ML19341C541
| ML19341C541 | |
| Person / Time | |
|---|---|
| Site: | 05000124 |
| Issue date: | 02/25/1981 |
| From: | Curtner A, Parkinson T VIRGINIA POLYTECHNIC INSTITUTE & STATE UNIV., BLACKSB |
| To: | |
| Shared Package | |
| ML19341C538 | List: |
| References | |
| NUDOCS 8103030719 | |
| Download: ML19341C541 (60) | |
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O i t I f I e ANNUAL OPERATIONS REPORT j Nuclear Research Reactor Virginia Polytechnic Institut) and State University January 1, 1980 - December 31, 1980 by AlanP.Curtner,pupervisor Approved by i Dr. T.F. Parkinson, Director : [h .m 1 4 i I 4. I e 4 f I L ..
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1 i , TABLE OF CONTENTS page I. Reactor Operations ... . . . . . . . . . . . .. ... 1
, II. Tabulation of Cascheduled Scrams . . . . . . . . .. .. 1 , ;- III. Reactor Operator Changes . . . . . . . .. . . ... .. 1 l '
- IV. Personnel Changes . . .. . . . . .. . . . . . . . . .. 2 V. Quarterly Scram Time Tests . .. . .. .. . . . .. .. 2 l r
- VI. Health Physics ............... . ... .. 3 i VII. Control Rod and Fuel Inspections .. . .. . . ... .. 3 VIII. Amendment No. 4 to facility License No. R-62 . . .... 3 IX. 13 Plate Fuel Element Tests . . . .. . . . . . . .. .. 3
- 1. Reactor License R-62 Renewal . . . . . . . . . . . ... 3 XI. Facility Changes .............. .. .. . . 4 i
j XII. New Experiment - Shield Tank . . . . . . .. . . . .. . 5
, XIII. Equipment Failures ... .. . . . . . ... ... . .. 6 1
APPENDICES Page
- A. Amendment No. 4 to Facility License No. R-62 . . . . .. 7
- 3. Measured Reactivity Effects - 13 Plate Fuel Element . . . 21
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C. Back Flow Preventer Valves .. . . .. . . . . .. .. . 44 i i
*- D. New Experiment - Shield Tank . .. .. . . . . . . . . . 47 i
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l I. Reactor Operations Reactor operating parameters for 1980 were as follows: ( Quarter Jan-Mar Apr-June July-f igt Oct-Dec Kilowatt hours 22873 23997 14535 8873 Hours critical 268.5 282.3 230.9 123.4 l l Ar-41 Released, mC1 36894 38511 23328 14241 , I l Number of startups 59 52 51 43 l l? Unscheduled shutdowns 2 1 4 0 I: Yearly totals I Kilowatt hours 70278 Hours critical 905.1 Ar-41 released, mci 112974 number of startups 205 unscheduled startups 7 II. Tabulation of Unscheduled Scrams
- Momentary loss of power to the building (electrical storms) 6 Spurious 5 second period, electrical noise within the intermediate range instrument 1
, III. Reactor Operator Changes i Mr4 Leo Eskin successfully completed a senior reactor operator exam
' and was issued a license effective November 21, 1980. Mr. William J.
3ryan and Mr. Douglas H. Rockwell successfully completed reactor operator exams in October 1980. Mr. Bryan was issued a license effective November 21, 1980 and a license for Mr. Rockwell has been withheld pending a final
- review of his hedical Examination Report.
l l The following individuals have permanently lef t the facility: Mr. C. 3. Holland, June 18, 1980
- hr. J. L. Nelson, June 5,1980 r
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l. IV. Personnel Changes Mr. Christopher 3. Holland, who was emploved as a half-time staff reactor operator, per=anently lef t the facility in June 1980. Mr. Leo Eskin vro was also employed half-time, assumed a full time
, reactor operator position in June 1980.
I Mr. Eskin resigned his full time position in September 1980 to begin graduate studies; he plans to maintain his senior operator license while at Virginia Tech. f, Mr. Dennis Prater was hired in November 1980 as a full time senior 1 reactor operator, filling the position lef t vacant by Mr. Eskin. He is currently in operator training and a licensing exam is scheduled for April 1981. V. Quarterly Scram Time Tests (seconds) Quarter June-Mar Ap r-June July-Sept Oc t-De c Average Safety #1 0.46 0.42 0.48 see 0.45 Safety #2 0.51 0.53 0.53 note 0.52 Shim-Safety 0.43 0.42 0.43 below 0.43 Note: Oct-Dec scram time tests were not performed. i The tests were schedu.ed for December 20, 1980. The last reactor run for the year was performed on December 19, 1980. The reactor staff I was on vacation and the reactor shutdown for the remainder of the year. l_ The tests could not be scheluled earlier due to special testing with a 13-plate fuel element. (See secticn Ij[ of this report.) Annual in-core maintenance is scheduled for the first week of January 1981 and scram time tests will be performed after that evoluation and prior to the next reactor run. 3 L. 2 .
.-, - , a .- - _- , _ , - _ _ . - - . - . . , , . . - , m .
VI. Health Physics Area surveys and wipe tests were made on a minimum of a quarterly basis. There were no significant _aanges in observed radiation levels during the year. p- The only radioactive vaste released to the environment was Ar-41 through A the ventilation stack. The total amounts released per quarter are shown on the operations su==ary. VII. Control Rod and Fuel Inspections The control rods and reactor fuel rere inspected during March 1980. There was no evidence of cracking, pitting, or unusual wear or corrosion. Control rod worths, stringervorths, and reactivity input rates were measured. Nu significant changes were noted. VIII. Amendment No. 4 to facility license No. R-62
- An application for an amendment to Appendix A, Technical Specifications, license No. R-62 was submitted on September 17, 1980. The NRC granted a temporary amendment from Deceuber 1 to December 31, 1980 (see Appendix A).
IX. 13-Plate Fuel Element Tests Experiments with a 13-plate fuel element in various core locations were performed from Dece=ber 15 to December 19, 1980. Measured reactivity
. effects are su=marized in Appendix 3. A complete report will be available in first quarter 1981.
X. Reactor license R-62 Renewal Reactor license R-62 expired on November 16, 1979. .sn application for license renewal was submitted in a timely canner allowing continued . operation on the existing license until a review of license renewal i documentation is completed by the NRC. e 3 1
r 1
'/he reactor facility submitted the following documents to the NRC in August 1980:
A. Financial Qualifications of the Applicant
; B. Ensironmental Impact Appraisal C. Safety Analysis Report D. Technical Specifications E. Reactor Facility Emergency Plan F. Operator Requalifi:stion Program G. Physical Security Plan The NRC indicated by recent correspondence that a review of these documents would be conducted in early 1981.
3 XI. Facility Changes A. Back Flow Preventer Valves Two parallel BFP valves were installed on the 2 inch inlet cooling line to the reactor en July 30, 1979. Subsequent operational problems developed (Appendix C). The BFP valves were removed from the system on February 5,1980 by the Physical Plant Department and replaced with straight pipe, The ear.ter opersed '..1 problems have been alleviated. f
- 3. Reactor Control Console Changes
.- The following changes were made during February 1980:
- l. The secondary pressure low warning light has been disconnected l and replaced with a encondary pressure low annunciator alarm I
on the new annunciator panel. t l l l 3 l 4 \ a.
9
- 2. The negative pressure in the reactor room is sensed and an annunciator -
now alarms on the reactor control console. This alarm will detect a failure in the reactor room ventilation fans. There is a 30 second delay between loss of negative pressure and the p-annunciator sounding to provide for normal entering and exiting of the reactor room during reactor operation.
- 3. The water in process pit warning light has been disconnected and replaced with a water in process pit annunciator alarm on
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l the new annunciator panel. The alarm will sound when the sump in the process pit is approximately 75% full. C. Cooling Tower A Marley Cooling Tower model no. 8805A has been installed on the roof of Robeson Hall in November 1980 as part of the continuing 500KW power upgrade program. Piping and electric.al conduit instal-lacion was begun in December. The existing primary and secondary i systems will not be modified until complete review by the Reactor Safety Committee, and if appropriate by the Nuclear Regulatory Commission. D. Physical Security System A new intrusion detection system with an alarm at the Campus i i Security Division Office has been installed in December. This system fulfills new requirements of 10 CFR part 73 (73.67) Physical Protection of Plants and Materials. XII. New Experiment - Shield Tank A new experiment was approved by the Reactor Safety Committee and initial
.. measure =ents began in September 1980. The purpose of the experiments is to measure the integral spectrum in a " dummy" fuel assembly which consists of 5
_ -. .-. ,.,.-+.,,,.-..-,,y. -
L circalloy -4 clusters of UO3 (natural uranium) fuel rods contained in
... tubes. See Appendix D for details)
XIII. Equipment Failures Tabulation of Failures by Month Jan. - Regulating rod indication meter sticking - repaired I Regulating rod feedback control potentiometer - replaced Feb. - Power level chart recorder, pinched cable harness - repaired Chemistry Laboratory heating system pipe rupture, water entered nuclear instrument cableways, dried cables, replaced damaged connectors l May - Radiation Monitoring System, vacuum tube failure - replaced Power level chart recorder, pinched cable harness - replaced June - Solid State Relay failure, safety system - repaired Intermediate Range, compen volt. power supply - replaced Source Range preamplifier - replaced July - Source Range Fission Chamber - replaced Air Particulate Fission Product Monitor, vacuum pump - replaced Shim Rod up/down switch - replaced Sep.,- Safety Rod #1 and .2 up switch - replaced Nov. - Reactor Coolant Flow Monitor, transformer in amp. module - replaced Radiation Monitoring System, alarm reset switch - replaced f
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t 4 i I' i 9 4 r APPENDIX A Amendment No. 4 to Facility License No. R-62 t I a. 4 o S e 6.e 7 .
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l t i 6 Ds COLLEGE OF ENGINEERING VIRGINIA POLYTECHNIC INSTITUTE AND STATE UNIVERSITY NQw hisburg. l'i*ginia 2406l NuctaAa Encantaminc Caour September l7,1980 J' Mr. James R. Miller, Chief
! Standardization and Special Projects Branch Division of Licensing U.S. Nuclear Regulatory Commission Washington DC 20555 Re: License No. R-62 , Docket No. 50-124
Dear Sir:
In our preliminary analysis for operation at 500 Kw(t), we have determined that the present fuel element configuration will need to be modified to add more reactivity. Our current technical specifications limit us to operations with no more than 12 fuel places per element and with a maximus k (excess) of 0.6%. We request a temporary waiver of the technical specifications to permit tests of one 13-plate element without exceeding the limit on k (excess). Testing would be carried out at low power (< 1 kw(t) ) and should not require more than 30 days. If possible we would like to perform the low-power physics tests during the period from Dec. 1-31, 1980. The specific tests to be done would involve measuring the change in k gp when the 13-plate element is exchanged for a 12-plate element. Sub-critical multiplication measurements will be made during the approach to critical to insure e. hat we do not exceed the limit on k (excess). One or more dummy fuel plates will be substituted in the 12 plate elements as required. Since we will not exceed the k (excess) limit of 0.6%, we believe that no unreviewed safety hazard will exist. This measurement is classified as a "Nev Experiment"; therefore, it will be submitted to the Reactor Safety Cocmittee for their prior review and approval. The attached Safety Analysis indicates that if all twelve 12-plate elements were replaced with 13-place elements, the thermal utilization
! change, af /f, would be 0.0039. Assuming thot the non-leakage probability does not change, ak, , ak , ,
, Sterefore, on the average, sub-k k= f e
stituting one 13-place element for a 12-plate element would result in U e = 0.0039= 0.0325% k 12 e I For the exchange of a particular fuel element, the abeve average change would be modified by the statistical weight, 1 3
l 4 Mr. James R. Miller, Chief Septembe r 17, 1980 i Page 2 2 ij W= -- 2
,wherei ore is the average thermal flux in the core and j # core y,
$ is the average flux for a fuel element located at the (1,j)th j ebbrdinate. The objective of the proposed tests is to validate the 8- . calculation of the change in k with a 12-place element locate 8g atwhena13ggatoelement the (i.j) coordinateisposition.
exchanged If you have any questions about this request, please call me at ,! (703) 961-6510. Sircerely yours, _e ' { Vb' . G w < - T. F. Parkinson, Director Nuclear Reactor Laboratory Attachment TFP/nf1 cc: Reactor Safety Committee A. P. Curtner (Reactor Supervisor) Mark Embrechts s 4 9 4. I t u 9 .
N Safety Analy:1s
. . Reactivity Change with 13-Place Fuel Element Mark Embrechts 6
i s The method of calculation is that used in reference 1; we wish to { , calculate the change in thermal utilization, f, if all the 12-plate fuci elements are replaced with 13-plate elements. Let f and f' be jrespectively, I the value of f with 12-plate and 13 place fuel elements. f i Ia f og (r)dv Then f= y/,. (1) Is t .' p y(eldv + Ia; 2(r)dv 8 where the subscripts 1 and 2 refer to the lael and moderator, respectively.
.For a unit cell in the plate-type fuel asscebly, v, Ia 3
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f =1+ #
, F + (E-1) (2) 1 '#1 i r
k where F = 92 /# #
" <'t R1 e th e R y (3) j and E = <2 (R - R t) coch (R 2 -R)y (4)
In equations 3 and 4, R and R y are the half-thicknesses of the fuel plate
! and cell, respectively; i . 9 S *
< { = Ia /Dy and<}=Ia/D' 2 2
( Reactor Physics of UTR-10, Report ASEA-19, Advanced Technology Laboratories, June 1957. D4
, 10 -
I f* The values of the constants are collected in Table I. i $ i Table I r Thermal Utilization Constants
! Cons tant 12-Plate Elemen2 13-Plate Element l
R 1 0.040 in. (0.1016 cm) 0.040 in. (0.10?5 cm) s . R 2 0.240 in. (0.6096 cm) 0.220 in. (0.5500 cm) 2.6532 cm
-1 -1 f Ia 1 2.6532 cm 0.03857 cm
-1 Dy 0.03857 cm-1 0.022 cm -1
-1 Ia 2
0.022 cm D 2 0.7767 cm~ 0.7767 cm -1 c 0.1024 cm
-1 0.1024 cm g 0.0171 cm" 0.0171 cm-F 0.9997 0.9997 E 1.000025 1.000920 f 0.9601 0.9640 Thus the reactivity change if all 12-plate elements are replaced with 13-plate elements would be
^k ak k
= 0.0039 k, E t
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I, UNITED STATES f,s %, NUCLEAR REGULATORY COMMISSION 1 ; ,v y , vi g - 3 .j WASHINGTON, D. C. 20555 g y , s $, $ ~ (I s
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P i Docket No. 50-124 NOV 21 1980 Mr. T. F. Parkinson Director, Nuclear Laboratory Virginia Polytechnic Institute and State University Room 108 Robeson Blacksburg, Virginia 24061
Dear Mr. Parkinson:
The Commission has issued the enclosed Amendment No. 4 to Facility License No. R-62 for your Argonaut-type reactor. The amendment consists of changes to the Facility License and ic Appendix A of the Technical Specifications in response to your application dated September 17, 1980. This amendment allows testing of one 13-plate fuel element between December 1 and December 31, 1980, providing the power level remains below 1 KWt and the maximum excess reactivity is limited to 0.6% sk/k. This is a temporary amend-ment which expires on December 31, 1980. A copy of the related Safety Evaluation is also enclosed. Sincerely, t-James R. Miller, Chief l Standardization and Special Projects Branch I. Division of Licensing
Enclosures:
- 1. Amendment No. 4
- 2. Safety Evaluation cc w/ enclosures:
See next page Gd e
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t I 4 ! Virginia Polytechnic Institute ;- . .. l ccw/ enclosure (s): I [' Mayor of the City of Blacksburg City Hall
- Blacksburg, Virginia 24061 l t ;-
l Mr. J. B. Jackson , Comonwealth of Virginia Council of the Environment i l 903 Ninth Street Office Bldg. Richmond, Virginia 23219 ta k i i t 8
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- VIRGINIA p0LYTECHNIC INSTITUTE AND STATE UNIVERSITY 00CXET NO. 50-124 AMENDMENT TO FACILITY LICENSE
!~ Amendment No. 4 l License No. R-62 j 1. The Nuclear Regulatory Commission (the Commission) has found that:
A. The application for amendment by Virginia Polytechnic Institute and State University (the licensee) dated September 17, 1980, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Comission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in confonnity with the application, the provisions of the Act, and the rules and regulations of the Consnission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the common defense and 2ecurity or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
. F. Publica*. ion of notice of this amendment is not required since it does not involve a significant hazards consideration nor amendment of a license of the type described in 10 CFR Section 2.106(a)(2).
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. . . 2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, 8
and paragraph 2.C.2 of facility License No. R-62 is hereby amended to read as follows: . 2.C.2 Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 4, are hereby incorporated in the license. The licensee shall operate the facility
.. in accordance with the Technical Specifications.
1 3. This license amendment is effective as of the date of its issurance and shall expire on December 31, 1980. i' FOR THE NUCLEAR REGULATORY CCMMISSION
- 3//r, .
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James R. Miller, Chief Standardization and Special Projects Branch Division of Licensing
Attachment:
Changes to the Technical Specifications Date of Issuance: November 21, 1980 i 0 1. f I e Sm 15
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%u **"* .- Sn.'ETY EVALUATIOl? BY THE
, OFFICE OF tiUCLEAR REACTOR REGU'_ATION Supp0RTING AMEN 0 MENT NC 4 TO FACILITY LICENSE Q. R-62 p l, VIRGI?.IA p0LYTECHNIC INE ITUTE AND STATE UNIVERSITY COCr.ET .0. 50-124 Jn_troduction By letter dated September 17,1980, t* Virginia Polytechnic Institute and State University (the licensee) requ- ted an amendment to Appendix A of Facility License No. R-62. The request:d e end. ment would grant a temporary t waiver of the Technical Specifications n permit tests of one 13-plate element. The prop::ed testing would be carried c at led power (power level less than 1 KW:), the reactivity would not ax.:aed -he technical tpecification limit of i t.k/k of 0.6". t.k/k and should not require more than 20 days (1 to 31 December, 1980). The objective of the s:ecific tests involved is to measure the change
- , in Keff when the 13-;1 ate element is charged for a 12-picte element. These
- terts will provide data for preliminary analysis for operation at 500 KW, as i
tns present configuration will have to de modified by adding an additional
- plate in eac subassembly in ordar to operate at the higner power level.
{ Backcround i j The current Technical Specifications for the reactor cces (3. 4, No . 6.1.1 ) require that "standa-d fuel eleman s shall be of the fl-t-plate type with 4
, twelve plates to each element" and (p. 4, . :. 6.1.3) " Maximum excess ! reactivity above cold clean critical shall be limited to 0.6% ak/k including 1- positive reactivity frcm experiments." These mea urements are classified as a "New Experiment." One or more duc.my fuel plates must be substituted in the 12-plate elements as necessa.y to satisfy the excess reactivity 1 recuirement.
Evaluation ? To suppcrt the recuest for operation with one 13-plate subassembly, the
, acclicant pe-formed an approximate calculation of the reactivity char.;e (based on thermal utilization only) associated with a loading consisting of 12 subassemblies of 13 plates each. This calculation indicates that the total reactivity worth change should be less than 0.4%. This amendment will allow the determination of the reactivity addition due to one 13-plate D**D *D TY 16
element at various positions in the core. The licensee will remove the appropriate number of plates elsewhere to counteract the reactivity increase. In addition, suberitical multiolication measurements will be performed during the approach to critical to insure that the limit on k(excess) is ot exceeded. Curing the proposed measurements, the reactor will be cl vated at pot.er level less than 1 KWt, hence it is not r :ected that any otiver technical specifications related to the thermal
- quantities will be exceeded.
l ! I~ Environmental :nsidera tions We have determir d that this amendrent will not result in any significant environmental impact ar. that is does not constitute a major Comissien action significantly af ecting the quali y of the human environment. We have also deternined tr.ct this action is not one of those covered by 10 CFR Section St.5(a) or (b). Having made these determinations, we ' nave furtner concluded thrt, purs :nt tc 10 CFR Section 51.5(d)(4), an environ-mental impact tatement or environmental impact aporaisal and negativ3 declaration need not be prepared in connection witn issuance of this mend-ment. g *usions We conclude that the proposed technicul sp.cifications waiver for .he experiment with one 13-plate fuel subassembly is acceptable. We b 3e this conclusion on the following: (1 ) the ak/k (excess) will n.t be exceeded from that required
, by the technical specifications; (2) the pcwer o. ring the : .asuremeni.s will be less thc abot.t ! l KWt; 1
(3) the technical specification waiver is of limited duration j ; and applicable only to these measurements; and (4) subcritical : Jitiplication measurecents will t madt during
, the aoproach to critical .
We have further determined that this amendment will not increase the orobability of occurrence of an accident analyzed in the safety Analysis l Report, nor does it increase the consequences cf such an accident. We
. . have concluded, based on the considerations discussed abc.e, that:
I' } because the amendment does not involve a sier.ificant increase in the probability or consequences of accidents previcusly considered, and does not e
.8
_. - -. - _ = - - . . . .._.- . _. . _ P J ir.volvt a significant decrease in t sa'ety margin, the amendment does not 7 ir.velve a sign!ficant hazar.: censideration, (2) there is reasonable assurance i . that the heal'.n and safet;. cf the publi will not be endangered by operation ir the :roposed ma.- er, a'id (3) such ac-ivi- ts will be conducted in compliance with the C. :inic 's reoul tion: and the issuance of this amendment will not l be inimical to the ecmcil defense and security or to the health and safety of a tne sublic. 7 l i l l 4 e r i t. Da ed: tiovember 21, 1980 4 se f Pi 64 oam A l k"/AL _ 18 [
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** ATTACHMENT TO LICENSE AMENDMENT NO. 4 FACILITY LICENSE tiO. R-62 I' DOCKET NO. 50-124 1-r Revise Appendix A as follows:
j Remove Page Add Page 4 4 I i I n i . l !. 4 i 84 f I.. b .
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6.C Reactor Core 6.1 Fuel : 6.1.1 Standard fuel r:1. .ents shall be of the flat-plate type
- with twelve plates to each element. For a period of 30 days i.e., December 1 to December 31, 1980 one fuel
. assembly may contain thirteen fuel plates. Each fully-loaded fuel plate shall be approximately 26-inches long ' by 3-inches wide by 0.030-inch thick, which includes 0.020-inch aluminum cladding on each side, contrining a nominal 22 grams of U-2?5 as ur:nf um-aluminum alloy.
The fuel platas sh:11 b :eparated by 0.40 inch and mechanically joine: at the top and bottom. Half and quarter.'ead fuel -lstes and aluminta dummy plates may be used : 3djust the cora loading. 6.1. 2 Low power fuel ele-wts shall be the same as the standard fuel element: ex:c : for 0.010-inch cladding and a 4 di'ferent s ft.ng attacnment. The low power elements shall not be op ated above 100 watts. 6.1. 3 Twelve fi. il elements, loaded six to each core tank, shall
- ..ake up a core loadi g. Maxicum excess reactivity above cold clean critical shall be limited to 0.6% delta k/k including positive reactivity from experiments.
6.1.4 The r.tderator temperature c: S'ficient of reactivity shall be negative gnd have a minirs- absolute reactivity value of 6.8 x 10-3 delta k/k/*C at cr.y operati g temperature. The moderator vc i coefficient of reactivity shall be negative and have a core averaged value of 0.184% delta i/k per 15. void.
, 6.2 Control Rods 6.2.1 Four 1/8" thick boral control rods, 2 safety, I shim and ' 1 regulating of the windowr de type sball be positioned ; in slots machined in the graphite external reflector adjacent to the outside face and near each outside corner of the core. All control reds shall be inspected for cracks and deformations at least once a year. , '!cis autnorization shall expire on December 31 , 1930. I l
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F9 i i J APPENDIX B 1 13 Plate Fuel Element Measured Reactivity Effects
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i W l i d e.. f I e 21
4.0 Measured Reactivity Effects 4.1 Approach to Critical Experiments A thirteen plate fuel element loaded with 7 dummy aluminum plates and 6 fully loaded fuel plates was first placed in core location E-4 4
. A centrally located fuel element was replaced since it is in the most i reactive region of the core. Shifting the 13 plate element to any other i
core location after it is fully loaded should result in a decrease in total axcess-K for that particular location. A summary of the U-235 loading in all fuel elements is given in table 4.1.3. and 4.1.4. II.1 Approach to Critical Procedure in the Reactor Procedure Manual was followed. One fully loaded fuel plate was added to the element between each critical mass determination. (tables 4.1.1 and 4.1.2, figures 4.1.1 through 4.1.9). The existing source range fissica chamber and a BF3 detector located in the north beam port were used for obtaining count rates. Criticality vts predicted and did occur with the addition of the eleventh fuel plate. The approach to critical procedure was followed when adding the twelf th and thirteenth fuel plate to insure 0.6% ak/k total excess was not exceeded. (figures 4.1.8 and 4.1.9) Graphite stringers were removed as necessary which prevented exceeding the total excess-K limit. 4.2 Critical Experiments Reactivity measurements with the fully loaded 13 plate element were performed in core locations E-4, E-3, E ', '4-1, and W-2. The total net gain in excess-K by replacing a 12 plate with a 13 plate element was measured. The total actual excess-K with the 13 plate element in
~
each location was also measured. The total theoretical gain in excess-K
! if no graphite stringers had been removed was calculated. The results e
22 , m.
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are suma.rized in table 4.2.1. The total worth of the shim rod and the regulating rod were measured with the element in each core location (cable 4.2.1). Integral rod worth was measured at 4", 6", 8", 10", 12", and 16" withdrawal for the reg. rod and shim rod in location E-4 and for the shim rod in locations E-5, E-6 W-1 and W-2. (figures 4.2.1 through 4.2.6). f; The original plan was to measure diagonally opposite locations in the core. Five out of six of the locations were measured. I'
! The experiment was terminated on location W-3 when it could not be assured a total excess-K of 0.6% aK/K veuld not be exceeded. The best extrapolated estimate that could be obtained when approaching critical indicated a possible total excess-K of 0.601: Ak/k. Reactivity measurements were performed to an accuracy of + 0.001% ak/k using a micro-processor based reactimeter.
All removable graphite stringers were out of the core. Reducing the existing reactivity of the core would have required disassembling a
" hot" 12 plate fuel element and replacing one of the 12 plates with a dummy aluminum plate. Therefore, it was decided to terminate the exper-iment due to time constraints and the cedious process that would have been encountered in manipulating " hot" fuel.
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1 . w APPROVED /HEVIEWED l U Date
- N /L L u, g 11.1 Approacli to Critical Procedure Data Slicet .Cffahabn,. Hadiation Saroty Committeo M d Total Counting Rate C'D Reciprocal Hultiplication Fred. Next #
Run Fuel Fuel U ^f Spare Crit. Fuel Rodi
# Addition Loading Fiss.Ch --ppe. Fina.Ch J' d '
.upa Hass Add. Out I)e t ec to r v= NN .
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- o. .u II+ll 2 All T
0 0 0 U 1 2 t_] All M 64 m b h 3, C
- T A D Lli % f. 2.
$ T dl C
Log of Tuel in Reactor listed qugntities in grams
, Slab "A" Thermal Column
~
Plat 4 Mass U" N~Y U~[ W~b TDE-2cl22.46 11-7 21.39 9-12 21 .7 9-4 21.7[ 10-1 22.04 4-27'21.49 l 6 TDB-05 21.53 13-1 22.67 12-31'22.38 4-31 21.52 10-2 22.22 13-27 22.58 i TDG-08l21.50 11-9 21.39 12-2 22.27 12-23 22.46 10-4 '22.13 4-16 21. 51 TDB-17 22.42 13-2 22.61 2-31 '21.48 2-20'21.48 10-5 22.04 13-28'22.57 TDA-06 22.37 11-12 21.37 12-3 22.23 12-26'22.41 10-7 ' 22.22 4-21 I.O TDB-30' 22.4 2 13-4 '22.61 3-1 22.0] 12-28 22.43 10-9 22.32 12-1 22.61 TDG-06 21.87 4-29 21,56 12-19 22.06 2-24121.48 2-30 21.48 9-28'22.32 TDS-14 22,35 .13-8 '22.62 12-8 22.26 9-3 21.75 10-12'22.22 12-9 '22.55 TDE-04 ' 22. 25-'4-28 21.53 9-2 21.75 12-30l22.44 10-14 22.13 11-16'21.38 EDB- 23 22. 27 13-11'22.63 12-15 22.27 2-25'21.48 10-18'22.22 12-13'22 47 EDB-22'22.24 11-10 21.62 10-10 21.76 12-4 '22.35 10-19'22.22 11-13'21.3] ! EDS-11'21.81 13-12'22.61 12-16 22.26 12-5 '22.33 4-23 21,34 12-22 22.42 Al 265.49 A2 264.61 A3 264.50 A4 263.9 AS 264.78 264.58'A6 E-I E-2. E-3 E-4 E-T l*- 6 13-21 t22.58 11-19 21.39 12-6 l22.31 4-15l21.47 11-21 121.38 10-22 22.22 I 3-2 22.06 13-31 22.64 2-26'21.48 I 10-24'22.22 13-14 22.55- ?.3-13'22.63 3-3 22.01 4-8 21.42 12-7 22.36 g 4-17 21.47 11-5 21.48 10-25 22.13 3-4 22.08113 22.51 2-27'21.48 13-15'22.59 13-18'22.64 1 10-26 22.22 3-5 t22.21 4-9 '21.47 10-31 23,13 4-19 pl.50 11-18121.43 10-27 122.13 3-7 I 22.14 13-5 22.60 2-28 21.48 13-20 I22.56 13-19 I22.61 I 4-14 21.48 3-8 22.04 4-10l21.42 12-18 22.40 4-20 21.49 11-30121.45 10-29 22.22 1 3-9 21.99 13-7 '22.58 3-6 '23.24 2-32hl.48 13-23'22.62 10-30 22.04 3-10122.12 4-12 21.44 2-29121.48 13-22 22.60 11-28121.55 10-11 ;22.04 3-12 I22.01 l13-9 I 22.58' I 12-21 2.38 I 4-22bl.47 13-2522.6210-32b2.13 3-13l22.03 10-29 22.04 12-25 22.35 4-26 21.48 ! 11-31 f21. 55 3-19 21.95 3-14 I22.14 13-10'22.53 12-27 22,35 13-26 I22.58 11-30 22.61 I 3-20 21.97 h1 26; 24 265.41h2 264.62 33l 266.4a 34 is 264.37 B6 264.75 Slab "B" Shield Tank TQ3LE Ql3 v - 26
9
. 13 PLATE FUEL ELEMENT U-235 LOADING Plate No. Serial No. Grams, U-235 1 TDA20 21.95 2 TDE21 22.56 3 TDJ07 21.87
, 4 TDH04 21.92 5 TDE07 21.88 6 TDK09 22.68 7 TDK06 22.54 8 TDG17 22.55 i
9 TDJ05 21.83 10 TDB14 21.97 11 TDC23 22.52 12 TDA04 21.96 1 13 TDE26 22.50
., 288.73 gr. totcl
+
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l l t 9 APPENDIX C Back Flow Prevencer Valves t 41 i e i O a . t 6 e 44
l
, . . . - - . - - -- ~ ~
13, -.3x4. 3
,_, COLLEGE OF ENGINEERING
$ J VIRGINIA POLYTECHNIC INSTITUTE AND STATE UNIVERSITY T
.. ns--
Blacksburg. l'irginia 2406l Nin::.zAa AcTrVADON ANALYMS tA8044704Y January 4, 1980 1 Mr. James R. Clemens - Physical Plant Department Mechanical Services 6A Maintenance Building Campus
Dear Mr. Clemens:
On July 30, 1979 Physical Plant installed 2-parallel 1 inch back
. flow preventer valves on the 2 inch inlet cooling line that supplies water to the heat exchangers required to remove heat generated in the operation of the nuclear research reactor located in Robeson Hall.
The installation of this size back flow prevencer has adversely
, affected the operation of the nuclear research reactor when operated at 100% power. The back flow prevencer valves have sufficiently reduced the pressure and flow of secondary cooling water to the reactor such that temperature limitations imposed on us by our technical specifications were reached requiring us to reduce power to less than 95% on certain days during the remaining summer months. 100% power operation is re-quired for performing Neutron Activation Analysis, a major on-going analytical service used by many researchers on campus and numerous other outside organi:stions around the country.
I anticipate we will not be able to operate the nuclear reactor 3 at its full potential to provide needed research services during the months of May through August because of this reduction in cooling capability. I would desire that the back flow preventer valves be completely removed, restoring the system to its original configuration. However, it is my understanding now that they are required by new federal
; regulations. So I would like to propose an alternate solution.
i I respectfully request that the 2-parallel lh inch back flow pre-t venters be replaced with 2-parallel 3" or greater back flow prevencers
,, before May,1980 in an attempt to alleviate this problem.
l , I might add we noted a temperature increase when a back flow pre-venter was installed on Robeson Hall water supply earlier in the year. This was before back flow pcevencers were installed on the individual l , supply of water to the reactor. This building back flow preventer pushed the temperature to just below our operating limit. !Ihen the ; l i - other valves were installed in series with the building valve, i t c at..- pounded the problem and increased our temperature beyond the limit we 45 l l
i , _ _ . . --_ . Mr. James Clemens January 4, 1979 Page 2 e t could satisfactorily operate. ., e~ / Your attention to this problem would be greatly appreciated..
, Sincer y yours,
/ n /
i-Jut Alan P. Curtner Reactor Supervisor APC/fbl ec: T. F. Parkinson A. K. Furr A. H. Krebs J. B. Jones i 4 d 46
i 4 APPENDIX D New Experiment in the Shield Tank f 4 l l e a f sa e 47
, .u ..
,, COLLEGE OF ENGINEERING M . ,,
J '
.. h1RGINIA POLYTECHNIC INSTITUTE AND STATE UNIVERSITY
.. A ,5/
Blacksburg. I*irginia 2406l Necma .icnnnow .\w Aum Lasoe uen April 8, 1980
, To: Reactor Safety Committee From: T. F. Parkinson and H. J. 3oado
Subject:
Safety Analysis of New Experiment in the Shield Tank I. Introduction It is proposed to perform a series of experiments in the reactor shield tank. The purpose of the experiments is to measure the integral spectrum in a
" Dummy" fuel assemt,1y which consists of clusters of UO2 ("#""#81 "# "i"") f"*1 rods contained in Zircalloy-4 tubes. Initial experiments will be done with D 0 moderator while subsequent experiments will utilize other moderator-2 coolants such as H 2O and benzene (C66 H ). Integral spectra will be deteruired by activating bare and cadmium-covered foils inserted inside a few of the fuel rods and in the moderator. The types of foils to be used include such materials as In, Au, Lu, U-235 and Pu-239, i.e., both 1/v and non-1/v detectors.
The experiments will be under the supervision of T. F. Parkinson. Co-investigators include H. J. doado (Research Associate) and R. P.ogow (Graduate Research Assistant). All these personnel have passed the exam to qualify for working with radioactive materials. II. Description of the Exoeriments A drawing of the D 20 tank containing the Dummy fuel asse=bly is shown in Fig. 1. This tank will be posicianed adjacent to the graphite duct in the l shield tank as shown in Fig. 2 and Fig. 3. . iter an irradiation is ce=pleted, the D 0 tank will be placed on a platform suspended f rom the side of the shield 2
.. tank wall. (See Fig. 2) . Fuel rods containing foils will then be removed, after monitoring the radiation levels from the t0 fuel. Individual foils 3
will then be recovered and counted using either a NaI (Tl) scintillation detector, 3 Ge (Li) detector or a C-:t counter. 48 _ _ __
I i III. Safety Consideratio.is . A. Effects on Reactor Operations Since fission neutrons will be produced in the fuel assembly, it is necessary to insure that the experiments will have negligible effect on reactor operations. Prior work using the pulsed neutron technique has shown that k,gg is low for the fuel assembly (1) . 'lalues obtained were 0.30 1 0.03 and 0.111 + 0.004, respectively, for H O and D 0 2 2 moderators. The fuel assembly is isolated from the reactor core by
- 2 meters of graphite so that any coupling with the reactor core will most likely be negligible. A reactivity balance will be performed to insure that the coupling is negligible. However, the location of the Dummy fuel assembly adjacent to the reactor neutron detectors =ay cause sufficient flux perturbation to af fect the detector readings. It should be noted, however, that prior experiments ( } have demonstrated that very strong absorbers placed next to these detectors did not affect their readings. Nevertheless, measurements will be made of detector readings at several power levels with and without the Du=my fuel assembly in the shield tank. From a comparison of the readings of detectors in the shield tank and in the thermal column, it can be determined whether the flux perturbation due to the Dummy assembly is significant. If significant differences are noted then a graphite slab will be inserted between the Dummy assembly and the reactor neutron detectors as shown in Figs. 2 and 3. The ressurements will then be repeated and the thickness of graphite necessary to isolate the Du==y assembly from the neutron de-tectors will be determined. All subscquent experiments with the Dummy
. assembly will then be done with the appropriate thickness of graphite.
e 43
, . . . , . _ . . _~ _ . - . . . . . . . _
B. Radiological Safety Radiation levels due to the irradiated Dummy assembly will be carefully monitored af ter each irradiation to insure that personnel exposures will be minimized. Typical irradiations are expected to be for 30 minutes at a power level of 100 KW or less. A " worst-case" calculation has been performed of the radiation levels to be expected f com the irradiated UO2 pellets. .e assumed that the fuel pellets were irradiated for 1 hour at a power level of 100 KW, corresponding to an incident thermal neutron flux of 4 x 10 10
~
cm' sec . A one-hour wait time was then assumed. The estimates of the gamma activities and radiation levels resulting from fission products and from Np -239 were made for a single UO2 pellet. Each fuel tube con-tains 50 pellets and the entire DUMMY assembly contains 37 fuel tubes. In order to evaluate the total activity, it is necessary to know the average flux in the assembly. This was esti=ated from the relation z=L r=R 7 2" e o e ds Jo g rdr Z=o /t=o
}~
(1) 3R L ^ where R = 5.3 cm (the radius of the UO, cluster) 7 R = 18.0 cm (the extrapolated radius) L = 70.0 cm ( ff = 0.091 cm'{he extrapolated (the length) length) inverse relaxation e = 2.404 The detailed calculation.of the activity and dose rate from a single UO2 pellet are given in the Appendix. The results are A 7= 32 mci and (DR) = 3.4 m Rem /hr at 10 cm. Then the total activity and dose rate uere evaluated from equations (2) and (3): O
e -. s A - r-Tot = At CL, 3
$o (DR) Tot = ( }1 --
N ho s here N = the total number of pellets and 3/4 = 0.15 from egn. 1. The final estimates of total activity and dos,e rate are: A To t = 14.4 Ci
, (DR) Tot = 0.94 Rem /hr i
1 In order to validate the calculations, a single UO 2*11*C 2 4 will be irradiated for a known time at a known flux. The radiation levels from the UO 2 Pellet will then be measured for several wait times and compared to the calculated dose races. i
References:
- l. Lolich, J. V., and Boado, H. J., '*:teasurement and Calculation of the Eigenvalues 's' and 'k' for a Highly Suberitical Assembly" (Submitted to Atomkernenergie)
- 2. :11nutes of the Radiation Safety Committee.
i
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T. F. Parkinson
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~. -- -_ - - . - . -
COtLEGE OF ENGINEERING h VIRGINIA POLYTECHNIC INSTITUTE AND STATE UNIVERSITY l 3
'Blachtburg, l'irginia 24061 Necuna Acmanow AN AMM tAWANY .auly 21, 1980 e
TO: .leactor Safety Committee
- FROM: T. F. Parkinson and H. J. Boado
SUBJECT:
New Experiment in the Shield Tank
References:
(1). Memo (April 8,1980) from T. F. Packinson and H. J. Boado to Reactor Safety Committee (2). Minutes; Reactor Safety Committee Meeting; May 19 , and 23, 1980 I. Introduction
. In reference (1) a general description and a safety analysis of the new experiments were presented. The RSC approved the experi-ments at its May 19 and May 23 meetings subject to the preparation of more detailed procedures for the experiments. A question was also raised about disposal of the radioactive vastes after the experiment is dismantled. The purpose of this memo is to respond to the comments of the RSC.
II. Waste Disposal The major radioisotope waste produced will be fission products from the natural uranium dioxide fuel pellets. Most of the fission products will be retained within the pellets. Gaseous fission
, prcducts will be retained within the Zircalloy housing tubes except
! for ti.e small amounts released when activation foils are removed from the fuel rods after each irradiation. For this operation, a glove box will be used to handle the irradiated fuel tubes and pellets. At the completion of an irradiation the fuel tubes will be loaded into the Al tank and it will be stored in the shield tank.
The duration of these experiments will be at least two years and
. possibly as long as 5-10 years. The fuel rods were obtained on loan from CNEA (Argentina) and will be retained for an indefinite period.
After our experiments are completed the fuel rods can be stored in one of the fuel storage pits. We do not anticipate having to return the fuel rods to CNEA within the next 10 years. By this time the
, fission product activity will have decayed to very low levels.
Some solid radioactive waste will be induced by activation of the Al double-walled tank. This tank was fabricated from higft-purity Type 1100 aluminum to minimi::e induced activation. At the
.. conclusion of the experiments, the tank will be stored in the shield tank or in a shielded container to allow the activation products to decay. The tank will then be shipped off-site to a low-level waste disposal ground in accordance with NRC-DOT regulations.
55
n .. . l - Page 2. !aw Experiment in the Shield Tank I ] The Al tank will be coated with clear Krylon paint ta minimize I corrosion which could possible cause the release of radioactive corrosion products into the shief i tank water. 5
- III. Determination of Neutron Detector Effects and Reactivity Ef fects of the Al Tank and Fuel Assembly l
A. Neutron Detector Effects Two of the neutron detectors for the reactor instrumentation j system are located in vertical tubes placed at the end of the shield tank adjacent to the reactor. These detectors are : (1) a compensated i ion chamber (CIC) which supplies the signal for the Intermediate Flux ; I Monitor, and; (2) a compensated ion chamber which supplies the signal l for the Keithley picoammeter. Prior experiments have demonstrated that l very strong absorbers and voids placed next to these detectors did not affect their readings. To confirm that the double-walled tank has a negligible effect, a set of readings on all neutron detector channels will be taken at several power levels (e.g.1 watt, 10 watts, 100 watts, 1000 watts) before the Al tank is inserted in the shield tank. The I same set of readings will be repeated for the following conditions: (1) Al tank filled with H30 but no UO2- ~ (2) Al tank filled with H 0 and 7 UO2 fuel rods. (3) Al tank filled with H O 2 and 19 U0 2 fuel r ds. (4) Al tank filled with H O 2and 37 UO2 fuel rods. . If the presence of the Al tank shows any significant perturbation a ' on the neutron detectors, then a graphite slab will be introduced be- ! tween the Al tank and the detectors to insure that there is no effect. " l B. Reactivity Effects 1 During the experiment sequence of Section III A. , the control
- rod settings will also be recorded. From this data the change in k will be evaluated. If positive changes are found, k g, will be pIbbtedvjt.numberoffuelrodssoasnottoexceedthemaximumallowed limit of 0.60" ak/k . Independent measurements have shown that the UO 2 assembly has higher kegg with H 2O moderator than with D 20 moderator.
IV. Review of Experimental Results A summary report of all the above measurements will be submitted to the Reactor Safety Committee for their review. fp :;Q :s_
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