ML20198N726

Forwards Response to RAI Re Section 2.5 of SAR for Mcclellan Air Force Base non-power Reactor License Application Dtd 971010.Info Requested Is Contained in NUREG-1537,Part 1, Chapter 2

ML20198N726
Person / Time
Site:	University of California-Davis
Issue date:	10/29/1997
From:	Richards W AIR FORCE, DEPT. OF
To:	Weiss S NRC
References
RTR-NUREG-1537 NUDOCS 9711060054
Download: ML20198N726 (33)

Text

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6

, "I a DEPARTMENT OF THE AIR FORCE HEA00VARitR$ SACRAMENTO AIR 400 stICS CENTtRIAFMC)

McCLtLtAN AIR F0hCt SA$t.CAllFORNIA 29 Oct 97 i Shl ALC/T1-1 5335 Price Avenue hicClellan AFB CA 95652 2504 Dr Seymour 1-1. Weiss U.S. Nuclear Regulatory Commission hiail Stop 0-1 iB20 Washington DC 20555 Ref: Docket No,50-607 Dear Dr. Weiss In response to the letter request for additional information for Section 2.5 of the Safety Analysis Report for the hicClellan AFB Non Power Reactor License application, dated 10 Oct 97 (Atch 1), the following information is provided:

a. The information requested comes directly from NUREG 1537 Part 1, Chapter 2, Section 2.3 (Atch 2).

b. The information provided from the hicClellan Safety Analysis Report (SAR) Chapter 2, Section 2.5 can be found at Atch 3.

The information provided in the hicClellan Safety Analysis Report does not specifically address each of the points provided in the NUREG 1537 Part 1, Chapter 2, Section 2.5, guidance document. Nevertheless, the information provided does meet the intent and depth of detail implied by the following guidance statement found in Chapter 2, Section 2.5.

" The degree of detail and extent of the considerations should be commensurate with the potential consequences of seismoiogical disturbance, both to the reactor facility and to the public from radioactive releases."

7D f0 '

The worst potential consequences of a seismological event is analyzed in Chapter 13, Section 13.2.3 (Atch 4).

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The analyses describes the complete failure of the reactor tank and the "INSTANTANiiOUS" loss of all water from the reactor tank, regardless of the magnitude of the 9711060054 971029 PTA ADOCK 05000607 e ,

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s,eismic event. The analysis shows that with the i CCS this event can be mitigated to produce no public health problems due to radiological releases.

In conclusion, the information submitted integrated with the analyris of Chapter 13, Section 13.2.3, should provide a sullicient basis for this non power reactor license application.

Please contact me if further discussion is needed to resolve this issue.

Sincerely e.Au p t '<N$4 WADE J. hlCllARDS I Chief, Nuclear Licensing and Operations 4 Attachments:

1. NRC Request,10 Oct 97

2. NUREG 1537, Part 1, Chapter 2, Section 2.5

3. h1NRC SAR C.. apter 2, Section 2.5

4. hiNRC SAR Chapter 13, Sectim 13.2.3 cc: Warren Eresian, NRC

f ATTACIIMENT 1 14RC REQUEST FOR ADDITIONAL INFORMATION I DATED 10 0CT 97 l

avr m o4 WM W

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REQUEST FOR ADDITIONAL INFORMATION FOR SECTION 2.5:

REPORT FOR THE MCCLELLAN AFB NON POWER REACTOR LICE  ;

N 1.

Provide a description of the structural geology at the site and its relationship to the regional tectonic structure, 2 ..

Provide a list of all instrumentally recorded or historically reported earthquakes with magnitude greater than or egual to 3 or modified Mercalli intensity greater than or equal to IV within 200 kilometers of McClellan AFB. If available. provide the latitude, longitude, depth.

magnitude, maximum each earthquake. Modified Mercalli intensity, and total felt area for epicenters. Also provide a clear map showing the earthquake 3.

Provide an updated list of the largest earthquake associated with each geologic structure and tectonic nrovince within 200 kilometers of McClellan AFB. If available, include an isoseismal ma) for each earthquake. .For each geologic structure, such as the r oothills Fault System and the Calaveras. Hayward, and San Andreas faults provide an assessment of the maximum earthquake potential based on fault length, fault displacement and earthquakt history. For each tectonic province.

any part that is within 200 km of the site (such as the Central Valley),

identify the largest-Dotential earthquake within the province and estimate the earthoua te return period. At a minimum, this event should be the largest earthquake to have occurred in the tectonic province.

N 4.

Using the latest attenuation relationships, assess the ground motion at McClellan AFB from the maximum potential earthquakes associated with each tectonic province or geologic structure, site amplification effects. Provide an evaluation include a discussion of with respect to the s xctral accelerations predictedofbyyour design basis the most recent attenuation relations 1ips.

5.

Provide a descr;ption of the foundation materials at the site and an assessment of the liquefaction potential.

ATTACHMENT r

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. 1 ATTACllMENT 2 NUREG 1537, PART 1, CHAPTER 2, SECTION 2.5

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NUREG - 1537 PART 1 1

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Office of Nuclear Reactor Regulation  ;

Division of Reactor Program Managment  ;

l Snr ChiudstKs vicinity from routine releases during normal operations and from postulated releases resulting from accidents. The analyses of potential doses from normal and j

accident releases should be placed in Chapters 11 and 13, respectively. The i

i meteorological information used for both long-term and short term dispersion calculation:, along with a description of the technical bases of the dispersion model should be summarized. The continuing onsite measurements program or an 4

altamative source of meteorological information (e.g., National Weather Service station) should be described; and plans for access to meteorological information during the license period should be described. Desciption of the meteorological

' program should include measurements made, locations and elevations of measurements, description ofinstruments and their performance specifications, and calibrations, type of data output, and data analysis pro vnres.

i 2.4 Hydrology In this section, the applicant should give sufficient information to allow an independent hydrologic engineering review to be made of all hydrologically related design bases, performance requirements, and bases for operation of structures, '

systems, and components important to safety.

Sufficient information should also be given about the water table, groundwater',

and surface water features at the reactor site to support analyses and evaluations in Chapters 11 and 13 of consequences of uncontrolled release of radioactive material from pool leakage or failure, neutron activation of soils in the vicinity of the reactor, or deposition and migration of airbome radioactive material released to the unrestricted area.

The effect of potential floods on sites along streams, rivers, and lakes should be analyzed. Effects and consequences of a probable maximum flood, seiche, surga.,

standing water, drainage or seismically induced flood (such as might be caused by dam failure) should be considered. Hazards of tsunami, river blockage, diversion in the river system, or distant or locally generated " sea waves" should be described to establish the suitability of a site. The detail and extent of the considerations  ;

should be commensurate with the potential consequences to the reactor and to the public, the environment, and the facility staff.

2.5 Geology, Seismology, and Geotechnical Engineering In this section, the applicant should detail the seismic and geologic characteristics of the site and the region surrounding the site. The degree ofdetail and extent of ,

the considerations should be commensurate with the potential consequences of seismological disturbance, both to the reactor facility and to the public from : '

radioactive releases; i

REv.0,2/96 25 STANDARD FORMAT AND CONrENr

CHARTER 2 2.5.1 Regional Geology The applicant should discuss all geologic and seismic hazards within the region that could affect the facility, and relate them to the regional physiography, tectonic structures and tectonic provinces, geomorphology, stratigraphy, lithology, and geologic and structural history and geochronology.

2.5.2 Site Geology l

The applicant should discuss in detail the structural geology at the facility site, including the relationship of site structure to regional tectonics, and should pay particular attention to specific structural units of significance to the site such as folds, faults, synclines, anticlines, domes, and basins. The applicant should also discuss the geologic history of the site and should relate it to the geologic history of the region.

2.5.3 Seismicity The applicant should list all historically reported earthquakes that could have reasonably affected the region surrounding the site. The list should include all earthquakes of modified Mercalli intensity greater than IV or magnitude (Richter) greater than 3.0 that have been reported in all tectonic provinces, any part of which is within 200 kilometers of the site.

2.5.4 Maximum Earthquake Potential The applicant should note the largest historic earthquake associated with each geologic structure or tectonic province. If the earthquakes are associated with a geologic structure, the applicant should evaluate the largest earthquake that could occur on that structure on the basis of such considerations as the nature of faulting, fault length, fault displacement, and earthquake history. If the earthquakes are associated with a tectonic province, the applicant should identify the largest historical earthquakes within the province and, whenever reasonable, should estimate the return period for the earthquakes. Alt,o, isoseismal maps for the earthquakes should be presented.

2.5.5 Vibratory Ground Motion The applicant should proceed from discussions of the regional seismicity, geologic structures, and tectonic activity to a determination of the relation between seismicity and geologic structures. The earthquake generating potential of tectonic provinces and any active structures should be identified. Finally, the applicant should assess the ground motion at the site from the maximum potential NUREO 1537,PAaT 1 2-6 REY. O,2/96 O

4 sntcurauimmes earthquakes associated with each tectonic province or geologic structure and i

should consider any site-amplification effects. Using the results, the applicant should establish the vibratory ground motion design spectrum. )

2.5.6 Surface Faulting

) The applicant should discuss any potential for surface faulting at the site, and 1

should list all historically reported earthquakes that can be reasonably associated y with fauhs, ar.y part of which is within 8 kilometers of the site.

i 2.5.7 Liquefaction Potential The applicant should discuss soil structure. If the foundation materials at the site adjacent to and under safety related structures are saturated soils or soils that have a potential for becoming saturated, the applicant should prepare an appropriate i

state-of-the-art analysis of the potential for liquefaction at the site. The applicant should also determine the method of analysis on the basis of actual site conditions, l l the properties of the reactor facilities, and the earthquake and seismic design j requirement for the protection of the public.

2.6 Bibliography American National Standards Institute /American Nuclear Society, ANSI /ANS 15.7, "Research Reactor Site Evaluation," 1977.

American National Standards Institute /American Nuclear Society, ANSI /ANS 15.16, " Emergency Planning for Research Reactors," 1982.

Intemational Atomic Energy Agency,IAEA TECDOC-348, " Earthquake Resistant Design of Nuclear Facilities With Limited Radioactive Inventory," 1985.

International Atomic Energy Agency, IAEA TECDOC-403, " Siting ofResearch Reactors," 1987 U.S. Nuclear Regulatory Commission, NUREG-0849, " Standard Resiew Plan for the Review and Evaluation ofEmergency Plans for Research and Test Reactors,"

1983.

U.S. Nuclear Regulatory Commission, NUREG/CR 2260, " Technical Basis for R.G.1.145 Atmospheric Dispersion Models," 1981.

U.S. Nuclear Regulatory Commission, Regulatory Guide 1.145, Rev.1,

" Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants," 1982.

REY.0,2/96 27 STANDARD FORMAT AND CONTENr

ATTACHMENT 'a MCCLELLAN SAFETY ANALYSIS REPORT CHAPTER 2, SECTION 2.5

. I 2=23 Rev. 0 8/20/%

Neither of these rivers presents a flood hazard to the MNRC facility. He nearest 100 yr Doodplain 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 j

De probability of an accidental release of radioactive liquid effluents from the MNRC in surface waters is extremely low. Two (2) MNRC systems may contain radioactive liquid:

De reactor primary and the water purification systems. All of the components for these systems; reactor tank, pumps, heat exchangers, filters, resin tanks, valves, and piping, are locawi within the MNRC reactor ar.d equipment rooms. Any contaminated water leakage from this e,ciipment will be wiped up and disposed of as discussed in Chapter 11. The only other areas w are contaminated water may be encountered is in the radiography bays and the mer ; wrbwm. The radiography bays have a drain system that leads to a sump in Bay 1. i Any wat::r collected in the sump is pumped into an above ground liquid storage tank. The decontamination shower located in the men's washroom also drains into the storage tank.

There are no floor drains in the men's washroom that lead to the industrial waste. Any water entering the tank, even if other than the reactor systems, will be analyzed for radioactive materials. If radioactive materials are found it will be disposed of as discussed in Chapter 11, 2.5 Genlaey. Seinmalnav and Gaatachnical Fnoinnering The Sacramento area is located in Seismic Zone 3 of the Uniform Building Code. In general, seismic activity is not as great in the area as it is in the coastal areas (Refs. 2.10, 2.11, 2,12 and 2.13). Based on a review of historical records, the maximum-intensity earthquake in Sacramento in historical times has been about VII on the Modified Mercalli scale (Refs. 2.12 and 2.13). This intensity was the result of earthquakes centered about 20_mi (32 km) west of Sacramento with an estimated magnitude of 6.0 to 6.5 on the Richter scale. Earthquakes of the in'ensity of VII are characterized by collapse of weak chimneys, nederate damage to masonry walls, fall of cornices from high buildings, and fall of some nonstructural, unreinforced brick walls (Ref. 2.12 and 2.13). However, earthquakes of higher intensity

- could have occurred prior to the coverage of the historical record, and higher intensity carthquakes are possible in the future. Figure 2,12 is a historical summary of the seismic activity in the area.

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. EARTHQUAKE EPICENTER MAP OF CALIFORNIA (PARTIAL)  ?*

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Rev 0 8/20/% 2-26 California contains innumerable earthquake fault! 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 by adequate explorations. The fact that no surface faults appear on the map in the Sectamento or San Joaquin Valleys may only indicate that sediments laid down during late geologic time cover the fault scars. On the other hand, rock or the firmer sediments usually found in the hill and mountain areas retain the evidence of faults over long time periods, l

As shown in the figure, surface faulting has been identified in the Bear Mountain fault zone some 25 miles east of Sacramento and in the Rumsey Hills area west of Woodland. A number of subsurface faults have been found during explorations for gas near Sacramento as reported by the Division of Oil and Gas of the California Department of Conservation. Such subsurface -

faulting is reported near Freeport and Clarksburg just to the south of Sacramentol in the

~ Todhunter 1.ake area a few miles north and east of Davis; and in the Rio Vista area, to identify '

a few areas near Sacramento. Data are not available to indicate the existence of subsurfa faulting nearer to or within the City of Sacramento, j

Geologic investigations to date have not discovered evidence indicating movement on subsurface faults in the Sacramento Valley more recent than Eocene time, about 40 million years ago. Eocene rocks extend generally from the surface of the ground to 0.5 to 0.75 kilometer depth. One fault in the Folsom area, recently mapped by the California Division of Mines and Geology, has been interpreted as having moved during the Quaternary Period.

One conclusion based on this evidence is that except for the possibly more recent movement on  !

the fault in the Folsom area, there has been no near surface fault displacement in, or within <

close proximity of Sacramento during the past 40 million years. The focal depth of California earthquakes (the depth below the surface of the earth to the start of the rupture in the rock that provides the energy for the quake) ranges from a few kilometers to 15 to 20 kilometers, and therefore earthquakes of a i,maller magnitude could have originated here during the past 40 million years, but the faulting might not have extended into or through this layer of post-Eocene rocks.

A second conclusion is that faulting did extend to the surface, but the evidence for this surface breaking either has not yet been found or is undiscernible in the sediments which fill the valley,

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. . A SACRAMENTO AREA SIGNIFICANT FAULTS FIG. 2.13

Rev. 0 8/20/96 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 Sacram(mt: 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 appears that Sacramento has t not experienced violent earthquake motion of a nature compared with that experienced by several other areas within California.

Probably the greatest amount of earthquake shaking experienced in Sacramento during the recent past occurred on April 21,1892, his earthquake produced extensive damage to towns some 25 miles west of Sacramento.

As noted above, the April 21,1892 earthquake, along with the quake two days earlier, i

probably produced the most vigorous earthquake shaking in Sacramento .furing recorded '

history. Dere is some evidence that the epicenters of these shocks were in the area between

! Winters and Vacaville. Both of these towns, as well as Davis, Dixon, and Woodland experienced significant damage to many structures. Although the location for the fault responsible for the 1892 earthquakes is not known, the California Division of Mines and Geology and the U.S. Geological Survey have recently found (May 1972) that the Green Valley fault, west of Fairfield, is showing active fault creep or slip movements just to the south ofInterstate 80 highway.

A lineament on the east flank of the Dunnigan Hills has been mapped recently by the U.S.

Geological Survey. It may be the surface expression of a fault that has moved recently.

In recent time there was about $10,000 damage t.t the Sacramento Filtration Plant resulting from the Dixie Valley earthquake, east of Fallon, Nevada, December 16,1954 - a Richter magnitude 7.2 cartnquake. This was about 185 miles northeast of Sacramento and clearly indicates that the long period earthquake waves resulting from distant earthquakes can have definite effects upon structures or their contents. Damage also occurred to the digestion tanks at the Sacramento Sewage Treatment Plant and to a clarifier tank at the Campbell Soup Company.

There appears to be a strong northwesterly structural " grain" to California geology.

Earthquakes having epicenters towards the west have not affected Sacramento in the past to the same exterA as those centered east and south of Sacramento. The 1892 Winters earthquake appears to be an exception to the general statement. To explain further, the April 18, 1906, San Francisco shock of Richter magnitude 8.25 with its epicenter about 80 miles west of Sacramento was probably felt in Sacramento with about the same intent.ity 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' Augrst 12,1966, 95 miles northeast of Sacramento was strongly felt in the Sacramento area as well as the above mentioned Dixie Valley earthquake 185 miles northeast of Sacramento. -

- 2 29 Rev. 0 8/20!%

i The University of California Seismographic Station Reports that since 1932 there have been approximately 700 carthqur. N of Richter magnitude 4 and greater in the area bounded  :

between longitudes 118'W ac.J 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 and from south of Fresno to Redding. Also within this area there were approximately 90 earthquakes of magnitude 5 and some 15 earthquakes of magnitude 6 during this period.

A As noted above, the distance of the closest fault to McClellan AFB far exceeds the siting requirements of ANSI 15.7, Section 3.2, which states 'no proposed facilhf shall be located '

closer than 400 meters from the surface location of a know capable fault.'

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ATTACHMENT 4 CHAPTER 13, SECTION 13.2.3 LOSS OF COOLANT ACCIDENT (LOCA)

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13.2.3 Loss of Coolant Accident (LOCA) l

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[ 13.2.31 Accident Initiating Events and Scenarios Loss of coolant from the MNRC reactor could occur primarily through one of tivo scenarios,  !

! pumping water from the reactor tank or reacto tank failure. These scenarios are analyzed part of this section. ( as I ,

13.2 3.2 Accident Analysis and Determination ofConsequences.

13.2.3.2.1 Pumping of Water from the Reactor Tank l

4 l

The intake for the primary-cooling-system pump is located about 3 ft below the normal tank 1

l water level. In audition, the line is perforated from about 8 in. below the ncrmal tank water I

level to the intake line entrance. The intake for the purification-system pump is through a short I flexible line attached to a skimmer that floats on the surface of the tank water. However, the l

length of the flexible line is such as to cause loss of pump suction if the tank water level is .

[

lowered about 4 ft. Thus, the reactor tank cannot be accidentally pumped dry by either the l j

primary pump or the purification system pump. Also, it is not possible for other cooling system i

{ or water cleanup system components to fail and syphon water from the tank since all of the 4 1 primary-water-system and purification-system piping and components are located above the i I normal tank water level.

l i I The tank could be pumped out with a portable pump, but this would require deliberate action on I

the part of the operators and it is inconceivable that such an action would take place while thel reactor was operating or at any other time withoc removir.g the fuel and taking numerous other l precautions. However, if the reactor were somehow pumped dry while the reactor was shut i- I down, the fuel temperature obtained would be considerably lower than for a loss of_wa:er while i j

the reactor was operating, and this unlikely error would not cause damage to the fuel eleinents. I Similarly, the dose rate from the uncovered core and the water radioactivity concentration I would be less than that shown in 13.2.3.2. I1 and 13.2.3.2.12.

l 13.2.3.2.2 Reactor Tank Failure l

I A hole in or near the bottom of the reactor tank could cause the water level to drop below the top of the fuel elements. This event could occur either during reactor operation or while the j

reactor was shut down and unattended. There are no nozzles or other penetrations in the reactor tank below the normal water level, so the only mechanisms that could cause tank i failure are corrosion of the tank or a mechanical failure. Leaks caused by corrosion would I

unquestionably be small leaks, which would be detected before the water level had lowered significantly. In such a case, makeup water could be supplied by the auxiliary make-up waterj system (AMUWS) until the reactor had been unloaded or the leak repaired. l

__ __ __ - ~ -

.l- Rey, != --4/14/97 13 12_

[ -

Provisions to monitor for and collect tank leakage have been incorporated into the facility

[

l

- design First, the tank is surrounded by cormgated metal. The corrugations provide a path to j

the bottom of the tank for any water leabge from the walls. Second, a drain, see Chapter s,-

j within the bulk shield surrounds the bottom of the tank This drain will collect any water that may leak from the tank walls or bottom. Third, a duct leads from the drain to Radiography l . Bay I and the exit of this duct is periodically monitored for water leakage. Ifleakage is j

detected;the water could be easily colluted at this point or diverted to the liquid holdup tank outside the building.

e Consequences of a slow tank leak would be minimal r.nd would require collection and containment of the water which leaked from the tank. This would be easily accomplished by _  :

using the exist.ng liquid effluent control system described above. Small tank leaks due to corrosion are normally repairable using conventional techniques for patching aluminum, and thus it is expected that a leak could be located and fixed before there would be any significant loss of wa'er from the tank.

An earthquake of much greater intensity than the Uniform Building Code Zone 3 earthquake l appears to oc the only credible mechanism for causing a large rupture in the tank, since the tank l when supported by its associated biological shield structure was designed (with an importance i factor of 1.5) to withstand this magnitude of earthquake. Even if such an event is assumed to

] cause very rapid loss of water while the reactor is operating at peak power; a reactor shutdown would be caused by voiding of water from the core, even tf there were no scram due to dropping of the control rods.

l _ A large rupture of the tank would obviously result in a more rapid loss of water than a leak due l to corrosion or a minor mechanical failure in the tank wall. The MNRC reactor tank has no l breaks in its structural integrity (i.e., there are no beam tube protrusions or other discontinuities -

l in the reactor tank surface). In addition, the reactor core is below ground level. Thus the l- potential for most types ofleaks is minimized.

I l Part of the 2 MW upgrade to the reactor included a new cavity (Bay 5) cut into the biological

-l shield. This cut exposes the reactor tank wall below the reactor core level, and this introduces

-l an increased possibility of draining water from the core area. While steps have been taken to

_l mininuze the probabihty of a tank rupture in this location, and it is believed that the likelihood of ,

l such a rupture is very low, an unplanned occurrence could neverthekss initiate such a event.

l Therefore, an Emergency Core Cooling System (ECCS) has been installed to cool the core until

-l the fuel has decayed to a level where air cooling is adequate to malatain fuel temperatures l _ below the design basis limit (see Chapter 6 for details of the ECCS design and operation).

l l - An analysis detailing the cooling capabilities of the ECCS is described in the sections which l follow. This analysis does not postulate the occurrence of a particular initiating sequence of l events leading to all fuel elements in the core being uncovered. Instead, it simply assumes that I the tank has ruptured and all the water is lost. Such an event has several different l consequences. First there is the possibility of fuel clad rupture should the fuel temperature

l 1312a Rev.1 4/14/97 [

i l

exceed design basis valuesJ This event is covered in the analysis that follows, and focuses on -

l the action of the ECCS to prevent fuel temperatures from reaching safety limits. Second, .l the '

is a possibility of personn;l exposure to rediation from the uncovered reactor core due to the .,

, l

! direct beam from the core or from radiation scattered from the reactor walls and ceiling. I -l Finally, there is a chance that the lost water could cause ground water contamination. -

e l

[ 13.2.3.2 2.1 Description ofECCS and Assumptions I

i I I

l

' A loss-of-coolant accident (LOCA) is postulated for the MNRC in which the reactor pool is l

rapidly drained of water during operation at 2 MW (it is assumed that the rt actor has been i I running at 2 MW for an infinitely long time). Because the LOCA uncovers the core quickly, 1 the fuel clad temperature in some of the centrally located fuel elements could exceed I the d basis temperature limit of 930'C.

l l

When the reactor tank water level drops below the normal operating range (typically a loss I of approximately six (6) inches of water) a tank low-level alarm sounds. This alerts the operator l that action must oc taken. Depending upon the rate of water loss, the suspected cause of the l

loss, and other considerations, several different actions may be taken by the operator in l response to a reduction in the tank water level. One such action could be activation of the ECCS. l l

1 Upon activation of the EC,CS, cooling water from the domestic water supply will be introduced I into the reactor tank and maintained until the fuel no longer contains sufBeient decay heat l to present a threat to the fuel cladding or water is restored to a level above the core. If the tank I

water level has dropped to less than about two (2) feet above the core, water from the ECCS l

will be sprayed onto the top of the remanung water column above the core; however, if the tank l

water has dropped below or partially below core level, the ECCS water will be sprayed l dire onto the core. During this time, the decay heat will be removed by the remaining tank wateri or by the water spray and the maximum fuel temperature will be reduced rapidly from an elevated I

operating temperature down to about 200'C and then gradually to 100'C with continued j spra cooling.

1 I

At the end of spray cooling, natural air convection will be established in the core. DuringI this cooling phase, the temperature of the fuel will rise slowly over several hours to a maximum and l

then decrease with continued air cooling. The maximum fuel and cladding temperature isI controlled by the length of spray cooling and by the natural air cooling. Under the preceding l conditions, no fur' clads will be ruptured.

I I

The detailed components of the core cooling system to be used to maintain fuel temperatures l below the design basis limit are described in Chapter 6. Basically the system consists l of a q connect system for coupling to the domestic water supply, sensing devices to indicate the need I

' to initiate emergency cooling water flow, a nozzle to distribute the coolant flow over the core, l a

chimney mounted above the core structure to provide a sufficient channel length for maintain l sufficient air flow through the core, and a ventilation system to provide air circulation through I the reactor room.

I

o,- , ,

- l Rev. - 1_ ' 4/14/97 13 12b

-l It should be noted that in a TRIGA8 reactor, loss of reactor coolant water will automatically j cause a complete reactor shutdown even without a control rod scram. Experiments with the i GA subcritical assembly have indicated that the reactivity worth of the water in the core is on i

' l the order of 10% (more than 13 dollars). As a result, were the reactor to be operating during a 4

l catastrophic event in which the cooling water were completely lost, the reactor would l automatically shutdown (even without insertion of control rods) once the water level dropped a

I few centimeters below the upper grid plate.

!4 l

l 13.2.3.2.2.2 Spray Cooling i l 1 A considerable amount of experimer tal data has been gathered on the efBeacy of spray cooling j

l for a system of heated cylindrical rods in bundles. This data (Reference 13.7a) indicates that the l amount of heat that can be removed by a water spray without the rod's wall temperature.

I exceeding about 100'C is simply the amount of heat that would increase the enthalpy of the i sprayed water from its inlet enthalpy to the saturated liquid enthalpy. The experiments were 2

l conducted for heat fluxes up to about 4 W/cm , which is larger than the maximum heating rate 1 in the hottest fuel element during the loss-of-coolant accident. Even if the initial surface l temperatures were very high (~900*C) before the spray is initiated, the surface temperature l would be very quickly reduced to about 100'C if sufficient water is provided to remove the I heat without increasing the coolant temperature to the saturation point (Reference 13.7b). The i spray flow rate required to cool the fuel to 100'C from 2 MW operation corresponds to 12.3 l gpm through the TRIGA@ core, including the consideration of peak power in the core.

l l Measurements have been made to determine the actual flow rate required to fulfill the 12.3 gpm l flow requirement through a TRIGA@ core (Reference 13.7c). These tests indicated that both-l the nozzle type as well as its location and orientation are important in order to provide the l required cooling spray. Results also showed that a total spray flow of 20 gpm from the nozzle l speci6ed in (Reference 13.7c) located approximately 2 ft. above the top grid plate will assure

-l that adequate core spray cooling is available to meet the requirements above. Provisions have l .been established to ensure that sufBeient spray cooling water can be supplied to the reactor core l when needed from the building domestic water supply.

l-l 13.2.3.2.2.3 Air Cooling I

l The relatively small size (-7500 cu. ft.) of the reactor room can affect the convective air -

l cooling of the reactor core after spray cooling ceases. In the small reactor room, hot air from

• l the core is expected to overload the air conditioning system and raise the ambient air l temperature. Since this is the air that is available for cooling the core, this situation was l analyzed in detail.

l l The air flow in the reactor room during normal operation is the following An exhaust flow of l 800 cfm passes through absolute filters on the way to the stack. Of this 800 cfm,500 cfm

) comes from the air conditioning system (1100 cfm outgoing,1600 cfm returned) and 300 cfm

____._m__.____ _ . . _ . _ _ _ _ - . _ _ _ _

-. Q-t1 '13-12c _Revf1 4/14/97 l

4-comes from leaks into the reactor room from around doors or other leaks in thel reac 1

' enclosure. Appendix D provides schematics of the reactor room and the exhaust and supply [

ducts. '

. l 3-l

' Although 1100 cfm is withdrawn from the room by the HVAC, and is refrigerated, and returned l with an additional 500 cfm of air at ambient temperature, it will be assumed that during thel LOCA event, this air flow continues but that the refrigeration fails due to an excessive heat l

load; (Note that if the HVAC fails, the reactor room exhaust fan will still be able to draw at -

1 least 500 cfm of ambient air in through the open HVAC damper,) Thus,500 cfm (from the air l_

conditioning) plus 300 cfm of air (from in-leakage in the reactor room) are continuously I

supplied to the reactor room at an ambient air temperature (~80'F) to match the 800 cfm

_l f exhaust that continues during the accident. To ensure a continuous air supply to and from l-the reactor room a backup power supply has been provided for the reactor room exhaust fan _

j. (EF-1). l

}- l l

13.2.3.2.2.4 Assumptions Made for ECCS Operation.

l 1

The following assumptions are necessary to initiate and evaluate ECCS operation:

l f .

i 1.

The ECCS will be initiated by the reactor operator if the water level drops to a level- l that requires the system to be turned on. (Operator action and manual operation of l

[

4 the ECCS is considered sufficient since at least 20 minutes is available for initiatio l after an instantaneous loss of the tank water before sufficient heat will build up in the l fuel to threaten the safety limit, See Reference 13.7d.)

l

2. I-If the reactor room continuous air monitor (CAM) actuates the recirculationI mode ventilation for the reactor room due to elevated radiation levels following tank water l L loss, the reactor operator will assess the situation and then switch the room ' ventilation l

from recirculation back to the manual ventilation mode (See Chapter 9).

l

3. l Based on assumption number 2, the reactor room exhaust fan will continue to extract l

800 cfm from the reactor room (typically 500 cfm from the top of the reactor and 300 l l cfm from near the ceiling).

4 l

I 13.2.3.2.2.5 Performance of the ECCS l

4 l

! Because of the relatively small reactor room, it is necessary to consider for any air cooling l portion of the loss-of-coolant accident that the initial conditions consist of an air filled reactor l

tank containing a hot core near its bottom and surmounted by a small reactor room (7500 cu.

l ft.). Hot air rises (-227 cfm) from the core in a plume, part of which is removed into the 500 4

1 cfm exhaust duct at the top of the reactor tank. The remainder of the hot air plume rises into j

- the reactor room, mixing with the room air (aided by the 1600 cfm from the inlet air duct.) l

'Near the top of the reactor room 300 cfm of mixed air is exhausted. Ambient air at 80'F comes I

j_ into the reactor room at 800 cfm.

I I

4 l Rev.I 4/14/97 13 12d

-l

{

At quasi equilibrium, the mixed air in the reactor room, including that near the top of the reactor tank, is warmer than the 80*F ambient air from the outside. This mixed air flows in a near i

annulus down the reactor tank adjacent to the tank wall as the hot plume from the reactor core l

flows upward in the center of the tank. The downflow air partially mixes with the hot air plu jI the rising from reactor the core and increases in temperature. This downflow air then enters the bottom of core. ,

i l

An estimate of this air mixing using boundary layer analysis for the mixing region indicates that i

the temperatu e increase of the downflow air is approximately 10% of the difference between l

the downflow air temperature and the upflow average plume temperatuie.

l l 13 2.3.2.2.6 Thermal Model for Natural Convection Air Cooling l

l A thermal model was constructed to assess the fuel temperatures for the LOCA event after the l

l termination of spray cooling and with subsequent natural convection air flow through the core.

The TAC 2D general purpose thermal analysis code (Reference 13.7e) was used to calculate the i

l maximum and and cold regions average of the core. fuel temperatures for typical fueled channels representing hot, avera l

l l

Four flow cham.cls were used to represent the natural convective flow past these three (3) fueled regions and one (1) unfueled region. No cross flow was considered between the various l

}

tiow channels. One flow channel represented all the flow channels in the cooler F and G ring l

and one flow channel represented all the flow channels in the average powered D and E rings Individual flow channels were modeled to represent the locally different flow channels l

surrounding the hottest fuel element. To complete the surface boundary conditions for these I latter two flow channels, it was necessary to include in the thermal model a fifth flow channel.

l i

This channel was used to represent the temperature response of two adjacent unheated gra elements in the C-ring and the adjacent central in-core experiment facility.

l l

Decay heat is removed from the reactor by radial conduction to the surface of the fuel ebments l

l where it is removed by convective air currents driven by buoyant forces generated by the reactor natural convection loop. The resulting peak and average fuel temperatures were i

calculated for the hottest element as a function of time. The natural convection flow I dependent on the pressure balance in the system. The buoyancy driving head for the natural l

convectica flow is the difference between the density head of the cooler downflow and the l

l density head of the hot upflow. The subsequent analysis shows that a chimney two (2) fe provides adequate buoyant driving head.

\ l l 13.2.3.2.2.7 Reactor Core for LOCA l

t l The 20E Core (see Chapter 4) with the central experiment facility containing the aluminum and i

graphite plugs in place, all control rods fully up, and 101 fuel elements was chosen as the l

LOC A core configuration. The axial power distribution (Reference 13.8), with rearrangement

. 4 J

l13-12e-

{, ,

Rev.1 4/14/97 Tl l

required for the TAC 2D code was used for power density calculations. The results are shown .-

3 m Figure 13.2. l ll l F -

I i - Account was taken of the five fuel .~ollowers on the control rods. For the LOCA event, the I

1. control rods are fully inserted into the core. This means that the five fuel followers are l

suspended below the bottom grid plate. Each of these fueled sections is located within a_guide [

tube that has 12 openings in the surface. The 12 individual openings provide 24 in t of surface

-l

' area and are situated symmetrically around the device to provide adequate cooling air for each l element. '

. ]

I 13.2.3.2.2.8 Mixed Air Temperature in the Reactor Room l

j l For the design case (3.7 hcurs of spray cooling,2-ft chimney) the highest average temperature -

j l in the hot air plume from the core is approximately 1360*F. At this value the plume density I is

! very low and, consequently, the mass flow rate is low relative to the other air streams in the

l room. The volume of cir flow in this plume is 227 cfm. It is assumed, on average, that about I i 100 cfm of this plume is swept into the 500 cfm duct at the top of the reactor tank and that j I about 127 cfm ofhot air rises into the reactor room. (See Appendix D for details.) It is [

j assumed that 300 cfm of air at.80*F is continuously supplied to the reactor room and that the

! I duct near the top of the reactor room exhausts 300 cfm of mixed air. It is further assumed that i l the reactor room is small enough and the air cooling time is long enough (several hours) that la quasi steady state condition exists. That is, the temperature of the mixed air in the reactor room i- I is simply the mixed mean temperature of the plume from the core and the incoming air streams l on a mass flow basis, and further that constant specific heat and ideal gas behavior for the air -

l streams can be assumed.

l l

1-

. For 80'F air inlet to the room and an average plume temperature of 1364'F, the mean l

} temperature of the mixed air is about 138'F. Even if all the hot air in the plume were to rise

l into the reactor room rather than a ponion being drawn off at the top of the reactor tank, the l mixed air temperature would only rise to 180'F, a value that does no. alter significantly the l cooling conditions of the reactor fuel.

l L \

In addition to the above consideration of the mixed mean air temperature in the reactor room,' l

[ . there is the additional consideration that the 800 cfm rate of room air exhaust will proside about l 6.6 changes of room air per hour. During the two hours during which the peak fuel temperature [

- exceeds 900*C and the average plume temperature exceeds 1290'F, the ventilation system -l changes the reactor room air more than 12 times while bringing into the room 80'F air at 800 l

4 cfm. This fact provides additional rationale for a quasi equilibrium condition with mixed room l

air at relatively low temperature.

l I

i=

i 1

l Rev.1 4/14/97 13-12f

' l

.l 1.3 I i i l l i

!  ; c

! l i '

l 1.2 r

M  :

i 1.1 4

1

/ <

0.s =

1 L

0.8 )

)

1 -

0.8 -

Top of Fuel moetem of Fuse

! l i I i

i  ! l 0.4 0 2 4 4 8 10 12 14 it Distance, In.

l AXIAL POWER DISTRIBUTION FOR FUEL IN CORE 20E l FIGURE 13.2

,ji 1 1312g Rev.1 4/14/97

, l

( -

The calculation of the mixed air temperature in the reactor room is conservative. It is assumed

[ that the hot air plume has its maximum temperature even at the start of the air cooling cycle l

l Actually, the plume temperature starts at 212*F, reaches 580*F in a half hour and is below l

1000*F for the first hour of cooling. Uader these circumstances, the reactor fuel would be

! l cooled more efficiently since the inlet air at the bottom of the core would be lower. However, I to be conservative, it is assamed that the hot air plume has its maximum temperature during l th

entire air cooling cycle.

1l I l-13.2.3.2.2.9 Results of ECCS Calculations j i- I Although it is recognized that the ECCS system when hooked to the domestic water supply I i

should be able to deliver an infmite supply of water, should the domestic water supply not l-be l available, the ECCS function will be supplied by the backup system, the auxiliary make-up l water system (AMUWS). Since this system has a limited water supply, considerations of a

l finite water supply with transition to air cooling were utilized in this calculation.

i I l l Using the preceding assumptions for the reactor core and for the temperature of the cooling ai

i. available in the reactor room, the TAC 2D code was used to evaluate the cooling requirements l

! l in order to maintain fuel temperatures below safety limits. Figure 13.3 presents the peak l and average fuel temperatures in the hottest fuel element during the air cooling cycle after spray I F cooling for time varying from zero to four hours (with a chunney height'of two feet). From l

' Figure 13.3 it may be noted that spray cooling for three hours will lower the resulting average l temperature in the hottest fuel element to 886'C, well below the safety limit of 930*C. From l l the discussion in the following sections, it will become clear that to maintain cladding iintegrit is really only necessary for the average temperature to be below the safety limit, since the colder I

' sections of the fuel will act as a sink for any free hydrogen released from the hotter sections.

l i

Figure 13.3 also illustrates that with a two foot chimney and slightly more than 3.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> of I

spray cooling, the peak fuel temperature in the hottest fuel element will not exceed the safety l limit of 930'C.

, l l

Figure 13.4 demonstrates the time dependent fuel temperatures (peak to average) during I the a i- cooling cycle with a two foot chimney after spray cooling for three hours. This graph shows I that the peak and average fuel temperatures reach a maximum at about 4.1 hours1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. While not I

shown in this graph, it is also clear that the six inches of the fuel in the hottest fuel element j (

which is cooler than the average fuel temperature (886'C) has temperatures that are far below l

, the applicable safety limit.

i I I

TAC 2D calculations were also made to illustrate the effect of chimney height on maximum

{

' and average fuel temperatures assuming three hours of spray cooling (or more in the absence l

of any chimney). These results are shown in Table 13-2a. Table 13-2b shows the spray l cooling requirements if the chimney height were three feet rather than two feet. As expected, l the performance of a three foot chimney is better than that for a two foot chimney. However, l the specifications for the location of the spray cooling nozzle (Reference 13.7c) suggests [

4

I Rev.I 4/14/97 1312h

'l 1400 1300 1200 '

N

\ Peak Fuel Temperature 1100

. \ /

1" 7

Safety Limit N

N

$ '....... .... 4.

\

.00 '

N l

Average Fuel Temperature N

700 11t. Chimney

^^^^

000 ^^^^ ^^ ^ ^ ^ ^^^ ^

^^ -^^^

0 0.8 1 14 1 2.8 3 3.8 4 Spray Cooling Time, hrs i

i MAXB1UM AND AVERAGE FUEL TEMPERATURE DURING AIR COOLING CYCLE FOR VARIOUS SPRAY COOLING TIMES l l FIGURE 13.3

- J

13 12i Rev.1 4/14/97 l

I 1200 i

[

Maximum 1000 9 . ; _ . . _ . l/. g . e .: . _ . . _ . .

Safety Limit #! ,.**"~.............

j , y soo '

.u kl i/ ,

,a' Average

/ ,' I .

I soo / ,!

i

/*

• / ,'

b m f ,'

/,.

4oo  ! .'

I,' i I*'

2 fL Chimney

~  !

I O

O 1 2 3 4 s Air Cooling Time, hrs MAXIMUM AND AVERAGE FUEL TEMPERATURES l FOR THE HOTTEST FUEL ELEMENT AS FUNCTIONS OF TIME l AFTER END OF SPRAY COOLING FOR THREE HOURS l FIGURE 13.4 i

.,./*

l Rev.1 4/14/97 13-12j

.l l

l T(hr) Chirnney ~,

l spray cooling T,,,

(ft) , C)  !

('C) l 3 l 3 900 l 3 813 2 982 886 (MNRC) 3.7 2 930*

3 845*

1 >l220** >l112**

3 0 >l451** >l355**

10 0 >l209** >l122**

l 48 0 953*** 873***

I

• fuel temperatures takenfrom Figure 13.3 l *

• temperature stillinsing aper 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> ofair cooling l * *

• terusperr w peaks at about 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br /> and then decreases I

l Table 13 2a Maximum Fuel Temperatures with Various Chimney Heights

{

\ T(hr) Chunney T, l spray cooling T.,,

(ft) ('C) i

('C) l 2 3 941 850

{ 3 3 900 813 l 3 2 982 886 l

l Table 13-2b Comparison of Cooling Results with 2 ft. and 3 ft. Chimneys l

jl suggests that the nozzle should be 26 inches above the top of the core. This height is I

compatible with a two foot chimney, but could be a problem with a three foot chimney.

l Finally, additional elements of conservatism not mentioned earlier in this analysis provide furthe I assurance that the cooling of the fuel will be at least as effective as described above. For l

instance, no account was taken of the added cooling provided by the conduction of heat from l - - the fuel elements to the cooler portions of the fuel assembly, Similarly, the radiation of heat l

l (especially when the fuel temperatures have reached the higher values) to the cooler parts of th system outside the core was not included in the transient heat flow considerations.

I Furthermore, the TAC 2D calculations reported herein assumed that the temperature of the inlet

1. air at the bottom of the core was 300*F. This is somewhat higher than would result from the l considerations in sections 13.2.3.2.2.6 and 13.2.3.2.2.8. With those earlier results, the core I inlet air temperature would be about 260*F (138'F [ mixed air in the reactor room] plus 122*F l

j

[AT from additional mixing in the tank]). If all the hot plume were to rise into the reactor l

room, the mixed air temperature in the room would be about 180*F. In this unlikely case, the l

core inlet air temperature would then be about 298'F (180*F [ mixed air in the reactor room]

{

plus 118'F [AT from additional mixing in the tank]). Both of these estimates of the core inlet air temperature are less than the more conservative value of 300*F used for the TAC 2D l calculations performed here.

.,.ia 1312k Rev.1 1/14/97 I

132.32.210 Cladding Stress Analysis i

I In Figures 13 3 and 13,4, it is shown that spray cooling for only three hours with a two foot I

chimney will assure that the average fuel temperature in the hottest fuel element will not exceed l

886'C, although the corresponding peak fuel temperature will reach 982'C. Figure 13.5 l presents clad strength and applied stress from equilibnum hydrogen dissociation Ipressure plus any other gas present within the clad as a function of fuel temperature. Early in theI fuel life, there is residual air backfilling bu' relatively little fission gas. Both the nitrogenI and oxygen form metal compounds afier the fuel has been operated at full power for a lperiod of tim in the effective fuel life, the air disappears as a gas leaving only hydrogen and fission I gas.

I For Figure 13.5 to be valid, all the fuel within the clad must have the same temperature l and be at the same temperature as the clad. In this case, the Safety Limit is the crossover of the Clad l

Strength Curve and Gas Pressure Curve. Except for a time duration very early in Ithe fuel element life before the air has been absorbed, the Safety Limit is about 930'C. In a real fuel l

element during a LOCA much of the fuel has a temperature lower than the peakI temperature For this case, the excess hydrogen gas from the hotter portions of the fuel element will 1

disappear into the sink created by the cooler portions of the curve.

I l

In the current example using three hours of spray cooling and a two foot chimney, l the hottest fuel element ranges in temperature from 603 *C at the bottom to a peak temperature of 982*C I

near the top. The average fuel temperature is 886'C with the bottom six (6) inches of the 15 I

inch fuel having fuel temperatures considerably less than the average fuel temperature. I The c temperature is a few degrees (6-8'C) cooler than the adjacent fuel temperature. There is thus a l

small area along the clad in which the clad temperature reaches 975 *C. The icurve in Figur 13 5 shows that the resulting clad strength in this region of the clad is about I 35 Mpa. Th temperature of the fuel slowly rises (over a 4-hour period) from 100*C to 980*C.

l Consequently, the excess hydrogen gas due to dissociation has time to be absorbed i in the c fuel sections without raising the pressure substantially above that characteristic of the cooler i

section (603'C-700*C) of the fuel element. The resulting gas pressure will be less than 0.8 l

Mpa (700*C) which is considerably less than the 35 Mpa strength of the hot clad. There is th i

no danger of clad rupture during the ai cooling portion of the LOCA scenario when the fuel is 1

previously spray cooled for only three hours. However, as a further conservatism, the MNRC l

will spray cool the fuel for a least 3.7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> and will therefore experence an average 1 fuel temperature in the hottest element of approximately 845'C and a maximum fuel temperature l of 930*C during the air cooling phase (See Figure 13.6). Since these temperatures I are lower th thcac used in the preceding example, it is clear that at the MNRC there isI even less da clad rupture during the air cooling portion of a LOCA l

l 13.2.3.2.2 11 Ground Water Contamination i  ;

As a w; ult of activation ofimpurities in the primary cooling water, the water will contain small amounts of radionuclides depending on reactor power, reactor operating time and time since reactor shutdown To characterize the radioactivity expected to be present in the MNRC primary coolant at 2 MW, measured values for the predominant radionuclides were adjuste

, . s ,~

l Rev.1 4/14/97 13 121 1000 100 A

04 SS UI.TIMATE STRENGTH} ,... ..... . ,..:.. ,..

10 -

....-.,t....--

I_

... .s.. .... .

. Mgg.. air and Anmon gas ,; ;. .. '

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. . :. . . ..i... . . :. .

. . . . ..n.

nl .

g. . .

.... . . .. .. r .. .

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. . ... ..... . .,. . ..... ... . . . . . Hydmgen and Reaien gae . ..

toI _

.... ..s

. ,.c . . . . ... .. . . ... . ..... . . . . .

tool .

. ... ..c .. . ... . ..s

..i .

.. ... . ~ . . . . ... . . ... . ,

. .%.. .. o.. . .%.. .. ...

...... . . . . . . ..s.. ... ..a..

. . .. . . ' . . . ..t... . . ' . . , . 8.. ... ..

..i... . . . ' . . .

.. .. ..d. . .g ..

.. .g .... . ... ..... . . ... . . .,,,,

M M 390 M M tm Dagenamvc ,

1 l CLAD STRENGTH AND APPLIED STRESS RESULTING FROM EQUILIBRIUM I HYDROGEN DISSOCIATION PRESSURE AS A FUNCTION OF TEMPERATURE FIGURE 115

l e

.,O-1 13-12m Rev.1 4/14/97 l

. l l

1000 l

- *-~

s '

p , ... --

800

/ ,,

! * ~

700 / -

Ay,r80s of the 1 '

Houset Element

/ ,*

u 300 -

e , i t

j

.00 -

> *o j

]

S U // a ti. chimney

' , ,. 3.7 hre Water Coeung

,!l'

' 100

• C inNief Fuel Tempereews See l*

-f 20e j-  !

o P

P 100 --

l 0 -

O 1 2 3 4 4 MAXIMUM AND AVERAGE FUEL TEMPERATURES l FOR THE HOTTEST FUEL ELEhENT AS FUNCTIONS OF TIhE Ar IcR l END OF SPRAY COOLING FOR 3.7 HOURS l FIGURE 13.6 l

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