEo CONTROL Roo PNEUMATIC TUBE NEUTRON SOURCE 7\

\

v MixJ REFERENCE CORE FIG 4.13

Rev. A 7/5/96 4 30 O

\

@ fi~ -;:

..... g9..) .. ..

4:e.:

- e..

@f: iiii; - i@ '

Pi ._.

/ (jgs'i) .

. :g .:: 4.S,E. : ~ .

• • , 3  :.

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ii di.:{i Q[  :!:E..jj:ef,:

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43.: .,

wi ,6 j

M M ELEMENT

.. ') TnANstENr noo 1 p

PNEUMATIC TUBE { NEUTRON SOURCE L/ v 20E REFERENCE CORE FIG. 4.14

4 31 Rev. A 7/5/96 m

r:

( -

The fact that all the fuel used in the neutronics models was unburned makes the safety-related predictions conservative. This conservatism leads to overestimates of the power peakitig compared to the expected situation where, within a fuel zone, the most burned fuel will be loaded in the innermost rows. The leakage-related performance parameters also.t6nd to be underestimated by calculating all-fresh-fuel loadings, furthermore, therflux in:tliscentral experiment fecility is overestimated. j Osh7 Another factor also makes the predicted peaking k?$?%. factors $'$hi n (con predictions made here are approximately 10% higher th$ those madEwitNiilid6%Bdent m[d61s (Ref. 4.5). The reason for this has not been identifie is nopdde to diffull6Edi8&ridFfuel composition) but if the predictions here are in error, t Norfilin the conserva'NUfiection, wmpm%

For the hiix J core loading there are no restrictions on .~

. operation in regard to the configuration of the(I~e., central water @gendp

~ ~

re irlsdl5Neiffacility 4.5.5.4. IiiE6nbrime'nt, or other filledraluminum materials, all of which graphite plug asworthfM require reactivity described in Section;Naents). F T DOE core the central corp ,rradiation facility filled witfdIe/ fuel temperature measurement du performed in the anal [ maxim 0iiU%% ore position (C07) prior to routine operations to assure that fuel tenfperature1Tmits ar6not exceeded.

n A!/ // D~

I

) There are, however, requirementsINd restrichns for loking hiix J and 20E cores regarding fuel types A and theirAlocation

?f in N[ core as . IfowsN:f['

Mix J Cornbing qiWA gh[k MHlf;$

h Ms%[

em. , w wmA

• No fuelwillbgloadelinto:H6xsRing A or B.

wz;%A wem M923% N3f?!f A FliGKid6ated'inan 8.5 wtiC20

• wt % fuel environment must contain 8.5 wt %

y y fucV NNAWW&p -

, nmy f/The initial load . 'thst program to determine the excess reactivity of a Kiix J core or

% variations to hii4J shall be conducted with the central irradiation facility aluminum

1. 3 graphite plug,iblace, m jy fur?M /3 m MMAuninim6th'of 94 fuel elements including FFCRs is required for operation at power hMIEvels

%y ab5ve 0.5 h1W.

20E Core Loading

• No fuel will be loaded into Hex Ring A or B.

l

• FFCRs may contain either 20/20 or 8.5/20 fuel.

i Rev. A 7/5/96 4-32

. A minimum of 94 fuel elements including FFCRs is required for operation at power levels above 0.5 MW.

. The initial load and test program to determine the excess reactivity of a 20E loading or variations to 20E shall be conducted with the central irradiation facility al(minum graphite plug in place. /"

Variations into cerner positions toonly20E and having 20/20 graphite dummy fuel eJejiints in Hex in thM s. The Ring Gfr$

recommendation regarding the performance of fubl temperature measur@ots also= "

- applies to these variations to the as-analyzed 2

$Iion. y

<r-m For the safety and perfortnance analyses to be applicab}&$gpot necessary that the actual' Reasonable variations on ._,

loadings precisely match MixJ or 20E or a direct path ~

. these loadings will have comparable safety and perfon, nance Mcs.

c G~uidelines

> on tail in R' >

wha *. constitute reasonable variations are discuss h

d. 5.4 Reactor Core Parameters jM w8 a

4.5.4.1 Reactor Fuel Temperature

, tty, The basic safety limit for the TRIGMi'eactor stem is fuel temperature; this applies for both the steady-state and pulsed [th$de of " tion.gjf

& 4Y s Limiting temperatures fofMtwqmodes o_gn:rak!on are of interest, depending on the type of TRIGA'pfyij%The TRigfg.{hich is considered low hydride, that with a H/Zr ratio of less than 1.3Qisylower temp'erature!!imit than fuel with a higher H/Zr ratio. The MNRC reactor utilizesi6tt5Eh41/Zr rah.ktNeen 1.6 and 1.7. (i.e., greater than 1.5). This allows operation at,aMgikrJuel temperathipiimit. Fig. 4.15 indicates that the higher hydride compositWare sin'skylinhnd are not subject to the large volume changes associated with the phasiftransformatio'diEgidoximately 530*C in the lower hydrides. It has been noted in RefM6 that the higher hydddes lack any significant thermal diffusion of hydr 6 gen. These twS% cts preclude concoinitant volume changes. The important properties of. delta phase U-ZrHfare given in TabisA4.

[Nih M hbfh?k5 ene e q3yis O

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

4 4-33 Rev. A- 7/5/%

950 850

<-HYDRIDE (tetragonal) 750 Zr (p) ZII0)

• B-HYDRIDE .

~

Zr (a) /

650 -

y_ Zr (p) ,.-

a .

U 550 -

.E i 5 '  !

%. N'Z r 8 J (a) g 450 Zr (a) + B-HYDRIDE 5 f (cubic) 350 - i

% i  %

50 - +8 j

' ' ' ' ' ' ' I -

0 O 0.2 0.4 0.6 0.8 1.0 1.2 1.4 1.6 1.8 2.0

HYDROGEN, CONTENT (H/Zr) l km PHASE DIAGRAM OF THE ZIRCONIUM-HYDROGEN SYSTEM FIG. 4.15 1

f

. - - - , + ,

Rev. A 7/5/96 4-34 TABLE 4-4 PlfYSICAL PROPERTIES OF DELTA PHASE U-ZrH Thermal conductivity (93'C - 650*C) 13 Bru/hr ft *F 2

,f, Elastic medulus: 20'C 9.1 x 10' dils /k

, euere 650'C M,6.0 x 10' psinh s t alu .

Ultimate tensile strength (to 650'C) 43; 24.000 pdkA;&m #A ,

- 4 n.v : a &

Compressive strength (20'C) O! 60.000 psi V M W pt 7 Compressive yield (20*C) hN E3k000 psi m

NN

, lleat of formation (6117298'C)  %> Q37;72 kcal/g-mole Y Yhk NMW hydride with water Among the chemical properties of U-ZrH f.

and Z_rdm(e:mthem reaction ratio "

is of particular interest. Since the hydriding WEction isnexothermicgwater will react more readily with zirconium than with zirconiur6Tiydride$ifitc69 Zirconium is frequently used in contact with water in reactors, and the dEoniumd/ater'rdsSiddis'not a safety hazard.

Experiments carried out at GA Technd$gies sbdi$ that thEMiffonium hydride systems have a relatively low chemical reactivitygrespeg'jb water aN1 air. These tests (Ref. 4.7), have involved the quenching with watebof both powders and' solid specimens of U-ZrH after a Uf solid phldEfter heating to as high as 200'C Tests heating to aghigh have also,been made to deb as 850*C,p$(ine the extenIENNf6h fission products are surfaces $f NM6el eler5@iN$om tempDMiiYe, Results prove that, because of the high resistance'iklAsdShg, a larYetMic'tidn of the fission products are retained in even completely unclad U-ZrlFfEl%

_;sp;; a NNY@b

%y37 mran .

At room,teji$bisiir6[Ntijshidride is~iike a ceramic and shows little ductility. Howe elevated t6mperatures 6fst26sf for pulsing, the material is found to be more ductile. The effect" cit very large therNiitNss on hydride fuel bodies has been observed in hot cell obhations to cause relitivIly widely spaced cracks which tend to be either radial or normal tait 66 central axis and d6 not interfere with radial heat flow. Since the segments tend to be dih$ gonal, their relatiYe positions appear to be quite stable, r .- p s-van w w w u ThWlim_itide,ffeEt of fuel temperature is the hydrogen gas pressure causing cladding stress.

FigM16% rates equilibrium hydrogen pressure in a Zr/H mixture as a function of temperature for three different H/Zr ratios.

The main concern regarding hydrogen pressure is to ensure that the cladding ultimate strength is not exceeded by the stress caused by the pressure. The mechanisms in obtaining

4-35 Rev. A 7/5/96

/~N b/ temperatures and pressures of concern are different in the pulsing and steady-state mode of operation, and each mechanism will be discussed separately.

The MNRC fuel consists of U-Zril with a H-gatio between 1.6 and 1.7 and with Se uranium being 8.5 or 20 wt % enriched in U to approximately 205U-235? The cladding is 0.020 in thick stainless steel and has an inside diameter of 1,43 in!nfile feat of the discussion on fuel temperatures will be concerned with fugl[liaving @ZIllios, greater than if (i.e., single phase and not subject to the large volume cjianges associp$dyth phase g transformation at approximately 530*C in the lower hydrides). Fut{her, ifjil,1]specificallg#

address fuel with an II/Zr ratio of 1.7 since this is the4 hi ghest ratid fuel to be'esedKthV MNRC and will produce the highest clad pressure andistrEif fdr'a given temperlitiuiriF Fig. 4.17 shows the characteristic of 304 stainless steely 006hd to yield and ultimate strengths as a function of temperature. kwy

{"NDOA 4#iWIh. is given by:

ressure)

The stress applied to the cladding d3MA from the internal < hydrogen gas p%

S = P r/t , [ N hjh)hg VW;sp gf (1) where

/w> Ygw r

im i

S = stress ir)psy

!/ W gr 4,

O P = internalpisssureja, psi, fT r = radius of the stalidess steeliylinder, A t = Willitiickness dfRhtsih15s steel clad, j?k f$5Q %5 W Using tlielparanisters given aboFe:u;w,L

+%w@p% TM $~

,x5 ~

TBS:.=%36.7 e a.

i3h 5 N h sa PM@h For safety lconsiderationsPit is'necessary to relate the strength of the cladding material at its operatid(temperature t'olthEsk$s applied to the cladding due to the internal gas pressure assfuited with the fuelIhipfr'ature. Fig. 4.18 gives the ultimate cladding strength and the

, stress applied to the clallding as a result of hydrogen dissociation for fuel havirg H/Zr ratios of

,fand 1.70, both asifunction of temperature. This curve shows that the cladding will not fuel with Zr/H i7 if both the clad and fuel temperatures are equal and below about

' is Si5erv'ative since the cladding temperature will be below the fuel t ,,

is establis'ies the safety limit on fuel temperature for steady-state operations.

Th Niliteady-state peak fuel temperature at 2 MW will be below the limiting maximum measured fuel temperature of 750*C The remainder of this section dea!s with the safety limit for transient operation.

/

k t

Rev. A 7/5/96 4-36 3

10 I/

ZrH 3,7 ZrH l.6 E

ZrH l.5 e -

/

/

f

= -

5

// e 7 j/

l

/

3

$ 10 -

s~

~

///

/// DATA FROM GA-8129 AND NAA-SR-9374 0 i i i iii 10 600 700 800 900 1000 1100 1200 1300 TEMPERATURE (*C)

EQUILIBRIUM HYDROGEN PRESSURES OVER ZrH, VERSUS TEMPERATURE e

FIG.4.16

t' l

l o

4-37 Rev. A- 7/5/96 _.

10 5 p

.i-

~

ULTIMATE TENSILE

, 0.2% YlELD i

e G

a.-

~

m 10 4 -

O i

IO

~

Y m

103 -

400! 1500- -600- 700 800 900 1000 1100

-TEMPERATURE ('C).

k ---

STRENGTH OF TYPE 304 STAINLESS STEEL AS A FUNCTION OF TEMPERATURE FIG. 4.17 -

Rev. A 7/5/% 4-38 9

30 5 ULT l HATE STRENGTH "I.7 304 ss p

~

l /

'/ /

_ / / ZrHl.6

/

/

, /

0 10 - .

/

/i

=

/ '/

E -

/ /

O E / /f

! stress induced by

'o'

[ / h'ydroken pressure

/ / ,

\

l/

l/ i

/

g I I ' 'I i '

10 2 j 1200 500 600 700 800 900 1000 1100 TEMPERATURE ('C)

O STRENGTH AND APPLIED STRESS AS A FUNCTION OF TEMPERATURE FIG. 4.18

4-39 Rev. A 7/5/96

,m (e )

In transient operation, it s3 necessary to account for the difference in fuel and cladding temperatures to estaolish a safety limit based on fuel temperature. Additionally, the diffusion-of hydrogen reduces peak pressures from those predicted at equilibrium at the peak temperatures. The net result of these two points is that a higher safety limit exyfor transient operation. An analys.is of the two points is given in the following twojubsections.

4.5.4.1.1 Fuel anu Clad Temperature jkg } NlNAW

.m JMb wwmh .,

For the steady-state safety limit, it was assumed that ladding,atid'fuelMMpare ,

temperature is the same.

maximum The following fuel temperature after a discussion pulse. This allows shows

$ that the el _ %ijtieOafety limit on ture WEA

. The radial temperature distribution in the fuel elemenhmm

/)Y$$ .

_ ing a pulse is very similar to the power distribution shown in Fig. 4 A9aThis initi , mal gradient at the fuel surface results in some heat transfer duri e put at the true peak temperature does not quite reach the adia ( . ge temperature gradient is also impressed upon the clad which esult i ,

om the clad into the water. If the heat flux is sufficiently , Olm ling ~ _

r and form an insulating jacket of steam around the fuel el per ng the claii temperature to approach the fuel O)

( temperature.

Thermal tra lent calculat_ N T computer code. RAT is a 2D transient _

port f repedmadg@[3or to , fluid flow and temperature d material , Caleb that if film boiling occurs after a pulse, it may take place eithe Igle of t flux from the clad, before the bulk temperature of the coolant hasig h preci

^

may take place at a later time when the bulk

y appmac ed the saturation temperature, resulting in a reduced temperature p' _ __WData obtained by Johnson 3etal Ref 4 8 for transient he threshold fo,r film b611i@lshdwiid burnout fluxes of 0 9 to 2 0 MB ribbons'in from[to 90 millisecon 100*F waterT[OrI the other hand, sufficient bulk heating betwen tuel elements can take place in several tenths of a ycond to lower the departure from r

(g$didate boiling following analysis! (DNBf" that the second mode point is thetomostapproximately likely; i.e., when film boiling0.4 MBtu/ft occurs, -hr. I s$hM. git pigunder' essentially steady-state conditions at local water tempera ionny

%3g9 A value for the temperature that may be reached by the clad if film boiling occurs was

obtained in the following manner. A transient thermal calculation was performed using the radial and axial power distributions in Figs. 4.19 and 4.20. respectively. The thermal bv

, l

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Rev. A 7/5/96 4-40 .

u a

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Rev. A 7/5/96 4-42 resistance at the fuel-clad interface was assumed to be zero. A boiling heat transfer model, as shown in Fig. 4.21, was used in order to obtain an upper limit for the clad temperature rise.

The model used the data of McAdams (Ref. 4.9), for the subcooled boiling and the work of Sparrow and Cess (Ref 4.10), for the film Soiling regime. A conservative estimate was obtained for the minimum heat flux in film boiling by using the correlations of Sp;cigler cLal.

(Ref. 4.11), Zuber (Ref 4.12), and Rohsenow and Choi (Ref. 4.13), tofmd,tinminimum temperature point at which film boiling could occur. This calculation #gWejarupper limit of 760*C clad temperature for a peak initial fuel temperature 4f4000*CIEliliown in Fig. 4.22.

Fuel temperature distributions for this case are shown ingif4.23 an:dgeikf]ux into the water from the clad is shown in Fig. 4 24. In this limiting case, DNB occurriGnly 13 ~

milliseconds after the pulse, conservatively ca:culated 5sluming a'Meady-stais correlation. Subsequently, experimental transition andjil}n boiithg data were IUUNlW ave been reported by Ellion (Ref 4.14), for water conditioidfliiiiiilhtte those for the TRIGA system. The Ellion data show the minimum heat flux, $$dfiiithe; limiting calculation described above, was conservative by a factor of 5. A9ippF65Fisi8borrection was made

' which resulted in a more realistic estimate of 470*C a's the maifaT6hicInd temperature expected if film boiling occurs. This result is ingjE5 Inept withIidiEnmdital evidence obtained for clad temperatures of 400*C to 500*CifoitTRIGA Marl ((fuel elements which have been op: rated under film boiling condiffons (lyfM15jhBastd on this analysis, the peak cladding temperature will be 470*C forgiransient!fuefTeiriigkMEii6 of 1000*C. Further analysis shows that this peak clad tempflature is% lid forghiglier peak fuel temperature.

Ad W pf The preceding analysis assessing [tbaximuTriIclad tenip5ratures associated with film bo assumed no thermal resistance at[ tile fuel-cisdiinterfaR Measurements of fuel temperatures as a function of steady-state piMcNevel prohENEeYde that after operating at high fuel

[-

temperat,u' ei,Tperman6isjt iM $ nien the fuel body and the clad. This gap exists at itlijchipEiatures b'616Etl@

maximum operating produce $iiexample, Fig.16 in temperature (for Ref. 4.15).'Thfg@hickne5Talie@with fuel temperature and clad temperature: cooling of the fuel or ovejl6tidfBfsthe clid@idsTo widen the gap and decrease the heat transfer rate.

Additional,th6rmEl? rbsis&due tE6xide and other films on the fuel and clad surfaces is expected $IIxperimerifiiYifiNeoretical studies of thermal contact resistance have been reporld Refs. 4.16-4.ll@j nk$ provide insight into the mechanisms involved. They do not, howeyer, permit quantithiivQrediction because the basic data required for input are presently n%Ily known. Instedd, several transient thermal computations were made using the RAT co4Qvarying the effe~ptive gap conductance, in order to determine the effective gap coefficient tWMhich departurMom nucleate boiling is incipient. These results were then compared with higifient;filyboiling conditions of the 1000*C peak fuel temperature case.

Q,5 y For convenience, the calculations were made using the same initial temperature distribution as was used for the preceding calculation. The calculations assumed a coolant flow velocity of I ft per second, which is within the range of flow velocities computed for natural convection under various steady-state conditions for these reactors. The calculations did not use a l

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4-47 Rev. A- 7/5/%

M i ij complete boiling curve heat transfer model, but instead, included a convection cooled region (no boiling) and a subcooled nucleate boiling region without employing an upper DNB limit.

The results were analyzed by inspection using the extended steady-state correlation of Bernath (Ref. 4.19), which has been reported by Spano (Ref. 4.20), to give agreement gvgh SPERT II burnout results within the experimental uncertainties in flow rate. A, ff'

, J2&%if The 375, andtransient 250 Btu /hr-ft thermal cylculations

• F. The resulting were performed wall temperndistribujifn(Qinspected to using eff determine between the local thecomputed axial wdl position surface heat fluxand time and the ISNB after the accordilighB heatJlu'x pulserwhich I gave the The axial distribution of the computed and critical heaiilujesf6r each of the thFee:casis at the time of closest approach is shown in Figs. 4.25 througd%27i91f the minimum approach to an estimate of the effective DNB it, corrected to TRIGA Mar $ F conditions and crod$166$1[io%u

. gap conductance of 450 Btu /hr-ft' *F is obtained for'iricipie2ildit 500 is thought to be representative of standard TRIGA' fuel. T${M)

&%yf!' W The surface heat flux at the midplane of thshekd in Fig I.28 with gap conductance as a parameter. It may be otNerved tidlEn5Ihti$f heat flux is approximately proportional to the heat t.ansfer coefficient of theMap, antl%1iIne lag after the pulse for hN which the peak occurs is also incredd!"by aboit'the same# factor. The closest approach to DNB in these calculations did not'Meissariljl8ccur at th'ese times and places, however, as initial DNB point occurred near the indicated on the curves of Figs.[Iji5 througilh.27.[h *F according to the

~

ore outlet for a local heat 4flhx'of about 40 kBhiW conservatidliernath corfel5tidS at a locIlWalifft"emperature approaching saturation.

%$$ Ni@&

From this aidIhnjaxirddhikingture for the clad during a pulse which gives a peak adiabatic fuel temperatore of 1000*0 isistimated to be 470'C. This is conservative since it was obtained b)fasii$$iQno tgerMalIesistance between the fuel and the clad. As was show above, a, vain of 5605tu/hi-fts? 'E ior the gap conduction is more realistic.

[ &$f As cari be seen from FigM17, the ultimate strength of the cladding at a temperature of 470*C is)9[000 psi. If the strbss produced by the hydrogen over pressure on the clad-is less than psi, the claddig will not be ruptured. Referring to Fig. 4.18, and considering i,7 fuel with[a' peak temperature of 1000'C, one finds the stress on the clad to be qignalysis in the next section which considers diffu.csi will show that the actual pfessure produced in a pulse is less than the equilibrium pressure for the peak temperature. This allows a safe limit on fuel temperature to be 1100*C, TRIGA fuel with a hydrogen to zirconium ratio of at'least 1.6 has been pulsed to temperatures approaching 1150'C without damage to the clad (Ref. 4.21).

p d-

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7 0 9 10  !! 12 13 Distanc e from Bottom of Fuel (In.)

SURFACE HEAT FLUX DISTRIBUTION FOR STANDARD NON-GAPPED FUEL ELEMENT AFTER PULSE h,,, = 500 FIG. 4.25

t, 4 49 Rev. A 7/5/96

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- L - SURFACE HEAT FLUX DISTRIBUTION FOR STANDARD NON-GAPPED FUEL ELEMENT AFTER PUL.SE, h,., = 375 FIG. 4.26

Rev. A 7/5/96 4 50 0

ITU 0 .! P 5 i i F ' E. L .' E a

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SURFACE HEAT FLUX DISTRIBUTION FOR STANDARD NON. GAPPED FUEL ELEMENT AFTER PULSE, h,,, = 250 FIG. 4.27

7/5/% 4 52 (o) wJ Rev. A 4.5.4.1.2 11ydrogen Pressure in TRIGA Fuel Elements To assess the effect of the finite diffusion rate and the rehydriding at the cooler surfaces, the following analysis is presented.

As hydrogen is released from the hot fuel regions, it migr to the ons and the equilibrium pressure that is obtained is characteristic of m r than the maximum. To evaluate this reduced pressure, diffusio ryisu the rate which hydrogen is evolved and reabsorbed at the fuel ace, Ordinary diffusion theory provides an expression for the time depe of gas from a cylinder:

e-e .

y t , g .

c; - ct n=! hj2][kp, (3)

,y p ""W&y,,,p r

p rN // x

= the aye%e, th id ial, and final gas

( Where e, cg, eg, L gconcen ion in thef linder, respectively,

& jf e' Toots of

!?NE&

'E

• geittion J0(*)

= 0, 4 j.. =4**Q M$pdh

'hAlfissign coef0clent for the gas in the D$[WPgak= Q; cylinder, T,

d "#"'@$Nd%e radius of the cylinder, f?.rf 4e yta 7/yg 0 t =7;qby4 ti me.

Al H S$hing the term on the'right hand side of Equation 3 equal to x, one can rewrite Equation 3 As Q @a jy

@mww x .ec7 = + (1 - ege;) x TgWgjy w ~s.., y

-c/cg eg/c; g) and the derivative in time is given by

,s d (c/c;)

'( ) =

(1 c/c;) dK

(_j' di dt' (5)

4 51 Rev. A 7/5/96 0

106 .

, . i l i i .

ETTECTIVE HEAT TRANSFER -

- COEFFICIENT IN GAP, BTU /HR FT2 ..y N 500 375 250 -

, ( 105 _

+ _

s q

E 5

d 4 -

y -

U O ~

E k _

" to FLOW VELOCITY = l FT/$EC ~

~

GAP THERMAL RESISTANCES ARE

~ REPRESENTATIVE OF CONDITIONS AT END OF PULSE (1.E. TIME = ZEROL

~

i I . t i e go) .

1,0 0.01 0.1 ELAPSED TIME FACH END OF PULSE (SEC)

SURFACE IIEAT FLUX AT MIDPOINT VERSUS TIME FOR STANDARD NON-GAPPED FUEL ELEMENT AFTER PULSE FIG. 4.28

4 53 Rev. A 7/5/%

/

(a- nis represents the fractional release rate of hydrogen from the cylinder, f(t). The derivative of the series in the right hand side of Equation 3 was approximated by U = - (7,339ed3" + 29.88c -2W) g A di dt N (6)

A

[Af

/ew pg n. ;;.y;am Where e = Dt/r0. ((

m

$Nk b A f;;

a:- '

j,wf % s%qf;'

g sTf hump The diffusion coefficient for hydrogen in zirconium hfdi&1nk<fhich the 1.56 and 1.86 is given by TYNN 44%A ,

ef7 ';% i (7)

. D = 0.25 e-17800/R(T + 273) ,

ffh Nf{l Sf where R = the gas constant M M  % %Y;+

@g/mlb.Cyg n a T = the zirconium bye @ w ture in C.edW sr V O Equation 3 describes the escape of skiftom a cy inder through diffusion until some final (d concentration is achieved. Actualiffin the obsed syste8 considered here, not only does the hydrogen diffuse into the fuel-c$1' gap, bulaiso diffuM back into the fuel in the regions of lower fuel tetuperature. 'IMgin'diffusesjhrcsih'fWclad at a rate dependent on the clad temperatur(Although tiinsprikto reduce.hhydrogen pressure, it is not considered in this analysiskWhen'the diffusioti rates ^are equal, an equilibrium condition will exist. To account lacing the concentration ratios by the ratio of the hydrogen pres @sure intbrium ~

(Se hydrogen gap topressure, the"for this, Ph /P,. n us.EdS n 5(Mas m

, g n"gg%s, c V{c'/cg) d

= (1 - P h(t)&,) dx gf,pF*%;%d f(t)l=,

is pc dt dt (8)

W QNT ff p n kr -

P (t) = the hydrogen pressure, a function of time and h

?e Wn. f MiPggthe equilibrium hydrogen pressure over the zirconium hydride which is a w%mpfu'nctioa ma of the fuel temperature and Zril ratios.

ne rate of change of the internal hydrogen pressure, in psi, inside the fuel element cladding is dP h 14.7 f(t) N h 22.4 T + 273

(%

q di

=

6.02 x 10 23 y 273 (9) m -

Rev.A 7/5/96 4-54 where N. = the number of molecules of 11 in the fuel, 2

T = the gas temperature ('C),

f(t) = the fractional loss rate from Equation 8, V = the free volume inside the fuel clad (liters).

H

&e 3 '65 and Zrli 7 For is 19.9aand fuel20.6 volume of 400 cmThe moles, respectively. , the 2moles of 11 available from fuel witlt.Zrli 'i a c free vc!ume is assumed toi:orisist

?

volyme, at the top of the element,1/8 in, high with a diambt clad. The effect of changing these two assumptions wiitested by.ciilcula'tions in Which theV gap volume was decreased by 90% and the temperaturfe'$[Eh equal to the maximum fuel temperature. Neither of thN pressures different from those based on the original asslirnhiidns'although the initial rate of pressure increase was greater. For these conditions 7tjlWjh

,y q Ph = A x 103 p(T oa +,273), I f(t) v I .c

' j7' A er m Y4 (10) gP"% f where A = 7.29 for Zril l.65 and 7.53 f6EZril; w if fD(Q [

m. 4

{ldti '

The fuel temperature used in Equi 7 to e ate the d usion coefficient is expressed as:

T(z)4

= t<0

%f hh 8f T(z)g&h =.fl/TNgl;::'- T ) e s [2.45(z-0.5)]Qt((T; t 2 0, (11) m O V . 1. ' IN whereT PR %rN theida'lifuel temperature (*C),

mp ,A ;m7, <x t ms-Ty '=pyihe clad temperature ('C).

v, n y y g-1 . ..

[' z

! ;i

=J;lNethe axial distance expressed as a fraction of the i 3 t j= the time after step increase in power.

7: q g C. L% iV lt wasiissumed tht the fuel temperature was invariant with radius. The hydrogen pressure obr'theiiitonism hydride surface when equilibrium prevails is strongly te mperature depen-dein'as sh~own in Fig. 4.18, and for Zrli, can be expressed by:

9 (12)

P, = 2.07 x 10 e 'I'974 x 10 + 273)

O

1 4 55 Rev. A 7/5/%

D ne coefficients have been derived from data developed by Johnson (Ref. 4.22). The rate at which hydrogen is released or reabsorbed takes the form:

[P,(z) Ph (I)] f(t,z) A+

s(t.z) = -

ar Pe(2) gf

/:hgf gai k (13) jjM QiW where f(t,z) = the derivative given in Equation 8 with espect to time;evakat the axial [

position z, 9 9 p (t) = .

Yi.,1N h the hydrogen pressure in the gap wggE t, Wy33 '

l',(z) = the equilibrium hydrogen press fibegemperature at position z.

The internal hydrogen pressure is then b 3~

Ph (t) = A x 10 g

/ xmW

/ry O, This equation was approxim .. t_ by:p

/

Np 5

fo. x) 3 kkh('i)gNh@0 10 I~

" h)

I 4! P i*'

k %w NQM{@f qv x f(tg,z 6f8k g7 (14) 4.a

_B.

w& 3gpdMA-where disinner summ . 'is"dver the fuel element's length increments and the outer sumniition is over time;!)V ie cases where th aximum fuel temperature is ll50'C for Zrli and 1100*C for D , the equilibdim hydrogen pressure in Zrli is 2000 psi, which fe'aYs to an internal

,. jsf[ Using Equation 14, it is found that the internal pressure for both 2r11 7 increases to a peak at about 0.3 sec, at which time the pressure is about one. f the equ;ilibrium value or about 400 psi (a stress of 14,700 psi). After this time,

. the pressure slowly decreases as the hydrogen continues to be redistributed along the length of the element from the hot regions to the cooler regions.

n

.l Calculations have also been made for step increases in power to peak Zrlil I"'I k temperatures greater than ll50'C. Over a 200'C range, the time to pressure the pedand 5 the

Rev. A 7/5/96 4 56 fraction of the equilibrium pressure value achieved were approximately the same as for the ll50*C case. Similar results were found for fuel with Zrli g below about 500'C, the internal pressure thatstress the yield would in the produce,7.

clad Thus (35,000 psi) is about 1000 psi and the corresponding equilibrium hydrogen pressure is 5000 psi. This corresponds to a maximum fuel temperature of about 1250'C in Zril i g and ll80*C in Zrlig 7. Similarly, an internal pressure of 1600 psi wouldproduce'a stress equal to the ultimate clad,streng (over 59,000 psi). This corresponds to an equilibrium hydrogen pressure of 5 x 1600 or 8000 psi and a fuel temperature of a,b.out 1300[C inlZril l and

, 1240'C in Zrli 1.7 s/

/, /My l 1 % .65 n a 8

Measurements of hydrogen pressure in TRIGA fuel el r nts durjng steady hkko rItl have not been made. Ilowever, measurements have bynisadi[during transienPopera'tibns and compared with the results of an analysis similar to that described;here. These measurements indicated that in a pulse in which the maximum temperafurKhithe' fuel was greater than 1000'C the maximurn pressure (Zril was only ab6ut 6Wof tiie Equilibrium value evaluated at the peak temperature. hab)lations of the pressurelhsuiiing'f using the methods described above gave calcula(ed p'ressitte values %ut'three times greater than the measured values. g"W h we f

h y us n w-An instantaneous increase in fuel tempernture will produccith,e;m/ ost severe pressure conditions. When a peak fuel tempsathre is re$Ehed by Nreasing the power over a f' mite period of time, the resulting pressbre dill be no greater than that for the step change in power analyzed above. As the temperdtufe rise tiines become long compared with the diffusion time of hydrogengthe pressure willjehome incicasiliglj, leis than for the case of a step change in the prc'singin the clad element results from the hot fuel power. Thenascm for dehydrillhig' faster than the cooler:

thliis th$t, fuel rehydrides (takes up the excess equilibriuni with tlE hydrog'e'n'civerlp'ressure in the can). The slower the rise to peak temperature, tlie loper'the presidie because of the additional time available for rehydriding.

p ,

y y;

w. ~ . ,,a 4.5.4.1.3fZril Fuel Temperature Limits

?

y 2 The foiegoing analysis gives a strong indication that the cladding will not be ruptured if fuel temperatures are never greater than in the range of 1200'C to 1250*C, providing that the cl' adding temperature is less than about 500'C. Ilowever, for fuel with a Zrli j7 a conserva-tivsshfety limit offl00*C has been chosen for this condition. As a result, at tfus safety limit tempeinture the pressure is about a factor of 4 lower than would be necessary for cladding failures 71;li factor of 4 is more than adequate to account for uncertainties in cladding strength andYnanufdcturing tolerances. As a safety limit, the peak adiabatic fuel temperature to be allowed during transient conditions is considered to be 1100'C for U-ZrH fuel with ratios up to 1.70.

Under any condition in which the cladding temperature increases above 500'C, the temperature safety limit must be decreased as the cladding material loses strength at elevated

4 57 Rev. A 7/5/%

b temperatures. To establish this limit, it is assumed that the fuel and the cladding are at the same temperature. There are no conceivable circumstances that could give rise to a situation in which the cladding temperatures was higher than the fuel temperature, in Fig. 4.18, the stress imposed on the clad by the equilibrium hydro ' re as a function of the fuel temperature is plotted. Also shown is the ultimate streng tainless steel at the same temperatures. The use of these data for establis Jhe sa . r conditions in which the cladding temperature is greater than 500'C i j fled as: ' " ' "

7 a, the method used to measure ultimate strength requ the irsposition of a longer time than would be imposed for accident cotu N

b the stress is not applied blaxially in the ultimate s ents as it is in the fuel

. Clad.

m The point at which the two curves in Fig. 4.1 ifitdiiFct r Zrli} j lithe safety limit, that is,930'C for conditions in which the claddi $Tn* is abog500'C. At that temperature, the equilibrium hydrogen pdsEure wo6fdCTiitress on the cladding equal to the ultimate strength of the clad. //

                                                                /

jY The same argument about the redi tionp 'the hyd n within the fuel presented earlier is valid for this case. In addition,[at' elevated'teinperatures the cladding becomes quite permeable to hydrogen 4Thus not onl d pressureMeome will ,hydt escape'"{$[hydrogda

                                                       $5i1lthIs'ystem entirely,r lSth% tite itself within wgh qqg               -

The use oYtkhiti$hte stren[diiditSi c) adding material in the establishment of the safety limit under these c$istiti6 ins isjustif$d,tisciuss'of the transient nature of accidents. Although the high cladding tem; imply sharply reduced heat transfer rates to the surroundings (and consequentif longeIdebling times), only slight reductions in the fuel temperature are necessary to red,uce the stress shaiplyn Fof a fuel with Zrli; , a 40*C decrease in temperature from 930!CTto 890*C will redue(the stress by a factor o 2. g N n r l ,bove analyses and limits are generic. They establish the bounds of the element's 1ity; the limlt(fre not related to any specific fuel element power or fissionable material

                 ?tThey relate to the temperatures in the element, to the properties of the fuel, and to t              "pf a'nd the stress on the cladding that can be allowed without cladding rupture.

Th 'sillinits are thus reaffirmed; they continue to be fully applicable to operation of the MNRC at 2 MW.

 's

Rev. A 7/5/96 4 58 4.5.4.1.4 Performance of liigh Uranium wt % Fuels A substantial review and evaluation of the performance of 20 wt.% uranium fuel elements was conducted based on the information provided in References 4.23,4.1 and 4.24. , The basic conclusions are: The performance of these higher uranium contentfuelsh}ubsta$&$$ dependent o uranium content up to 45 ut.%. The 20 ut.% fuelillNiistinguishhb$ e 8.5 ut.

         % fuel. Fuel granth is as predicted; there is limit}ihermaljihatid '                          ,

there is no pressure buildup inside the cladding &upffcceeds; at . product release (fraction) from high-burnup elemefinnot sigm'ficantly d(lf rom freshfuel (Ref. 4.23). From these studies, the release .* offission products were observed not to be related to uranium content lation serves to describe the gas release behavior over a broad tetNferatu  ; . 4.24). The basic releasefractionforfuel temperatures lesshan 00 C  : assessed (1.5 x 10'). These studies covered burnups up to $4 . pranlu5235dontent. y

                                              /g P

i rties, irradiation In summary, the prompt negative temperature coef performance, behavior under pulse healing, and,effect ol en disassociation on the fuel element safety limits, for fuel contalidEg up 19p5 wt.% wanium, were all found to mirror that of the reference 8.5 wt.% fuel. 7 g'f g// ill rearly double the energy production Increasing reactor power leit! mIM\h M in individtfalTRIGA 9 erforrllanc fuelh(dsserve review. The effect on fuel integrity du on power the inwli hn[p$ rating nI8$31scussed next but the effect on fission andisieso postponed untllieciiE4.5.5.7MPif? JG @n . Q1:if

                 ,.   -ui               V TRIGA fyls f 20,;mg    y Awt. % uranium,19.7% enriched, were irradiated in the O Ridge,liescarch Reactor1 (ORR) and thoroughly examined (Ref. 4.23), (Ref. 4.25), and eva$ led (Wf. 4.1). Tabic;4'S presents a profile of the irradiation conditions of these elements.

vl} Q

   %performancepaibfor the higher wt % LEU fuels irradiated in ORR encompasses burnu than tiO% of the contained U-235, exposures as high as 919 full power days and fast to greater;gof 5 x 102'n/cm n'esit6dj 2

The maximum fuel temperature in the ORR irradiations was' comparable to the worst case fuel temperature predicted in Section 4.6 for the MNRC (620'C in ORR compared to 630*C in MNRC). O r

4 59 Rev. A 7/5/% g) ( \'~'/ Similarly, the maximum linear power density during the ORR irradiations was comparabic to the maximum predicted in Section 4.5 for the worst MNRC case.* TABLE 4 5. ORR IN PILE IRRADIATION PARAMETERSt

                                                                                          #jd-20 Wt.%,U     30Wt-%U       45 Wt %U Contained U 235 per 22 in, fuel rod (g)                            336 M % 55               jf Vol % U(19.7% En.iched)                             M[lf' 7                      N 20       &

Max Calc Rod Power Generation (kW) d 8 4711"D$@j!h["I ' Initial Configuration Full Cluster Configuration MA[ V41K4., 36 43 k'48 5 45 Wt % Only Configuration  % .4 55 Time at Power (FPD)  ; Mhh

     .             Initial Configuration (Dec 79-Nov 80)            '0 %)%27.8                    278 Full Chuter Config (May 81 Nov 82) 45 Wi % Only Config (July 82-Nov 83 ed!A0 f p 295 Q295)>                        295 328 45 Wt % Only Config (Aug 84)

Target Burnup of U-235 (%) Ml3% Sy)$1ffiW9 40 3[30 0 18 50 Final Bernup Range (%) /i , ds51ES 47-57 60-66

                                               $                     N experienced

('y it is stated in References e4.23 andA25 temperatures that.n$ by the LEU fuel during the ORR irradiatbas ranfeffrom 25fC to 650'/C. The upper end of this temperature range exceeds that predicted folIMW oMMI MNRC. In addition, the performance of these fushnder extenited %ulse heating has been reviewed and evaluate ({Ref/hl). Tlidlthe%malcycig rmal cycling specimens were cycled 100 times out of pile and 2 then 32 timOysifeutron flufof3p10u /cmn /sec over the temperature range of 500' to 725' C. The f$elihOlayed ou"tstahlEdintegrity and stability. My A 45 wt.ItN,4W; %.$Ib MU fSI rod tii$t was instrumented for measurin pressureTas subjecte(to a serids of 30 power pulses in a TRIGA reactor to maximum temptures in the rangyffl050' to 1100' C. Only very modest (generally less than 2 psi) pressure pulses were measured in the rod as a result of the pulsing. This is in agreement with us data showing 'n'egligible hydrogen release during the pulsing of 8.5 wt% uranium fuel 9 quatures up 161150 C. All surveillance examinations showed no rod deformation. 1 %ve;show(that the pulse response of uranium-zirconium hydride TRIGA fuel is i n of the uranium content of the fuel and is dominated by the behavior of the zir inf(ydride, along with the prompt temperature coefficient of reactivi fuel does not necessarily have the benign response to power pulsing that was demonstrated in (n) *ORR maunnun was 1.26 nmx/ avg x 55 kW/55.9 cm = 1.24 kW/cm. MNRC maunum is 1.33 max / avg x

       - 33.2 kW/38.1 cm = 1.16 kW/cm

l i Rev. A 7/5/96 4-60 , these tests, liydrogen migrates to the fuel pellet periphery during burnup and a strong pulse under these conditions can produce excessive hydrogen pressure and cladding deformation. The pulse analysis in Section 13.2.2 predicts that highly irradiated fuel can be subjected to a reactivity pulse as large as $2.12 without damage which justifies the $1.75 limit specified in Section 4.7. g

                                                                                                    .h It was concluded in Reference 4.1:

Aj qMis [N ' Tests of uranium-zJrconium hydridefuels # have showit 1NNt the the operation of TRIGA fuels is independent of urahm content,'up to '45 MQ (3 {-JR jffc r

                                                                                                        % q%[

Q mv Both the 8.5 and 20 wt. % fuel variants can be expect 6d toipeiform reliably and Mil ~a the new power rating. NM$$

                                                             ,h%d&

, 4.5.4.2 Prompt Negative Temperature Coefficient V Ng( 4% k$7 The basic parameter which allows the TRIGA. reactor system to operate' safely w.ith large step insertions of reactivity is the prompt negativihiiifdature"$befficienfissociated with the TRIGA fuel and core design. This tempcNture co[ffibidtMallNs a greater freedom in steady-state operation as the effect of a6cidentaldactivity cha65Es occurring from the experimental devices in the core is.grfftly re uEd. GA Technologies, the designerp each have de ped techniques to calculate the temperature j coefficient acetratelj and, thsrefMEMdict the transSnt behavior of the reactor. This tempe'rit0rgcoeffici$ishides primarilfifkhlfa change in the fuel utilization factor resultinffMS%ir:heatin'gTMHranium zirconium hydride fuel moderator elements. The coefficientls)rosMgcausIkfudis intimately mixed with a large portion of the moderator, thuizfuel Mixt solidE6dkad temperatures rise simultaneously. A quantitative

                                          ~

calculation of thMmpeiMure coeffici8nt requires a knowledge of the energy dependent distributi[dTtherrifalydtironjux in the reactor.

           $                 Wh%f Thegtksic physical procMf(Ehich occur when the fuel moderator elements are heated can be dedribed as follows: tN rise in temperature of the hydride increases the probability that a t!Iednal neutron in the lf$el element will gain energy from an excited state of an oscillating lijdr5 gen atom in thiiattice. As the neutrons gain energy from the ZrH, their mean free path isMnecd appre'ciably (Ref. 4.26, Fig. 4.15). Since the average chord length in the fuel c1(minthsomfarable with a mean free path, the probability of escape from the fuel element befdre rcapture is increased in the water, the neutrons are rapidly rethermalized so that the capture and escape probabilities are relatively insensitive to the energy with which the neutron enters the water. The heating of the moderator mixed with the fuel, thus, causes the spectrum to harden more in the fuel than in the water. As a result, there is a temperature dependent fuel utilization factor for the unit cell in the core, which decreases the ratio of absorptions in the

1 4 61 Rev. A 7/5/% fuel to total cell absorptions as the fuel element temperature is increased. This yields a loss of reactivity. The temperature coefficient then, depends on spatial variations of the thermal ngujron spectrum over distances of the order of a mean free path with large chinges "^ o{the mean free path occurring because of the energy change in a single collision. tve description of these processes requires a knowledge of the differential sloFneutron~ ~ ~ ' ansfer cross section in water and zirconium hydride, the energy deppEe of the~ oss section[ of hydrogen as bound in water and zirconium hydride.fthe energygpen cagtu[e

                                                                                                             ~

and fission cross sections of all relevant materials, and ggy gictor description which allows for the coupling(multi ppup of groups up transport b)M@l as well as by 1(l(lskr[wIiig down.

                                                                      $f$$Al%
   . 4.5.4.2.1 Zril Model                                                   K:$

Qualitatively, the scattering of slow neutrons Q.- ydride% described by a model in which the hydrogen atom motio t monic oscillator with energy transfer quantized in multiples .14 e ,

                                                                                      , the  SUMMIT model uses a frequency spectrum with two b            ches,         r th     _

modes for energy transfer with the bound proton, and the ot the stical es for energy transfer with the d lattice as a whole. The optical 0.14 eV, and whose width is ad are dtof' sented cr broad frequency band centered at tion data of Woods claL (Ref. 4.27) . The low fre ncy acousti ' es are al ave a Debye spectrum with a cutoff of 0.02 eV Ight d an e 'Ipitgifmass of 360. _.o m 9 This structu ws a A l ow down by the transition in energy units of ~0.14 eV as long as nerg'y3dMisEt0.14 eV. Below 0.14 eV the neutron can stil; Vse energy by ss o$iting acoustic Debye type modes in which the hydrogen atoms m , m phas airconium atoms, which, in turn, move in phase with one ano These modes , correspond to the motion of a group of atoms whose mass is m reater than that g ogen, and indeed, even greater than the mass of zirconium, se of the large effg6tive mass, these modes are very inefficient for thermalizing neutrons, r neutron energini below 0.14 eV they provide the only mechanism for neutron slowing ithin the 7,rF (in a TRIGA core, the water also provides for neutron thermalization Od4 eVh in addition, in the Zrli it is possible for a neutron to gain one or more e 0.14 eV in one or several scatterings from excited Einstein oscillators. Since the @f Mro excited oscillators present in a Zrli lattice inc. ases with te process of neutron speeding up is strongly temperature dependent and plays an important role in the behavior of Zrli moderated reactors. O O

Rev. A 7/5/96 4-62 4.5.4.2.2 Temperature Coefficient Calculations - Calculations of the temperature coefficient were done in the following steps:

a. Multigroup cross sections were generated by the GGC-3 code (Ref. 4.28) fona-homogenized unit cell Separate cross-section sets were generated for escl/fiel element temperature by use of the temperature dependent hydride kernels /5dheIdjiler broadening of the U 238 resonance integral to reflect the proper te&jbraturei(WQat, room temper-ature was used for all prompt coefficient calculations s gggk;gg h f W

b. A value for k, was computed for each fuel elemerit,,tegpera3ure by transpug _lts to calculations, using the P approximation. Compar"isonslhave shown S4and Sg resu 3

be nearly identical Group dependent disadvantage'fsctors ivere calculated for each cell re c

     #gion       fuel,
         / $5 (region      clad, and water) where
                       / cell).                          9'                                          the disadvantag6
                                                                                                    '%   k                     fa'lorLis d gh                                               T Jb

c. The thermal group disadvantage factors wersiised ninput for h spnd GGC-3 calculation where cross sections for a homogenized 6or6Ycire" genhiated whidh balance as the thermal group portion o'fIhe discNte celicaiculaIIon.ga

                                           !/        $                              %l

d. The cross sections for an equivjden homogerilred cor6:were used in a full reactor calculation to determine the cdntribution atdthe tempdr'ature coefficient due to the increased leakage of thermal neutrons,lato the reficctor with1ricreasing hydride temperature. This calculation still requires seyMal thermalMr6'upis$but transport effects are no longer of major _c'ondern. Thus, Yeactidi,ty calculations as a function of fuel element temperature hav6been done on the critire tsactor with the use of diffusion theory codes.

Vv A s ' w%

                                    / ' JA m

Indidate that more than 50% of the temperature coefficient Results fromcoretifejbdicalculatiorj[frbih for a standard TRIGA" comes the temperature-depe effect," anxi approxiitiately1.0(cach 2 from Doppler broadening of the U resonances and tempe,rature dependent leakage'fr'om the core. These effects produce a prompt negative tempdature coefficient of approximately 0.01 %/*C (1 x 104 Ak/k/*C) is rather constant with teniperature. The temperature coefficient is shown in Ref 4.26, Fig. 4.18 for a typical high-ifdride TRIGA core y

  ,-                    ,3 N146.3 Codes wlised for Calculations g gg#    "

Calcuhitional work on the temperature coefficient made use of a group of codes developed by GA Technologies: GGC-3 (Ref. 4.28); GAZE 2 (Ref 4.29); and GAMBLE-5 (Ref. 4.30), as well a: DTF IV (Ref. 4.31), an S multigroup transport code written at Los Alamos. Neutron cross sections for energies above tSermal (> 1 eV) were generated by the GGC-3 code. In this code, fine group cross sections (~ 100 groups), stored on tape for all commonly used isotopes, are averaged over a rpaci ;ndependent flux derived by solution of the B 3equations for each

i 4 63 Rev. A 7/5/%

 \   discrete reactor region composition. This code and its related cross section library predict the age of each of the common moderating materials to within a few percent of the experimentally determined values and use the resonance integral work of Adler, liinman, and Nordheim (Ref.

4.32), to generate cross sections for resonance materials which are properly averased over the region spectrum. t ' 8 Thermal cross sections were obtained I in essentially GC-3 code. the s Ilowever, scattering kernels were used to describe pro [tlie inter neutrons with the chemically bound moderator atoms (Zril). und by gen hydrogen in the water were generated by the TIIERM cod ((Ref. 4.3 thermalization work of Nelkin (Ref 4.34). Early ther work by McR (Ref. 4.35), on zirconium hydride has been greatly ext _ Technologies (Ref. 4.36), and work by Parks resulted in the SUMMIT (Ref. 4.3 .i was used to generate the kernels for hydrogen as bound in Zrli. These scatte g mod 6tgy n used to predict adequately the water and hydride (temperature de ) spectrrs ! red at the GA Technologies linear accelerator as shown in i. 4.16 ai 7. N 4.5.4.3 Cross Section Generation f. Q Q A new broad group cross section sjg improve accuracy. One problen4kt. deve dfor analysis. The motivation was to ie 43[ group E , CELL cross secti first addendum is that the fine sg,ep libra ed t nerate them are so old that it is not clear what data base theygym Re Efinite lattice problem using those cross sections d to uq uous en nte Carlo show less good agreement than might bF ,it wa. e generate new cross sections using the WIMS D4 code (Ref. 4.3P)' DF/B ;pross section data base. WIMS is a general lattice cell program that _ and collapse of neutron cross sections using transport theory-based cell s be 'infd extensively for many light water reactor systems, including /litIGA rea_ " tions for the Reduced Enrichment Research and Test Reactor (RERp[ Program at d tional Laboratory (Ref. 4.39). It was decided to collapse to 7 s, rather than 4 prder to reduce substantially the running times for the full reactor ations. tions weJe produced for every cell type that could be anticipated for the 2 MW 441d all. The dimensions and compositions used in the cell models were the

     !                 Il obtained from GA. The coolant area for the pin cells was based on the hex pit           new MNRC grid plate. Unlike the EPRI CELL model, the central Zr rod in the fuel cells was not smeared throughout the fuel. For each of the four fuel cell types, an infinite lattice of that cell type was used. Non-fuel cells, e.g., the axial graphite reflector cell, were driven by a homogeneous core composition, with a water buffer region separating the core

'v;

from the cell. Room temperature (300 K) was used, which implies that the cold to-hot reactivity swing must be accounted for separately. The collapsing spectrum was generated

Rev. A 7/5/96 4 64 with a near-critical buckling. The energy boundaries of the 7 groups were chosen to be as close to the standard 7-group GA structure as the 69 fine group library structure allowed. Several tests of cross section accuracy were made, using continuous-energy Monte Carlo solutions produced with the VIM code (Ref. 4.40) as the standard. VIM provide)s essentially an approximation free solution to the Boltzmann neutral particle trans@t equallon limited only by statistical precision, the reactor description and thepasic neutron *cr(wi section database. It has been used extensively in the analysis of pitihal expklmespui for testing of de'erministic methods (Ref. 4.41). The cross sections p%ced withjVI upplied a diffusion theory code (see DIF3D description in the " t subsectdrii to .' results. The results from one of these test problems, finitr/Mtice of 2 without leakage, is shown in Table 4-6. The determit ~ uvalue is slightl 16#iMe main contributor to this is under-prediction of ruU fissidri 5. Also, absorption in Zr in the hydride is over predicted, probably because a WIMfp . Inade it necessary to use unbound Zr data. Overall, the agreem:nt is quite go6if' Fo. , leakage was areficct eeping the array introduced by making the fuel cells infinite in the hex plane. The deterministic e,lgenvalue38.1 cm tall (andp$githin ork", ard de Monte Carlo value,1.2480 compared to 1.2f72T'.0020s , ax,ial'rellector and steel end ! regions were added to the model. For this, case, tt4@ igenvalue is slightly low, ! 1.3034 compared to 1.3056 .0011. ghas corgil6ded fidii@ill' ese results that the accuracy of the fuel element cross sections ted WIM Inore than adequate.

                               /A                       M CROSS'ASBGTION ACCURACY TESTSMP! NITE LATflCE OF 20/20 Quantitpph. N fyfMNatue                  WIMS Value                   WIMS Error Eigenvalue $ [ ",              174200if.)0d11             1.4175                       -0.0025 Italance _M% .W\                    Y absorption           n (1,0000i.0015                 1.0000                        0.0000 Nission src           .
                               # J 1.4190 i.0025               1.4169                       -0.0021 5 'n 2n               d        0.0005 i.0000              0.0004                         0.0001 3 fission       ,#          0.5813 .0010               0.5804                        -0.0009

[dahure? 0.4I87 i.0005 0.4196 0.0009 Absorption U235 0.7036 i 0012 0.6982 -0.0054 U238 0.0893 .0003 0.0889 -0.004

4 65 Rev. A 7/5/% i m  :

       )

J

                                                                                                                     ~

Quantity VIM Value WIMS Value WIMS Error l Absorption Er166 0.0042 i.0000 0.0042 , jj!d000 Er167 0.0086 1.0000 0.0085 /$ h [ 0.0001 Zr (rod) 0.0006 i .0000 0'.0608

                                                                                         .                  3     0$hh.0.0002                  y Zr (ZrII)                          0.0228 i.000i                f$6.0266 ,8 ' G j.'9018 ,,f7
                                                                               >s                                                                     .

11(Zr11) 0.0278 i.0001 p,0.0277/u +Ny4+.000]Py U234 0.003I i .0000 \M.[0030\ -0.0001. U236 0.0010 i.0000 [0.06(d$k 0.0000 Fe 0.0202 i.0001 og 0.0205\!M(e: fp- 0.0003 11 (1120) 0.0303i.0001gh k M 0302 h -0.0001 Fission f.[ [YIMM7 U235 0.5786t(Obl0 p hd0 -0.0046 { s U238 0.0dOf.'0000/[ ,/5.0046 0.0002 U234 ,, A h0 lt.0006 Mr; ggY 0.0001 0.0000 Spectrmn$k Ml5fh D)V Gr N IBins$(cy) %_h. . j h. ,,., 3 0 gl'.4 m ,,,19878.i

                                    ~-...:... x   0.0005 i 0,0000                           0                                -0.0005
                          ,            _      ,,         s I   A

• 1.000B7/1 %,9.3332I0.0005 0.3373 0.0041 2]f 5.000E5 l ,$If0di97i0.0002 0.2173 -0.0024 I#

             /s)               9.118E3 /          0.2643 1 0.0002                    0.2 643                                  0.0000

[bk4 1.123E07 0.0273 1 0.0001 0.0272 0.0001 dhiGn.,, A000s1

                     .                            0.0341 i 0.0001                   '0.0336                                  -0.0005 kn Mj U~l.400E1                  0.0722 1 0.0002                    0.0721                                  -0.0001 7          - 5.000E 2            0.N86i0.0001                       0.N83                                   -0.0003 O
 /    s 1

L _ - __ _ _ _ _- _. ._

f Rev.A 7/5/96 4-66 A test of the cross sections generated for the boron poison region of control rods was also made. Self shiciding in the boron is extremely large, which made it difficult to get a well behaved solution from WlhtS. A comparison of the effective cell average microscopic B 10 absorption cross section by group is shown in Table 4-7. The errors are statistically significant for most groups but they are not large. Thus, these cross sections should be adg$ ate for

                                                                                          ,A computing rod worths.

A  %

                                                                          /IO          f.T TABLE 4 7. / 7 TEST OF *B CELL AVERAGE hilCROSCOPIO%j,BSORITION CR hyQ K

K [

                                                                                                     ,       . IONr Group                         VIM Value                 y/IMS.Valde'                 \VINi$Ert0I
                                                                       ,m         .

1 0.35210.4% N' 0.344 !s 1 2.3 % 2 2.I16i0.8% /[3.1M gA 3.5 % 3 19.06 i 0.9 % ,, pg 19.82 h @) 4.0% 4 36.07 i 3.7 % / iJ[ hh2.98 )[y 8.6% 5 31,62i3.157 fjfMIMY -0.2 % 6 29.18i2.'d% f 28$ ^ -l .0 % 7 24.69$E % // /E4.88 0.8% gg A. '

                                                        /

4.5.5Ip%hto? N Physics Ardlysis(RefererIcesore's s 4.5.5.1 Ilskonsi Z? Calculational Model y_] y:: y

• An entirely ~nek 2 MW calculatiorial niodel of the h1NRC was constructed. A few assumptionsiad to b'e mide about graphite components
• but the new model rests on a reliable operational and information base.

3 p r De$cliptions of some of'the more important cell types associated with elements are shown in l T ble 4 8. liere FFCR stands for fuel-followed control rods and TR stands for the transient rod.iThe outermostTc' ell region in each element cell contains water and its outer boundary is a inexagoriwith ajitch of 1.714*. hiaterial descriptions are given in Table 4 9. Tne elemental weight fractions for the two fuel types were obtained by averaging values from fabrication data ! sheetsta 31 elemcnt sverage for 20/20 fuel and a 4 element average for 8.5/20 fuel. The l

• All tie graphite was assunxd to hase no inpurities of trutronics significance and to have de typical density of stactor grade paphite,1Je g/cc. The dianeter of the paphite in dummy elements was assumed to tie 1.435*, which is larger than actual.

l

4 67 Rev. A 7/5/96 p uranium isotopes were obtained as follows: U 235 was from the fabrication data sheets, U.234 and U-236 were from an assay of 12 fuel samples, and U 238 was the residual. Erbium isotopes for which no cross sections are available were ignored, not replaced by Er-166. The boron carbide description was taken from the previous MNRC model and it is nefrly consistent with recently obtained manufacturing si,ecifications, f l

                                                                                                                              ?

A hex mesh with a 1.714' pitch was superimposed on tl , plahl "* NRC reactor. in the reactor grid region, this process is simple, since esh bou i[ patural, alb[ artificial. At the edge of the grid and beyond, the vol fractio: mal aneaab j hex mesh cell were determined using reasonable appr tionaf or cells ar . Md outer edges of the radial reflector, the approximation few percent) w .L lace a hexagon divided by large radius arcs, with an equivale e divided by chords. For the reflector inserts, with their beam holes, the appro shift some materials a

• fraction of a cell width and to replace circular hole e s sect tangles having nearly equivalent areas.

Aside from core loading variations, there r hex) plane layouts where this process had to be carried out amp s are the core, axial blanket and c!cment end fixture level the c muu s of the beam holes make O the hex planes different throughou eigh urall ge of the ap m holes, flowever, imation of a small number of unique discretization into axial nodes I , hex planes. Those planes diffe yat 11 of hex cells where the slope causes the hole loc ons to step as the increases, in order to reduce the model e 'y fu beam e represented as having a 20' slope, even though Slp has

                              <  am 4               S.
                        ,f)             bik,                                                                                                                 ,

o ye p4 .V

Rev. A 7/5/96 4 68 TABLE 4 8, - DESCRIP'TIONS OF SOME IMPORTANT ELEMENT CELL TYPES Cell Type Radial Regions From Center Outward (OD in inches) Axial Reflector era,Shite (1.353), void (1.438), 304SS (1.418),1120 A-Fuel . od (0.225), void (0.250), fuel meat (1.435), vold (IM38), 3(54$S (1.478), 111 0

                                                                                             .4,       ji $tQ'.

m FFCR Fuel Zr ral (0.225), vold (0.250), fuel meat (1911), void (1pl4)l304SS (1.354), f 112 0 [/ ,g/ Nf7h ,,9 FFCR IMC IMC (l .300), void (1.314), 30455 (1 h54)s i .l2O[' Y;NF TR IMC IMC (1.187), vold (1.194), 6061 A1 (l')25dN 2 H' 0,(1.399), 6061 Al (I300),1120 Dummy Graphite' graphite (1.435), void (1.438), 6061 A[(I?N5j, inh Open 'Ihimble 1I20 (1.344), 6061 A1 (1.500), il2OE, M,O Rabbit Tube void (1.050), 6061 Al (1.158)N6O (Id30), 6061 AIlh485),1120 d'? .

                                                                           //         4{/%53hrf
                                                                                          %p;29*
 'After the analyses were completed it was determined ilIat the clslMalf thickness is actually 0.030", not 0.020". Accordingly, the OD for the pahhite and,vfO shown idic and used in the analyses are too large, but the elfeet on parameters of interce phould ig. negligible. //'

s j d.\ .d M (lE s'dfhg,,,h w Q, , ., i%' 4 w% a i ~,

                        %7            t, w._       t.m                 c.' .

( h.A 'ti .' h- , j b,

               ,s .
                    .r                     V(

i'] f = 'v! ' ' i :. f l, lr 8

             .+
      ,    '[                ,.e.,,
      }}T' h               :t :

w....._ O L

4-69 Rev. A 7/5/% p ( . ! FA13LE 4 9. COMPOSITIONS USED IN Tile IlEX Z MNRC MODEL Material g/cc Element (Weight Fraction)' 20/20 Fuel Meat 6.5474 11 (.01337), Zr (.75411)! Er (.004ili U (.19495) (.03.859)} U236 (.00037), [U234 (.00044), U238 (.15555)], impur U235,ities'(.03286) neglected 8.5/20 Fuel Meat 5.9643 11 (.015ff)IZr ( 89067)[d[(l08335) [U234 ,j (.0001))[ (.066(0)),,impurkE U235 (.01640), (.01012) Ngleesd U236 ;V (.000f 6) Zr Roi 6.506 Zr(1.b00IME 3NSS 8.032 Fe (.7,0lNNfUOpi5), Cr (.1900), Mn (.0100),

           .                                                                Si(.0050)        QKh 6061AI                       2.70 4 Af,006b));Qu (.0030)lneglectedg(jfA 112 0                         1.00          /[ llhiiiNh488k p               B4C                          2.49         3((            ,'d(.216),MM) [ BIO (.I56), Bi1 (.628)]                                     ,

Graphite 1.60 d# ,d2 C (l.000'Y )

                                    ,         AW mmman'compositled
                                                               / 4?    y~w ers are:
              *(net Equivalent          Valiacs of morp{fMtib(pre wt o(fdileineit below                           d whereas    hydre weight fractions in the table are relative to gross weight (Ile![afiier s hydridlig%[y?%,

20/20/kel meal U-235~snass =97.436wtd U235 in U = 19,79, wt% U in net wt= 19.76, wt% Er in net wt=0.478, II/Z%ratid= S.5/20fuct maw U% 37.7g, manip,1.605t%@6 wt9 U235 in U= 19.70, wt% U wt=0, il/Zr atom ratio 41623;%s CN' B4C *M N 19.Vwi% fL S Jk ri B10 in B. h%f K'[ j compositions the usgfortouching radial re0ector, and positions positions of theor beyond the radial reflector MNRC hex plane are discussed in mo i ti rESoction3d.3 of Ref 4.4, The reactor grid is represented by 121 hex cell positions I ( jli.34.8MThe outer boundary of the model in the hex plane was specified so as to omit

             - regions ofTo significance to neutronics performance. There are 18 hex .ings in the model but only the first 15 rings are complete. It was demonstrated with test cd ulations tvi completing these rings has no significant effect on any quantity of interest.

( ( Many variations on this model were used in the analysis but they differ only in the reactor grid area. The meanings of reactor grid position symbols are ghen in the next section. Programs

Rev. A 7/5/96 4-70 were written to modify the reactor grid loading easily and verify that the intended loading was specified. Although the use of diffusion theory is a source of error, the errors are acceptably small. The three-dimensional diffusion theory code used, DIF3D (Ref. 4.43), employs m efn nodal diffusion theory, which provides an accurate solution within the diffusj6n the Tramework. It has been used successfully countless times for reactor design calculfdolhird critical experiment analyses at Argonne National Laboratory and cisswhereRPdEge dominated performance quantities, such as beam tube flux, diffusig[n'tifeory the relative merits of various loading options but not giv accurateMiiIoluYe ' ice cank levels. Diffusion theory should be acceptably accuratfor the,qfiiittitles of jance, such as excess reactivity, peak fuel element power and)6diskths. When tested ta variational nodal transport solution, the DIF3D diffuslEt%gediction of peak power was found to be in error by less than 1 %. /[${JRj

neglect of one Test calculation results indicate small errors of the fuel meat impurities (carbon), replacetagint B crystalfronpoumpther ap "'sMM bea stainless steel, and the spatial truncation errdfrEii nodeff hex in the radial plane.

A calculation using Er-166 as a substitute %r the rpR Risoedfes in 20/20 fuel, indicated that neglecting these isotopes could mal $f the caldifated eiijigr(activity for an all-20/20 loading too large by about $0.7. p f J 4 The combined error from all so. mp' s-m g, s and data - is indicated by comparisonstbetween calc

• I and me$sistiliiB escribed in the next section.

b[hr 0 N!!$ V 7 Model Analysis 4.5.5.1.*14Validahon ofh sh 5 Wh Excess reactivityjtOkworthhM'd in several loadings of two different TRIGA reactors were compared'ioyAlulls compute 71?pitfi the new model and cross sections, in addition, the measured nd calculalifd"pdfuel temperature was compared in one case. A series of three loadingsiuilt during thYialtialffartup of the Cr:igladesh reactor were modeled. These loadini,s had fresh 20/2 prehbus MNRC loading (J6efin an hexagonal reactor grid. The

                            , the original Core B loading and the January 1995 mixed-fuel Idadhlg, were the othebloadings used for the comparisons. The meanings of the symbols used are gwen in TableAjl0. Measurements in the Bangladesh reactor were made by GA during tlAciddmissynitig%f the reactor, and the measured data were supplied by W. L. Whittemore (RefE4;43).MFdr more detail of these modeled cores, see Section 4.3.3 of Ref 4.4.

my O

4 71 Rev. A 7/5/% '

                                                                                                 ~

V TABLE 4-10. SYMBOLS USED IN REACTOR GRID LOADING DIAGRAMS A central thimble if in Rou A, otherwhe aluminum dummy element # ' C fuel-followed control rod (FFCR) , fin 8 D graphite dummy element ,dh

• Y$% j E experimental in< ore irradiation facility ff[ .6 # N[DA ,[

F 8.5/20 fuel element hk[ %l,M[ 11 10/20 fuel element WW

                                                                        .c.

L low worth, steel control rod [ Y UN P pneumatic rabbit tube ,p m N@[p S source holder /h%$$k 'bf T transient rod (TR) jfk [d3h%'![ W water-filled (vacant) GIN! positlodI E*#

wt if '

i t f

T/ ,

p // While they give a reasonableti ndication of tional accuracy, none of the comparisons below givedi A unambigudUsIndasure of dbrtodi$ predictions of zero-power excess reactivity)Q and Q1ljdere essenlisll[ water reDected, which is unlike any loading of direct inteiestbThe Bang 100'idadel has the MNRC radial reDector with its inserts, rather than the Ban 31hdesNEffector iDIEN its'INnope production facility. That this difference can be significant is illAtraneil 6) the $1'93Ahrease in calculated excess reactivity when the reDector inserts were'$dded'tdiprevhsly ci6an model of the radial reDector. The Core B and Jan '95 loadingsNere actually'injajimular grid but they were modeled in a hex grid. More importantly, all fuel is modeled as fresh, whereas, in reality, the Core B fuel was slightly f buried and much of theian*95 fuel was quite burned.

     $?k                     n The experimental and 6alculated results for zero-power excess reactivity are compared in Table El16The errors' sire relatively small for the water-reDected cases, Bang 50 and Bang 71. The NsD95'briIdis firi larger than the others, probably because burnup was not accounted for in thWaisuintiod. As an attempt to generalize, it appears that calculated predictions are high by roughly $1.0i0.2 for graphite-reflected loadings with little burnup. Depending on how burned the fuel is, the prediction could be high by as much as another dollar or more.

(D

Rev.A 7/5/96 4-72 TABLE 4 11. COMPARISON OF MEASURED AND CALCULATED EXCESS REACTIVITY OF HEX-7 MODELED CORES Zero-Power Excess Reactivity (S) g Loading Measured Celculated j ( fj$rror e Bang 50 0.19 p.48, /;A y;m,,A 0.49 f) J.. Bang 71 8.94 /j.,89 jj $N.dN.05 Bang 100 10.27 [$1,1.40 AV 'Tijjjj$,;f m.. m ,,. m.- Core B 6.75 W7.62% 0:87

 , Jan '/5 6.95               487/8L %y                 1.83
 ^
                                                                        ~         vw w AIh                 h The experimental and calculated results for rod worthiaretshown i?iTable 4-12. It appears that predictions of individual rod worths arigneralikssufate sto githin about 10%. Rod worths are measured one at a time, by m$fing thelinglO6d"aisirst a bank of all the others.

Correspondingly, the worth of each indifidual rdwas caMulat6d from the difference between a rod-up model and a rod-down modi}i where kil other rdds were banked at a constant, near-critical elevation. For the design #Eciding calElations id5ection 4.3.3. of Ref 4.4, only the total rod worth was calculated ddthis wa.EdSte(d2 dues in the last two and all rods (down. Compsilori"of the cajeulMe shows thaftni@ total rod"w$rth<ls'about 13Y$lkl$r than the sum of individual rod w

                   %g%[(Q.

th 'q77aix C ' U.Q

                              . N!                          _%

{

                            ,          a               e. 1 jh ' Y                               s
                                    .g          -

t ,,  : ,6

            ,                          %    ..:?
         .'*5

[ [ [? Oj bd

    /j - A                       , is Il' A                   ,y '
    * }j((

m O

4 73 Rev. A 7/5/96 / ) - \ l

 %d TABLE 412.

COMPARISON OF MEASURED AND CALCULATED ROD WORTilS OF llEX-7 MODELED CORES Rod Worth ($) ,. V Experimental Calculated f. t- 4[ C/E Core D /b/ \ N g DIO FFCR 3.69 jY 3.98 [ VW1.0,8 ,,,fr# Bang 100 h([ \eMY D01TR 2.24 D2.3$h 1.04 1.00 D04 FFCR 2.82 [ '2il[A

                                                                                            >       0.99 D07 FFCR                              2.73      ,   a        2.70 % O [

D10 FFCR 2.78 /OU. $;$$?.2.73 M 0.98 013 FFCR 3.%7 , /%I2.30[ ' 0.91 Q D16 FFCR ,4[I2 f[ h2Y5 O.88 sum individual /hd.75f[ [ 16.11 0.96 all simultaneously j @ [fb M 18.20 AQ fi 4% V My* A test of%k ti, ratuYpr6ddlons was5fde using a measurement performed in the Bang 100 Idading' png E1:2 Min the nat'ral circulation mode. A neutronics calculation of th 'er distributioh was marie and then these data were input to a thermal hydrauliss% intion. TWaciuracy of the prediction cannot be determined precisely, because theltheriii6 couple! wis in a sfeep therma! gradient, although near the center of the fuel. Still, the'igreement seb{ns' to beircasonable. The measured temperature was 415' C, the calcu$ted centerline tenip.eratuie is 440' C and the calculated temperature at the estimated thermocouple location li425' C. This result constitutes evidence that the neutronics calcu.iations of power dis'tributions are reasonably accurate. li$ N 4}5,.512 ReferenceLCore Mix] - Mixed Fuel Loading fy x;yW TNhid IEdindfor the mixed-fuel reference, MixJ, is shown in Fig 4.29 (the hex grid numbEri'rfg systems; Rows A, B, C, D, E, F, G and Hex Ring A, B, C, D, E, F, G are synonymous). It is a nearly symmetric, two-fuel zone, annular core design. The outer zone has 49 20/20 elements Olling all of Row F and most of Row G. The rest of Row G contains nine

,G        graphite dummy elements, the neutron source and the rabbit tube. The inner zone has 50 8.5/20 L     j    elements, including four FFCRs, which fill most of Rows C and D and all of Row E, Two Row
  #      C positions have graphite dummy elements and two Row D positions have air followed control rods. The central region (Rows A and B) contains the plugged experiment facility.

Rev. A 7/5/96 4 74 O m.,_p3- _ ,g g._

                                        .:g .::: ::: ::: :::. iiii: ph di...::((!!!.Gi!Mi!)ii!!!i ii:i:

i!!!j)!!!:i!:@l

                             ^

ij: Q Il #@ :li: iji:!!::ij:!!!)(!!!!i

                                                                                  @ g Gili-
                     .iis. :iii gj ii                            iii. h :ii: o ee!!gy::.<@!                           g!!: !!! !!! e e
                     ;;::. .i.!:: :                 gg                        gg R

e:!!!.!!:1eiiiviiiiie i!!@h,.y:!!:%g!:!.!:

f~ ..

                                                                  ..w          :.

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s

                                     .h !!;i: :i!Mdij.9 i!;i: !:!!!

h -if i : !!!!!! hh W' TRANSIENT R00 20'20 FUEL ELEMENT

                .~

lfs. LOW. WORTH S.S CONTROL R00 O

               .=.-

';~ OR FUEL FOLLOWEO CONTROL R00 &

I':::? c.V10 FUEL ELEMENT Q FUet Fou.owEo coNrRot 3 Roo aRARNiTeoUuuYELEMENT PNEUMATIC TUBE NEUTRON SOURCE p . v REFERENCE CORE MIXJ 9 FIG. 4.29

l 4-75 Rev A 7/5/% T T U The calculated MixJ zero power excess reactivity is $8.24 with all fresh fuel. Recall that the calculated excess reactivity tends to be overpredicted by a dollar or more. More fuel may well need to be added, depending on the actual amount of fuel burnup. Fuel would have to be removed if the excess reactivity is so large that the 1/3 down guideline would be c(geeded, There is ample flexibility to adjust the excess reactivity by fuel-dumm stit s in Rows C and/or G. The approximate reactivity change produced by such a subs a function of grid row is shown in Table 4 13. In accordance with the t . ne la ubstitution the ot: ' he exact involves 20/20 fuel in Rows F and G and involves 8.5/2 worth depends on the specific grid position (with highe rth on th frat ' ._. Am on t corner) but the table provides adequate guidance. B ofa that mgg mused A ni ith fuel wash QNMand power might the peak power wentbedosaffected, negligibly. Thereplacement peak eleme of the two Row % ith all fresh fuel'O'rd' control rods full up is 27.6 kW. ,- TABLE 413.' FUEL-DUMMY SUBSTITIffiRIMORTI Positio gp 2 % sl S h )ydtth($) Col [ $RURY OA8 CO2 fN Y [f 0.S$ [}\

                                               &Y

^ U D02 fj ff 0A0 i E02 4 (( k/$ 1(7 0.30 gN Folh y d D$ # 0.21

                   tE@%G02 VU@s                                                         0.18
                         %-5 ;.2:R 6                    % xfI The calculaIhYxicN;4INxi wort (s are shown in Table 4-14 The individual rod worths corresp6ii3 to raising thA' 6d in hilestion while all other rods are banked in the full-up position.

Thegrth of the transieQ li $2.05. A new rod type may be introduced for use in the sixth rod position. It has the pme external dimensions as a FFCR and the same void follower at the bd0dm, but the fuel secdon is replaced by a solid Type 304 stainless steel section and the boron dime section is repikted by anotJer void follower section. This steel control rod has a low

         $6fth'(50 169), making it useful for fine reactivity adjustments.

p;\Mif#7 It hgtowortNh that large rod interaction effects are predicted, This is evidenced by the 37% ditTerence between the sum of individual rod worths and the worth of moving all rods simultaneously. Another manifestation of the interaction efTect is the dependence of a rod worth on the bank position of the other rods. For example, the worth of raising Rod D10 when all p) ( other rods are fully down was computed to be $5.79, more than twice the worth computed when all other rods are fully up ($2.62 from Table 4 14). The computed reactivity change from all

Rev. A 7/5/96 4-76 rods up to all rods down except for DIO is $11.39. This reactivity change is related to the shutdown reactivity with the most reactive rod stuck in the full-up position. The computational approach took proper account of rod interactions and it is encouraging to observe that the resu is between two alternative estimates obtained from the Table 4-14 data: 1) om!tting DIO from the summation ofindividual rod worths yields $9.96, and 2) subtracting the DIO v4rth from the worth of moving all rods simultaneously yields S14.56. In any case, therpe,,is ade$a'te reac control and shutdown margin. W f A @ ey% . It is not necessary that a steel control 8.5/20 FFCR or by any of a number of stationary elem rodin,c!(dl$g*it Iypes, be used. gga'Ihis Fpd80sition e c dummy or an aluminum dummy. The $0.69 loss in re ty contr6ffrom

                                                                         "              using "

(to use 20/20 fuel id clement clearly would be acceptable. The only restrict position. i M.- y.... r as presented in Table 4-Calculated rod worths in MixJ with a fuel followed con 14. A$% d MA kiWV

                                                   /Riidlyg                  W
                                                })    E 4-j hh CALCULATED RODgORTHss                          J MixJ with Stainless Steel Control R6d[in D13 y        MixJ with Fuel Followed Control Rod in D13 y/ orth ($) fN            / Rod (s)                   Worth ($)

Rod (s) D018.5/20ff,CR ,#)gM2.43 hp; lIE)I 5.5r0 FFCR 2.43 D04 TR($;k  %; i 2 05

                                                 *     'D04 TR                              2.05 D07 8.5/2dNdtk               M N$h                  D07 8.5/20 FFCR                     2.52 DIO 8.5/20 FFCR                     2.62 D10 8.5/20 FfC}th [ Sh                362)#

[ Vig 0.69 2.27' Dl3 SteeIOE 3 D13 8.5/20 FFCR D16[5/20 FFCR jO[ 2.27 D16 8.5/20 FFCR 2.27 suii[ individual i 12.57 sum individual 14.16 a!'l'Aimultaneously 5 171.8 all simultaneously 17.34 " r N$w .m

           $ % ?,?

• Aisuised Evest rod worth replacement of steel CR with 8.5/20 FFCR

  " Value for 20E which contains five 8.5/20 FFCR's The shutdown margin for MixJ having an excess reactivity of $8.24 and with a stainless steel rod in D13 is 50.71. This was deterW.ned using the most reactive control rod worth in the up position of $2.62, and the higher worth experiment being $1.00. The shutdown margin for

4-77 Rev. A 7/5/%

    -m

[V) . MixJ with a FFCR in _D13 is $2.30. For a loading of Mix] having $9.50 excess reactivity with a FFCR in D13, the shutdown margin is $1.04. Both loadings of MixJ satisfy a minimum shut down margin of 50.30. A total rc,d worth of $12.57 was used which is the sum of the individual 6-rods. ere com vater Two alternatives to elementthe pluBFed replace the Al/C plug results in a large drop in excess reactif kW) change in peak power 6 central but there exper is a element. The power peaking is still acceptable. Repla the plug;f .- silicon results in a small drop in excess reactivity,50.2 d a safall&. arse ent power,1.8 kW. This substitution is benign. Jy-The performance results for MixJ and changes relativehk g. '95 loading operated hypothetically at 2 MW are presented in Section 4.3.Yof Re Q 4.5.5.3 Reference Core 20E , The grid loading for the 20/20-fuel refeyeace, 20 s . 4.30 (the hex grid numbering systems; Rows A, B, C, , F, G d He , B, C, D, E, F, G are n ( ) synonymous), it is a nearly-sym , singl el-zone . ular core design. As many grid V positions as possible in Rows D G ' . filled w el. There are only two fuel elements in Row C, wPh the re that r with graphite dummy elements. The central re ' ontabs the exper I ,

                                                        .u-     -

l The calhzero - s reactivity is $9.35 with all fresh fuel. If more fuel

1. ip losses, dummy elements could be replaced by elements ne" ded fuel at corners Cal ye confirmed that this does not produce a higher j

lements for fuel in any row could be made to reduce peak power 3 y. t dum the excesafi5ctivity3 il [ app ate reactivity cll l g' Wows C and G would probably be tk bes sduced by a fuel-dummy substitution as a function of grid row is n in Table 4-15. L spin, the exact worth depends on the specific grid position, f TABLE 4-15. AFUEL-DUMMY SUBSTITUTION WORTH IN 20E Wonh (S)

          $b@S$F Position

, lD# C09 0.68 0.49 l' D12 EIS 0.25 O l F18 0.21 l-0.20 l G21

1 i Rev. A 7/5/% 4-78 O S: -: . . ::.

$$. - (8* . e-$: 5:-:-:::- $$:-

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                                                                                                              ~
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                                                                                                    ,           2S20 FUEL ELEMENT Wb ,

TRANSIENT ROO

                                                                                              %                  GRMTE DUMM ELEMM O FUEL FOLLOWEO CONTROL ROD PNEUMATIC TUBE                                             1                    NEUTRON SCURCE v

o REFERENCE CORE 20E FIG. 4.30

4-79 Rev. A 7/5/96 The peak element power in 20E is 34.8 kW (in C07) with all fresh fuel and control rods fully up. The fueling of all available positions in Rows D through G was done to minimize the peak power. This step has only a weak effect on the peak power, since, for a given overall fuel distribution and total power, the peak is inversely proportional to the number of,clements. There almost certainly would be no motivation to take this step if partistly burd"d e fuel were used in Row C. Thus, it should be acceptable to have somesdumm n Row G, for example, if burned fuel is used in Row C.  % The calculated control rod wonhs in 20E are shown in T e 4-16. trac I worth is $1.95 limit. Four of the FFCRs shown here h 3.5/20 1. These 1 eventually reach their end oflife and be replaced by 20 I . As shown in 16, this substitution has very little efTect in an all-20/20 core. nge was made in the model, the excess reactivity went up 8 , the peak element powg , kW and the worth of inserting all rod simutaneously was virtually unchangerrhere interaction effects, as was also the case for MixJ. This is th: cause ofAas:iuge di n the two

                                                               -~^

predictions of total rod worth in the table. T rths in le were computed with the other rods fully up. The worth of: m J with the other rods fully down. The result, $4.51, is about twice rge as Ia,le. The worth of dropping all rods from the full-up position but R D07 n[ovin eled explicitly and was g found to be worth $12.83. By an ese estistates, th . worths are adequate for control (V) and shutdown. 5 fh ,$6B r ~ 15,

                                     *=

N45% <=1 ATED ROD WORTHS IN 20E TiMSRod(s) %$li% Worth ($) D018.5/20 FFC(.$$$$3, N[ 2.11 D04 TRy7 Mdh.qh 1.95 D0'7M/20 FFCR ((U[ 2.36 - jbM8.5/20 FFCR 2.31

                                //

lN3 8.5/20 FFCR jf 2.18

                            ?                                                            2,12 6iifkESOTFER sNi[!511fidiial                                                                13.03 all simultaneously                                                             17.34 p

(

l l l Rev. A 7/5/96 4-80 The shutdown margin for 20E with excess reactivity of $9.35 is $0.32. This was determined using the most reactive control rod worth in the up position of $2.36, and a highest worth experiment of $1.00. A minimum shutdown margin of $0Q&is satisfied. A total rod worth of $13.03 was used which is the sum cf the individual rodP.#@ A% Two alternatives to the plugged central experiment facility were compdted. Oi$fation at full power with 20/20 fuel in Row C and the central facility 9 fill d with vpifdrIwd$alculated to quantify the consequences. The excess reactivity decreasdI2.08. #dtiedita 4% (1.5 kW) ost[peakingMkEdelement. decrease These results do notin peak indicate element that the power peaking power is a a and a:mmw 3% increase in p restricted to corner positions. However, because the sparability.assumptionsyydy analysis are approximations, fuel temperature measuremE9iratlihe anticipated hbt" should be made if this mode of operation is attempted. Replacli$glugged central facility with silicon results i.1 a small drop in excess reactivity, 50.07?akdianihil decrease in peak element power,1.8 kW. As in MixJ, this substitution is beni[iE m% ] [M M Q h The performance results for 20E and change g $M Januai%ploading, operated II hypothetically at 2 MW are presented in NlT$31of4tef..Y.4h" MMMW5 wmm 4 4.5.5.4 Reference Core Considerati pff W y? Three con 6gurations of the central regi'on w

                                                   #9 computed /lThe central core irradiation has a boundary defined by a 10.1$2m 4") @l tujepith        e    a 0.254 cm (0.100") wall. The reference loading calculati.on{hh an a(lum $ ipa @isphite (Al/C) plug f slice thro      g                              QRiT4'.31. The plug has two components, a central aluminuTN h}gpluggedTMtp;is shown of these pl gd be repla[cidgan experiment but no specific c

calculated. A seconildinfigurationkwks calculated for the reference cores is with the plug removed and 1%$Mhfilleditti water. The results for peaking factors are shown in Table 4-17.} The finYcalduliitidconfiguration has the plug and tube replaced by a 10.16 cm (4") diamejefcolumn of siliEA%nited by aluminum rails that occupy 10% of the remaining water gapgolume in Row B. This; approximates the anticipated configuration for silicon doping irtadiations in the centerjiore irradiation facility. These results were stated earlier for the MixJ add 20E referenc: cotes: Mg f r%a The neuggn%mct parameters used are the efTective delayed neutron fraction, p,e, and the c'ffe@ye prompt neutron lifetime. The p,e is essentially 0.007 for all the $.ariations in fuel type and 15iidirig arrangemant. p,e was used in this chapter to convert from calculated Ak/k to

• In the scoping studv of the previous subsections, the diameter of the aluminum rod and

he ID of the graphite anm as were 1.78 cm smaller.

1 l 4-81 Rev. A 7/5/96 7 I t dollars. The prompt neutron lifetime varies from 53 ps for aa all 8.5/20 loading to 32 ps for an all-20/20 loading. This parameter is used only in Chapter 13. 4.5.5.5 Power Peaking Factor Analysis g ik N The thermal-hydraulic calculations of peak steady-state temp,eraturesjM54.6 and the accident calculations of peak fuel temperatures in Chapteyljrequirrpgeaking factors as input. The required peaking factors Since the predicted temperatures determine safety limi a reaso were obtained fromiead l(amo0Et was built into the peaking factors. A conservative co ijion thay[ runs throug'kg mg nics calculations er factors is that fresh (unitradiated) fuel was used in all conservative conditions are noted below.  % As many as four factors contribute to the total power ing ' Jch is the ratio of the e fuel, r"ated factors are highest power density to the average power de displayed in Table 4-17 for several variations ference gs. The core radial d powers. It depends factor is the ratio of the peak to average v_ hinte experiment facility and strongly on the core loading but weakl e tsg g the control rod bank position. The ax' i ctor di g_ trongly on the control rod bank position but weakly on the c . ding the con $dits at the center. Based on these ( lyses are the 20E loading with

  'v   considerations, the conditions ch           rth mperatur .

control rods down 1/3 of their trfiEl (A ain 't$e excess reactivity is that the rods will not have to be, inserted modth$il1/3 of t uring operation.) For the'fidM.h kkh N factors frodig$, steady: state;aemperature prediction, the yere usI % $

                                                 $ilt condition is used. To be conservative, the last two has          slightly d

peaking factg3 $in"iddip,secti higMMidking factNt%(h total pe 6f Table 4-17 were neglected, because those factors tend to reduce the peak st$dy.statertemperature.

            - 4F WMP Wau Th       rd and fourth facedrip6 count for the power distribution across the cross sectiona' area
                                                                                                ~

o fuel element. ThEpower density increases with radial distance from the center of the nt because the p$ulation of neutrons that were thermalized in the coolant water s with radiidThe peaking factor that accounts for this (pin radial) was computed with

                   . _ _ dEio code for an infinite lattice of 20/20 fuel cells. The fact that the fue!

onuniform core, rather than in an uniform infinite lattice, alters the power dis across the element. The peaking factor that captures this effect (pin tilt) was computed for the highest power fuel element using the RCT code (Ref. 4.44), which reconstructs the flux within the homogenized node using the DIF3D nodal diffusion theory p flux solution. NI

4-82 Rev. A 7/5/%

                                         .je,-bb 3-                                         .       .

9,. k N Plug ,

                                '                                                                                                                                }

y ar i Graphite y pgug s .s AlTubes M

                                                                                              @eriment Tube;         OD 4.00' ID 3.80*
     ., OD 3.34' 10 2.334'                     OD 2.00                                        Outer Clad for Graphite; OD 3.64' 10 3.44' Inner Clad for Graphite; 0D2.234' ID 2.034' l

f I \ 9 CROSS SECrlON OF CENTRAL EXPERIMENT FACILITY WITH Al/C PLUG FIG 4.31

4-83 Rev. A 7/5/96 t s i d -

   ' _./
               'Ile peak-power element is in Row C and its pin tilt peaking factor is influenced primarily by thermalized neutrons coming from the central experiment facility. The water is more effective at thermalizing neutrons than is the Al/C plug, making the pin tilt factor larger with the water hole. A much stronger influence from water than Al/C was observed in node-aye! rage detector
                                                                                                         ?

fission rates in both the MixJ and 20E loadings but it is not apparent in'the 20E sin tilt factors. The difference is less pronounced in the 20E pin tilt factors because oEdiffe$$ces in the local geometry; only two of the six hex faces are exposed to modI$tator in OENNJ peak element (C06), whereas four of the six faces are exposed to eithe(Ufe'centralg[ikMjacent graphi[ dummy elements in the 20E peak element (C07). h,, f7' YEM._, y, %w :- ==n.

                                                                                                                    ,7 r

Jih J/ MMinF For the pulse accident analysis, the four factors from CisMaivere used. (i.e., tHectal aking factor is 4.86). The third and fourth factors were incluWdflEEbeer.ise the power peaki tg at the outer radius of the fuel element raises the peak fu 8 wm NI a ,

             -  assumption of the analysis. Case 7 condt rst-case  represents ons,             the wo,el., th[ir A                  $$}Y TABjdpish                      "

POWER PEAKINGiFAC'IORS. a === Case Conditions 1 2 73 94% jf5/ 6 7

         ')       Loading                     Mixj       Mixj7  Mixjf     20E/ 30E                  ,         20E Center Contents             Al/C    [il       Hid       Al/dE    Al/C         1O            H2O Rod Elevation                 uphs [d1/3     ((uph    ~luh       d1/3          up           d 1/3 Peaking Factok            $$ (               EW g ,      2             -
                                      'V      1.37 % J1:.~36h    1.35     1.76      1,68        1.t>8          1.60 core radial V axial          , 25     g  $h25      '132N 1.27     1.23      1.33        1.25           1,35
                                      ,'                                                        1.33           1.33 pin radialj '             '"3       m
                                                       ~

1.33 1.~ 3 pinii!E 1.63 1.69 1.68 II.32( 1.50 1.64 Gdtiibined Factors  ;: 1.71 1.80 1.71 2.16 2.23 2.10 2.16 hsEtwo f M .- , third and fourth 7;s 2.18 2.17 2.25 2.23 v.u: y all fourF v" 4.72 4.84 4.72 4.86 a

Rev. A 7/5/96 4-84 4.5.5.6 Neutron Flux Analysis The rates of production of the radioactive gases Ar-41 and N-16 in the core are computed in Appendix A. These computations require values of core-average fluxes as input. The thermal flux is needed for the Ar-41 calculation and the flux above 0.6 eV is needed for tfie N-16 calcula*. ion. These fluxes were obtained from steady-state neutronics c&lculatidsI

                                                                                                   .$5hW The core-average fluxes over energy ranges of interest ar                                   en in Tib        8 for five different these data that        core the thermal    loadings.

flux levelThe dependsthermal fluxprogions strongly Mthe boundary ch8!5 usedWre d2 is 1.le fuel types in the core. This is because the erbium anditA highpcentratiohbfi6(iin,hj,fh the 20/20 fuel deprest 'he thermal flux. These results i dNY6tfesh fuel and th2ifIdiNN flux depression would decrease as the erbium and uranium kdEph. The flux above 0.6 eV has a relatively weak dependence on the fuel mix. V ~? . gN jyjfgghyp4 Future fuel loadings will begin at about 50% 20/30fuckand wilfNy-being entirely 20/20 content, onservative to use fuel. Since the flux values decrease with incr i values at the all-8,5/20 fuel limit. f)dl(t$i Accord,ifealiW20g}origended ues to use in Appendix A are 2.0x10" thermal flux for Ar-41 pr6Juction a5dk2x1. .._ bove 0.6 eV for N-16 production. // M3}7 e I n . 4 TAB 4-18 A EDhAVE . ElfiRON FLUXES. Q[jQFiie, led Elen'ie:F Core-Average Flux (10 3 n/cm:s) [ k i1 Il[Ing[Sh 83/SIh , 20/20 'Riermal Above .6eV Total

   @lMW                 [N jdh,                     DY
                                                       ~
                      #     Y              IM89                            0             0.98          2.09             3.02 Core B f JanI95                                   .c 163                        30             0.75          1.96 . .,        2.66
   @[25tW                                4 MB                            S                 89                       0              1.97         4.18             6.04 30              1.50         3.92             5.33 Y1539 h -e M                                    63
   $5xE $ $ #                                     48                      48              1.38         3.70             4.99 MixJ                                           50                      49              1.36         3.71              4.97 4                      97             0.87          3.57              4.35 20E 0

4-85 Rev. A 7/5/% i 1 ~ L.) 4.5.4.7 Fission Product Release Fraction Considerable effort has been expended to measure and define the fission product release fractions for TRIGA LEU fuels. Data on this aspect of fuel performance are re;4ned in References 4.23,4.24 and 4.25 and evaluated in Reference 4.1. 6 # Mh47 Using these data, GA developed a conservative correlation'f3n fissichkN case, lsON$1- l ReleaseFraction = 1.5 x 10%lIY3600eL I [ MNIkI d i' W ijf! M hk Y where T = fuel temperature in degrees Kelvin. In cha?" the conservatism, it is stated on page 35 of Reference (Ref. 4.24), "At noriaal TRigA' ' temperaisres (<750'C), ~) there is a safety factor of approximately four betweelFpredi the above equation and experimentally deduced values." The same obastmation is tsgg Ref.' 45. This correlation was adopted to predict the rel es andipspf-volatile halogen fission products to the fuel-clad gap. -- [ It is generally accepted that the solid ion pr ts(thdb6hkhllow volatility, such as Cs er rat N and Sr) are released at significantg (* The appropriate temperature to use in the A9 trel " is the fuel temperature averaged over the irradiation history. Th(Ifuebn be c . having two separate temperature histories; , rage tQ",the fueQ$eed during its steady state irradiation and the tem fuel rnayL .icnce during the accident that is presumed to lead to a cladding ru ~ nduceY '9the fuel temperatures must equal or exceed the safety limits. The fining hriate fuel temperature, and thus the release fraction, is given in Re Ch. 5,'apfollows: AI ~ r$74 yy*The release frah!$ accident conditions is characteristic of the normal pf operating temperaturp7not the temperature during accident conditions?This is because the fissi$iIt products released as a result of a fuel clad failure are those that have bee 511ected in the fuel-clad gap during normal operation." res used to compute the release fraction are those predicted for the worst-

                                                                  " fating conditions. Of all the projected normal operating conditions, the hi          inent power occurs at position C7 of the 20E (all-20/20 fuel reference) loading when the control rods are banked 1/3 down from the full up position. This element's power was used in the thermal-hydraulic model of the 20E loading (see Section 4.6) to predict the radial and axial Let temperature distribution in the hottest channel.

v

Rev. A 7/5/96 4-86 The gaseous Ossion product release fraction was determined by integrating over this hottest element's fuel. The release fraction at a given fual node was found by evaluating the correlation function at the node-average fuel temperature. Weighting by the volume fraction of the node and summing over all fuel nodes yielded 7.7x10-8, the fraction of gaseous fission products released to the fuel-clad gap. This is the value recommended for use ina$e acci analyses. g

                                                                ./p
                                                                    .s t

h&TW[ 4.6 Thermal and Hydraulic Desien ik erhormed M]fgYN focoperaNoe$A

cfhMN at The a nominalthermal 2 MW and and hydraulic maximum required 2.3 MW power using analysis has been p%REL,Ap5/

(Ref. 4.46). The RELAPS code was developed for thsU$NR'C'by the Idaho Nhiihiiil Engineering laboratory (IN'L) to analyze transients ariiMEnt< in light water reactors. The ret AP5 code is highly generic and can be used toIQ' ^hide varietj~of hydraulic

                                                                                                          ~

• and thermal transients involving almost any user defirEd nucfiR[djs&Wg y by the N'RNnd a consortium of The MOD 3 version of RELAPS has been de i several countries and domestic organizatio, ta M(of tgYnternational Code

! Assessment g .T[re!qgD3 development progf.un and Applications Program (de res)ulttof assesteentsagainst small-brea included many improvements based on and operational transient test data.g7 gV jf[ A RELAP5 model consists of aqstem of ol volu;nes connected by now junctions The Guid mass, omentum, and y equa #Nfith the appropriate equation of sate are solved er defiggIdyt%try. Th$@P5/ MOD 3 code uses a full non-quilibriun equation, two-fluid model for transient simulation of two-homoge ,g t structures are used to simulate the reactor fuel rods, phase system . User' , lieat transfer are co appropriate for the channel flow and Gufd state. A space indepyndk%[gtics'n%1 is available for reactivity transients. 8 WSh Somepfue RELAP5/M D31fistures important for simulating a natural circulation reactor likgNRC m include: b[7 p{s$momenturnigf ability to ationcomp [6Te the system density distribution Ebd my compute implicitly the local pool or convective sub-cooled boiling, which is ideo,wn to occur in TRIGA reactors

• a new critical heat flux correlation for rod bundles based upon an extensive tabular set of experimental data
• temperature dependent material properties

4-87 Rev. A 7/5/96 / i V

                                                                                          ~

• special cross flow models that allow simulation of the two dimensional Dow due to radial power differences in the core While no references to application of the RELAPS code fer analysis of a TRIGAdeactor could be found, analyses of many different systems have been reported in the(open IMfature. Many of the system transients analyzed were at low pressure and with naturAIRcdation

                                                                                     '           flow. The RELAP5 code selects the heat transfer correlation to be udbas:d kioji thifgall temperature and local Dow and Guid state. The critical heat flux correlation also uliciflodal' conditions ang implicitly accounts for axial power distribution. The critical heat fliffcolisliidiid' corrected for potential errors if the correlation is enterMUwith CoPa are not in the dominant regions of the data base. The %fEode can thus beidsed:for     "

analysis of the MNRC thermal and hydraulic performaE6EQ

                                                                   %                    .-c
                                                                                                               .g' -
      , 4.6.1. Thermal and Ilydraulic Analysis                  [M5N%j@;, pix              "

N?Mik As poker in the MNRC core is increased, nucleaQon ns to okkOfe fuel rod surfaces and fully developed nucleate boiling occurs. Af @surfhde, heat flurt,emains , below th cat linha eg;ggfpreciable heat flux (CilF) it is possible rod surface temperature. If the CHF isfeee to increase ti6R ecchrs and the surfaceincrease in fue t@dedJi1 temperature increases almost immedialeth to a rud6h highj95f86s and fuel rod damage will i p) occur. The safe operation or the reictIr is defEdent upon the operating heat flux in relation V to the critical heat flux. to the peak core heat flux is a measure of the safety margin. YThe ratidaM5e %3 #n critsi$l heat fl[r A ~ mAE= The RELAP5'model uspiri MNRC anyali shown in Fig. 4.32. The model speci0es pipe, grid plate br(icti$iiifsiggle ahdhupper water vol@3uirthdbThese components are connected required. lieMFrict0rcs are dsgiifdsbimulate the fuel in the average core and hot channel, pTssumed to be connected only to rods with the hottest The hotpoyfei#PIR@nts fuel rod fluid chithedR$onservad@E ar divided into a user specified number

                                   ~

core regidh where the hxial'diidribution is important, pipe components with 9 axially distribsted volumes wcE6i$!Nor the average and hot channel regions. Branch components conlaih a single volume $itlfa user specified number of junctions connecting to other i c " nents. BranchesTere used to model the unfueled rod regions directly above and below _ ive core. SingI[ volume or pipe components and single junction components were used

                                 ~

alance of_ die system.

                       ~w loop, including the N-16 diffuser, was also modeled. Time dependent junctions were use to model the Dow from the upper reactor tank and the return flow to the diffuser and lower tank. The diffuser How was assumed to be 20% of the total primary flow. A time dependent volume was used to reference the entire model to atmospheric pressure.

O) ( The net driving force for flow within the MNRC tank is the difference between the net buoyancy of the water heated in the core and the friction within the flow paths. Both are

Rev. A 7/5/96 4 88 computed implicitly by the RELAP5 code. The friction losses consist mainly of the wall friction within the fuel pin now channels and form losses in the upper and lower grid regions. Friction in other flow paths are computed but are small due to the low velocities. The wall l friction is computed directly within RELAP5. The form loss coefficients for the upper and lower grid regions are supplied as input to the code and were computed from dapta' presented! handbooks for similar geometries. The calculated loss coefficients are signi6cantry larger than J those usc.i by General Atomics (Ref. 4.47) in their analyses. The cou$futedSil ms were used AratA"caIEulation was for the using performed reactor thermal the General Atomics and hydraulic loss coefficients analyses toh8Ide a in this sectio the fission gas transport calculations in Appendix A of tius addendunn 9.MESBN f!$ The steady state fuel temperature depends strongly uppthe,therrnal [ khNN#

• resistanceythe%f cladding interface. The resistance was assumed to be zerog

                                                             % iEQ d                . . ~                  N The buoyancy of the water in the core hot channel canNIntlucii6ed by the cross' flow between

' the hqt and average channels. Traditionally the hot a'n3 averaRchanhels have been assumed

                                                                     ~

to be completely separate (no cross flow) because;of3heqery narr6EMng between the fuel rods. The RELAPS code provides a means for"i:Isiliinstmgsthe effedti'o~ fiross flow between the hot and average flow channels. The crossllMEtidEU&ef@ected tofe very small, and it is impossible to assess the accuracy of comgFed crossffilSEQ1$[ calculations with RELAP5 showed cross flow to have no'effect on f6el terhperiture and to slightly increase the critical heat flux ratio. Thus, cross fi6s is consIrvativelp$egiected in this analysis, sfrp a AV P d f jf 4.6.2 Steady State Results @&

               +

A w w $m$% s wmwd The RELpPimodel desciitMI ove waspedwevaluate the thermal and hydraulic performance!df7the MNitG'ddinriksteady stiiliperation. The power distribution in the model correspod@6%Irst case Ediiditioditas described in Section 4.5.5.5. This is referred to as kiiEIthe control rods lowered 1/3 of their travel from the Case 5, the 20ETeference loadii@N i full up positionMiingpowe 4.33, was us65."The:axialjeaking% factor was 1.33. The corerom the 3-D was assumed neutronics to have 101 fuel calcu elements,idith the hot'fnro{op,erating at 34.8kW for a radial peaking factor of 1.68. The total ppking factor (axial?ggliil) was 2/23 which is higher than the 2.0 assumed in prior SARjtnalyses. The radkij power distribution in the fuel was conservatively assu'med to be unifsrm. The temperatGre dependent fuel thermal properties were obtained fr'o~m Reference 4%59 Two ca!culationswere performed. The power level and core inlet temperature was 1ed to be at theJafety limits of 2.3 MW and 35'C respectively for the first case and at

   @Moperatieg' nc                        conditions with power of 2.0 MW and 32.2 C inlet tem s6coiiTeasci,W
       %gge 9

l I 4-89 Rev. A 7/5/96 U

                                                                                                          ~

RELAP5Comoonents k' 102 tank outside core barrel (11 volumes) 105 non fuelod rod region 124 -_-_l 106 average core region (9 volumes) 107 hot channel reglon(9 volum6s) g 108 upper non fusted rods 125 109 upper tank over diffuser (3 volumes) l 110 connecting junctbn 3 120 N16 diffuser regMii-  %, l 121 connecting junction l 122 time dependent d;tfuser temperature i 123 diffuser injection] Unction I 124 time dependent pressure volume 125 connecting junction l 129 primary retumtemperature i 109 I O l ( l I i 121 l 122 ] Primary 1.oop p 123 l 110 1 1M l t l

                                                                                                                                 *+-
                             .,L       n     _
           ,      102     '

p l y fuel 5 106 105 129 h 128 MNRC RELAP5 MODEL l FIG. 4.32

t Rev. A 7/5/% 4-90 1 l l l I j .4 s u i g u a u i 4 ' ' ' ' I 'I ' ' ' ' 3 ' ' ' ' 3 ' ' ' ' 3 ' ' ' T"' y

                                                                                                                                                                                                           #.              . . _~

1.2* - 3 p . h 0.8 * - 0.0*

                                                                                                  . , , , i . . . . i . . . t . . >             ae       a t . i a i ei a > a ia                         i =

0 5 10 15 20 25 30 35 40 99V8 tion. Cm

                                                                             +

e . -. MNRC POWER DISTRIBUTION 0 FIG. 4.33

4-91 Rev. A '7/5/96 m

             \

The steady state results are presented in Table 4-19. TABLE 4-19. HEAT TRANSFER AND HYDRAULIC PARAMETERS FO E ON AT 2.0MW WITH 101 FUEL- ELEMENTS.- At Limiting inlet Tengiesture *#NoininalInlet Temperature (35'C) and 2.32Y/ gi@M),and 2.0ME? Diameter of Fuel Element 3.75 edQ ,y NsM.tmeg7  ; length of Fuel Element 38.1 crialk$2L7 38dhE Flow Area 546 cm 'j$h 2 546 cm2

                                                                                                            - ~

flydraulic Diameter 1.86 cdf  %{$$ h .86 cm

                                                                                                % [M)
                                                                                                                 ~
                                                                                                               "4.53 m2 Heat Transfer Surface Area                            $5( &

Inlet Coolant Temperature _ jfj;j3MM5h g 32.2 *C Exit Coolant Temperature [ 10d' CME %sf/ 103 'C Upper Pool Temperature ,g gis C WW 57'C O) i

      .\ j Coolant Mass Flow                  /y           f*/.7 kg/secg -                            6.7 kg/sec Avg Fuel Temperature             ,Y          /j373'C (hothin)                          341'C (hot pin)          0
                                                                /;213$@gepin)                            254*C (average pin)
                                .A.             efh  !.

Maximum [C6dSurface1(apiiissiin, M9N1"46*C 144 *C Maximuni'PiddjTpalperatureMfd A 705 *C 630 *C Avg Ileat Flux %QQ yijf 50.8 w/cm2 44.2 w/cm2 Max Heat FhixWQQQ 7 Il3w/cm2 98 w/cm2 Hot C,hsh5el Outlet Void @?@ 4.0 % , 2.0 % Cudutlet subcooling ))Y 8*C "-l1*C

                     )41ainium CIIF Ratio //                                 3.13                           ~ 3.ti8 AV p

min %gr$7imumYritical heat flux ratio of 3.13 is much higher than the values calculat

                                   ~
                ' sa li0ilys'es because of the very conservative correlations used in the past. The current value indicates that a significant margin exists between the proposed operating power (2 MW) and the power that would result in exceeding the critical heat flux. The magnitude of the
        . ..       critical heat flux is dependent upon local fluid conditions as well as channel inlet conditions and power. The change in magnitude as power increres is, thus, not linear with power and L V/7 l

the critical heat flux correlation cannot be used directly to determine the CHFR. It is not L ..

Rev. A 7/5/96 4-92 practical to perform numerous RELAPS analyses to determine the power level at which the CHFR would exactly equal 1.0. An alternative approach was chosen in which a calculation was performed to show that film boiling will not occur at a power significantly above the new operating power. A calculation at 3.0 MW, resulted in a CHFR of 2.0 and a maximum ^ fuel temperature of 870'C. This calculation demonstrates that even when operatingjt power which results in the steady state fuel temperature above 750*C, some CHFRhargin-exi'sts. Operation at 3.0MW, is clearly not acceptable even though film boilligModfi>redicted to occur. The predicted outlet fluid temperature ati3.0 MW Mtat sat 0rit$ddad a void fraction of approximately 15 % occt'rred. This is expected to rejdilYiii fluid cNNdI2Ehgging", wh' is known to cause power fluctuations in TRIGA reactoipf g *$$4 r (d m ef WMhrY vgwnry All other reactor parameters in Table 4-19 are also acce,ptable@*The predicted f0Sli6t is well below 750"C. The calculated coolant temperatur5)ifydinid distributions in the hot

                                                                                                                                                          ~

channel for both the nominal and limiting cases are shov/N%Eidf4434. TheTriuid subcooling

 . and low channel voids are expected to result in condeiIation'Ohd$kimrite$                                                         li     ately after detacliing from the fuel rod surface, thus the process;of>* heat terEi@byldatural convection is assured and fuel temperatures will be well belon7thGifety    "                     limit. Chuffing and the resultant 81f chugging were to power       fluctuations are, therefore, not expect $130 N$u'r!htSft4.48)hnd the occur, it would be detectable below thepYei of an[sM{jyg reduced to eliminate it; experiments conducted bfGA demonstrated that there was no fuel damage from deliberately induced cfnyging w1$se assocj$jed power fluctuations were readily observed (Refs. 4.49 and 4.50). g             37                                jf d       M                                 At 4.7 Onerating             Limits Xi s hiMY gy%[d W W" MZ%           ,w@rm               6-4.7.1 em%

Op"erating' m M g$h T k

                  @"vcMP  arameters5@nmme The maitt safetymosid'etation is to mamtain the fuel temperatures below the value that would result in fuel;%NDErfuel teY$hfature is controlled by setting limits on other operating parametedd.e., linlitigshfStysystem s   settings). The operating parameters established for the MNRQ[ reactor are:

Wsy -

         $                    WW L;j: a. Maximum lignsed steady-state power level Eh                    lY f$cMb. Fuel tempture measured by thermocouple fs$W4'wan w unccMax:wimum
         %B;dr reactinty worth of transient rod

d. Core inlet water coolant water temperature O !

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I l Rev. A 7/5/96 4-94 4.7.2 Limiting Safety System Settings The limiting safety system settings given in Table 4-20 are defined to assure that the safety limits in the design bases will not be exceeded for normal and abnormal operations. f? TABLE 4-20. 4  ;[ LIMITING SAFETY SYSTEM,5EITING h , Parameter Limited Ssfety Setting 9%), @N$ wm s, unction sf// m a Maximum power level at steady-state Q2 hjf 'R)Sctdr[50im Maximum measured fuel temperature Rea'Elfr' Scram

                                                                                %d.30.iC..

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• Maximum worth of transient inserfiIn of $1 $NhN  %

y 44 em*'/

• Reactor tank inlet water t yture less than V7&40o b k@p.

The $1.75 reactivity insertion liinifis justified by the difalysis in Section 13.2.2, which shows that there will be no damagdifIn'sertions Odie'islinad $2.12. These safety settings are conserva,ti(iiltth.ec, au - sense'th.h v - SIhey are

                                                                       -    adisd. nthe consequence of normal or abnormal operation;would.be    w%

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7/04/96 CHAPTER 5 i l l 1 1 REACTOR COOLANT SYSTEMS,U- m l I I m M 4 1

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Rev. A 7/5/96 7m f \ i  ! TABLE OF CONTENTS .

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REACTOR COOLANT SYSTEMS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-1 5.0

                                                                                                                                 ..... 5-1 5.1 Summary Description . . . . . . . . . . . . . . . . . . . . . . . . . . . M . . .                            5-1 5.2 5.3      Reactor Tank . . . . . . . . . . . . . . . . . . f. . . . . .5-3. . . . .[.

5.4 6 & @ g ....... 5-5 Secondary Coolant System . . . . . .MI . . . . ,, . . 57bOMA . . . . . . 5 5.5 Primary Coolant Cleanup System . . . . Y ). . . . 5.6 Primary Coolant Makeup Water Systemi) Nitrogen-16 Control System . . . . . . i. . . . . .p/. . . .

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Rev. A 7/5/96 l_lST OF FIGURES 5.1 Reactor Ta nk . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 5-2 5.2 Reactor Primary Cooling System . . . . . . . . . . . . . . . . . . . . . . . . .. . . .. 5-4 5.3 ... . . . . .. 5-6 5.4 Reactor Reactor Water Secondary Purification andCooling Makeup System System . . . . . .. . .. . . . . . ........ . . . . . . . .j.5-7

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5-1 Rev. A 7/5/% q - i'~') 5.0 REACTOR COOLANT SYSTEMS - 5.1 Summary Descriotion 5.2 Reactor Tank g

                                                                                           'A The reactor core is positioned near the bottom of an open alumi                     ~ W/2 ft in diameter by 24-1/2 ft high, Fig. 5.1. The tank contains approximatalp .000%1' s f high-purity water so the core is clearly visible from the top. About 2DYof wateriovs(tRiop of the                       .

FdIthe reac is reactor imbedded in core a massive provides concrete structurebiological which providshiriding shieldi for personne)EJ YET in biolo surrounding areas.

                                                                                                                         ~4 Pipe assemblies welded to both the inside and outside o condnuous), slightly above the reactor core, form ondit 6 m

(n the tank wall is tubessFlanges have been welded to the pipe stubs on the inside of the . gndare h the in-tank section of the beam tube, see Section 9.2. Cle . Acen pro tween the pipe stubs outside the tank and the reactor bulk shiel ' ctural ising of the tank wall from thermal expansion. An aluminum e use I storage racks, underwater lights, and other equipmc locat oun top. The exterior surface of the tank is d wi k poxy af r-saturated roofing felt to prevent corrosion. The felt is applied i ouble _ _ ness using a bituminous material. In addition, a corrugatedjiner, approxi 1 in. inL L f*klocated between the tank exterior and the concreWahield. e The ,ed liner; , y *a path for water to drain to a collection point undjoif k sh overflow or leak. A drain around the base of the tank is designed te wat . .. corrugated section. The drain is installed so that it can be routinely { r evi kage. y nne ass V v support for the control rod drives and the tank covers. It

                                         . jdes A   c_ enter,ibove the top is locat,ed                           stinctor tank directly over the reactor core. ThEassembly                    4 con , of structural ch9              tovered with plates.                                   .                       ;

p of the reacto nk is closed by aluminum grating covers that are hinged and installed ith the floor bucite plastic is attached to the bottom of each grating section to prevent

                              '%ntering the tank while still permitting visual observation.                                 .

i 1 s, welding procedures, and welder qualifications were in accordance with the l ASME code.- The integrity of tank weld joints have been verified by radiography, dye i penetrant checking, leak and hydrostatic testing. ( 10 t / v

l Rev. A 7/5/96 5-2 f(AC;C2 R0bt FLOOR - V4 E2 LEVE

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53 Rev. A 7/5/% p V

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5.3 Primary Coolant System The reactor core is cooled by natural circulation of the reactor tank water. The tank water temperature is maintained at approximately 110'F by the primary cooling system.

A> The primary cooling system, Fig. 5.2, is designed to continually remove 2 MVfibf heat from the reactor tank, it contains the necessary equipment and controls told,6uladSp to 1000' gpm of tank water and maintain the temperature of th'epater Jjpo the tank at about 90' (104*F). Instrumentation is provided to moniddlie systenlopehti6n, water /> temperatures, pressure, How, and tank level. Tank buyater outlefanf ~ ' rat are continuously recorded on the DAC. [k j7 WW

                                                                                            &g
                       - This system is operated and monitored from the reacto                     m. Theremote controls            .Q
                     . and monitoring instrumentation located are in the reac t
                                                                                                -i y           -

The system is regulated to maintain the primary er than the secondary system pressures. This pressure d stem pregg&S6fctivity from entering prevenFt the secondary system especially the coolin; . [d i op between the two systems. The differential pressure is estal# hed utetlon and ensured by starting

                                                                                                " simultaneously.

Q G' and stopping the primary and second $ stem a lat L With the exception of pressure, parargeters are ryad out in the reactor control room. Alarms are provided on the rea control ole,gw, tank bulk temperature, or tank water level e3ceeds preset Syste es have local readouts. There is also a tank leve tor on 11 that g om the reactor room. All syste ts that 41)e aluminum or 1. Tne%g, primary hange water is a plate-type with theare primary normally water flowing made within the pl,aggg p

                              ~f                "WMWh The entra'nce to the pu~~yiildddst line is less than 3 ft below the normal tank hiter level. In A add     15 the line is per ~ Yrom about 8 in. below the normal tank water level to the e      cc. Should a pri         system component fait downstream of the pump, the tank water would lower to th^e"first perforation, about 8 in. At this pdnt the pump should lose and quit pg'hi'ng. However, in no case can the pump lower the water level beyond pump snction line, less than 3 ft. Even if the water level lowers to the mp sur son line there will be approximately 16-1/2 ft of water above fuel ele"moti0$Y2 e core. Tnis feature prevents the loss of a signincant amount of tank water should a leak develop in any of the primary system components when the pump is operating.

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5.4 Secondary coolant System The secondary cooling system is capable of continually removing 2 MW of heat from the primary system during normal weather conditions. The system circulates approgately 1000' gpm of water from a cooling tower through the primary-to-secondary pt exch, anger and back to the cooling tower, Fig. 5.3. The pressure of the seconday systemnshf a;dtiined higher than the primary system to prevent cross contamination ognd uld a leak develop in the heat exchanger. To ensure this differentialpressure, d second y I" # water system pumps are started and stopped simultane Iy. Water chemistry, conductivity, and pH, is monitored ?maisiined by an autQnater conditioning system that adds chemicals es required. f

                                                                                                - -gj.                 .d 5.5 -Rea.ctor Water Purification systems                    []          _

4 and optical clarity, The reactor water purification systems maintal Fig. 5.4. There are two separate systems t _ mu p water'M [siently inde ,or can be cross-connected to operate as one unit. One s "inisut 9 TC l 7 ate matter from the O surface of the reactor tank and the othep'ystem nizedg% ter to maintain the purity. V en h The filtration system uses a drun1Mee skimmer that fl6 hts near the surface of the water in the reactor tank. A pump movf7ater fro [* e surfa& skimmer to fiber cartridge filter elements, se filter ele ' mmove - bris from the reactor tank water by mechanical ring i iwater [ing n to the reacte- tank. The system be used " e filt e reactor tank or, through a series of valve gAirectly manipula _ __ _ Jend tl ter through the deionizers and then back to the reactor tank. The sy'%fgped to su ionizing resin bed during extended shutdown period when the pri stem ' _ operational. A set _ <lonizing e res _(four) are supplied from the primary cooling systhm (outlet of the- exchanger) at a 1 flow rate of fifteen gallons per minute (15 gpm). The resin nsists of four fibyrglass, throwaway, canisters of mixed-bed resin. Two of the canisters armally on-line asd the other two canisters are in a stand-by condition. A g7

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Rev. A 7/5/96 58 Two conductivity cells are used to measure the conductivity of water entering tne resin beds and the conductivity of the water exiting the resin beds subsequent to entering the reactor tank. There are local readouts of the conductivity near the resin tanks and remote readouts and alarm functions located on the reactor console. Pressure gauges are located within the systems to monitor the overall performanpf the systems. p jf jmd 5.6 Primary Coolant hinkeun Water Syltem yf h- (jh#$wwA A 300 gallon plastic tank of demineralized water # upY Tis avalli c le to make,pw elving 11 or system water lost by evaporation or other means. A r tank g'with a positiv connection to a delivery truck. The makeup system i [' , [ inent ' pump and resin canister of the same type that are used 7@J\h 't eation systern:, The outlet flow of the niakeup system discharges 0 to the 'purl 0cationsymhq 5.7 N11togen-16 Control System N O !? W Yf R ~ % A diffuser has been incorporated into the sysd t'tdj'iducie nhe N-16It the tank top The diffuser discharge is located about 2 ft ab@'e the re# Mdfgabout 70 gpm across the top of the reactor (see Figuro%2). (fu gTes anytime the prir ary pump is running. (( / 5.8 Euct Storage Pit water svatem a A P :Vd%pp4'j ' The fuel s pit wate[sk. .n.is used I ing of stored fuel elements is required, level is coy ha0 jater supply valve. Each pit subsystem contains a pump and a three wayjyaijo;)q the p, ump; discharge line. This configuration allows for once-through, recirculation,,or; feed'.a5d: bleed einstf6n depending on fuel element shielding requirements. t Wheqo(>eniliig reactor tank. mMthS%'p@ro, ugh'or feed and-bleed modes, excess o

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e i , i i I , 4 Chapter 6 - Current Valid Pages  !

                                ~l   Rev.1 -/-/97 61 Rev.1 --/--/97 i

62 Rev.1 --/--/97 i 11 i .i i i l v , I l i l r 4 ! l i I a f 4 j

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TABLE OF CONTENTS  ; t 6.0 - ENGINEERED S AFETY FEATURES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-1  ;

6. I '- I nt roduction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 > l j 6.2 Emergency Core Cooling System (ECCS) . . . . . . . . . . . . . . . . . . . . , . . . . 6 1 l- l i

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                                                                                                                                                                                                )
                                                                                     - LIST OF FIGURES                                                                                    l     !
                            ' 6.1
                                           . Emergency Core Cooling' System (ECCS) . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . 6 2 - .I                                    l

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REFERENCES I j 6.1 Emergency Core Cooling System Requirements for the McClellan Nuclear Radiation l l Center 2 MW TRIGA@ Reactor, GEN-52, General Atomics, January 1997. ( { t l . i b P .. t 4 6 5 6 4 . ) i

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l 6-1 Rev.1 - /--/97 l l O  ! Q 6.0 ENGINEERED SAFETY FEATURES l 6.1 Introduction l During the design of the h1NRC and subsequent analysis for safety considerations for the MNRC, l the only requirement identified for an Engineered Safety Feature was for an Emergency Core l Cooling System (ECCS). This feature is required for operation of the MNRC at 2 MW. Previous l analysis has shown that an ECCS was not required for the MNRC since at 1 MW, even an instantaneous loss of the entire tank water would not have resulted in fuel temperatures which l would have threatened the fuel clad. 6.2 Emergency core cooling system mccs) l The emergency core cooling system (ECCS) supplies water to the reactor core in the event that l the reactor core was to become uncovered. The water is supplied from the domestic water l supply. l l Water is supplied to the reactor core by hooking up a hose between the domestic water supply l and the aluminum piping to the reactor core area. The domestic water su0 ply and hose are l located on the roof area outside the reactor room. All hose connections are quick-connectors and l require no tools to attach. The aluminum piping that goes to the reactor core area has a nozzJe l that is positioned approximately two feet above the reactor core so water will be dispersed over l Os the reactor core (Reference 6.1). The nozzle rests on a two foot high aluminum chimney that l surrounds the upper grid plate. l l The system is actuated by the reactor operator using the following indications that water is l leaking from the reactor tank. l l

1. Reactor tank low level alarm l

2. Primary systeniIow flow alarm l

3. Demineralizer low flow alarm l

4. Reactor room radiation area monitor alarm l S. Reactor room radiation criticality alarm and evacuation alarm l

6. Possible reactor room and stack continuous air monitor alarms l

7. Visual with the remote television camera l ,

l The ECCS contains pressure and flow gauges to verify sufficient wcter flow is maintained l (Chapter 13, Section 13.2.3.2.2) for the duration ofits use. l O )

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1 i l l l CHAPTER 7 1 j

                                                                                                                                       - -m . .              .Q INSTRUMENTATION AND CONTI[gL                                                                                            l d

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Rev. A. 7/5/96 l ( ) TABl.E OF CONTENTS 4 INSTRUMENTATION AND CONTROL ....... .......... M ... 7-1 7.0 7- 1 7.1 Introd uction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . !" . J . . . . .

                                                                                                               .........         73 7.1.1System 7.1.2    Design          Basis Design    . . . . .. .. . .. .. ... .. .. .. . . . . . . . [. . A   A. ........
                                                                                                                     . . .2*.7)  . . 5[

7.1.2.1 NM 1000 Safety and NeutrofMonitorips ..... 7.1.2.2 NPP-1000 Safety Channe T. . . . 45. .. 7-4 7.1.2.3 Data Acquisition Comput .. ....... . 7-6 7.1.2.4 Control System Computer A............ '... 7-6 7.1.2.5 Reactor Operating Controls . . . . . . . . . . . . . 7-7 i 7.1.2.6 Reactor and Facility Dis ............ 7-12 7.2 Reactor Protective System . . . . . . . . . . . . . . . Al  :: L '...... 7-15 ,-

     .          7.3 Rod Control System . . . . . . . . . .                             .....                ......... 7-19 7.3.1 Control Rods . . . . . . .
                                                                            ~         "
                                                                                         ,.....         . . . . /.' . . . . . 7-20 7.3.2 Control Rod Drive Asse                   ies .                            . . . . . . . . . . . . . 7-2 0 7.3.3 Transient Rod Drive                embly              ..                .............. 7-23 O                                             N _lST                  TABI                                                                  i Table 7-1        ical CRT i

Window l f y........................ 7-15 Tabic 7- ICRTi indo ay.......................7-18 Table 7- xCRT w ow Display . . . . . . . . . . . . . . . . . . . . . . . . . 7- 19 aa . LIST OF FIGURES

                        " ~ MM.g     i.

hNsf-h 72 Fig.J(1 Block Diagrg(Instrumentation and Control System . . . . . . . . . . . . . . Fi '2

              . Neutron Chyet Operating Ranges . . . . . . . . . . . . . . . . . . . . . . . . . . . 7-5                               ~

Fi .3 M N RC Reactor Control Console . . . . . . . , . . . . . . . . . . . . . . . . . . . . . 7-8 4 M ode Codr'ol Panel . . . . . . . . . . . . . . . . . . . . . . . . . . . . c . . . . . . . . 7-9 5 Rod Cdtr'ol Panel . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7- 10 l $fCRT Display of Reactor Operational Information . . . . . . . . . . . . 7-13 cator Panct . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7- 14 I* Fig. 18 Protective System Scram legic . . . . . . . . . . . . . . . . , , . . . . . . . . . . 7 16 Fig. 7.9 External Scram Inputs ....'.............................. 7 17 Fig. 7.10 Fuel Follower Control Rod Shown Withdrawn and Inserted , . . . . . . . . . . 7-21 Fig. 7.11 Rack-and-pinion Control Rod Drive (Typical) ................ . . 7-22 Fig. 7.12 Adjustable Fast Transient Rod Drive Assembly . . . . . . . . . . . . . . . . . . . 7-24 g h -1

                                                                                                     )

Rev. A. 7/5/96 REFERENCES 7.1 " Microprocessor Based Research Reactor instrumentation and Control S idm,' INS 24, GA Technologies, Inc., May,1986. .,

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f 7.2 'Qua ifk; Nation Tes.t Re%portd M-1000 eps4hTfiMonit Igita  ;

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() O - l x 7.0 INSTRUh1ENTAT10N AND CONTROL 7.1 Introduction 4 A ./  ! The instrumentation and control system (ICS), Figure 7.1 for the hiNRCiTRIGA reactor is a computer based system incorporating the use of a GA-deg,ji tped, m#ii7, NM 1000 *

      . microprocessor based neutron monitoring channel and a@PP-1000 aniif$                                        tron            >

monitoring channel. The Nht 1000 system provides a$fety chatff(per'  : scram), a wide range log percent power channel (beloOurce svel to full po _ l indication, and a multirange linear power channel (sotrky NPP-1000 system provides a second safety channel for rp - g full power), R 7:1. The (cercent power with scram). In the pulse mode of operation, the Data AcpTt ger (DAC) makes a gain change in the NPP 1000 safety channel to provide NVand N Mtion along with a peak

    . pulse power scram. The Nhi 1000 is essential                                                     been initiated.

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   . The Nhi 1000 digital neutron monitor sy                    was                                clear power industry                       ,

and is fully qualified for use in the dem .ing anc s ions of a nuclear power  ! generating plant, Ref. 7.2. The syst,e s basedi a s -designed, fission chamber and low-noise ultra-fast pulse amp , TheEPP-1 ety channel was designed to the same high performance criteria p e Nht- channe*. , i gr The controp stem logic pned in x Montrol System Computer (CSC) with a color grg., ..hplay. rmatioM Iie Nhi 1000, NPP-1000, and fuel temperatiif < is pr . isplayed by the CSC, each is direct wired to its own output displ ~ gsafe nnects directly to the protective system scram circuit. That is, signal - gaJn circ ot processed by the Data Acquisition computer or the control co _ Mear information goes directly from the detectors to either the Nht-1000 or.#EP'-1000 whM ssed. The processed signals connect directly to the scram circult% itches. Fuel ' re information goes directly to " action pack modules

• for a icram circuit switches. The ability of this configuration to meet

(' fication and then t stent of protection-system requirements for reliability, redundancy, and independence for type reactorsfiEs been accepted by the NRC. -

                              .8 na[ sill control rod movements, accounting for such thmgs as interlocks and

["l(ular operating modes, it processes and displays information on control rod posi ions, power level, fuel and water temperature, and pulse characteristics. The CSC performs rnany other functions, such as monitoring reactor usage and facility radiation instruments, and storing historical operating data for replay at a later time. A computer-based control system has many advantages over an analog system: speed, accuracy, reliability, the ability for self-calibration, improved diagnostics, graphie displays, and the logging of vital

     - information.

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l 7-3 Rev. A. 7/5/96 7.1.1 Design llasis 0 1 The instrumentation and control system for the MNRC reactor is designed to perf Mm the I following functions: 9 1

                                                                             '*sC',
                                                                                ~

• Provide the operator with information on the statusgthe rea" ility.
• Provide the means for insertion or withdrawal of control rods ~

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• Provide for tutomatic control of the reactor po OY level. sF I s to
• Provide for detecting overpower conditions and matjEdlly inserting t$

terminate the overpower condition. Y

• Provide for the storage of data for later retrieva A scram system is included as part of the instrumentat on and m@c scram .,-

 . system is designed to meet the single failure critpd{fpplied to                  tors a,nd is independent of the normal reactivity-control s,.

7.1.2 Instrumentation and Control Sys esigp/T l 7.1.2.1 NM 1000 Safety and n Mo ring Ch[nel The NM-1000 nuclear channel the mul ik etiontepability to provide safety 6.am) action as well as nchtron monit "r a wid ydge from a single detector. The functions are the folfdiiv$g: $t .E # 1. Nkhb Percen powhlbwith sc i%

2. Wide-r(afg'ejng7po Npf f 'er. gQ 3- POW OIbh. A 4.

          /y Mtilbrange             1inear,n[MW
                            \ powery
           /                 nww Fogthe MNRC ICS, thd NM*1000 system is designated to provide the wide-range log power fu6ction and the percep;jpower safety channel with scram (linear power level from 1 % to l
  %{20%. The wide range log power function is a digital version power system to cover the reactor power range from below source level to 150% power a6dpifeyIde:ayri6d signal. For the log power function, the chamber signal from startup (i fulsticounting) range through the Campbelling [ root mean square (RMS) signal processing]

range'8Mers in excess of 10-decades of power level. The self-contained microprocessor combines these signals and derives the power rate of change (period) through the full range of power. The microprocessor automatically tests the system to ensure that the upper 6 des are operable while the reactor is operating in the lower decades and vice versa when the reactor is at high power. The output signal from the microprocessor goes directly to the scram circuit switches and the direct reading bar graphs on the console.

Rev. A. 7/5/96 74 l For the multirange function, the NM 1000 uses the same signal source as for the log function. However, instead of the microprocessor converting the signal into a log function, it converts it into 10 linear power ranges. This feature provides for a more precise reading of 1,inear power level over the entire range of reactor power. The same self-checking fgatures gincluded as for the log function. The multirange function is auto ranging. The NM 1000 syr,tr , is contained in a National Electri [AMnufach lation (NEM enclosure located li. sne reactor room. The amplifier ly con _ ' ug in subassemblies for pulse preamplifier electronics, band filter 'RMS'8' conJitioning circuits, low voltage power supplies, det Itage powert ital diagnostics, and communication electronics. The pr mbly is made up ular plug in subassemblies for communication electronics lifier and processor), the microprocessor, a control / display module, low-volta ' , isolated 4 to 20 mA outputs, and isolated alarm outputs. Outputs are Cl IE as s IEEE'323-1974. M Communication between the amplifier and pr mblics ed palj shielded cables."The amplifier / microprocessor circui *s the I neepts Tri automatic on line self-diagnostics and calibration ve lon, cceptable circuit per-formance is automatically alarmed. Th stem y calibrated and checked (including the testing of trip levels)' " to o Ion. Wieckout data is recorded for future use. The accuracy of the c is is e Itoor er than 3% of full scale, and trip b3 V i settings are repeatable within 1 11- nput. - tector uses, dard 0. r ny fission chamber that nas provided The neutt reliable the p . i howev improved by additional shielding to provide a greater' isera noise construction of the chamber assembly allows the system to res Jow re_ wn level which is subject to being masked by noise. An illustration n clia rating ranges is shown in Fig. 7.2. 7.1.2. NPP 100'(f ' " i nnel Th P-1000 system ~. Vides the redundant percent power safety channel with scram. The # a fled signal from tifs channel goes directly to the direct wired % power indicator and the circuit switchgin the pulse mode of operation, the DAC makes a gain-change in the safety, channel to provide NV and NVT indication along with a peak pulse power 1000 system is an upgrade of GA systems which have been in use in TRIGA i _ w _ orld wide for many years. It employs state-of-the-art electronic components and concepts. The nuclear detector for the NPP-1000 is an uncompensated ionization chamber. NPP-1000 systems are utilized at the Sandia National Laboratory, the AFRR1 reactor at Bethesda, MD, the University of Texas, and at GA's facility at San Diego, CA. o (V)

Rev. A. 'll5/96 1000 MW

 '                                                                                                l 100 MW     ~                                                                               l l

10 MW B C A ' ' - -- -- - 100% 1M - -- ---7.. - i- - 1Fe 100 kW

                  -                             --               --                     1%

10 kW 10 -I% 1kW ._.

                                                      % POWER                        -

10 -2% 100 W CHANNELS 1kW - 10 -3% 10 W INTERLOCK 10 9 . IW - 10-4. 0.1 W

                                                                                     -  10N.

0.01 W SOURCE LEVEL 10 -7% 0.001 W - SOURCE INTERLOCK TRIP 10N. 0.0001 W A = W10E RANGE LOG CHANNEL ' B = WIDE RANGE LINEAR CHANNEL

          -                 C = MANUAL, AUTOMATIC, AND SQUAREWAVE MODES O

TYPICAL NEUTRON CHANNEL OPERATING RANGES FIG. 7.2 i

Rev. A. 7/5/96 76 ,Q ly The NPP 1000 is located in the reactor room in the DAC assembly cabinet. The cabinet that houses the NPP 1000 has a heat detection and a halon fire suppression system. If the heat detector is activatcd, a 'DAC HALON RELEASE

• warning appears on the reactor control room console. After a short time delay, the electric power to the cabinet is turneFoff. The time delay is long enough for the operator to confirm reactor shutdowrih [

dGW TN 7.1.2.3 Data Acquisition Computer jD y 4]? . As indicated in Fig. 7.1, the Data Acquisition - N .-l M 0Compu converts from analog.to-digital form or digital to-analy' essociated with and NPP-1000 as well as from numerous other instru operations. The processed information is then transmit priate, to the Control System Computer (CSC) the NM-1000 or the NPP-1 transfer between the DAC and CSC is by high speed data transmitter. N >7 ~~ '; 4 In theiulse mode of operation, the DAC m ge in tl P-1000' channel to

                                                                                     ^#

provide NV and NVT information along a am. The control and transient rod drive cop l 1 sig produ the CSC are processed by the DAC prior to being sent to the de k ( V) The DAC is located in the reaci room a -lhousagd the same enclosure as the NPP-1000. 7.1.2.4 ys ter/Pri The Contro" L :' mputeK ovides all of the logic functions needed to control the reactor and au[ _ safety 6 monitoring for (dWiddiroperating characteristics. Information "" 4 ' Jer is layed on CRT monitors for ease of comprehension. Essentialij f the hem logic contained in prevloi 5 TRIGA reactor control systemiris incorporated thsCSC. I JV ver, instead of usufg electronic circuits and electrical relay circuits, the lodic is mmed into thejdmputer. The availability of the computer allows great versatility and in ope ally related activities aside from the direct control of rod movements. ns are performed by the CSC, such as monitoring reactor usage, monitoring r _ ment, storing data, and logging operator identity. A rod-drop timing circuit and a c isplay, capable of time measurements in the 10 ms range, are provided within the CSC and CRT displays. The computer samples all operational data in the steady state mode every 30 s and stores this data. The memory can hold 9,000 such samples or 75 hrs of operational data, in the pulse (o mode, there is enough storage for 10 pulses with all parameters.

77 Rev. A. 7/5/96 O Operational data can be printed in the same format as displayed on the console CRTs. ' nils includes all real time and archival data. The CRT displays can be reproduced in graphic and print form only. hk /j9' The computer is located in the reactor control console, Fig.g.3. The cogl' room contains a halon fire suppression system activated by product of compustion detectors.'h M Ap s p%. p/ m .s ,a N 7.1.2.5 Reactor Operating Controls # huw n nd The MNRC reactor can be operated in four modes: manual automatic, square wave pulse. The operations are controlled from 3the mode cong;@l Fig. 7.4 an control panel, Fig. 7.5. g/ g%jg jg 4, The manual and automatic modes are steady-stateer6iic' lot conditi%Ohspuare wave and pulse r~ nodes are the conditions implied by thc($ists ad! require tifu'se of the pulse rod. b The manual and automaticodreactor 'Ypw4.,fiNbr tlIN contrhm,$7 E operation from source level to 100% power. These two m are us Tor marishlIdctor startup, change in power I level, and steady-state operation, g

                                                            /

A captive keyswitch, magn poj$r, loca the r6d' control panel, controls the current to

                                 }gnets. Tl             ich must be in the "ON" position for any the controlp transient t rod moveirnbetions.     /Atig                     lborrent has been removed, this switch must be "ithe mag "pk            Jhen back to the "ON" position for the magnet current to turned       t8th3Rty'keyswitch be restored.NThis

• REACTOR ON" lights to be illuminated throughout the

                 %2"A            %Y MNRC.             sem plPW% .

Manual rpd control is"acctmi ed through the use of pushbuttons on the rod control panel. The tofrow of pushbutidhE(plig); iriit net) is used to interrupt the curre if tidrod is above the dhivhlimit, it will fall back into the core and the magnet will augsatically drive toje down limit, where it will again contact the armature. - SHO V Theiniddle row of hshbuttons (up) and the bottom row (down) are used to position the Off6dscIMessing the pushbuttons causes the control rod to move in the direction indigedydni6 hocks prevent the movement of the rods in the up direction under the following conditions: O

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MNRC REACTOR CONTROL CONSOLE FIG. 7.3

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1. Scrams not reset.

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2. Source level below minimum count,

3. Two UP switches depressed at the same time.

4. hiode switch in the PULSE position.

5. Mode switch in the AUTOhiATIC position [servocontrolled y]. ro[d:(tIf

6. Square wave mode - switch aepressed or lighted.

jr y 9 N$nm is gy gg%. A y M. There is no interlock inhibiting the down direction of L control ryds except le servocontrolled rod (s) while in the automatic mode. g y ;l Nhn: Automatic power control can be obtained bytoswitching tilual operation automatic frbMh.l operation on the mode control panel. All the instrumen@fonisafofygand interlock circuitry V described above app'ics in the operation of this mode Howenifjljhegocontrolled rod (s) 1.1 Q . (are) controlled automatically to a power level andJiiilod"' signal. Nils re$ tor power level is coropared with the demand level set by the mode c'6 fir 61 panelfand used to bring the reactor power to the demand lege[oi~a pjeo feature is to maintain automatically the set pop I

                                                             " I,delagjo[ng          term power runs.Yhe The square wave mode allows the, jf          W uickly raised to a desired power level.

or powei broughttot be q$' criticality below one kW in the In manuala squarc mode, leavingwave operation, the transl56I rod partia the rdtlir is fg(llygddore. The desire the reacto ~ rator using; "ver deminilselecypt' located on the mode control panel. All is in o;5rsijoii7 The transient rod is ejected from the core of the s,s MAnslent'g instrtf ' ' ' by mear power level "the dema"rM:It%pushbutton located on the rod control panel. 1[hJs maintained in the automatic mode. Reactor ??h contdis ths p\ysmg,m\hif odfconsists of manually establishing criticality at a flux level below c14kW in the stsady3 tine mode. This is accomplished by the use of the control rods, leaving the transient rod eithe6fu,lly or partially inserted. The pulse mode selector switch locasd on the mode codtfoyinel is then depressed. The MODE SELECTOR switch auMhatically causes thd.DAC to make a gain change in the NPP 1000 safety cha'nnel to r@ tor and record pe$k flux (NV), energy release (NVT), and to provide a peak pulse power KraAThe pulse liinitiated by activating the FIRE pushbutton. Once a pulse has been Inhisedlnd it ifdstected by the DAC, the NM-1000 safety scram is bypassed. Pulsing can b'sinMijMm either the critical or subcritical reactor state. The rod control panel contains a manual scram switch and a switch to acknowledge warning information that appears on the CRTs. The mode control panel contains comrols for inst ument po ver, prestart checks, and reactor scram test. Controls for the MNRC entrance ga'es and the CCTV cameras for the gates and the reactor room are located on this panel.

Rev. A. 7/5/96 7 12 -(N 7.1.2.6 Reactor and Facility Display Equipment Reactor and facility operating and monitoring information is displayed on two colgr CRTs and a bar graph indicator located on the reactor control console. m 47 e The high resolution CRT, Fig. 7.6, displays important reactor oper mation. This CRT has a scram / warning window which indicates the calis7ef th ng when a scram occurs or a predetermined limit is reached. This$IIidow is but ch to red when under a scram / warning condition. An audble alarm h[i8und scram / warning condition exists. Tables 71 and 7 21 the eters that the window, one at a time, if more than one limit is reac ~ ' " st-in will be di . Once e, time of day, operating acknowledged and cleared, mode, and demanded power are also displayed. TheJa the next parameter will generated information, ap "ii: ton by the control system computer, is displayed on this CRT as .7 % 4

• tinear power
• 1eg power
• nnels .
• Percent Rod positionpower from (resolution of both safety,f in.)
• Fuel temperature J,

O

• Tank water temperature The seco T is used ly react i information.

A '* Thre, t matior, T available for reactor operator use: scram, warning, and status. Th avai lay for each of these three categories is shown in f[ Tables 7-1,7-2 spen be displaye,dfl goh kp e console keyboard is used to select the cat the scraufsetpoints w 'ed in the order in which the setpoints were exceeded, first-in. A(noted above, th meter to cause the scram is indicated in the scram / warning wi won the high res CRT. The scram indication will remain on the display until it i 4 cared, f/ arning category is selected, the parameters in Table 7-2, that have exceeded the ill be displayed. The display on the high resolution CRT, the order, and c tMame as for the scram category. He third category that may be selected is System Status. The parameters listed in Table 7 3, with the current reading, will be displayed. The bar graph indicator panel displays information important to reactor operations. The information displayed on this panel is shown in Fig. 7.7. The reactor power and period g( Qt

l l i / ,13 Rev. A 7/5/% l O l l l i I Tue Oc t 4 20:45:00 !??O

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i I I i 1 l 1 l  ; \ TYPICAL CRT DISPLAY OF REACTOR OPERATIONAL INFORMATION e; [ FIG. 7.6 . l l

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7-15 Rev. A. 7/5/96 information displayed on this panel is directly from the NM-1000 and NPP-1000 safety O channels, it is hard-wired and does not come through the control system computeJ. ll Included on the indicator panel is a single pen recorder for wide-range /linearpw :r.

                                                                 -,       g!0ift@

7.2 Reactor Protect.ive SyMcm - N'sausgL M 5h y N f wmW JD I The reactor protective system scram logic is shown in fG;. 7 setpoints of any of the parameter shown in Tt.ble 7-1 Ec4ehed. The externalNEiinputs are shown in Fig. 7.9. 7

                                                                   -4 p'/pr%5;le}

1?na, TABL Q DISPIXYYN'QM[j7IEA 7 ~. TYPICAL CRT SC g 1; 9Af4 gass .:#'

                                                               ,     I

1. Scram - Console Manual i 14. YpeM - NPP-1000 Power Hi l

2. Scram - Reactor Room M 15. AScram - NM-1000 Power Hi

3. Scre.m - Bay Rip Cord , 16 g7 Scram - NM-1000 Hi Voltage Lo

4. Scram - Fuel Temp 1 I; gyh l Scram - NPP-1000 Hi Voltage to

5. Scras (p WWJM18. Scram - Keyswitch off

6. ScS$$,- Fuel Te External . Mk Scram - Please LoS i n l -a mpy -:=.c-....

l

7. rnal #2Qy.,_,x Scram - Net Fault, Please Reboot 8.

9. SctmL(CSGQlS64 Scraii - TI-T 21. 22. Scram - Database Timeout Scram - NM-1000 Comm Fault Scram -%Ad[niS(>4 TirY

10. Scra3nf/DSCyjHsf3og Fa - 23. Scrata - NM-1000 Data Error Scram - DOM 32 Fault

11. Spam - CSC .imeout 24.

l , . . l 12. AScram - DAC Fault 25. Scram - A1016 #1 Fault 13.dflScram - DAC V/AicIx! g Timeout 26. Scram - A1016 #2 Fault l ll - n L@(

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i Rev. A. 7/5/96 7-18 f3 ) TABLE 7-2

        '(

TYPICAL CRT WARNING WINDOW DISPLAY  ;

1. Pulse Not Detected 22. Rx Room RAu ,

2. Demand Power Not Reached 23. DemineralEMIRAk

3. High IC-Net Comm Fault 24, ipmeEIIIM ArWA0 Mg'j/ [

4. Low IC-Net Comm Fault 25.

5. Power Too Hi to Pulse 26 taging$f Stagingf ea #f ~

, 6. Trans Rod Air Must Be Off 27 Area #4 RA Particulate -

7. Period too Short to Pulse 28

,            8.      Line Printer Not On Line                     29.               oole Gas

9. Rod Withdrawal Prohibit 30 Jne

10. Rx Tank Return Temp Hi 3. Ba .~

                                                                                             ~

31. Magnet Supply Voltage

{lrounded-Hi Side Stack ck h ,k9YGas"$c ~

12. Magnet Supply Voltage n Grounded - Low Side , on 13.- Primary System Flow 36. CAM Fault

14. Demin System Flow 37. tack CAM Fault

15. Secondary System Flow 38. Bay CAM Fault p/

        \    16. Demin Inlet Cond                                    Rx Rm CAM Alert

17. De utlet Co 5. . Stack CAM Alert Rx Rm CAM Alarm 18.

19. ater ter 12 h 41.

42. Stack CAM Alarm

20. Coo ater - 43. Fire in DAC

21. Coolin R ter le /

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7-19 Rev. A. 7/5/96 TABLE 7-3 O TYPICAL CRT STATUS WINDOW DISPLAY STATUS WINDOW ek l

                                                                           - yg.
                                                                      #1              000 mR/hr        l Primary System Flow              000,0 gpm              Stagi               ,

000.0 gpm Stagm Area #2 ~ ' mR/ia l Secondary System Flow Demin System Flow 00.0 gpm Stapijg Areaf EAM ' gRJ Ise pm Demin Inlet Condtyty 0.0 uMOllS Rx2ntPariptilate Demin Outlet Condtyty 0.0 uMOHS Rx$ge Gas M% cpm 00.0 C Rx R$1coint 0.0e+0 cpm Rx Tank Temp Hx Outlet Temp 00.0 C StacfFP# 0.0e+0 cpm 00.0 C SGIk Nob s, 7 .0.0e+0 cpm m Hx Inlet Temp ~ . Rod Drop Timer 0.00 sec q Reactor Room RAM 000 mR/hr .j3a;g[ Argon '0.0e+0 cpm Demineralizer RAM 000 mR/hr ,Bjt g late 0.0e +0 cpmEquipment Area RAM 000 mR/hr/y j' Bay Tdgon 0.0e+0 cpmOne Kilowatt Interlock Yes j Rod Withdrawal Prohibit No i g Jf 1 fjnitiate scram action in two ways. First, the The majo ested in . in-series the paramg!8, will i ' ~ fthe rod drive magnet current causing a scram if the p" hown 67ig@MSecond, the SCRAM signal is detected by the computer int [Ev and the coni , ates d^rMMgQscram signal that opens the four parallel relays p8, K6, K7) interrupting rod magnet current. These (RLY08-3 K2, KKKf(fand ned by RLY@NEomputer generated signal whenever th same parametersparal,lelIHi%fer 1Isted in Figi7 @j8MgComputer Generated" exceed th 8 KBWi/ Scraftfconditions are autydSEally indicated on the CRT monitors and there is an audible an6diciator. A manuallcram may be used for a normal fast shutdown of the reactor. [bk Y 785hRod ControIIkkstem h f iW Thei(g!N f6f the MNRC reactor is controlled by WA6.six control rods. The control and transient rod drives are counted on a bridge at the top of the reactor tank. The drives are connected to the control and transient rods through a connecting rod assembly. The following sections describe the control and transient rods and their respective drive assemblies. 6

{ Rev. A. 7/5/96 7-20 O () 7.3.1 Control Rods Reactor core loadings utilize fuel followed control rods, i.e., control rods that have a fuel  ; section below the absorber section. The uppermost section is 6.5 inch-long air-fill 6d void and the next 15 inches is a solid boron carbide neutron absorber section, l$mediafefy below the absorber is the fuel section consisting of 15 inches of U-ZrHu whose%rini$5 is enriched in 20, depending on the core loading. The bottom sectio Me approximately 6.5 inches long. The fuel and absorbe tions are[EaledIQMhiess P steel tubes approximately 43 inches long by 1.35 inch ]anster. y

                                                                     %1 One control rod containing a stainless steel neutron absdFeiniititian and an air follower may be included in core loadings that have only four fuel-fgMW                        ods. The low reactivity worth of this rod allows very fine reactivity control. 'This rodM
                                                                                       '         ~

411nensions as a M The design of this

     . fuel foll. owed control rod of 43 inches in lengthgkinches iA rod differs from that of the fuel-followed Tp3 @;jy two wg1) the 56ron carbide neutron absorber is replaced by as 15 inc              solii            staipfess steel section, and 2) ir-filled $                  has a 22 inch in length the    fuelsteel stainless     follower     section top section           is replaced by 'n and a 21 incidi                         length     l$r follo i the bottom  section.

aw The fuel-followed control rods addh prese the sta s steel followed control rod, pass

 /^\

V through and are guided oy 1.5 i$f{liameted control rods ith fuel fo hown i s in top and bottom grid plates. A typical Iwn and inserted positions in Fig. 7.10. g i p.pg The trans a sealedA44:251in long by 1.25 in, diameter tube containing ;olid boron carbide as a' lbbsorber351* min [ iib a follower. The absorber section is 21 in. Iong ar.d the follower is'ayhTt$iately 2f$M The transient rod passes through the core in a

                                      }upe. Mtube receives its support from the safety plate and its perforated lateral positioning frohaldiiM6I(M@,hp plates. It extends above the top througlithe tube is provid6d.byt large number of holes distributed evenly over its length. A locidnh device is built iSEM lower end of the assembly,                                                    A h0                    $

Control Rod Drive Assemblies (For Transient Rod Assembly see Section 7.3.3) W roll% positioned by five standard TRIGA clectrically powered rack and pinion kb@$$11. vges One rod is designated as a regulating rod and use automatic power control. All rods anc rod drives are exactly the same and operate at a nom-inal rate of approximately 24 in, per minute. The rod drives are connected to the control rods through a connecting rod assembly. These assemblies contain a bolted connection at each end to accept the control rod at one end and the (n) v control rod drive at the other. The grid plates provide guidance for all control rods during operation of the reactor. No control rods can be inserted or removed by their drives L a sufficient distance to allow disengagement from the grid plates. j a l s

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RACK-AND-PINION CONTROL ROD DRIVE (TYPICAL) FIG. 7.I1

7-23 Rev. A. 7/5/96 O Each drive consists of a stopping motor, a magnet rod-coupler, a rack and pinion gear system, and a ten-turn potentiometer used to provide an indication of rod position. The p(nion gear engages a rack attached to a draw tube which supports an electromagnet. The , magnet engages a chrome-plated armature attached above the water level to the end offconnecting rod that fits into the connecting tube. The connecting tube extends dogo the %$1[M. The magnet, its draw tube, the armature, and the upper portion of the, connecting r$dMiigmsed in a tubular barrel. The barrel extends below the control ro(dfive mountfij$$ith the lo end of the barrel serving as a mechanical stop to limit tg downwgd' trave!%g(g drive assembly. The lower section of the barrel contalas:an airf$nubber to dahipenNi6 k of the scrammed rod. In the snubber section, the cont $iYdire decelerated thN5jh'a length Edi$idSched to the connecting rod by of 3 in. During this length, air is compressed under a p'i$13Nf"" 1; an a the weight of the control rod and is slowly bled to atmai electiohagnet is ~ valve. The control rod can be withdrawn from the rel$ tor co?' energized. When the reactor is scrammed, the eleclHiinapnet is %b3d and the armature is relefsed. A W:/ M]NNE1?dsk The rod drive motors are stepping motor [ iven b) $ e speed of the rods is settable and are normally set to insert 4IwithdraEthe co'ndfl7dds at a nominal rate 24 in./ min. The unique characteristEi'of a ste%ing motM/ translator system are used to provide fast stops and to limit c dj or o " avel. 4 e

                                                         ~

These rod drives have the; ~A ity of w mg the rods in as little as two seconds (7.50

                             ~

in/sec).goMe[, the s _ fail-safe, EAWmultiple system failures are required to get uncontrolled %ithdr;1wal cIfIhhjidirat the n.aximum speeds.

               $hNEk                     N!Ik                                                d bottom Limit switches'isoGr t of travel andpn@# ed'on each'dr$iissembly stop the rod drive mot NitEhing for'Eoltsole indication which shows:

4? kWE%

1. AWhen the magnefisiithb up position.

2fylWhen the magneiMiifthus the control rod) is in the down position. _ 3h When the controfrod is in the down position. Q M fi ocked l switc!Nn the reactor console power supply prevents unauthorized operation of allTcMtiol; rod'diifes. Sleicfr

      %.;qug These r"od drives were first developed in 1959, and have been modified and improved a number of times. The design has proven to be reliable and has been used in more than 60 TRIGA reactors containing more than 160 rod drives.

7.3.3 Transient Rod Drive Assembly The MNRC adjustable fast transient rod drive, Fig. 7.12, consists of a combination of a standard TRIGA rack-and-pinion control drive, described in Section 7.3.2, and a standard

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RDCTOR 820CE ADJUSTABLE FAST TRANSIENT ROD DRIVE ASSEMBLY (TYPICAL) FIG 7:12

7-25 Rev. A. 7/5/96 TRIGA fast transient control rod drive, both of which have been modified. This combination transiegt rpd drive can be used to fire low-level pulses and keep the pulse rod totally out of the core 6PMdkt-:msM pulse. This combination drive unit was chosen to tak($dvantage of iniotrdrive and on the extensive operating experience gained on bo the standard fast transient rod drive. This combination driye unit has%xt osfv'e operating i experience at the Japan Atomic Energy EY Institute NS (JAERI)fnd:Sandih aH r i The standard fast transient rod drive portion of the asseNbly consi[bo to drive the transient rod out of the core and a dashpotMdecelerate the transientMifaisf system at the end of the stroke. The total length of th53AsierTt rod travel is 21MIiffiiluding a 6 in. deceleration length. yhwsm , ymm . The pneumatic cylinder is single-acting and has a maximum stro{eyf;22JJi.Tlearance is provided to ensure that the dashpot will bottom putfefo(e the uppglimipbf the pneumatic cylinder. The cylinder is equipped with lowJrickin%so

                                                   ,                that th'ipt will df6p freely back i

into the core after the transient. A piston ppsition s%iti:hiiprovided to indicate when the piston, and therefore the control rod, isjiEihe full 45wWdsiGEsthe dashpot is at the lower l end of the cylinder assembly. This ass &Inbly hasjits own {iisforfand bleed parts and is designed to decelerate the transientfoland elidiinate any!h'ard stops. fh !Y The standard fast transient rod dnve is providhd with;its own pneumatic cylinder, accumulator so tank, presspfe,. regulator,[iybe%noid accumuladiciaintwhichd valini.QTli(561enoid lpressuriddWtE150 psi. Aluminum tubing 7/8 in. diameter valve ac is used Efei$$h$ll: Ling rMnMdiiitthe pneumatic cylinder, which is mounted on the rod drive bridge, and'fhitMwat rod.VNb NSNb mwah DM2 me i The standard 1ast tragsidlifod driG6;is thoroughly developed and tested. These drives have been instalI$d on the TRIGAtrbactors at GA Technologies Facility, University of California at Irvineffandia :4ationalM1 Airy, University of Illinois, Japan Atomic Energy Institute (JAE$1), and the dual-cNeVklGA res Trch reactor in Romania. _ M kasic conceptn usehl1 the MNRC drive is to modify the standard fast transient rod drive to haivea' portion o thibiston assembly extend through the top of the drive. This extension of thEtEalifent'rtid(yliton engages a yoke mounted below the armature. The rac d'riN[MEdlitTd slightly above and to one side of the transient rod drive. When the rack-and-pinioDNN is driven up, the yoke moves under the transient rod drive piston extension and moves the piston assembly, and therefore the transient rod upward with it. The rack-and-pinion drive position is read out on the console and the drive can be stopped at any position of its travel by the reactor console operator. Upon firing a transient, the transient rod will move until it is completely out of the reactor core. During this portion of travel, the piston rod extension will slide freely in the yoke mounted to the rack-and-pinion drive and no movement of the rack-and-pinion drive will be required.

l RevcA. 7/5/96 7-26 O ( Upon completion of a transient, A%2r-r 2, both of the coupled drives will scram and the transient rod will fall completeiy back into the reactor core. The yoke mounted on the armature will fall about the same speed as the piston rod assembly attached to the, transient drive. The rack-and-pinion drive.is capable of moving and holding the transie at any position between full out and full in. The rack-and pinion rod drive i - a Af moving the rod approximately 15 inches of travel, the same as the travel of ' control rod drive, and of scrame.ing and dropping the transient fjr'om an . W  % In order to combine the operation of the rack-and-pini< drive - e tra N e rack-and-pinion rod drive has been modified. The dri' gs same rack- , sembly assembly, magnet and armature connection, and modileasesson of the lower as the standard control rod drive. The lower barrel as rter and contains a slot on one side for the yoke assembly. The lower i;arrel ass I , terminates in a large , Ethh bottom of --a

                        - heavy flange. A bearing housing with a double set o all the lower barrel and an actuator shaft passes                    ing                   e top end of the actuatoTshaft contains the magnet armature,                    sembi            lted to the actuator shaft just below the magnet armature. Th           ing                      a rigid and accurate parallel path for the entire rack-and-pi      rod dr The entire assembly consisting of          nd        ntrol r    ive assembly, the modified lower idly        to a      rt which runs parallel to the h)

( barrel, and the bearing housing transient rod air cylinder. < f& k

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7/05/96 i l I CHAPTER 8 i } i

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