ML20043A041
| ML20043A041 | |
| Person / Time | |
|---|---|
| Site: | University of Missouri-Rolla |
| Issue date: | 05/08/1990 |
| From: | Bolon A MISSOURI, UNIV. OF, ROLLA, MO |
| To: | Alexander Adams NRC |
| References | |
| NUDOCS 9005170239 | |
| Download: ML20043A041 (29) | |
Text
.. e
-e W Nuclear Reactor FacNv II UNIVERSITY OF MISSOURI ROLLA Nuclear Reactor Rolla MO 05401 0249 Telephon* (314) 341-4236 May 8, 1990 i
Public Document Room U.S. Nuclear Regulatory Commission ;
Washington, D.C. 20555 Docket No. 50-123
Dear Mr. Adams:
It is my pleasure.to inform you'that Mr. David W. Freeman has reported to work as the Reactor Manager at the UMR Reactor effective Tuesday, May 1, 1990 We feel that his prior experience at the University of Virginia Reactor and in industry should be of great benefit to the facil-ity. He had a senior operators license at UVAR and will become licensed at UMRR as soon as reasonably possible.
He has been informed that he is the third person to serve as UMRR's Manager in the 29 year history of the facility.
Sincerely, hc b Albert E.-Bolon Director l
l AEB/lp copies to: Alexander Adams, Jr.
A. Burt Davis - Region III Don L. Warner Signed before me this 8th day of May, 1990.
DLWL h% %u$ A.t~LM Notary Public
/
Deborah M. Middendorf Phelps County Missouri My Commission Expires 10/31/91
, 239 90050s
, Ihxxosoogg,we .,e-_,,_,,_
go
'It
s i
i 0 t F
h RESPONSE !
I TO REQUEST FOR ADDITIONAL INFORMATION l
t t
University of Missouri-Rolla Reactor (UMRR) ,
f-Facility License No. R-79 Docket No. 50-123 l
l i
l.
l l
l May 8, 1990 l
l i
t
o 1 l
)
- 1. The Irradiation Fuel Element described in Section 3.2.5 of the SAR has blank aluminum plates in positions 10 and 17.
There are no plates in positions 11 through 16. Positions 1 through 9, and 18 contain fuel bearing plates. The first sentence on page 3-19 of the SAR has been revised fer clar-ity.- Revised page 3-19 is provided on the following page of this response.
I
E 3-19 !
element, except that the fuel plates in positions 11 through 16 ,
Rev. 4 '
have been removed and the fuel plates in the positions 10 and 17 ;
have been replaced with aluminum blank plates. This leaves a l space about 3.5 cm (1.4 in.) wide to accommodate various samples for irradiation. A-computational analysis using a diffusion code (7) has been carried out to investigate the power distribution in y the vicinity of the irradiation space. To obtain the most con- f servative results the irradiation fuel element has been placed in the position with the highest neutron flux and adjacent to-two ,
control rod fuel elements. The total power peaking f actor was calculated to be 2.4 which is only slightly higher than the power peaking factor in a control rod fuel element. A further detailed l i
discussion is given in Sec. 3.4.7. ;
3.2.6 Core Support Structure i
The reactor core is supported by an inverted aluminum tower assembly suspended from the bridge which spans the pool as shown in Figures 16 and 17. The bridge is made of structural steel, approximately 3.3 m (11 ft) long and 1.35 m (4.5 ft) wide and is a wheel mounted on tracks located parallel to the long axis of the reactor pool atop the pool walls. The bridge can be moved along !
its rails for a distance of approximately 1.8 m (6 ft) from its
- normal operating position, thus providing water shielding between !
the experimental facilities and the reector core when required.
Mechanical stops are provided on the bridge rails to limit bridge travel within the pool area. An inadvertent movement of the reactor bridge causes the reactor to be-scrammed (see Section 1
3.5.8).
L i l
l
. 3 Response, cont.
- 2. Regarding the " negligible" temperature drop across the fuel meat and the aluminum cladding mentioned in Sec. 3.4.7 of the SAR, page 3-28, the temperature drop between the center of the fuel meat and the surface of the fuel element has been calculated to be 0.054 CO (or 0.10 FO ).
- 3. The " amount of heat stored in the fuel, meat and cladding" mentioned in Sec. 9.3 of the SAR, page 9-4 simply means the heat capacity of the fuel par ER, which is the sum of the products of the respective specific heats, masses and temper-ature changes between normal operating temperature values and those at the time of concern, that is following a loss of.
coolantacogdent. An order of magnitude calculation shows that 4.6x10 cal (or 1.8x104 Btu) would be stored in the fuel during heat up after a loss of coolant accident. (Because of the uncertainties involved in calculating that term, no credit for it was taken in the accident analysis.)
- 4. We gladly re-instate (Attachment A) the previous fuel-handling accident (as presented in the 1984 SAR submission).
Please note, however, that several fuel data values for the LEU fuel are different from the previous ones for HEU fuel as presented in Table XV on page 9-16.
- 5. The analysis presented'in SAR Section 9.7 addresses the failure of a fueled experiment which is ;. *eximum hypothet--
ical accident for the UMRR. Because the Ls Snt represents the design basis accident, we feel it is appa .ciate to apply 10CFR Part 100 criteria to postulated releasek.
It should be noted that the analysis presented is unchanged from the previously approved SAR and was deemed' appropriate by NRC at that time. Additionally, the analysis is not impacted by the conversion from HEU to LEU fuel. We have intentionally tried to limit the scope of SAR revisions to information concerning the fuel conversion.
6.
- a. See the revised pages 7 and 10 in Attachment B. Compliance with shutdown margin and excess reactivity criteria will be demonstrated as discussed in Attachment C, " Loading and Characterization of the LEU core.
, 4 -
Response, cont. __
- b. The licensee is not certain that any specific changes need to be made to the safety limit. The melting point.for pure aluminum is 12200 F (or 6600C) according to the American Society for Metals, " Metals Handbook", (1948). The solidus -
temperature for'the aluminum alloy 6061 is 1080 0F (or 582 0C),
according to the American Society for Metals, " Metals Hand-book (8th edition)", Vol. I (1961). Perhaps it was not clear -
that the aluminum alloy values were the ones being used, rather than pure aluminum. ,
- c. That was a typographical error which we thought had been
-corrected. The value should be 1700 m3 (See the revised specification 5.1.2.)
- d. See the revised pages 35 and 35A (Attachment B), which more fully describes the fuel, as well as the half fueled and the new irradiation fuel element.
- e. See the revised specification 5.3.4 (Attachment B).
- f. Revised Specification 5.4.1 is applicable to either the current HEU fuel elements or to the future LEU fuel elements, but not both simultaneously. We do-not anticipate having any difficulty in terms of defective fuel in the new LEU fuel --
elements. However, there will be two empty spaces in the racks in the storage area even after all of the current HEU fuel has'been placed there.
s should a generic problem arise with the new LEU fuel, storage racks could be designed and fabricated to hold that fuel in the open portion of the reactor pool. Any spare new LEU fuel -
elements will be kept in dry, secured storage prior to their being placed in the core.
- 7. The Physical Security Plan needs to be revised in order to '
show that the TRIGA fuel has been shipped away and that the new LEU fuel may be present in the Reactor Facility for up to one year before the HEU fuel is shipped off. Table 1 on page 14 will be revised to reflect this.
After the HEU fuel has been shipped from the site the requi-rement in sentence one of Section 4. should have the phrase,
" including the night watchmen." deleted. (Rev. 4) And the first two sentences of paragraph two of Section 4.1 should be removed. (Rev. 4) That is - "The watchmen of the University Police conduct periodic physical checks of the facility on an irregular basis. The watchmen are not armed but are in radio
-. 5 Response, cont.
contact with the University Police office and the patrol cars." (Rev. 4)
It is felt that in the absence of HEU fuel that the current intrusion alarm system will adequately protect the facility and its contents.
- 8. We can find no correspondence nor a certificate that indi-cates that federal funds will be provided to accomplish the conversion to LEU fuel and to ship the HEU to its proper destination.
The U.S. Department of Energy has been contacted ~regarding this matter and have indicated that they will provide such written assurances.
- 9. The proposed schedule for the conversion of the UMR Reactor from highly-enriched uranium (HEU) fuel to' low-enriched uranium (LEU) fuel is as follows:
July 1990 receive U.S. NRC approval of revised Technical Specifications,_ Safety Analy-sis Report, Facility License and Physi-cal Security Plan.
June-Sept. 1990 finalize plans and procedures for the actual fuel conversion.
Sept.-Dec. 1990 contact appropriate agencies for cer-tificates and permits needed.
May 1991 receive new LEU fuel from manufacturer and store in dry storage at the Reactor Facility.
June 1991 remove HEU fuel from the core and place in fuel storage racks; load LEU fuel in.the core.
July-Aug. 1991 characterize the new LEU core.
Mar.-May 1991 ship HEU fuel from UMR Reactor to Savannah River Plant tentatively via a GE700 shipping cask in three separate shipments.
., 6 Response, cont.
- 10. The current limits on our Facility Operating License No. R-79 allow us "to receive, possess and use up to a maximum of 9.9 kilograms of uranium-235 at various enrichments and up to a maximum of 200 grams of plutonium-239 in the form of sealed plutonium-beryllium-neutron sources in connection with the operation of the reactor". Furthermore, "Without exceeding the foregoing maximum possession limits, the maximum limits on specific enrichments of U-235 are as follows:
Maximum U-235 % Enrichment 4.95 kg < 20 4.95 kg .h 20" Thus, the most-recent revision of license R-79, (Amendment 8, dated April 16, 1985) was written in anticipation of the conversion of the UMR Reactor from HEU fuel to LEU fuel, and should not require any further revision in that regard.
- 11. Attachment c presents plans for fuel loading, approach to criticality, final core loading, control rod calibration, and power level determination.
i
o :
I 1
i a
,l l
3 d
Attachment A. Section 9.6 of SAR " Maximum Reactivity Insertion" i
t m
1 1
'l i
1 i
_ _ __ _ . _ . . _ _ . . ~ . _ _ _ _ . _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ . _ _ . _ _ _ _ . _ _ .___ ___ __- _
9-13 A-1 9.6 Maximus Reactivity Insertion In this section mechanisms which could give rise to a large reactivity increase (such as p > 8) are analysed in order to identify the maximum reactivity insertion f or which a saf ety 1
analysis needs to be perf ormed. The Technical Speci fications specif y the excess reactivity f or the UNRR as follows:
The fuel loading shall be such that the excess reactivity +
above the reference core condition will be no more 'than ,
1.5% delta k/k, except that the excess reactivity may be increased up to .a maximum of 3.5% delta k/k f or the . ,
I purposes of control rod calibraton only. This. increase in excess reactivity above 1.5% delta k/k will be permitted no (
more than twice a year and- f or no more -than five i consecutive working days' each time. The reactor shall be operated only by_ a li. censed Senior Operator when the. excess reactivity is greater than 1.5% delta k/k.
i ;
I in spite of extensive staff. discussions- and literature i research no credible accident scenario has been f ound which could possibly lead to a sudden release of excess reactivity-l larger than 1.5% delta k/k. Therefore, an instantaneous 1
insertion of the excess reactivity largek than 1.5% delta k/k has been excluded from f urther analysis. j i l c 1 l
! Experiments at the Curtiss-Wright Research Reactor (11) l t
have shown that the worth of a fuel element at the core i
. 1 periphery is less than 1.5% delta k/k. This is consistent with i
l experi ence at the UMRR gained with different core i i'
configurations. Depending on its position at the core periphery I
a standard fuel element can be worth between 0.5% and 1.5% delta k/k. -(The reactivity worth of a f uel element within the reactor l
u_i_____________.______m. .,__,.___,.._..,.._._,..._,....._...,_._.,...,_r -
- 9014 g_p cor e i a not well known since the Technital Specif acatiens ;
i preclude reactor operati on with an empty internal lattice j 4 posttion.)
A hypothetical accident can be postulated assuming that a fuel element has been placed next to a barely suberitical core, i
thus resulting in a positive step reactivity insertion of 1.57. !
delta k/k. 'A sudden reactivity insertion of such a magnitude would cause the reactor to become prompt critical with a ,
l subsequent exponential power increase. The reactor period at the beginning of the prompt critical power excursion can be approximately calculated (14) from the expression TL - #*
P, - 8 .
l where ip = prompt neutron lif etime (for the UMRR i t - i s 4. 5 x 10~'s) 8 = delayed neutron f raction l Using B = 0.0075 the reactor period corresponding to the above
?
postulated reactivity insertion is 6 es.
In the analysis of short power excursions the total energy release and the resulting maximum f uel plate temperature are two ,
of the most i mportant physical parameters. In order to establish a relationship between those two parameters for the ,
accident investigated in this work, a comparison with
- experi mental data was sought. Fortunately, a- large coll ecti on 4 of data from the excursion experiments perf ormed at the BORAX and SPERT f acilities is available. Especially, some of the _,
l-
9 15 A-3 SPERT-3 experiments using the DU-12/25 core are applicable to the analysis of the UMR Reactor since the fuel geometry and composi tion are very similar (9). A detailed comparison is given in Table XV.
A series of self-limiting power excursion tests was carried out in BPERT-1 using 5 c ore loadings. The input variable commonly ref erred to in these experiments was the reactor period i nduced by a stepwise reactivity insertion. At periods of the order of 6 x if* to 9 x 10~' sec the tests have shown some plate buckling and a ripple pattern due to thermal expansion stresses in the plate (10). During shorter period transients the plates appeared to have sof tened and remained in a plastic state f or several days. From the tests it was conc 1'uded that the mechanism responsible f or self-limiting the' power excursi on consi sts of fuel and moderator _ thermal expansion and boiling.
(The latter being the dominant shutdown mechanism.)
There was one experiment in which the reactor period was 6 x 10 -8 sec. The total energy released in the excursion was 13.2 MW-sec. Dnset of the self-limiting mechanism occurred when about 7.2 MW-sec of the thermal energy was generated. No damage to the f uel cladding was observed. The maximum fuel surface temperature recorded was 5600C (10400F) which' is well below the aluminum cladding melting temperature of 6600C (12200F). 1 t Thus, f rom the results of experiments with various stepwise reactivity inserti ons it can be concluded that the
A-4 9-16 .
Table XV. Comparison of Important Fuel Data I
UMRR UMRR SPERT-1 ,
Geometry Plate Plate Plate Rey, 4 LEU HEU Length (cm) 61 61 61 ,
Width [cm) 7.6 7.6 7.6 Thickness (cm) 0.15 0.127 0.15 i Water gap (cm) 0.49 0.31 0.45 Fuel :
Material U 38 0 -Al U 3 Si 2-Al U-Al Enrichment (%) ~90 19.8 100 .
l Weight fraction of U 0.36 0.48 0.24 i Thickness (mm) 0.51 0.51 0.51 l
Cladding Material A1 6061A1 A1 Thickness (mm) 0.51 0.38 0.51 ,
4 9
. 9 17
-A-5
.above-postutsted accidont- eeould -be saf ely tersinated by: this l se) f-limiting shutdown mechaniso.' This 1e a rather surprising f
resul t . Homeever ,' the short. time constant of'the thin fuel plates allows' a large amount of energy to - be transferred.'into- ;
the- water. channels even during very short reactor ~ periods. -!
Consequently, boiling becomes the rapid and dominating -shutdown 1
i factor. Such an accident can, thereiore, be terminated even if i F the safety instrumentation, e.g. both power . safety channels, !
were inoperable.
In spite of this UNRR saf ety f eature, administrati ve steps have been established in the Standard Operating Procedures which are designed to prevent a fuel handling accident:
r W
(1) All. fuel handling is done in accordance with written procedures.
l.
i !
1 (2) Loadings are planned to include the sequence of loading:end positions of individual elements. Also a loading schedule .
is prepared prior to commencement of loading. l u !
l '~ (3) Loading operations are done by qualified personnel under l
E the direct supervision of a licensed- Senior- Operator. :
l l (4) Fuel handling tools are kept locked with the keys secur ed to prevent unauthorized movement of f uel . f (5) Loading of the core is done from the inside to the I
'i L
L 1 L :
9-17b.-
i .. .
A-6 .j 1
l-p
.outalde.- l
'1 l- Finally, it should be pointed _out that' the assumptions ;
l L leading to this accident are very unlikely, and therefore it is I L .
not believed that such an accident would ever- happen. .The analysis, however, has been qui te usef ul in showing the inherent-4 safety- capacity of- the UPRR. Therefore,-lno effects on the health and saiety.of the.public~nor on the reactor staff are to ;
be expected from this type of accident.
i k
i i
I 3( 1 1
1 1
, -, ~ , ~,, , - - -- .. ,, . . . _ , - - . , , . , - , . . - -.,,-,n -
L;
>1 Attachment B. " Revised Portions of University-of Missouri-Rolla 1
' Reactor Technical Specifications" 1
.i
}
'I j
7 l
i i
l
, i 3
a q
iG
. 1 4
, , . ,I.
' B- 1 '
b d
APPENDIX A
.i
' FACILITY LICENSE NO. R-79' .l 1
TECHNICAL ~ SPECIFICATIONS ;
. AND BASES ,
FOR'.THE' UNIVERSITY 0F MISSOURI-ROLLA REACTOR'
, DOCKET NO. 50-123' i ..
)
L'
.. 1 Rev. 4, May.4, 1990-t i
i f.
r-
) f l.
s
,. B-2 6
. 1%1 delta k/k in the reference core condition and the reactivity worth of all experiments:is accounted for.
reference core condition - when the core is at ambient tempera-ture and the reactivity worth of xenon is negligible (<0.21%
delta k/k).
I regulating rod - a low reactivity-worth control rod used pri-marily to maintain an intended power l'evel. Its position may..be 0 varied either byfmanual control or.by the. automatic servo-controller.
reportable occurrence - any of the conditions described in sec-
-tion 6.5.2 of these specifications.
safety channel - a measuring or. protective channel in the reactor safety-system.
t safety limi6s (SL) - limits on important process variables which are found to be necessary to reasonably protect-the. integrity of certain physical barriers which guard against the uncontrolled ;
release of radioactivity.(6) (The principal physical. barrier is -
the fuel cladding.) {
scram time - the elapsed time between reaching a limiting safety Rev.-4 system set point and the time when a control. rod =is fully; j inserted. '
i secured experiment - any experiment,.experimentalefacility, or component of an experiment that is held.in a stationary position.
relative to the reactor by mechanical means. =The restraining forces must be substantially greater thanithoselto which the experiment might be subjected. ,
senior operator - an individual who is licensed to direct the activities of reactor operators. Such an individual is also a- i reactor operator. !
.. B-3 1
7-t
~
shall, should and may - the word "sha11"'is used to denote a requirement; the~ word "should" to denote.a recommendation; and >
the word "may" to denote permission, which is neither a require- ,
I ment-nor a recommendation.
shim / safety rods - high reactivity-worth rods-used primarily to ,
- provide coarse reactor control. They are connected electro-magnetically to their drive mechanisms.and.have scram capabili-ties.
shutdown margin --the minimum shutdown reactivity-necessary-to i
provide confidence-that the: reactor can"be made suboritical by 4
' ~
means of the control and safety systems starting from any per-missible operating' condition with the maximum worth scrammable :
rod and any non-scrammable. control rod in.their. fully withdrawn Rev. 4 positions and that the reactor will~ remain subcritical'without further operator action.
startup source - a spontaneous source of neutrons which.is used :
, to provide a channel check of the startup (fission chamber) channel.
surveillance time intervals -
two-year (interval not to exceed 30. months).
annually (interval not to exceed 15 months).
semiannually (interval not to exceed 7 1/2 months).
l quarterly-(interval not to exceed-4 months).
- monthly (interval not to exceed 6 weeks).
, weekly (interval not to exceed 10 days).
- daily (must be done during-the working day)i 4
true~value - the actual value of a parameter.
unscheduled shutdown - any unplanned shutdown of the reactor ,
caused by actuation of the reactor safety system, operator error, equipment malfunction, or a manual shutdown in response;to con-' !
ditions which could adversely affect safe operation, not includ- -
ing shutdowns which occur during testing or check-out operations.
., B-4 t
10
- 3. LIMITING CONDITIONS FOR OPERATION 3.1 Reactor core' Parameters
- Applicability:~ These specifications apply to the reactivity condition of the. reactor and the reactivity. worths of control rods and experiments.
Qbjectives: To~ ensure that the reactor can be operated safely l and to ensure that it can be shut down at- all times.
Soecifications: The reactor shall not be. operated unless the- >
following conditions. exist:
i.
(1).The fuel' loading shall be such~that the_ excess reactivity :;
above the reference core condition will'be no more than 1.5%
delta k/k, except that the excess reactivity may be increased up to a maximum of-3.5% delta.k/k.for purposes of control rod calibration only. This increase in excess reactivity'above 1.5% delta k/k will be permitted no.more than twice a year
! and for no more than five consecutive working' days each time.
i The reactor shall be operated only by a. licensed Senior Operator when the excess reactivity is greater than 1.5%. '
(2) The reactor shall be operated only when all-lattice positions I
j internal-to the active fuel boundary are occupied by either a fuel element or control rod fuel element or by:an experimen-tal facility.
l (3) The minimum shutdown margin under.any condition.of operation l with the highest worth control' rod and any non-scrammable p,y, 4
- _ control-rod fully withdrawn shall-be no less than 1.0% delta-l k/k.
L (4) The regulating rod. shall- b6 esof th no more than 0.7% delta k/k >
in' reactivity.
1 I
,e -
,- B-5 34
.5. DESIGN FEATURES only those design features,of the facility describing materials of construction and geometric arrangements, which if altered or modified would significantly affect safetyL(and which are not included in sections'2, 3 or-4 of the. Technical Specifications),
are included in this section'.
The Safety Analysis Report contains the details necessary for establishing criteria for;the!following Technical Specifications.
5.1 Site and Facility Description' -
5.1.1- The Nuclear Reactor Building-is located-on the east side.-
of the University of Missouri-Rolla campus in-Rolla, Missouri, near 14th Street and Pine Street.
5.1.2 The reactor is housed in a steel-framed, double-walled i aluminum building designed to restrict leakage. All air q and other gases will.be exhausted through vents in the 4 reactor bay ceiling 30 feet (or ll meters) above grade.
The Reactor Building free volume is approximately 1700 Rev. Al i cubic meters.
5.2 Reactor Coolant System !
4 5.2.1 The minimum temperature of the reactor pool should be no less than 15.50 C (600 F) when the reactor is operated.
5.3 Reactor Core and Fuel 5.3.1 Core Configurations Various core configurations may be used to accommodate exper- -
i iments, i
l 5.3.2 Fuel Elements i
i
... B-6 .h 35 l (1) Plate fuel elements of the MTR typeLare used. The overall-dimensions of each element are approximately 3 inches'by 3 j inches by 36 inches. The. active length of fuel is approxi- i mately 24 inches and the fuel is clad in aluminum alloy. The pey,4 ,
, fuel elements have 18 fuel plates joined:to two side plates. ,
The whole assembly is joined at the bottom to a cylindrical nose piece that fits into1the core grid plate.
The low-enriched uranium (LEU) fuel meat.is U Si2 3 dispersed-E*V*N in an aluminum' matrix and is enriched to.approximately 20% ,
^
U-235. The U.S. NRC has approvedithe specific design;of the.
LEU fuel elements.in NUREG-1313 and'that isothe basis'for-converting'to the LEUtfuel.
(2) There are control-rod fuel elements which are similar to the i elements described in (1) with the exception that'the center, eight plates have been removed andshave been replaced with l Rev.4 guide plates such that the control rod cannot:come in contact with fuel plates.
(3) There are half fueled elements, which hav- nine LEU fueled RevA plates (either the front ones or:the reat ,nes 'eus appro-priately marked) and nine dummy (or unfueled)' plates.
(4) There is an irradiation fuel' element which haslsix fuel - Rev.4 plate positions left blank (that is plate positions 11 <
through 16), plates 10 and 17 are dummy (or unfueled) and all l the others (1 through 9 and 18) are LEU fueled.
5.3.3 Control Rods (1) Poison sections of the three shim / safety rods are stainless j steel and contain approximately 1.5% natural-boron. The rods; I have nominal dimensions of 2-1/4 inches by 7/8. inches in' cross section and'are 29 inches long. -
n
. (2) The poison section of the: regulating rod is a stainless' steel- U oval-shaped tube, 25 inches.long, having a wall thickness of ,
65 mils, and is mechanically coupled to.the rod drive.
r l
l t
t g w a . ,a -
4
- - i
+ B-7
[
35A' l
5;3.4 Control Rod Drive Mechanisms f 1
L (1) The shim / safety rod drives-have a maximum vertical travel of 24 inches and'a withdrawal rate-of approximately 6-inches per minute.:The shim / safety; rods are magnetically-coupled to'the p ,y,4 l drive mechanisms and drop.into the-core, by gravity, upon a:
l scram' signal.
l-(2) The regulating: rod drive.has a maximum vertical travel 2of 24 L inches and a withdrawal' rate of'approximately 24 inches _per '
minute.. The regulating rod- is mechanically _ coupled _to its Rev.4- .>
l rod drive and'does not respond to a_ scram signal.
l (3) Lights.are provided on the operator's console to indicate r l-(Go to page 36.)
L -
1 l I
?
-i.
i 7
(
3 I
.- B4B '
36 upper limit, lower limit,--shim range, and magnet contact'for each shim / safety rod.
5.3.5 Start-up Source
-A neutron source is available of.such a strength as to satisfy' the' requirements that the. count rate is greater than 2' counts per' second during a cold reactor start-up. !
5.4 ~ Fissionable Material' Storage ,
5.4.1- The fuel storage pit, which is located below th'e floor of the reactor pool and at the end opposite from-the core,-will;be capable of storing _the complete fuel inventory of either highly-enriched uranium 1(HEU). fuel or of low-enriched uranium,.but not both. The neutron
~
multiplication factor of the fully' loaded storage pit
~
shall'not. exceed 0.9 under any conditions.-
-l l
l l
4
'j l
l 1
l l
.t-e 4
h l
i.
f Attachment C. " Loading and Characterization of LEU' Core"-
t
. t
-h f
V b
w w b F - - t v ~
- s. C-1 --
PLANS TO UNLOAD THE HEU FUEL AND LOAD THE LEU FUEL IN THE UMR REACTOR CORE Unloading thefHEU fuel will be performed by the following steps:
- 1. Review the Tech. Specs. and SOP's in preparation of unloading.-
fuel.
- 2. Conduct a reactor staff meeting to discuss the planned oper-ation and to delegate responsibilities.
- 3. Set up an auxiliary fission ~ chamber in.the fuel storage area and take several base-line background counts.
- 4. Perform the fuel unloading'in accordance with~ SOP 207, " Fuel.
Handling", while taking auxiliary sub-critical multiplication counts. Plot the 1/M rurve for the fuel' storage area as unloading proceeds.
- 5. Unload all of the half and standard fuel elements.before.
disconnecting the control rod drive mechanisms..
- 6. Disconnect th'e shim / safely rods and remove,s/s rods one at a time. Then remove the associated fuel element and put it in -j the fuel storage rack..
- 7. Remove the regulating rod and its-associated fuel element.
l Loading the LEU fuel to construct a new core will be done by:the following steps:
- 1. Review the Tech. Specs, and. SOP's in preparation of' loading fuel. i
- 2. Conduct a reactor staff meeting'to discuss the planned:oper-ation and to delegate responsibilities., .;
- 3. Set-up an auxiliary fission chamber near the core (at,'say position E9) and take several base-line' background' counts on
)
it, as well as on the. normal start-up fission chamber. a
- 4. Perform the actual fuel-loading in accordance with SDP1106,
" Critical Experiment Procedures", while taking sub-critical multiplication counts on both the normal start-up channel and ;
on the auxiliary fission counter. (Note: It may be neces-sary to activate tne <2 cps by-pass.) Plot the 1/M curves as loading progresses.
AEB 4/24/90 ;
-i
. /Mu
t
-.- -i C-2 Plans to Unload,fcont.
- 5. When'a critical mass has been achieved and the core loading _. t is as desired, take the reactor to a low power (say, 10W) and :i accurately record. critical rod heights,. check the reactor instrumentation,'etc. t
- 6. Shut the reactor down and secure'it according_to SOP 105.- >
t Before-operating the: reactor-on a regular' basis'go through_at '
least steps'1 through 7 of the " Characterization of the: LEU UMR Reactor Core".
Y j .:
i i
Y
1 1
, C-3 _;
i CHARACTERIZATION OF THE LEU UMR REACTOR CORE l The new LEU core will be characterized by the following steps: *
- 1. Load the fuel according to SOP 106, " Critical Experiment Procedures", plotting 1/M versus number of elements or mass of U-235 (225 gm/std. element). -Load the core while in the ,
W-mode to the nearest half element.
i
- 2. Perform' preliminary rod calibrations-using the rod-drop method, SOP 109..
- 3. Determine the excess reactivity of each rod based upon the ~4 positive period method. Determine the excess reactivity of the-core based.upon the sum of the excess reactivities of the control rods..
- 4. Determine the detailed integral' worth curve for the regulat-ing rod.
- 5. Determine the experimental value of?the effective delayed C neutron fraction, B egg.
- 6. IX) a power calibration following SOP 816.
- 8. Determine the excess reactivity oftthe: core and the shutdown margin (as defined in the Technical Specifications) toLassure l compliance with criteria ~ presented 11n Section 3 of the' Tech-l nical. Specifications.
I j 9. Determine the temperature coefficient, ET, in accordance with SOP 304.
- 10. Determine the void coefficient around the periphery of the core using SOP 303.
( 11. Determine the thermal and the total flux in the bare =and.the cadmium rabbit tubes using cadmium-covered and bare gold:
l i foils.
- 12. Determine the total flux distribution at the mid-plane of the core by using copper wires. 1 l 13. Determine the thermal and total-flux at the..mid-plane of the- 1 core in the positions at the periphery of the core.
AEB 4/24/90 1
t
~
m - .
4
. j: C4 s PLANS FOR. UNLOADING THE HEU FUEL
,; FROM THE UMR REACTOR FUEL STORAGE AREA AND INTO-A SHIPPING CASK ,
{
. Unloading the HEU fuel from the fuel storage area will be per- ,
formed by the following steps: .
- 1. Review the Tech. Specs, and SOP's in preparation of unloading fuel.
- 2. Conduct a-reactor staff Neeting to discuss the planned oper-ation and to delegate responsibilities.
1
- 3. Lower the single-element fuel transfer cask down onto the j
. bulkhead which divides the reactor' pool:from the fuel storage area. ' Release the tension on the cable' holding the-transfer..-
cask. With the manual fuel handling tool; move a: fuel-element- H from the. fuel storage rack-to the fuel transfer cask and 1 insert'it into the cask.
l
- 4. Raise the transfer cask to above'the pool wall'and' position H
it over the floor. ,
'l
- 5. Lower the transfer cask onto a-cart on the bay floor'.1 .
- 6. Roll or otherwise reposition the cart by the bay door'or onto the trailer holding the shipping cask.
- 7. Using a mobile crane transfer the fuel element in air into
( the uprighted shipping cask.
i
! 8. Repeat steps 3 through 7 until the shipping cask is filled I
with spent fuel.
! H l.
l l
i i
I AEB 4/24/90 <
k
.i
, . , . . , , -I