Cover Sheet for a Hanford Historical Document Released for Public Availability
Released 1995
Prepared for the U.S. Department of Energy under Contract DE-AC06-76RLO 1830
Pacific Northwest laboratory Operated for the U.S. Department of Energy . by Battelle Memorial Institute .... --.I!- --.I!- I DUN-2873 Dt u P+ i i ! I i
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,* MONTHLY REPORT
AUGUST, 1967 i DISCLAIMER
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AS DEFiNED 1954. UNAU
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RICHLAND, WASHINGTON 4
L DISCLAIMER
' Portions of this document may be illegible in electronic image products. Images are produced from the best available original document.
I DUN-2873 ba Hanford Codes C-57 & c-65
This document consists of WITH DELETIONS
September 18, 1967
MONTHLY REPORT
AUGUST 1967
I I n3-J’- . . il NOTUCNCl M)UGLAS UNlTED NUCLEAR, INC.
Richland, Washington
Work performed under Contract No. AT(45-l’)-1857 bebeen the \ Atomic Energy Commission and Douglas United Nuclear, Inc. DUN-2873
REPORT DISTRIBUTION c AEC - RICHLAND OPERATIONS DOUGLAS UNmD NUCLJ3A.R 6 OFFICE 3. J. J. Bombino i. 1-7. D. G. Williams’ 32. C. E. Bowers (Oak Ridge) 33 D. H. Curtiss AEC - WASHINGTON 34. R. E. Dunn 35. W. J. Ferguson *. 8-10. F. P. Baranowski, 36. G. C. Full-mer Division of Production 37 C. N. Gross 38. R. E. Hall A.EX - SAVANNAH RIVER 39. C. D. Harrington OPERATIONS OFFICE W. G. Catts 40. R. W. Hallet,‘Jr. ll. N. Stetson 41. T. W. huff 42. H. W. Heacock EA’ITELI;E-NORTHWEST 43. W. E. Ingram 44. . R. T. Jessen 12. F. W. Albaugh 45. S. Koepcke 13 J. M. Batch 46. C. W. Kuhlman 14. J. J. Cadwell 47. M. Lewis 15 F. G. Dawson 48. A. R. &&re 16. E. A. Evans 49. W. M. Bthisi. 17 S. L. Fawcett 50 J. S. McBhon . 18. D. R. de Halas 51. N. R. Miller 19. R. L. Dillon 52. H. C. Money 20. D. D.’ Laming 53. W. S. Nechodom 21. R. S. Paul 54. J. A. Nelson 22. E. E. Voiland 55 J. W. Nickolaus 23 D. C. Worlton 56. R. Nilson 57. R. S. Peterson ATLANTIC RICRFELD HAWORD 58: R. W. Reid 59 J. W. Riches 24. L. M. Richards 60. K. L. Robertson 25. R. E. Tomlinson 61. R. K. Robinson 26. J. H. Warren 62. 0. C. Schroeder 63. , W. Seeburger DOUGLAS UNITED NUCm 64. SC. H. Shaw 65. R. H. Shoemaker 27. T. W. Ambrose 66. J. T. Stringer 28. G. 0. Amy 67 H. F. Tew J. A. Q3@n 68. A. P. Vinther .A E. V. Padgett 69 DUN File 8” 29. R. S. Bell 70. p. DUN Record .. 30. J. R. Bolliger
-I f ii DUN-2873
'PABLE OF CONTENTS c Starting Page
SWY A-1 REACTOR PLANT OPERATIONS
B, C and X Reactors B-1 N Reactor BN-1
FUEL AND TARGET FABRICATION c. c. B, C and K Reactors c-l . N Reactor CN-1
TECHNICAL ACTIVITlES
B, C and X Reactors D-1 N Reactor DN-1 ADVANCED CONCEPTS AND PLANN7NG E -3. IREUDLATION SERVICES F-1 ADMINISTRATION - GENERAL G-l APPENDIX
A. Project Status Summary - Reactor Facilities E-1 B. Significant Reports Issued H-5 C. Employment Summary H-7 FEATURE REPORT Poison Spline System for Supplementary Control - 1-1 B, C and K Reactors
-3- DUN-2873
SUMMARY . REACTOR PLANT OPERATIONS B, C and K Reactors
Reactor input production (Pu) was 338.1 I(EIwD, 233.5 at the two K reactors and 104.6 at the two smaller units. U-233 input production was 4,587 equivalent .. MWD. Overall. time operated efficiency was 90.4 percent, averaging 89.1 per- cent at XE and KW and 91.8 percent at B and C. There was one fuel element failure.
The C and XE Reactors operated the entire month without an outage. At month: end, C Reactor had completed 46.1 days of continuous operation; this exceeded its previous record by 8.1 days. All. four reactors operated concurrently for 13.3 days, their longest period of combined uninterrupted production.
N Reactor
Total input production was 93.3 KWD, with a coproduct increment of 91.9 KMWD. Time operated efficiency was 77.0 percent and steam availability was 76.8 per- cent. Steam delivered to the Washington Public Power Supply System enabled the generation of 435 million kilowatt hours, a new high for a one-month period.
Reactor operation was interrupted twice during the month, both times by fuel. cladding failures. A third fuel failure was detected during the shutdown transient of the second (August 26) interruption; this failure caused further contamination of the primary loop, as the tube involved could not immediately be identified and diverted. The reactor remained down through month end for scheduled charge-discharge and maintenance.
All three of the fuel failures were brk I1 coproduct driver elements. The one which has been examined was in a column averaging.1465 WD/T, representing 70 percent of goal exposure. Cause of the failure is still unknown, but there is a strong resemblance to an early coproduct-type element that failed by corrosion of the inner cladding under a locking weld tab.
The first failure of a coproduct target element, which occurred in late July, was evidenced by two distinct peaks in primary loop tritium content. These are suspected to reflect independent failures of the two aluminum-clad lithium aluminate components within a single Zircaloy-clad target element. No radio- logical hazard was associated with the amount of tritium released to the crib. Attempts to locate the failed target(s) have not yet been successful.
Good’progress was made on Cell 3 cleanup and restoration necessitated by the June 21 fire. Cell decontamination is planned for September, and the required temporary piping is being installed. Preparations for Cell 3 steam generator retubing were on schedule at month end.
J A-1 DUN-2873
FUEL AND TARGET FABRICATION
B. C and K Reactors
Fuel production totaled 210.5 tons of natural uranium elements and 138.7 tons of 94 Metal elements. This production represented 103 percent of forecast.
Fuel core inventory at month end was 700 tons, a 2.2 months' supply. Finished fuel element inventory was 1,992 tons, a 5.6 months' supply. This high inventory of finished elements reflects the sharply reduced fuel requirements for non- .. defense plutonium; canning line operations have been curtailed accordingly.
Hot-die-size cladding of the inner rods for the'' one-inch overbore rod-in-tube fuel has been completed through end bond testing. Process parameters for nickel plating and sizing of the outer tubes have been developed, and end bonding . specifications are being established.
N Reactor Input extrusions of 85 outer Mark I billets of 125 &tal for standard tube-in- tube production, and 19 outer and 32 inner erk IV billets of 94 &tal for a production test, totaled 27 tons. This input production represented 62 per- cent of forecast. Finished fuel assemblies totaled 836, of which 792 were Mark I1 coproduct driver elements. This completes fuel and target fabrication for the coproduct program. Finished assembly production in August was 12.3 tons, or 41 percent of forecast.
TECHNICAL ACTNITDS
B, C and K Reactors
Samples from the downstream end of the Zircaloy tube that had been irradiated in'D Reactor for 12 years have revealed only scattered spots of hydriding. There is at present no satisfactory explanation for this tube having picked up so little hydrogen during its long service. Two half-inch overbore Zircaloy .. tubes removed from C Reactor were found to have accumulated hydrogen at about half the rate of K reactor tubes.
Twelve test fuel columns were charged into B Reactor to provide operating and post-irradiation measurement data to establish the physics of Np-237 buildup in natural uranium with a high U-236 content (400 ppm) in reactors of the B-C-K type.
One column of recycled thoria target elements has been prepared for charging . into C Reactor in September. The high-fired denitrate& thoria was vibrationally compacted into the aluminum cans.
Laboratory equipment has been assembled for the fabrication of thulium micro- spheres. Small amounts of these have been produced successfully in diameters up to 500 microns.
A-2 A new computer program has been written to calculate the coolant flow through process tubes following a crossheader failure. The program will calculate the hydraulic conditions which would exist following a failure at any point on the crossheader.
It has been determined that a laboratory estimate of the 1-13 in a sample of reactor effluent water contaminated with fresh fission products can be made within one and one-half hours. After continuous effluent samplers are installed, DUN will be able to make a timely determination, with acceptable precision, of. any'significant fission product..release to the river.
The irradiation of K reactor graphite samples in the GETR will be curtailed after completion of the current cycle, as required by GETR space reallocation. Irradiations planned for the higher temperatures will be shifted to the ETR.
A study has been completed for the B and C Reactors showing that substantial benefits can he expected from having the diesel pump flow augment the initial high tank flow, and from extending the high tank draining time by an automatic flow reduction.
N Reactor
Data obtained during on-reactor decontamination of 14 process tubes revealed that 99 percent of the CO-60, representing 57 percent of the activity, was removed. The Fe-59, Mn-54, and Ce-141, which accounted for 30 percent of the actiivity, were 97 percent removed. The Sb-125 and ZrNb-95, which accounted for 10 percent of the activity, were.90 percent removed. Calculations based on these results indicate that applroximately 180-200 curies would be removed during a full-reactor decontamination program. Examination of primary loop hardware components separately tested has shown no harmful effects from the decontamination solution.
A program was developed for the irradiation of Hanford-generated neptunium in N Reactor, based upon results obtained from the test irradiation of four kg of Np-237. A maximum of 3.8 kg of PU-238 could be produced from Hanford- generated neptunium by January 1972. The production plan would fully utilize the separations facility by varying the in-reactor residence time of the targets throughout the program. Longer residence times would be used during the later years when neptunium throughput would become larger. The quality of the PU-238 would vary from 90 percent during the first year to 85 percent during the last year of the program.
A parametric study of the thermal hydraulic fuel design parameters associated with boiling in the fuel subchannels was completed. It was found that the active zone pressure differential for a tube power of 6000 kw could be increased to 183 psi with subchannel boiling, compared to 145 psi when no subchannel boiling is permitted. This increased pressure differential could be reflected in an increase in maximum fuel weight for a given maximum tube power, or in reactor power level while maintaining a constant fuel weight.
In the past, estimates of fuel failure rates have been based on an expression .. which related the probability of fuel failure to fuel exposure level. A new
A- 3 expression.relating the probability of failure to surface heat flux and time__ . in reactor has been developed which may provide guidance in the design of
9. improved fuel elements. I Design is 90 percent complete on the facility to permit metal-water reaction and fission product release tests on irradiated fuel heated to above the melting point of uranium.
Erection of the 300-foot meteorological tower was completed at 100-N Area. Procurement of instrumentation to be mounted on the tower is in process.
ADVANCED PLANNING CAGE programming has progressed to the point where both the detailed allocation report and the Mark IIA version of the Hanford program are operational. The khrk IIA version permits flexible shutdom-startup of reactors. Minor definition problems still remain, particularly in the allocation report. Significant CAGE applications during the month were: the completion OFthe Pu/Oy study, the first phase of the peaceful plutonium pricing study, and the cases provided to AECOP to study the impact on the diffusion plants of 2lO.Metal usage at Hanford.
A revised production matrix is nearly complete. Entries in the new matrix are based on more sophisticated treatment than before, and represent firmer state- ments of product capability. Closely coordinated studies with AECOP on long- range PU-238 production capability were completed.
Efforts are underway to incorporate N Reactor into the Company's longirange plan, to provide further detail in the plan, and to recognize changes in market forecasts' and environment. .
IRRADIATION SERVICES
The first UO2-Mo cermet fuel pin capsule cooled with recirculating helium was discharged from a special test facility at KW Reactor after achieving its goal - exposure of 2000 hours. Fuel temperature was controlled at 2OOO-23OO F by regulating the speed of the recirculation pump .built into the upstream end of the capsule. A similar second.capsule was charged and is operating at design conditions. These experiments are part of a NASA-Lewis Research Center program for the development of a nuclear space propulsion system.
The test assembly for the NU-121-1 experiment was received from Atomics International. This assembly contains prototypic SNAP-8 fuel elements rather than sample fuel specimens as irradiated here in previous experiments.
FEATURE REPORT
This month the appended special report describes the poison spline supplementary control system which is being used very effectively at the B, C and K reactors. Included is a brief discussion of the possibilities for extending the system's use so as to further enhance the irradiation flexibility of these reactors. GENERAL
GE pension refund checks were received for employees who transferred to DUN and who were not eligible to vest their GE pensions. Packets were prepared and.transmitted to these employees, providing them with, an opportunity to endorse their checks for the purchase of AEC annuities or to reject the annuities and cash the checks.
There were no disabling injuries, and no radiation exposures exceeded operational.control limits.
Charles D. Harrington v President REACTOR PUNT OPESATIONS - B, C AND Ks
PRODUCTION
General
Reactor production, power levels, efficiencies and related statistics are -. tabulated on the next page. Overall reactor input production and time operated efficiency for the past six months are charted below:
100 80 U
4-01
I 20 8 z 100 0 0
Deactivation of the D Plant is continuing. The reactor building deactivation is about 95 percent complete and deactivation work in the power buildings is approximately 65 percent complete.
Nondefense Plutonium
The nondefense plutonium program continues on schedule.
OPERATING EXPERJBNCE
Power Levels
Power levels at the C, KE and KW Reactors were restricted by the 95 C bulk outlet water temperature limit. Power level at B Reactor also was restricted by this limit during the early part of the mnth. Reactivity control consid- erations associated with the nondefense plutonium program were limiting at B during the reminder of the month.
f PRODUCTION REACTOR STATISTICS - AUGUST 1967
-B -C m- -m Input Production KIND - Pu - 42.1 62.5 13.4 10201 - U-233 - Equiv. MWD 891 1053 1220 1423 Power Level (MW)- mximum 1880 2045 4325 4350 - Average 1626 2017 42 39 4m
Time Operated Efficiency - $ 83.5 100 0 0 100 00 78.2
Outage Time Allocation - $I . Charge -Discharge 1.2 0 0 14,5 Failed Fuel Removal 0.4 0 0 0 Water Leaks : 4.6 0 0 0 Tube Replacement '0 0 0 0 , Other hintexlance 6.9 .o 0 2.5 Standards Ch-ecks , 0,6 0 0 4.4 Production Tes%s J. 0 4. 0 '0 0 02 Project; Work 0 0 0, 0.2 Other 1.0 4. 0 0 0 -_u - I, .IU Total 16.5 0 @ 2i8 Nunher of Outages 2 0 0 2 Number of Start;iip JCn%erruptions 0 -0 0 0 Water Leaks - !Tube 0 0 0 0 - Van Stone l 0 0 0 New Tubes Jns-taU.ed 3. 0 C 0
l84,3 0 3.68 I. E62 .li.(l) 2590 5 (1) 34.0 4 4.7 0 1 176c 5 177 * 9 1 0 0 0
325 PI. ~369 3-69L, o 3.25 9 Water to Reactor Normal 0perat;ing W.cw - gpm 94 600 104. 000 214 000 23.6 ooo 628 600 J?H 6.61 6057 6,68 6 7J. - Dichromate - ppm 0.75 1.00 0~90 0,go
(1) Inc1.udes the 4.5.2 ton6 of special depleted uranium in each K reactor E-D loading (PITA-048).
1 Time Operated Efficiencies
There were four reactor outages as listed below:
Date Outage Hours -Down Reactor During August Causes August 2 B . 64.0 Removal of a failed natural uranium bumper fuel element (see page B-6)'. Miscellaneous maintenance work, reactor leak testing and Production Test work were achieved.
*8 Kw 117.1 Scheduled charge-discharge and maintenance.
11 B a.5 Termination of a temperature cycle 58 due to reactivity conditions associ- ated with the production of the non- defense plutonium. A small charge- discharge was completed and a rear Van Stone flange leak was corrected.
27 Kw 44.9 Reactor scrammed due to an unexplained short in the No. 2 low-lift pump motor at 190-KW. Miscellaneous maintenance work was completed.
The C and KE Reactors achieved time operated efficiencies of 100 percent for August. This is the second 100 percent TOE month at C Reactor this calendar year; at month-end, this reactor had completed 46.1 days of continuous operation, . exceeding its previous record (established in February, 1953) by 8.1 days.
The B, C and K reactors established a combined continuity of operation record for these four reactors, when they completed 13.3 days of concurrent uninterrupted operation.
VSR Failure Study - K Reactors Investigation of the vertical safety rod tip failures has continued. After determining that some rods were bottoming-out following a drop, and that this was resulting in forces in excess of the yield stress of the flexible type rods, investigation of the cushioning mechanism was initiated. Of the three tip failures at KE Reactor, examination revealed that in one.instance the air relief system orifice plate was missing; this permitted rapid initial accel- eration and prevented adequate final cushioning. Investigation of the other two failures showed that final cushioning chambers were not adjusted to the desired pressure of 80 psi. After corrections were made, further testing demons-krated proper deceleration characteristics for all flexible vertical safety rods. :. ..- ..- ... , ...... I
Retentioz LjasiS Repair - K Reac-Lors- The repsir of 107-KE and.IWJreten-tion basin ta&s was started. The work we5 about 10 percent ctjHlplete at month-end. .
High Speed Scarming Systm - X Reactors The high speed scanning syatea at KE Eeactor was placed in the safety circuit August 28, and a Process Stnndard for operation of the systems was approved for both ME and FA Raactors. !!%e scanner at KW had been placed in the safety circuit in December, 1966.
Detection of Bichromte Loss - K BewAors To prevent diodrcoaate from being discharged hto the weste system, pressure guages and presswe switches have been installed in each of the six dichromate lines going from the 183 Buildiqs to the 190 Buildings. Indicating lights have bees hstalled at 1834~~an3 WW for each pressure svitch; a line rupture will give LGW pressure annunciation in the 3.83 Control Room, and the indicating light will show the lbe involved.
Installation of D Riser Low fiessure Switch, EalL 3X System - K Reactors Engineerisg work has been completed and a Qesign Change is beirg prepared to add'a lov pressure switch to..tlae Ball 3 Systems at KJ3 and XW to provide an automatic trip bypass vhen low pressu??e is sensed in D Riser. This addition is to provide protection against ball dropping durhg periods when the reactor is opzrating xith reduced pmping, 2nd o.dy one pump is supplying water to the D riser.
VSR and ECR plSaintena,nce
During August, VSR chamel Xo. &-KV was enlarged to a depth of h0 feet and the biological shield opening of VSR chams1 No. k0-W ms enlsrged, The overall status of Phase 1 cbmmel' sleeving is: at XE, all l2 channel& have been completed; at KWj 9 chamels have besto coqleted, one charnel has been enlarged to a depth of 40 feet, and the biological shield opecirgs have been enlarged 00 the remining two channels.
Investigation of the problems reported last month wit5 No. 8 HCR at KW Reactor showed the channel to be blocksd with broken graphite at 42 feet in, as measured fron the near-side of' the unit. All slee~rsswere removed and the graphite blocking ths &amel was cleared by breaking and vacuuning. A complete new set, of sleeves then was ina"&lled. No further difficulty with the operation of this rod has bees experienced.
Process Punp 8 and &tors
Following an overhaill of the No d l~wlift pump End 'mot$? at lP-KW, excessive vibration vas nokiced. The equipment was left on the line&o permit closer monitoring. Eowevsr, a short occurred in the motor on AW't 27 and caused a 1
scram at KW Reactor. A1 spare motor was installe - sent off-plant for repairs.
At 190-C the flywheel coupling on No. 12 annex pump was found to have excessive vibration. Balancing the flywheel coupling.sufficiently improved the condition to permit operation of this pump unit until the remainder of the motor rewedging program is completed. No. 11 annex pump motor was then sent to the General -- Electric Shop in Spokane for rewedging; it was back in service at month end.
Isotope Counting Systems
Specifications for procurement .of four isotope coun$ing systems have been approved. These counting systems will be used in 100-B, 100-K, and 300 Areas for radioisotope identification and spectrum analysis. This instrumentation will replace existing equipment which is incapable of meeting changing monitor: ing requirements. Sensitivity to low energy particles and the ability to differentiate between alpha and beta emitters is required. Overbore Tooling - C Reactor Using available one-inch overbore Zircaloy tubing, the development and testing of tooling for Van Stone flanging.is in progress. At present, testing to determine the proper Van Stone die configuration for explosive forming is in progress. Also, a trimmer was developed.which squares the end of the tube, preparing it for making the Van Stone flange, and squares the tube ends for welding the sections of process tubes together. A tube annealer is being tested to determine if improvement can be made in forming the Van Stone flange.
PROCESS ASSISTANCE AND CONTROL
Process Physics
The changeover from weapons-grade plutonium to nondefense (high Pu-240) plutonium in B and C Reactors occurred in a manner such that the residual exposure in both reactors has become abnormally high. At the time semi- block discharges of 12 percent Pu-240 material are made from both reactors in September, the material in the alternate rows (the other semi-block) will average around 10 percent Pu-240. The high plutonium residual results in a substantially larger moderator coefficient of reactivity, an important factor in flux oscilliation tendencies.
Cycling of the temperature distribution in B Reactor reached sufficient . magnitude early in August to require a shutdown in spite of special physicist shift coverage given to the problem. Subsequent measures taken to "deflatten" the flux distribution both radially and axially, and to improve monitoring of cycling tendencies, were successful in stabilizing operation during the remainder of the month. Cycling tendencies at C Reactor, where exposure lags that at B Reactor by about two weeks, have been satisfactorily controlled by close monitoring and the application of supplementary control techniques. (See the Feature Report beginning on Page 1-1 for a more detailed discussion of these control techniques)
'1 Y D
Plans for taking about half of the B and C Reactor loadings to high exposures (approaching 2000 WD/2 goal for 12 percent Pu-240)pill be accommodated by staggering of the semi-bl4cke so that the next resi?@l in the future will not exceed the cwrent residdl; thm, cycling tendencies should not exceed those already experiencedo .. ECTs and flattening efficiencies during August were slightly reduced due to the control measures described above for ensuring stable operation with the high exposure loadings.
Operational physics data of interest are summarized below: Reactor -B -C -KE -Kw Effective Central Tubes (ECT) (1) 1501* 1599 . a46 a53 Flattening Efficiency (2) - August 0080* 0.84 0.71 0.71 - l2-lbnth 0.m. 0.85 0.75 0.73 Average'
Equilibrium Scram Recovery Time - mnutes (3) 2 40 None None .
* This reflects only the 15 days' operation after reaching equilibrium power following the August 13 startup.
Effective Central Tubes-Reactor power level divided by the average .(1) power of the ten most productive tubes which are representative of the reactor loading
(2) Flattening Efficiency--ECT divided bi the number of power generating ' tubes. (3) Equilibrium Scram Recovery Time--!l?he maximum time which could elapse between scram and first indication an$ still permit a. successful scram recovery using currently accepfable startup procedures. Production Fuel Performance
The production fuel element failwe at B Reactor on August 2 occurred in a column of four-inch bumpered natural uzaniun! elements. This is the first known four- inch 18.43 fuel element failure, The column exposure was 784 W/T. The failure . was assigned to the "hole manufacturing defect" category when the internal coolant passage was observe'd to be partially obstructed. The fuel lot was classified as failure-prone, in accordance with current criteria, but no discharge action was taken during the reactor outage. The following summary shows failure frequencies, as number/million elements discharged, for the 3-,..12-, and 24-month periods ending July 3: I- 3 hnths 12 hnths 24 Months Small Reactors - Natural U 43.4 17.1 13.9 - 94 &tal 23.5 9.8 23.7 ! K Reactors - Natural U 0 6.9 8.9 - 94 *tal 0 2.9 4.4 C Reactor - Nat. U (Overbore) 0 232.9 750 0 Reactor Effluent Activity Data
'Ilhe following table shows the total kilocuries/month disch-rge f r six significant radionuclides in the reactor effluent during July: Reactor AS-76- p-32 a-65 1-13 Cr-51 NP-239 B 0.9 0.14 0.19 0.009 5 *o 1.5 C 1.4 0.24 0.35 0.018 10.4 2.4 KE 1.9 0.24 0.49 0.012 5 *2 2.6 Kw -2.5 -0.21 0*3 0.068 -5.7 -3.4 Total 6.7 0.83 1.34 0 .lo7 26.3 9.9
The progress of experimental work on effluent activity reduction is described in the B-C-K Technical Activi$ies Section of this. report under Mission 10.
.'% . i 33-7 DUN-2875
REACTOR PLANT OPERATIONS - N
PRODUCTION
General
Input production, time operated efficiency,. and steam availability for the -. past six months are charted below:
100 0 *+, 80 -. Y 0 k a 60 -
Availability (SA) OrJ 20 - - 20 H8 1 I I I I 2R 0 I 0
Statistical Summary &put production (WD) - Total 93.3 - Coproduct 91.9
Power level (MW) -MaxFmum. 4000 - Average 3904 Time Operated Efficiency (5) . 77.0 Steam Availability ($) 76.8 Number of Shutdowns - Scheduled 0 - Unscheduled '2 Fuel Failures 3.
- -_T--.-,. . , .... c_11_ . ... ._IC I . .. - ~ Fuel Charge (Tons) - 94 Metal 4.1 - 125 %tal 1.4 * - 2lO Metal -217.5 . To tal 223.0
Helium losses (M cu.ft.) 93 Fuel oil usage (bbl. ) 15,849
OPERATING EXPERENCE
General
The reactor operated at 4000 Mw power level following the startup on July 26, and continued to operate in this mode until 0107 hours on August 21 when a manual scram was initiated following fuel failure indications in process channel U57. Plant operation was resumed at 1748 hours on August 22 and continued until 1340 hours 'on August 26 when the reactor was again manually scrammed due to fuel failure indications in process channel 2762. The reactor stayed down through month end, as scheduled.
During the August 26 shutdown transient, fuel failure indications were also received in process channel 2455. All three rupture suspect columns were discharged and the fuel failures subsequently were confirmed.
Reactor Outages
The two outages and their principal causes are listed below:
Date Outage Hours -Down Outage Qe During August August 2l Manual Scram 40.7 Fuel failure in process - channel 1157.
August 26 Manual Scram 130.3* Fuel failure in process channel 2762. Total Outage Hours 171.0
* The fuel failure outage was extended into a previously planned scheduled outage at 1600 hours on August 27. Primary Loop Contamination
The primary loop became further contaminated with fission products on August 26 when the fuel failure occurred in channel 2455 and the tube containing the failure was not identified and diverted immediately. Initiation of the outage w6rk scheduled for August 27 was delayed approxi- mately 24 hours while.the loop was flushed at maximum flow with maximum fill and spill in an attempt to reduce radiation levels. Even so, the contamination of the loop contributed to higher than normal radiation exposure rates for the charge-discharge operation and other primary loop associated work which resulted in further outage inefficiencies.
EQUIPMENT EXPERIENCX
Reactor Gas System
The daily helium gas loss rate has increased slightly. It approached 4000 cu.ft. per day in August compared to about 3000 cu.ft. per day in July. Attempts are being made to determine the cause of the increased leakage. However, the recent reactor crate repairs are still limiting gas losses to about one-fourth of ~ pre-July rates
Cell 3 Restoration
Replacement of the elastomeric coating damaged by the cell fire on June 21 was completed, as was also the wall cleanup and painting. &st scaffolding has since been removed, The principal current activity in the cell is the installation of the temporary decontamination piping, and other miscellaneous work necessary to permit the start of cell decontamination in September. . Primary pump inspection is scheduled to follow decontamination; restoration of miscellaneous piping and electrical. systems on the south wall will continue for the next severalmonths.
Cell 3 Steam Generator Retubing
Tube bending and panel assembly is progressing satisfactorily; all tubing has been received for bending. Twenty-five percent of the tubing fabrication has been completed. Work is scheduled to start in the cell in mid-September.
Low-Lift Diesel Driven Pump
The discharge line of No. 1 low-lift diesel driven pump was broken by an unexplained line surge which moved the pump, causiiig misalignment of the pump and diesel engine. The misalignment in turn caused a bearing failure on the engine. Repairs were effected and the equipment was returned to normal operational status,
Primary hop Pressurizer
Packing leakage on the pressurizer vent valves became excessive during the month and necessitated the use of addit'ional pressurizer heaters to maintain primary loop pressure during reackor operation., Approximately 60 percent of the heaters were in use prior to the shutdown on the 21st. The valves were repaired and the pressurizer heaters retuned to normal during that outage. However, during the period of high leakage, the temperature and humidity in the pressurizer penthouse were very high and.a number of pressurizer heaters appeared to have failed. ,After the humidity returned to normal, most of the heaters returned to normal. . . .. I. * .
Zone I Ambient Temperatures
Higher than normal air and concrete temperatures (up to 162 F) were observed at the 40' level in Zone I of the reactor building. Maximum airflow from the supply fans was directed through Zone I but the temperature remained high. Calibration of the thermocouple confirmed that the indications were correct. Additional thermocouples are scheduled for installation during the month end outage to provide further data on this problem.
Fuel Storage %sin Lighting
Five of the special underwater lighting fixture hangers have been fabricated and are being installed in the discharge chute in preparation for the next charge-discharge. Sketches are prepared and materials are on site for the segregation pit relighting.
Flow Monitor Bias Voltage Power Supply
A purchased modular Dc power supply was modified to meet the process tube flow monitor requirements, and is being tested to determine its reliability and stability. The modification enables the output voltage to remain constant during interruptions of the AC input voltage for periods of up to six cycles. The bias voltage supplies of the flow monitor must be stable for line interruptions of up to four cycles. .This is required to permit scram-free operation during BPA switching operations.
The modified module can be adjusted for use in either the 90-volt or the 120-volt bias supply. Other modules are being procured and evaluated in order to locate more than one source of supply. The manufacturer of the power supply furnished with the flow monitor system is no longer in business. Other manufacturers submitted "no-bids" when solicited for quotations on replacement supplies.
PROCESS ASSISTANCE AND CONTROL Process Physics
The traveling wire flux monitor continued to provide adequate information to establish specific power limits. At 4000 raJ, the reactor is more than 10 percent below limit specific powers and is restricted instead by flow monitor outlet temperature limits.
The reactor radial flattening has been in the range 82 to 84 percent during the month. This reduction from the 87 percent achieved at 4800 MW results from deliberate distortion of the tube power distribution in order to "fit" the distorted flow pattern generated by 5-cell' operation.
Production Fuel Performance
The fuel failure which occurred in channel 1157 on August 21 was a &irk I1 driver element. It was the third fuel element from the downstream end of the column, and the column exposure was 1465 MWD/T (70 percent of goal) The cause of failure is still unknown, but there is a strong resemblance to an early coproduct prototype element that failed by corrosion of the inner cladding under a locking clip weld tab.
Examination has revealed damage to the element which appears to have.originated on the inner cladding. Corrosion products buckled the inner cladding onto the target, and closed the inner annulus. The outer cladding and core were cracked near the supports from pressure of corrosion products.
No specific information is yet available on the two subsequent driver element failures (in channels 2455 and 2762).
Target Element Failure
The first N Reactor target failure occurred on July 2l, and detected by-a routine analysis of the primary loop water sample taken thatAs day. The history of this failure, in reference to the variation in tritium concentration in the primary loop, between July 20 &d July 29, is shown in Figure BN-1 on page BN-60 On July 30 the tritium increased again and peaked.for the second time (at O.O47’Ci/ml) on August 1. These two peaks in tritium concentration can be interpretkd as two target failures, or possibly a failure of two cores in a singJ.e target.
As can be seen in Figure BN-1, the tritium concentration leveled-out at a range between 0.0027 and O0O053,uCi/ml, which is indicative of some smaller continued release of tritium from the failures. . Coincidentally, this is a release of tritium about equal to the expected amount of tritium generated in a single target located in the maximum power region of the reactor. Attempts to locate the failed target(s) had not been successful as of August 3.
STEAM TO WPPSS
Despite,N Reactor’s outages during August, it produced enough steam to enable WPPSS to generate a record amount of electricity for one month--435 million kilowatt hours. The previous high was 3.l0,500,000 kwhr in April 1967. ., .050
045 sI\ I\\ I \ I \ .Ob0 r I I I BI I I 035 I I I I' I I 3 I I I .030 I I $ I 1 Area Under Curve Not E: I I Cross-Hatched = -Curies 0 I I 430 $ .025 I as I $ I cQ) I a" .020 I 0 Actual Decay u I I .#-I I d -015 .I & I .I I .010 4 ', .005
0
Figure BN-1. Tritium Concentra ion in Primary Loop (Normal = 7 x 10-z /uc/ml.) FUEL AND TARGET FABRICATION - B, C AND K REACTORS
PRODUCTION
AcceDtable Elements Produced Elements to Storage - Tons Yields - Percent Tons Unrestricted Upstream Current Tne -Input Use . I Use . . Total &nth FYTD AlSi-Bonded. Natural U
8" Regular 114.3 45.1 18.5 63.6 95.7 95:9 8" Bumper 28.3 3.8 3.8 95 *2 95 w.2 8" Sex-Support 92.9 96.6 17.4 U4.0 91.7 92.5 . AlSi-Bonded, 94 Metal
6" Regular - - - - - 100 .o 6" Bumper - - - - - 96.9 6" SeUtSupport 1%. o ~0.3 28.4 138 7 93.8 .93*5 Hot-Die,-Sized U. 5.2 . 1.1 - - ' 1.1 . - - ,.: Thoria 2..8 - - - e --
Procurement and Inventories
Tons. Tons Placed End-of-hnth Stock It em . Received in Process -Tons Mots. Supply Natural U Cores 295 06 236.2 . 492.0 2.8 94 Metal Cores 79.7 136 5 u7.0 1.5 Thoria Powder 3.5 2.6 13.0 - AlSi-Bonded Fuel
Production totaled LO3 percent. of the July forecast. Of the 349.2 tons produced, 285.6 tons or 81.8 percent were fuel elements with bumper or self-support rails attached.
Hot-Die -Sized Fuel
The plating station input totaled 3,108 cores, of which 256 were for development purposes. OPERATING EXPERIENCE
AlSi-Bonded Fuel
Time operated efficiency tias 97.9 percent. Of the downtime, 1.2 percent vas charged to operations and 0.9 percent to equipment. Canning line operations totaled 92 line-shifts, at the rate of fowr line-shifts per day,
The total manufacturing yield was 93.4 percent. Of the total rejects in each category, the percent reclaimed or restricted to upstream use was: marred surface - 11 percent, bond - 76 percent, closure weld 7 90 percent, rail weld - 94 percent, and AlSi slopover - 94 percent. Uranium utilization, exclusive of special development but including defective material, was 98.6 percent
Hot-Die-Sized Fuel
The inner rods of the rod-in-tube fuel designed for the one-inch overbore tubes have been processed through end bond testing. Development of the process for producing the outer tube is almost complete.
EQUIPMENT EXPERIENa
Installation of shear testers mounted on the second projection rail welder was completed. This allows removal of the separate rail shear tester, and makes possible the installation of the .equipment needed to increase projection weld capacity.
A mercury relay in the electrical control panel of a projection rail welder failed when the container holding the mercury ruptured. The panel and surround- ing area were cleaned and the welder was approved for regular production use. Because this was the first such failure in 20 welder-years, it is not planned to immediately replace the relays with another type although possible alternates continue to be investigated.
PROCESS ASSISTANCE AND CONTROL Fuel and Spacer Quality
Outgoing fuel element quality continued in good control, as indicated by weld integrity and external bond count; these quality indices were 97.0 percent and 3.0, respectively. In the anodizing of aluminum spacers, anodizing yield; film thickness, and film resistivity were all satisfactory at 99.2 percent, 1.50 mils, and 338,000 ohms per square inch, respectively. Deep Crack Testing
A prototype deep-crack core tester presently set up in the laboratory has been used to screen alloy and 125 &tal elements which were de-canned and DUN-2873 .
centerless ground to obtain a smooth surface for testing. The test is capable of detecting a defect 0.050 inch wide located near the I.D. or O.D., and surface cracks less than 0.050 inch in depth. Several pieces were found which contain confirmed internal cracking.
Further testing of the unit will be required to determine the minimum detectable crack size. It appears that it may be necessary to provide a somewhat larger transducer for the tubes of the overbore rod-in-tube cores. While it is likely that some possible defect geometry may exist which'the tester is incapable of detecting, the major defects generally found in incoming material should be ' detected by the unit. .., - ...... -. _..- . FUEL AND TARGET FABRICATION - N REAC'l!bR.'
.. .i , ' I.. i.. .~ ... . . )...... _. . -_- ,.. . I.'. ,- L _. .- ...... , .- ,- . ,car-.<. .. -- ...... PRODUCTION
Statistical Smary .-
Input (Billets Extruded) August. ._.__-. - - - ...... Mrk. I, Cuterj 12.5 &tal: , i .: : ._ 35 ... _...... -. .. -.., .. .-: ..: -,-ihrk.IV, Outer,..94 &tal ..-...... :. . 19 .. ~ ,...... - ... I. . &rk '&mer, 94.&tal: ._.. ,...... 32, :. -. IV,: ...- .. , '3. ;; ,'i.. .-, ..... - Total Extrusions -_ . . 136 .. ,, ._ Tons - Total 26.6 .. - $ of Forecast . _- 62.2 ... .. t:->,.*. .- .. .. -..!.-. .Output. .(Finiehed.Assemblies).- : ...... , .. .. -. ... . hrk 11, OGters;. 210,&ta1 ...... 1792 1..
,. . hrk,.lY., Outers & Tiulersj.:125 &tal . : 44. .- - ., ...... ,. -.*I' . ..
L '. .... Total Assemblies '. . ' 836 a , ._.. - .. .. I ~ -I .. ..
. /I -I Tons --Total . a. 12.3 . ... -:$. of Forecast .:...... 4100 ...... Uranium Utilitzation - $ 87.4 Coproduct Program .. The.'assembly of'-.&rk I1 fuel elements and LiAl.02 targets for the coproduct program was .completed on, August 14. All excess targets.have been boxed and placed in storage.
Standard Tube-in-Tube Fuels
Extrusion of Mark I 125 &tal outer billets was resumed after a lapse of'14 months in production of this model. This long lapse resulted'in increased process problems which. contributed to the below-forecast -production. ' Stored Brk I-spike.:fuels requiring r.ework were re-introduced into the. production strew to.reduce .the in-process inventory of tus model ,. ...*i . ..I. ,. ... , .. I. ' Six days.,of special: extrusiQns further r.educed cormal production...... I.... Mark IV Production Test i
The extrusion of ark IV billets was essentially completed during August. , Associated process problems and design problems with the buggy spring supports reduced the output of assemblies below forecast.
OPERATING EXPERIENCE Extrusion of I# Bare Cores The extruded I# bare core program work, completed over the past eight months, has produced approximately 1000 core blanks, most of which are now at NIQ, for finish machining and heat treatment. Sample lots of this material were retained at Hanford for machining and heat treatment to permit product evaluation during the course of the work. The work was successful in that a basic fabrication route for bare cores was determined,. expected tool performance was evaluated, and surface quality and dimensional control capability before and after heat treatment were determined.
Based on currently available data from the feed-site stream, a design uranium utilization from ingot casting through primary extrusion is expected to reach 81.5 percent. The design utilization for uranium from primary extrusion through extruded-to-size and end-squared bare cores is expected to be approximately 89.7 percent, indicating an overall design utilization of 73 percent.
Because almost all of the above experience is based on machined core fabrication, and the limits imposed by 1&33 fuel manufacturing needs are not well known, a test is being developed to manufacture 1843 fuel elements from extruded-to-size cores. These cores will be made from ingot casting through final extrusion in a prescribed route so that overall uranium utilization and process yields and capabilitie's can be accurately determined. It it expected that this test will start in November and be completed in February 1968.
Mark I Fuel Reworking
Approximately 800 Mark I fuels were removed from outside storage and fed into the process stream for reworking into acceptable fuel elements. To accomplish this rework and increase manpower utilization, a swing shift worked the last two weeks of August. r
EQUIPMEDE EXPERIENCE
Extrusion Press Improvement Program
Schedules have been established and work is proceeding on items not requiring engineering studies. Safety shielding of high pressure oil lines has been initiated. The engineering study encompassing eleven major items is scheduled for completion in late September. Work on the improved ventilation system for smoke control will start in September with completion expected by October 1. The instruments were recalibrated, the thermocouples repositioned in the furnace, and the furnace temperature raised 7.5 C. Production was resumed on the swing shift August 24, with subsequent audits indicating all fuels to be within operating limits.
I PROCESS ASSISTANCE AND CONTROL Mark I-C Shoe Thickness
Processing of the initial lot of fuels for conversion to &rk I-C required higher shoes than were available. To determine the actual shoe heights needed, a random sample of 104 Mark 1,fuels were removed from the finished inventory, deshoed, and measured. The average support circle diameter found was 2.6476 inches, with a three-signa range of It 0.0198 inch. Based on these diameters, the fraction of the total number of shoes needed of each thickness was determined. The thicknesses will result in an average support circle diameter very near 2.690 inches. .!be current process requires a shoe about 0.015 inch thick and various thicknesses of shims to obtain the required height. Mark I-C production will beg$& when shoes and shims are available. End Contamination of Mark IV Fuels
The majority of the brk IV fuels, both inners and outers, contain uranium end contamination above the acceptance limit. Normal practice with end contamination fuel is to refilm after 60-90 days when the Th-234 has decayed. Nearly aU. fuel is accepted when it is refilmed. Since the Zkrk IV charge schedule did not permit the decay period, end contamination was evaluated by alpha counting as a back-up for end filming.
All of the contamination on the film appeared to be the Th-234 type which decays away, rather than true uranium contamination, Alpha count results for the fuels making up the 44 assemblies shipped to the reactor in August were satisfactory.
Inner Tube Support Fabrication (Buggy Springs)
Buggy spring support fabrication has been delayed because of failure of pilot lots of springs to pass the tensile test. In tensile testing finished supports, the majority of failures occur at the bottom loops. This indicates that the bottom forming fingers on the first.hot forming machine are probably defective and need to be replaced.
x The operating sequence for the second hot forming machine c&Us for the operator to open the door on the argon evacuation chamber $hen the red cycle-completion indicator is illuminated. The act of opening the door. interrupts the process circuitry and completes this phase of the forming' operation. Since the rapidity with which the door is opened directly affects the overall height of the finished support, it appears desirable to automate this operation to eliminate the human factor. msEARCH IWD DEVELOPMEIiT
The progress on research and development work conducted by Douglas United Nuclear is reppted by Mission number and title, here and for N Reactor (beginning on page DN-l)O. Unless othelwfse noted, the Missions are as defined in the' July 1967 issue of this report. No coverage is accorded the two F&D -. Missions (6 and 9) in which DUN does not participate. Mission 1 - -sic Pmduction 1-A. Fuel Development
Fuel Support Height
The post-irradiat.ion measurements for FTA-059 (Trial Reduction of C5N Suppod Height) have not yet been obtained. The discharge data indicate ,that a temporary solztion to the stuck-charge problem may be realized by authorizing the use of 77-mil support heights and reducing the goal exposure of ~gmfie1 to 600 m/~~If post-irradiation examination does not reveal any detrimental effects on fuel performance as a result of using the lower supgort height, a change to the lower height will be recommended. The basic problem ap-gears to be related to uranium stability.
Bonding Layer Studies
Work done thus far on the evaluation and charactefistics hot-die-sized of ' . fuel bonb is describkd in BNVG-CC-1314, "'An Analysis of Hot-Die-Sized Fuel Emding," dated August 1, 1967. To permit testing and further evalu- ation of the bond, four of the intermetallic compounds havebeen prepared to test corrosion behavior and other properties of'thase materials.
Nickel Piatfng of Uranium
Studies are ccntinuing on the electroplating of nickel on uranium. Prior work on occlusion plating, to evaluate the feasibility of embedding finely divided silicon in the nickel plate *wing the plat-ing operation, revealed that 325-mesh silicon Farticks could not be held in snspension satisfac- torily. Tests have been initiated with 1-10 micron particle size elemen- tal silicon In concentrations from 25-100 -/liter, preliminary results indicate that ths, silicon can be deposited with the nickel plate. Current studies are being expanded to include an evaluation of the use of finely divided ceramic as occltided material,
Zircalov Platine: of Uranium
The molten salt electrolytic cell used for studies involving the electro- . deposition of zirconium on uranium is being modified to permit continuous purification*of the melt, The remov $,of impurities from the system has been fomd t $,e highly critical in9.?F btaining acceptable plates. Cell f ! modifications include a gas lift by which a portion of the molten salt is continuously removed for Processing through a H2-HF purification unit. Demonstration of this technique will provide the final data which will be included in a current progress report now in preparation,
Titanium Plating of Uranium
Examination and evaluation of test samples of titanium electrodeposited on uranium was completed and reported in DUN-2939, "Electrodeposition of Titanium on Uranium," The work was performed by the Bureau of Mines, Reno, Nevada in conjunction with studies they are performing with electro- lytic refining of refractory materials. An intermediate sub-strate was found to be necessary to deposit titanium on uranium. Bare uranium in the system was severely attacked by the molten salt. Similar to zirconium plating systems, purity of the melt and system was found to be highly . critical to the process.
1-B. Zircaloy Process Tube €Qdriding
Analysis of samples from the downstream end of Zircaloy tube 3586-D, which had been irradiated in D Reactor since July 1955, has revealed only scattered spots of hydriding. These analyses confirm the findings of two years ago when a four-inch sample was removed. There is no satisfactory explanation why this tube picked up so little hydrogen during its twelve-year life. The tube is being sent to Battelle-Northwest for complete analysis of radiation damage effects. Two of the fourteen 550-mil'overbore tubes recently removed from C Reactor were analyzed for hydrogen. Considering the length of time they had been in service, it appears they were picking up hydrogen at 'about half the rate of the tubes in the K reactors. Two tubes were removed recently from KE Reactor, and one from KW Reactor. These had been irradiated with different types of anodized dummies. Analyses are not yet complete.
1-C. High Pu-240 Plutonium Because of the higher natural uranium'and 94 Netal exposures which are desired in the E-D loads, new generalized speed-of-control curves were developed and published. As far as the depleted metal is concerned.. the~- maximum reactivity gain on water loss occuis for a depleted uranium exposure of about 2000 W/T.
Since operation with E-D loads is very close to the speed-of-control limit, the assumptions used in speed-of-control calculations are being examined very rigorously to remove unnecessary conservatism. In particular, since the exposure dependence of the limit is much more pronomced in the lower part of the reactor than in the upper part, the exposure distribution and the graphite temperatures for each operating period are being considered. This has involved frequent calculations by the process physicist in determining the speed-of-control limits. s The change in material buckling due to temperature, AG, between water lcss and fuelmeltbm is an important garmeter in totaLcontro1, In the E-D . loading, becomes very large as the exposure of the depleted uranium increases. For natural and enriched uranium mer.normal operating exposures, A$ is monotonically increasing between 20 C and the melting tqerature of uranium, U33 C. However, for the E-D supercell at high exposures, A@ reaches a maximm’ somewhere less than 1133 C. Because of this, it was necessary to calculate this maximum point and use these maximum values to be conservative. In conjunction with this, the cold E-D ratio was calculated as a function of exposure and published along with the AB$ curves.
1-D. Computational Techniques
The HAMMER code appears to be working Satisfactorily on the 1108 computer.. Mechanical and program problems of a month ago, such as internal looping, appear to be reasonably well resolved. Some minor problems remain, but are- either being solved or calculations are being performed in a manner to avoid these difficulties.
The random access version of EIANMER is progressing satisfactorily, but it is too early to know how much trouble (if any) and time will be involved in completing it.
Battelle-Northwest is presently trying to produce EIAMLEC cross sections from ENDF/B. The program will not work with isotopes which have negative reso- nances, but when corrections to the program are complete it will be relatively simple to get cross sections for any isotope in the ENDF/B library. This will greatly increase the usefulness of the IIAMMER program. The computer code, JENNY-2, was revised to allow either flux squared weighting or flux weighting of the average isotopic concentrations cbtained from MOFDA. The changes are not yet on tape but will be very shortly, 1-E. Criticality Control
Preliminary recults from calculations of 4.8gwtg enriched rods were ob+,ained. The computed critical masses curves fclbw the shape of those for experimental values very closely; however, the calculated masses were non-conservative since they exceeded the measured Ones by 10 to 15 percent.
Calculations are continuing on criticality control evaluation of autoclave buckets loaded with PbCd elements and 210 Metal fuel elements, and basic storage buckets containing 210 Me-bal fuel. The poison-to-fuel arrangements resulting in an optimum safe configuration for an infinite system of auto- claves contain 42 Azel elements lccated in two rings along the outer periphery, while the center region of the basket is filled by water and six PbCd elements e .. : ....
Mission 2 - Coproduct 2-A. 2.1E-N Test Irradiation
One loading change has been made in the E-N test block (FTA-054) at KW Reactor since the last report. In this, one column of thoria was replaced with natural uranium to help increase the tube powers‘in the fuel columns as they lose reactivity. On July 30, the highest tube power in a column was 133 per- cent (2400 kw), and the low st was 113 percent, with the average tube power being 125 percent of that mal for natural uraniy. Tube power varies slightly with daily operational changes. Outlet cwlant temperatures average about 108 C, with the peak outlet temperature 113 to ll5 C. The average 210 Metal exposure as of August 20 was 1525 MWD/T.
It is anticipated that no more loading changes will be made as the test block is scheduled for September discharge. Arrangements have started for isotopic analysis of both the fuel and the lithium-aluminum. The lithium targets will be analyzed at Savannah River, and the 210 Metal fuel elements will be analyzed here.
2-B. Full Core 2.1 EiN Demonstration Because of initial difficulties with the HAMMER code in this application, and the tight time schedule, the MOFDA code was usedto calculate lattice param- eters in the engineering analysis report. Some changes in the numbers in the document may be made as information becomes available from the HAMMER calculations. However, a revision is not expectedto be necessary since the MOFDA calculations are in the direction of safety.conservatism, and no new safety problems are anticipated..
Recent success with the HAMMER code is permitting performance of a pertinent calculation. For example, the following table of kw values gives a compari- son of the supercel1,calculations with the recently completed PCTR measurements:
PCTR FUW43R-MTERMINATOR Runs Test B.N.L. Cross Section B.N.W. Cross Section
. Dry case 0.9969 0 * 9799 0.9716 Wet case 1.0008 0.98118 0 9 97576 Akw -0.0039 -0.0013 -0.0042 2-C. Lithium Splines
The vendor has made a number of spline extrusions using LiAl alloy ingots which he cast. Two problems have been encountered to date. First, the clad is not uniformly distributed. It is thick along the edges and thin along the center. This can be overcome by adding more cladding at the expense of reduced LiAl. A more sophisticated extrusion die might also help. The second problem is blistering of the cladding. This is believed to result from dissolved gas in’the LiAl alloy and can probably be &krcome by vacuum casting the alloy. .t E’ rrp ...... A Mission 3 - dansplutonium Technolog# Nothing significant to report. kssion 4 - Pu-218 . 4-A. Test Irradiation
The quick turnaround neptunium demonstration test block (PTA-063) in KE Reactor is operating without incident. Tube powers are being maintained at approximately the 50 percent above normal natural uranium level. As of August 18, the test block status was calculated to be as follows:
Np-237 Fraction Tube m-238 'Total Pu Percent Burnup Np-237 Np-237 Charged -No. (ms) (gms) Pu-238 (gms) Burnup (gms) 2960 57.7 64.1 90.0 68.6 0.1445 474.8 2759 54.6 60.2 90.7 63.4 0.1335 474 ., 8 2955 40.9 45.1 90.7 47.6 0.1335 356.1. 3157 36.7 39.8 92.1 41.2 0.1156 356.1 2857 32.6 35.3 92.1 36.6 0.1156 316.6 'It is planned to discharge one column early in September.
4-B. High U-236 Uranium
Twelve test columns were charged into B Reactor on August 3, under authority of PTA-069, "Np-237 Production from Natural Uranium with a Zigh U-236 Content." The test is designed to provide operation and post-irradiation measurement data to establish the physics of Np-237 build-up in the B-C-K type of reactors. AU. test columns involve a striped loading with fuel elements containing two core types, one with 400 ppm u-236 and the other with 6 ppm u-236. Discharges will be at approximately 300, 600, 900, 1200, 1500, and 1800 W/T with several 24-piece charges scheduled for discharge at the high exposures. Several columns were fitted as spline tubes to permit flux traverses.
4-C. Np-237 in Graphite Matrix
The fabrication of neptunium targets in a graphite matrix is progressing towards an early September reactor charging. The Np(N03)4 impregnation of machined graphite cores has been completed., The NpO2-graphite powder compac- tion process was initiated when the Np02 became available about midmonth.
The graphite used in the impregnation studies is nuclear grade TSF having a density of 1.65-1.70 g/cni3 and approximately 26-27 percent void volume. TWO sets of four elements each were machined from cylindrical TSF stock. The first set were solid cylinders 8.500 inches long, with an O.D. of 1.500 inches. The second set were annular cylinclers measuring 1.500 inches O.D., 0.500 inch I.D.,, and 8.500 inches long. The impregnation process consisted of three steps: P 1) The graphite wasr@mersed in the neptunii&$:nitrate solution and a vacuum of 25-28 inches of mercury was applied. Vacuum was maintained for about five minutes and then released; this step was repeated four or five times to achieve the desired neptunium penetration of 40 grams per element. Normally about 80 percent of the voids were filled by this techni'que. A greater penetration ('90 percent) can be obtained by heating the solution to boiling under vacuum, but this requires a much longer time. .
2) The impregnated element was mounted on a mandrel and rotated on a hori- zontal axis. A heat.lamp was used to dry the rotating element.
3) The dried element was placed in a quartz tube and heated in a tube furnace under an argon atmosphere to 10x0 C to calcine the neptunium to . the' oxide. *
The entire impregnation cycle takes about 16 hours to carry out. Mission 5 - Other Isotopes 5-A. U-233
One column of recycledthoria target elements has been prepared for irradia- tion testing. The thoria was prepared by high firing the denitratedthoria at ll00 C to remove moisture and absorbed gases, and then vibrationally compacting it to 5.6 gm/cc in the aluminum cans. A pressed thoria wafer was placed on top of the compacted powder to isolate the weld zone from the loose thoria. A Production Test Authorization is being prepared to permit charging of the test column into C Reactor during September.
5-B. Tm-1'70
Laboratory equipment has been assembled forthe fabrication of thulium microspheres. Small quantities of microspheres have been fabricated, but further process development and optimization is required. Microspheres up to 500 microns in diameter have been produced. In larger sizes (~1000microns) the spheres are irregularly shaped and.lack uniformity. Process study is continuing.. Mission 7 - Target Space Enhancement 7-A. High Power Density Fuel
Work on aluminum alloy development, including the preparation of dual-alloy cladding billets, awaits the completion of casting equipment changes to permit continuous chill casting of aluminum alloys. Some mechanical problems were encountered in the casting equipment and are being corrected. In limited tests, a small quantity of a candidate alloy having acceptable second phase dispersion was produced. I
7-B. Reactor Modernization - Fuel Development
Irradiation of the third XMlN rod-in-tube fuel column under PPA-021 in channel KER-4 in KE Reactor is proceeding without incident. As of . August 20, the 28-elemerit column had accumulated an exposure of 580 Nm/T. Average operating conditions for this column are: tube power - 2820 kw, AT - 96 Cy and flow - 112 gpm. Outlet temperatures during the past two weeks have been in. the 1.23-1.24 C range. To preclude the possibility of exceeding the 130 C tube outlet temperature limit, as the river temperature is expectedto increase.1-2 C over the next month, the seven 51-element K5E columns surrounding test . channel KER-4 will be replaced with 38-element K5N fuel columns. This step will not be detrimental to test results in that the altered tube power of XER-4 will remain within a satisfactory range. Component Development
A combination overbore and channel straightening drill to enlarge the tube blocks for the.one-inch overbore facility is being tested. To date the results have not been as encouraging as had been expected. Using present drilling equipment, tube flattening occurs with additional channel distortion such that unrestricted fuel charging of fuel elements having present dimensions may be limited to two to ten years, depending on tube size. With nominal tube and fuel sizes a charging "life" of about five years ;Is expected. However, with tubes of minimum inside diameter, this value could drop to two years.
While the channel straightening treatment is somewhat similar to that used for the Zircaloy tube repxacement program at the K reactors, the larger tubes are more subject to flattening during bending. Fuel support height has a significant effect on fuel clearance. A reduction in support; height of lO.mils, for example, could extend the time of no restrictions to fuel charging by six years. Other possible solutions would be the use of mixer . fuel elements with reduced fuel support height and/or redesign of fuel supports.
Experience with the 550-mil overbore tubes in C Reactor suggests that the problem in-reactor is not as acute as indicated by laboratory data. For example, it would have been predicted that the overbore tubes recently removed for replacement with normal size process tubes would have been bound in the reactor while in fact, they were found to be free in their channels, and were remcwed quite easily. Mission 8 - Nuclear Safety
.&A. . In-Channel Boiling
A new computer program has been w .'$ten to calculate the coolant flow through the process tubes following a cr2 sheader failure. The program will calculate the hydraulic conditions existingrfollowing a crossheader failure at any point on the crossheader. It will. also calculate the pressure gradient along the crossheader and the coolant flow rates through the process tubes fed by. the stricken crossheader. It does not consider, however, the effects of two- phase flow. The calculated flow rates will determine whether or not boiling occurs in the channel, and burnout will be determined from experimental data.
8-B. Iodine Release Monitoring
It was determined that an estimate of the I-l3lin a sample of effluent water contaminated with fresh fissio$b products could be made by the analytical laboratory within one and one-balf hours. The sample had a concentration of a few nanocuries per milliliter I-13l;which would be representative of an effluent release of several hundred to a thousand curies'bf 1-131 to the river. After continuous samplers are installed, DUN wikl be able to make a timely determination with acceptable precision of any significant fission . product .release to the river.
Mission 10 - Colqmbia River Studies 10-A. Activated Silica Studies
The effectiveness of activated silica as a coagulant aid is being tested in the Water Treatment =lot Plant by addition of 1 ppm, as Si02, to the water after prior treatment wikh aluminum sulfate and sulfuric acid to maintain pH 6.6 and zero zeta-potential. This treatment resulted in an increase in the amount of floc that was settled in the retention basin; however, somewhat 5 shorter runs were obtained during filtration of this water. The concentra- tions of biologically significant radioisotopes increased in the effluents from both the aluminum and Zircaloytubes, with but two exceptions: a 30 percent reduction in the concentrations of AS-76 and P-32 was observed in the effluent from the Zircaloy tube. The magnitude of increase of the other radioisotopes was not large, averaging about 30 percent.
It is theorizedthat the general increase of effluent radioactivity reflected less efficient removal of the parent isotope precursors due to the more negative charge of the floc reflecting the influence of the added activated silica. The improvements in AS-76 and P-32 are assumed to result from a surface condition characteristic of the Zircaloytube and the quality of the water employed, by which retention of these anions is inhibited.
10-B. Radiation Resistance of Cat-Floc
The influence of radiation on the chemical stability of Cat-Floc (a pmpri- etary positively-charged polyelectrolyte) was investigated after various exposures with CO-60gamma rays. Any change in polymer composition was measured by the resultant change in surface potential of a suspension of kaolin containing 1ppm of the polymer. At this concentration, a zeta potential of ~20mV ks impartedto the clay particles. An absorbed dose of approximately 106 R completely destroyed the influence of the polymer, which resulted in a -19 mV zeta potential on the clpy. An absorbed dose of approx- imately 142,000 R significantly decreased thgeffectiveness of the polymer and this effect was magnified with additional time after exposure. Ten minutes after this exposure, the clay particles had a zeta potential of +11mV, which gradually decreased to neutrality one hour after irradiation. I
At an exposure of approximately 30,000 R, estimated to. be equivalent to .that, .- received: by a given volume of water passing through the reactor, the material maintained a zeta potential of a20 mV, which decreased to ~7.5mV five hours, after the irradiation. It appears, therefore, that this polymer might be suitable for providing a positively charged site for nucleation of negative species in the cooling water, and thus compete successfully with the film surfaces for these materials.
10-C. Water Filtration Mechanisms
The turbidity of effluent backwash water of pilot-scale equipment is being monitored to determine the optimum backwash flow rate for the process water treatment plants. Initial results have shown that it is possible to differ- entiate between backwash flow rates because of differences in the rates of. turbidity change with time. The optimum backwash rate for two laboratory . filters appears to be between 15.8 and 19.7 gpm/ft2 at a backwash water temperature of 20 C. Initial results also indicate that the optimum backwash rate is not constant, but rather begins at a high rate and decreases during the period of backwashing.
Because of extended backwash periods, the sand in the laboratory filters was disglaced up into the anthracite which resulted in rapid turbidity break- through during filtration. The columns were repacked with 3 inches of 0.45 mm sand and !2j' inches of 0.7 mm anthracite. The evaluation of the turbi.dimetric method of determining optimum backwash rate will be completed with the re- newed filters.
ENGINEERING AND TECHNOLOGY - REACTORS Graphite Studies
The irradiation of K Reactor graphite samples in the GETR will be curtailed after completion of the cu'rrent cycle, as necessitated by GETR space realloca- tion. The graphite irradiation program will .continue by irradiating only the TOO C samples in the GEFPR; the 850 C, 950 C and 1050 C samples will be switched to'the ETR. The ETR flux is too high for the 550 C sample, and the fringe of the EBR-1 is being considered as a substitute site. The 550 C data could be important to detedine whether a decrease in graphite temperatures below current limits would be advantageous from a reactor life standpoint. FTA-057 - Low Dichromate Test - B Reactor The goal exposure of the 28 natural uranium fuel control columns remaining in the reactor has been changed to 14-00 MWD/T. The twenty columns of high exposure fuel (about 1080 MWD/T) discharged in July are awaiting examination. _. Last-Ditch Cooling - B & C Reactors c .l'p' A study has been completed for the B and C Reactors showing the benefits to be expected from having the diesel pump flow augment the initial high tank flow, and from extending the high tank draining time by an automatic flow reduction. The first.change Mould increase the high tank cooling adequacy by about 30 percent while the second would increase the adequacy of the diesel pumps by about 35 percent. Both changes could be instituted at low cost, but use of the diesel pumps to augment high tank coolant flow requires obtaining additional reliability data on the diesel system.
Fuel Rupture Model
The fuel rupture model was developed primarily on the basis of hot spot failure data, and there is doubt that it applies to current operations. Battelle- Northwest has been requested to evaluate how well the model fits groove corro- sion failures which occurred after the model was developed.
Universal Flexible VSRs
Because of the failure of two universal flexible VSR assemblies 0n.B Reactor, an evaluation of the d-esign calculations was undertaken. Reexamination of these calculations indicates that the present design of the VSR shogd be satisfactory for normal operating conditions, with stress values in the rod approaching 20,000 psi. With the ultimate strength of the material reported to be approximately 100,000 psi at the time the rods were fabricated, this stress value represents a factor of safety of approximately 5. The calcula- tions also indicate that, if deceleration pressure buildup beneath the VSR piston is insufficient, the rods may fail due to impact against the seal housing.
Tensile tests were conducted on seven VSR material samples. Yield strength at 0.2 percent elongation averaged 73,400 psi. The average ultimate strength of the seven samples was 93,000 psi. The percent elongation averaged 1.82 percent. In short, the material as tested meets the requirements of the material speci- fication and compares favorably with test results obtained at the time the rods were fabricated.
Work is continuing at the 195-D test tower to verify experimentally the stresses calculated to be present during rod insertion.
Effluent Disposal Studies
Engineering in support of determining feasibility of alternate methods of effluent disposal has continued. With the deactivation of D Reactor, the D crib effluent disposal test has been terminated. Monitoring of test wells and springs is continuing, however, to accumulate data relative to recession of the groundwater mound formed during the test.
Design of a ground disposal test facility in 100-B Area has been completed. The facility will consist of a tie-in to one of the two 60-inch effluent lines upstream of the outfall structure, and installation of necessary piping and valving to deliver reactor effluent to the existing C crib. Excavation has been completed and piping installation is proceeding; >, I 4 . DUN-2873 Foam Encapsulation of Gases
Development has been initiated on materials for entrapment or enc of radioactive noble gases which could arise from abnormal reactor-operations. Initial preparations for the first test work have been completed. The present objective includes production of a foam capable of high expansion and resistant to deterioration by sun, heat, and wind. Twelve foaming agents were obtained from various vendors and on-plant sources.
.. The best foam to date has been a two percent Antaron FC-34, one percent glycerin, and one percent cetyl alcohol solution. This is a high expanding,.foaming solution and shows extremely little deterioration after four hours in the direct sunlight. The intent is to develop a foam which will last 24 hours under extreme conditions. Testing will be continued with other chemicals as they. arrive, and new stabilizing additives are being sought. . Iodine Release-to-River Monitors Installation of the iodine release-to-river monitors in the outlet sample building of the 107-C east retention basin is 75 percent complete. C-mma photospectrums were obtained at this location, and preliminary data indicate that a cave must be provided to reduce the high background level in the sample monitoring assembly.
The integrated radioactive iodine (1-13) release-to-river monitor is being developed to record dynamically the total quantity of 1-131 which might inadver- tently be released to the river in reactor effluent. Monitor capability will cover 1-131 release rates of one curie per minute up to 105 curies per minute, without responding to changes in radioactivity from sources other than 1-131.
Several approaches to the problem of separating the 1-131 component were explored, with the most promising approach consisting of a solvent extraction technique. This consists of pretreating lo7 Basin effluent such that radio- active iodine isotopes are extracted into the solvent which is then monitored by a gamma ray spectrometer. This effectively eliminates all interfering background isotopes other than iodine. Early experiments demonstrated the potential of the solvent extraction approach, but had not resolved the problems . of solvent, effluent, and chemical flow control, for a continuous on-line system,
Figure D-1 is a photograph of the operational prototype iodine solvent extrac- tion apparatus which has undergone successful testing at the 107-D Basin. In this facility, the iodine isotopes are separated fromthe sampled effluent by the following process: -
The precooled effluent sample is fed into the system by the peristaltic pump No. 4. While the sample flows into the extraction column, it is pretreated with NaN02 solution fed by pump No. 3 and mixed with an %SOb-KI solution fed by pump No. 2. This+results in the KI carrier being converted $0 iodine by the NaN02 acid condition. Pump No. 1 supplies CH CC1 ,“solvent directly into column where the iodine is a 22sor ed. ... UNCLASSIFIED i? DUN-2873 4:: ,?' I
Col.um n rec &S;ti rrer
C:on t rol Col
Eff
Int
Sol
1 31tor
Figure D-1. Prototype Iodine Solvent Extraction Apparatus 3. D-12 UNCLASSIFIED
. .-_-- ~ ..,___n . . , ...... - ~ . .- me CH3CCl3, having a higher density than'water; migrates to the bottom of the column where it is recirculated through the treated effluent-by ! ' a hollow-stem 'stirrer driven by an adjustable speed motor. The 'CH3CC13 component, containing' the iodine isotopes, flows from the bottom of the column to the NaI crystal detector assembly for automatic analysis. The. monitored iodine-enriched CH3CCl3 then flows through a solvent level,
control column and,then to a retention drum for future reuse., The . . -r effluent water from the top of the extraction column is transported to a drain.
Constant and proportional flow of effluent, solvent, and the effluent * treatment solutions is ensured by the four positive.displacement peristaltic pumps which are driven by a common electric motor via notched belts and pulleys.
Figure D-2 shows the response of the total system to an injection of Ir131 into the sampled effluent which simulated a. release-to the river in the 107 Basin'bulk effluent of 2.9 curies per minute for a one minute period. This one minute injection.of 1-131 was insufficient to allow the system to come to equilibrium, but the results indicate that equilibrium response would require approximately 45 minutes. On the basis of this test, a fingl system sensitiv- ity of less than 0.1 curie per minute in the bulk lo7 Basin.effluent is anticipated-.
Goo 0-01
/o/y
500
400 /O
300 - 0-O
I I I I I I I 0 10 20 30 40 50 60 70 Time After 1-131 Injection Began - Minutes
Figure D-2. Response of Iodine' Extraction Column . _..-., , .:
Ball 3X Bellows Modification - C & K Reactors In the report for May 1967, the outlook for difficulty with the Ball 3X bellows was defined, and the proposed solution by severance and shrouding of the bellows was described as the planned solution to the problem.
On August 28, Ball 3X channel No. 46 on KW Reactor was made available for demonstration of the bellows cutting and shroud application equipment under the authorization of FTA-079. The remotely-operated tools cut the bellows, applied . the preformed shroud, and clamped.it in place with no problems encountered.
RTD Response Time Test Unit A new method of determining strap-on RTD response time is under study. If successful, this method will be applied to all reactor strap-on RTD installa- tions.
The response time of an RTD is related to the heat transfer properties of the materials between the resistance element and the heat source. It is also affected by the heat sinks in its construction. The proposed method assumes that the response time of an RTD during step-change can be correlated to element change in resistance during heating of the RTD element by a current pulse and its subsequent cooling. A resistance bri.dge has been fabricated in which the RTD element resistance can be monitored by use of an a.c. signal while it is being heated by a doc. source. Initial tests on a copper heat sink showed that the response time curves are similar but of different magnitude than those obtained on a reactor "pigtail" assembly. The problem of correlating these data with pigtail response times'is being studied.
Columbia River Cooling Program
A liaison and data collection trip was made to Grand Coulee Dam on August 8. Forebay temperatures at the dam were surveyed and information gathered which indicated a zone of cold water to exist at a location advantageous to the performance of the annual cooling program. F?ith completion of the flashboard installation at the dam, and elevation of the reservoir pool to 1,290 feet msl, the cooling program was implemented on August 18 through selective discharge of waste water through tubes located at 1,035 feet msl. Project Engineering - Reactor Facilities Project Status Summary
The status of approved construction projects relating to B, C, and K Reactor . facilities is summarized in Appendix A. P 5 :. I
ENGINEERING AM3 TECHNOLoeY - FUELS AIVD TARGETS Recycling of Hot-Die-Size Cores
Multiple cycling of uranium fuel csres in the hot-die-sizingprocess,would result in significant cost savings over the current specification which limits the fuel core to one cycle before it is scrapped. *This limitation iias arbi- trarily assigned during earlier stages of process development, and was based. upon the lack of technical experience with multiple cycling. In the light of this background, tests have been conducted to determine the effects of multiple recycling on the metallurgical properties of the uranium cores. These tests were conducted in.the following areas:
X-Ray Diffraction Study
Following feedsite fabrication, uranium core blanks are beta heat treated to eliminate the crystallographic texture induced during rolling and to produce a randomly oriented structure; Multiple'temperature application during successive processing cycles could give dse to in-bernal stresses with which the grains might become aligned. With the graiils in aligned positions, the core properties. would' become anisotropic. This condition could result in significant dimensional changes- during irradiation of multiple recycled cores,
To evaluate the effects of multiple temperature cycling on the texture of hot-die-sized uranium cores, 32 cores were processed through end closure welding five times. For X-ray diffraction measurements, three cores each were selected at random from the virgin material, and after the first, third and fifth processings. The G-2 coefficients at points on the O.D., mid-wall, and I.D. are summarized in the fallowing table:
Longitudinal X-ray No. Times Measurement Diffraction Coefficients Processed . Point O.D. Mi M? I.D. 0 (virgin) Cap End 4,023 -.0145 -.021 2.001 1" below cap end +.006 -.020 -.021 +.Olg 1 Cap End 6.036 0.003 -.060 *.021 1" below cap end e.O5b +,008 -.024 e.029 3 Cap End 6.044 4.005 -.062 +.006 1" below cap end e.027 -.OO& -,040 1.025 5 Cap End 6.021 -.025 -.obi +..044 I" below cap end +.065. *.OOg e,001 h.063 The data show noesignificant texture changs or trends which would adversely affect the gro'tJth (length) of the fuel coree Also, when pole figures from the X-ray data were mapped out, no great tendency was shown for any one plane or'planes to attain an oriented stmcture such that when the core was irradiated the plane or planes would grow. . . Hardness Measurement
Hardness numbers are a measure of a material's ability to withstand indenta- tion, which in turn i'spdependent upon the crys-fal structure and stresses within the material. Multiple cycling of uranium cores through the hot-die- sizing process could alter core stresses and induce more stresses when the cores are sized and end-bonded.
To explore these stress effects, Rockwell hardness numbers were taken on the virgin cores, and after the first, third and fifth processings. The data are summarized below:
Times Hardness - Rockwell B Scale Processed 13. M1 M2 O.D. o (virgin) 9 14.5 16.5 18.5 1 1-3.5- 16.85 16.0 18.5 3 15.0 17.85 16.25 17.0 5 15.0 18.25 18.25 17.0 It appears that the increasing hardness numbers at the I.D., and the decreas- ing hardness numbers at the O.D., show a trend of increasing residual stress at the I.D. and decreasing residual stress at the O.D.
Electronmicroscopy
Cores fromthe virgin, first and fifth cycle groups were selected and samples prepared from the longitudinal surfaces at the cap end and one inch below the cap end. The surfaces were magnified at 4,500 and l5,OOO times, respectively, and photographed. From these photographs, no change in grain structure, inclusions, or alloying constituent precipitation at the grain boundaries were observed. Aluminum Corrosion Testing - Hot-Die-Sized Fuel An additional three tube charges of hot-die-sized fuel have been prepared for irradiation under ITA-060-A, to relate hot-die-size process variables to aluminum ledge and groove corrosion. These three columns will be discharged early and examined for corrosion which could cause failure before goal exposure is reached. Evaluated will be the effects of: (1) reduced core size and increased clad thickness, (2) annealing the aluminum cladding after diffusion bonding, and (3) end bonding with a shorter end-bond cycle while employing higher power and pressure,
The graphite embedded in fuel element aluminum cladding is also being investi- gated as a probable cause of ledge-and-groove corrosion. Graphite is used as a lubricant during fabrication of hot-die-size fuel elements, and galvanic corrosion due to aluminum-graphite coupling is ~OIMto be severe. Prolonged water immersion of as-produced hot-die-size fuel (without cleaning) has shown severe localized corrosion. Fuel elements cleaned of grapFite by etching off five mils of the surface are be%& prepared for irradiati&. The tube charges will be made up by alternating etched hot-die-size, unetched hot-die-size, and AlSi-bonded fuel. .. TECHNICAL ACTIVITIES - N REACTOR
RESEARCH AND DEVELOPMENT ' Mission 1 - Basic Production 1-A. Prediction of Fuei Failure Rates
In late 1966, -an expression relating the probability of eiid-associated ;dark I fuel failures to column average exposure was developed. The validity of this expression was questioned'when no failures occurred in . 202 2olumns of Mark I fuel irradiated to goal exposures between 3065 and 5514 MWD/T, compared to the prediction of 14 failures. Because this expression relates failure. probability only to tube average exposure, it cannot provide guidance for 'improving fuel design, nor predict differences in failure rate between the inner and outer fuel components. Although the specific power of the outer element is 2.6 times that of the inner element, and the exposure of the outer element is about 1.2 times that of the inner element, only two end-associated failures have occurred in outer elements, compared to seyen in inner elements.
Surface heat flux expressed as Btu/hr-ft2 relates specific power, geometry, and temperature factors for each fuel component. An expression was developed which related the probability of end-associated failures to surface heat flux, 4; and time of exposure, t: -
This expression predicts the proper ratio of end-associated failures between outer and inner elements, and only 3.2 inner and 0.8 outer failures for the 202 columns irradiated to high exposure. This expression may prove useful in the design of future fuel elements. I-B. Irradiation of Mark IV Fuel - FT-NR-94 Three columns of 125 Metal Mark IV fuel assemblies have been fabricated for an. irradiation test to simulate 4800 M'GJ operation. This fuel was charge'd into I? Reactor late in August.
Operating parameters for these Mark IV fuels were calculated and are listed in TABLE DM-I. The most severe operating case is seen to occur with the Mark XV fuel in a Mark I-C reactor loading.
\ 'a
1
DN-1 TABU DN-I
FIVE 'CELL TUBE PARAMETERS MAFtK I, MARK 11, MARK IV (Columns Operating At Normal Flow, Power)
IK IV MK IV in in 14K I-c MK I1 PK I-c I1 REACTOR REACTOR REACTOR REACTORm 6 Reactor Flow - 10 lb/hr 85.4 84.0 85.4 84.0 Riser/Riser AP - psi (Max. Flow and Power Tubes) 178.3 178.3 191.5 Tube Flow - 103 lb/hr (Max. Flow and Power Tubes) 92.46 91.0 105.54 110.02 Enthalpy Imb. Ratio 1.08 1.125 1.15 1.15 Tube Power - kw 5117 5013 5587 5050 Peak Spc. Power - kw/ft 212 205 .4 258.6 226.8 Flux Shape - Axial Peak to Avg. Fwr. 1.45. 1.27 1. 22" 1.15"" Column Length - rt. 34.8 30.92 26.1 26.1 Augmentation Dist. - ft. 1.636 5.47 5.986 7.88 Normal Ah - Btu/lbm 192 184.3 180.7 156.7 No. Elements in Column 18 14 12 12 Active Zone AP - psia 103 3 111 I- Fitting AP - psia 75 79.5
Inlet Temp . = 392 F Inlet Enthalpy = 368.9 Btu/lbm Flow L.T./Fnorm = 0.92 Rear Riser Pres. = 1438 psia low pressure trip point Max. Subchannel Enth. Imb. Ratios: Outer ann. Ratio = 1.06~ Inner ann. Ratio = 1.00 Hole Ratio = 1.15
+Same over-all cosine length as PK I-C. **Same over-all cosine length as MK 11.
.-.....' I
Analysis has shown that the 125 Ketal Mark IV fuel will .satisfy require: ments for single-tube protection with respect to the. zone temperature hlonitor and outlet tem;?erature limits for both low and high flow-monitor trip settings. then calculated on the basis of a normal axial. flG profile, 10 percent local peaking, and with flow and pressure kt their respective low trip points , the worst-case boiling burnout 'heat flux ratio' for i4&k IV fuel is shown in the following tabulation (Cosine Flux Profile): .. '.
Tube Ratio: Locai Heat Flux Fuel Type Power, kw to Boiling Burnout Heat Flux Mark I-C 5000" 0.18 .. Mark I1 5650% . 0.30 Mark IV** 5587 0.24
. "In unifofm loading, 4800 MW reactor. ""In Mark I-C reactor loading, 4000 MW reactoro
Under the conditions shown, the burnout ratio for Mark IV is smaller than the maximum experienced with Mark I1 fuel during the 4800 MW demonstration, With the assumption of an accidental multiple control rod swap, zone temperature,monitors on adjacent Mark I1 or Mark I-C columns would scram the reactor prior to boiling burnout in Mark IV fuel, . Considering the most pessimistic rear flux peaking ratio of 1.91 (peak to average) and a 26 percent increase in Mark IV specific power over that of adjacent Mark I-C columns, film boiling could occur over the downstream three feet of the Mark IV center channel at the Zone Temperature Monitor (Z!I'M) trip; this would produce a peak uranium temperature of no greater than 1620 F,'which is well below the temperature at which prompt fuel failure could occur. The trip'would occur prior to occurrence of bulk boiling in the Mark IV outer annulus. This would satisfy the safety aspects of the ZTM protection philosophy.
All analytical results reported are based up,on Mark IV characteristics predicted from-out-of-pile test data, Thermocouple trains will be installed in two of the first Mark IV columns charged into the reactor to enable adjustment of predicted characteristics if necessary.
1-C, Graphite Studies ? %.e Use of a silicon coatikg to protect reactor graphite from oxidation is under consideration, Initial tests indicate that such a coating could "provide an aaequate margin of protection. A small semple of TSX graphite that had been treated with tetramethylsilane, (CH3)4Si, reacted at 644 C in air at an initial rate only one-tenth as fast as an untreated sample, Although the protective effect diminished with time, the treated sample reacted two-thirds as fas r of oxidation. - . e...
?, ...e.-. - Current proposals suggest that a protective coating can be formed by utilizing the radiation sensitivity of 'the organosilane class of silicon- containing compounds. Since it is known that organosilic'on polymers undergo radiolytic degradation and crossl5nking, it seems feasible to expect that a silicon-containing coating will form on the graphite if an organosilane were introduced into the reactor gas system.
An experimental program was arranged with Battelle-Northwest (BNW) to furnish laboratory data on coatings obtained from thermal and radiolytic decomposition of two candidate organosilane--tetrathylsilane, and c6H5 Si (CH3)s. The degree of protection afforted by such coatings .. will be measured.
l-D, Advanced Technology Program
Analytical studies are continuing in determining and assessing the operating conditions and plant limitations associated with the antici- pated higher reactor power levels of the Advanced Technology Case. Some of the principal areas of study have been as follows: Thermal Hydraulics , Physics, Primary System, Secondary System, Core Materials Limitations, Building Heating and Ventilating , Bottom Thermal Shield, Confinement Requirements , Maximum Steam Generator Pressures.
Mission 2 - Coproduct 2-A. Rod Calibration Tests
Detailed analysis of the rod calibration test results yields a value of 53.6 mk for the cold,xenon-free excess reactivity for Mark I1 fuel in N Reactor and for the inserted rod pattern at that point. A theoretical calculation of that rod patterm, using the two-dimensional f our-energy- group model 9-AMGIE,, yields a value of 56.1 mk, or 4.7 percent high. The same model predicts the strength of the entire system to be 82 mk, which then normalizes to a 78 mk estimated strength. This value is one of several which will be compiled by analyzing various rod configurations, to obtain a minimum probable system strength.
Calculations indicate that the system strength will exceed the preliminary value of 69 mk which has been used in the interim for evaluating the reactor shutdown margin, and should approximate the value of 77.5 mk reported in the Coproduct Hazards Summary Report.
2-B. Shipping Cask Analysis - Mark I1 Fuel Calculations of the reactivity of the shipping cask loaded with 12 coproduct canisters yield a kerf of 0.85, or 150 mk subcritical. This value is very satisfactory provided that the theoretical models utilized can be shown to be accurate to better than 15 percent k.
r.$j ,r.. P DUN-2873
0 2-C. Mark I1 Target Rupture Investigation
Tests to determine if a Mark I1 target failure would be revealed by the presence of aluminum in the channel effluent were run. Test procedures involved connecting the 400-channel gamma analyzer to the gamma energy crystal detectors on the rupture monitor. The system is functional and results do yield a reasonable-looking background crud spectrum. Bowever, the tritium content in the loop indicated that the release had subsided by the time that the system was brought into use an& no aluminum activity was detecteci, Since tine background gamma level was low at the aliixinum decay ecergy (1.78 mev), a detectable signal should.be obtained if the corrosion rate is high, Mission 3 - Transplutonium Technology There were no significant developments on this program. .Mission 4 - Pu-238 4-A. Dresden Tails
National Lead of Ohio (NLO)-has resumed operations after the vacation shut- down at the end of July. It is understood that ingots made from the Dresden tails are being prepared for primary extrusion.
4-B, Yankee Tails
The complex nature of the U-236 management program has .caused NLO to request advice and assistance-in determining the best method of introducing Yankee tails into the production reactor fuel.stream.
Six alternatives for the production of N Reactor fuel have been outlined '. in a letter from NLO (NL0406-494) transmitted to AEC-RL.and reviewed by the U-236 Management Committee. The consensus was that the third alterna- tive outlined in TABIZ DN-TI was probably- the best approach.
In comparing the various cases shown on TABLE DN-11.1, the total neptunium production was calculated with the assumption that equal quantities of - were to be charged into the reactor, using as a base the maximum tonnage alternatives for both the 0.947 and 1.25 w/o U-235 materials. The differ .*ce from these values was assumed to be made up from 500:Series metal on hand at Hanford. For the 0.947 w/o U-235 fuel, the U-236 content was assumed to be 349 ppm; for the 1.25 w/o U-235 fuel, 140 ppm for the 500-Series metal on hand, Tne Alternatives 4A and 6A are additional cases in which N Reactor metal scrap was added as in Alternate 2. I?!!the calculation of neptunium production, it was as umed that the fuels wohd be fabricated in Mark IV geometry and that th ,&ford production yield was 75 percent. TABLE DN-I11 provides the composition4 of the charges and the neptunium productions. TABU DN-I1 ALTERNATIVE BLENDING PLANS'FOR YANKEE TAILS BLEND MATERIALS PRODUCT Alternative Uranium Source -Tons w/o U-235 ' 'ppm'U-236 3/0 U-235 TonsiL' Ep m U-236 KJS U-236 1. Yankee Tails 50.00 2.43 2000 Depleted U3O13 91.78 0.142 22 0 947 67.6 738 45.4 N Reactor Metal 10.00 0.947 1000 (Dresden Tails) I s ?-
. ... (Dresden Tails ) 10e 00 0 .947 1000 1.25 52-07 955 45.2 N Reactor Metal (~k11 Scrap) 3.56 2.1 150
3. Yankee Tails 50.00 2.43 2000
SR V-R Tails 374.24 ' 0.75 . 500 0.947 193 29 684 102.2 Reactor Metal '? (Dresden Tails) 10. 00 0.947 1000 ' 7 .. . 'Yankee Tails' 2, 50.00 2.43 2000 ev3- SR V-R Tails 118.00 0.75 . 500 1.25 74 29 946 63.9 *-- 5. Yankee Tails 50.00 2.43 2000 E=i SR Normal Tails 247.70 0.65 150 0 * 947 138,O 478 60.0 c N Reactor Metal (Dresden Tailsj 10.00 0,947 1000
6. Yankee Tails' 3, 50.00 2.43 2000 . I. I SR Normal Tails 98.00 0.65 150 1.25 65 46 77 5 41.6 R> ,I . -4zgs (1) First run casting yield. W (2) If N Reactor Metal (Dresden Tails andMK I1 Scrap) are added there will be 80.71 T of 1.25 wlo U-235 at 932 ppin U-236 (75.2 Kg U-236). (3) If N Reixtor Metal (Dresden Tails and IvlK I1 Scrap) are added there will be 71.88 T of 1.25 :I/O 11-23? at 775-pprn ~=?36(55.7 Ke; [I-236). . I. t , .j .: . $. # :.-
DUN-287 3
NEPTUlVIUl4 PRODUCTION .. -..Case Fuels - Tons a *- 50.7. Alternate 1 2.0 94.3 500-Series 0,947 2.6 95.3 500-Series SDike -240.3
2 14500 Alternate 3 5.5 w95.3 500-Series Spike 3 103,5 Alternate 5 3.3 41,5 500-Series 0 e 947 1.2 95.3 500-Series Spike ZGZ
1: 145.0 500-Series 0.947 4.0 61.4 Alternate 2 2.8 33.9 500-Series Spike 1.0 ZTZ . -7.8
5. 145.0 500-Series 0.947 4.0 88.7 Alternate 4 4.0 6.6 500-Series Spike .2 m3 -8.2
6 145.0 500-Series 0 947 c 4.0 95.3 Alternate 4A 4.4 _I_ 240.3 mi 7 145, 0 500-Series 0.947 4,O 77.3 Alternate 6 30O 18,O 500-Series Spike -.5 w 7.5 8 145 0 500-Series 0 .947 84.9 Mternate 6A 10,4 500-Series Spike 2in
NOTE: Data are for first-run castings. The internal recycle of metal scrap at NLO to produce second-run castings will provide approxinately 50 percent addi€ional fuels, and ultimately 50 percent more neptunium than shown in the Table, *!$E n. .. Information tabulated in TABU *%-I11 indicates that ody Cases 2 and 6. Alternates 3 and kA, need to be considered. The U-236 Working Committek has decided to recommend that NLo’s AlterGte 3 (blending Yankee tails with Savannah River V-R tails) be used. This was decided because neptunium will be available sooner if produced in the .947 w/o mterial than if’mde in the 1.25 w/o material. Information from AEC-RL indicates that additional alternatives have been submitted to the Savannah River Plant for recommen- dations. Details of these alternatives are not known. Presumably, the recommendations from both sites will be coq’aredand the final decision will rest on the conclusion as to which will”yie1d the most neptunium. 4-C Pu-238 Demonstration’Program A final report (DUN-2913) summarizing the ‘results of the a-238 demonstra- tion program (13-NR-68) has been drafted, in .which. the experimental procedures and conclusions of the irradiation testing are discussed; The exposure-dependent production data resulting f’rom measurements aken along the three neptunium target COh~msare given in ‘PABIE DN-IV, and the PU-236 contamination in the Pu-238 produced with various irradiation times is L plotted in Figure DN-1.
1.0
0.8 % ?Ja .f 0.6 %
6 I E 0.4 I 1 I 1 t I I 44) 60 80 100 120 140 160 180 Equivalent FLUPower Cays
Figure DN-1. PU-236 Contamination vs . Irradiation Time c TABLE DN-IV .. EXPOSURE-DEPENDENT PU-238 PRODUCTION PARMETERS
Equivalent NP Pu-238 Made/ Conversion Pu-238 Full-Power Daye Consumed Np Charged . . Efficiency Purity Measured along Column 0856 (70 g-Np/ft) 8.0% 7.3% 91.3% 95 2% 10.0 8.9 89.0 93.4 90.0 11.5 10.2 138.7 93.0 11.0 9.8 89.1 93.14 74.5 10.0 8*9 89.0 93.9 8.8 - 8.0 90-9 93.7 44.0 7.0 6.3 90,o 95.4 Measured along Columns 0858 and 0860 (45 g-Np/ft) 174.8 24,7% 18.9% 76.52 85.1% 23.3 18.0 77.3 86.0 162,4 23.0" 17.9*. 77.13* -85.9% 22.6 17.5 77.4 85.8 21.43 17 .0" 79*44 87 0% 135.00 20.9 16. G 79.4 87.2 19.6 15.6' 79.6 87.0 19 .6% 15 8% 80.6% 87.8% 17 -3 14.2 82.1 88.3 104.0 17 1* 14.2" 83.0s 89.0" 16.2" 13.6" 83 'g:* 89.4" 76.6 14.7 12.6 85.7 90.8. 13 .1% 11,2" 85.5' 90 2* 67.7 11e7* 10.2% 87.2% 91.4"
(l)The equivalent full-power days were calculated only on the elements containing flux monitor wires, The values marked with an asterisk are for Column 0858, The unmarked data are for column 0860. The experimental data obtained with 45 and 70 g-Np-ft have been used to derive production parameters for a theoretical target con-baining . .. 120 g-Np / ft . To obtain a maximum 31.8 kg of Pu-238 by January 1972, using only Hanford- generated neptunium, a variable target throughput mode of operation is proposed. This operating scheme, designed to yield a throughput quantity compatible with the separation capacity, uses a 45 g-Np/ft target exposed for 87 full-power days to yield 90 percent quality material for the first year of operation (October 1967 to October 1968). Starting in October 1968, and continuing through October 1969, the target design has been changed to contain 70 g-Np/ft with an exposure of 122 full-power days, The mode of operation for the rest of the campaign (October 1969 to January 1972) would be 70 g-Np/ft target operated for 225 full-power days to give 85 percent purity,
On the foregoing basis, the schedule for producing the 31.8 kg of Pu-238 in N Reactor would be as follows:
Calendar Np-237 Pu-238 Yea Feed, ks; Made, kg
1968 29.65 4.4 1969 , 22.44 6.4 1970 21 .20 7.7 1971 25 13.3 7 f Tot a1 31.8 - Mission 5 - Other Isotopes There were no significant developments on this program.
i.I p Xission 7 - Target Space Enhancemelit Advanced Fuel Design
A parametric study'was made to determine the possible rage of tiiemal- hydraulic fuel design parameters that would be associated with iI Reactor power levels of 4800'14W and higher, assuming that bulk boiling in fuel subchannels could be tolerated. Calculations were made for radial flatten- ing efficiencies of 80 and 85 percent. As illustrated in Figure DN-2 (80% flattening efficiency), increases in' the desired tube powers result . in significant decreases in pressure drop available to force coolant through the fuel columns with significant increases in the maximum outlet temperature. Results for the 85 percent flattening efficiency were always identical,
1n.addition to subchannelboiling, the system pressure (at the rear risers) selected for a given' tube power wes based upon the required tube flow rate, the corresponding system pressure drop and normal reactor operation with an 1825 psig pressure at the primary pump discharge. Previous studies considered no subchannel boiling and a fixed rear riser pressure conservatively selected on the basis of operating experience. The active zone pessure differential for.a tube power of' 6000 kw could be increased to 183 psi with subcliannel boiling, compared to 145 psi when no subchannel boiling is permitted. .The increased active zone pressure differential could be reflected in an increase in maximum fuel'w'eight obtainable for given maximum power level. Conversely, reactor power level could be increased while maintaining the fuel weight at the no-subchannel boiling design level.
' Plans are being formulated for a test to demonstrate stable operation with subchannel boiling and outlet connector boiling in%wo fuel columns in the N Reactor. i4ission 8 - Nuclear Safety Program
%A. Studies of Overheated Fuel
The project to equip BNW's 324-D facility to permit melting tests of irradiated fuel continues to progress. Facility design is now 90 percent complete.
Fission product release tests and metal-water reaction studies with irradiated fuel heated to failure temperature were to be resumed this month in the BNW high-level radiation cell, but facility modification .. is behind. schedule,
i - DN-11 Q Q) ,j78 3 03 576 574 120.I
LOO -.
80-
60 - Uet Temp. 400 F (150 psia Steam) Outlet Pressure 1600 psia Central Tube Flow = 1.06 ave = 0.8952 at LP Trip i max sat Low Flow = 0.94 Noma1 5 Percent Instrument Error Allowance \
2ol01 I I I I I I I 6000 6200 6400 6600 6800 - 7000 7200 7400 '600 Max. Tube Power - Kw
,Figure DN-2: hximum Tube Conditions With Subchannel Boiling (Flattening Efficiency = 80 Percent)
. ,. . ... -- .--.I ... . .
.5 $? -- :a ?
8-Be. Meteorological Studies Erection of the 300-foot meteorological tower vas completed at the PT site August 21. Bid letting for the sensors can proceed in about one month after the F'I4 radio frequency assignment for data transmission is a3proved.
.- . ENGINEERING AND TECHNOLOGY - I? REACTOR Fuel Spacer Redesign
An experimental program to evaluate the effects of design variables on apparent spacer motion was completed, Pertinent results, trhich are to be . reported in DUN-2948, are summarized briefly as follows:
e Mark I1 (post-coproduct ) sprcers produced significatly less' vibration in the section downstream from the fuel column tha? was experienced with Mark I spacers.
e Longitudinal alignment of spacer supports decressed vibration levels additionally,
Q Non-perforated, spacers vibrated at significantl3- lower levels than perforated ones ,
o Cruciform spacers produced virtually. no vibrztion.
o Spacers having supports with a large contact area (prototyp Mark 11) produced less vibration than those having small con- tact areas , or line contact with the pocess tube.
Test results provided little information applicable to the elimination of mechanisms responsible for spacer vibration. It was determined that the downstream charges were subjected to substantial axial loEdings which fluctuated by amounts up to 75 pounds.. Visicorder traces fromtne load cell correlated well with those fron; strain gage feelers a%tached to a spacer, and this has proved that the axial force fluctuation was related to lateral motion of the spacers, Additionally, the load cell results correlated well with those obtained from accelerometers zttachcd to the outside of the process tube. These results lead to the folloving conclu- sions :
o Unstable buckling action .of the spacer column is one of the mechanisms involved in the vibration. While it may not be possible to eliminate the forcing function, any design modi- fication that would damp lateral motion should reduce impa& forces on the process tube wall.
.t - ,__,......
e The fuel column may be subjected to substantial dynamic forces. due to spacer motion. These forces undoubtedly drive the fuel into vibration, with potentials for fretting at .the fuel. supports
' and for damaging the internal support systems ofthe f'uel; There is an 'obvious need for damping the spacer vibration as much as possible .
e Process tube vibration levels are directly related to spacer- ' , motion . The accelerometer arrangement is therefore adjudged to evaluate design prototypes ,
The results obtained from the recently completed testing program indicate that substantial improvements in spacer performance can be achieved through relatively simple modifications These improved results should serve as an interim solution to the process tube fretting problem while a more comprehen- sive study is being made of other concepts. It has been proposed that action be undertaken in the following order:
1, Additional post-coproduct (Mark 11) spacers ordered should have supports aligned longitudinally,
2, An investigation should be made into the feasibility of utilizing noii-perforated spacers in most of %he length of each spacer column. Factors to be considered are:
0 Procedural measures in charge makeup or design features which' would ensure that non-perforated spacers do not block connector ports
0 Requirement that enough perforated spacers be included in each column to absorb energy sufficient to prevent damage to the inlet nozzle or end cap in the event of an inlet connector failure severe enough to.cause upstream displacement of the fuel.
e Provision for additional testing to determine vibration characteristics of charge makeups compatible with Mark I, Mark I1 and Mark IV f'uel. Prototype non-perforated spacers with longitudinally-aligned supports are required for this test.
3. Vork should be started on the design. of a tubular spacer having the largest possible diameter that could be installed without hampering adequate removal of heat from the process tube wall., The concept behind this recommendation is that, by reducing the annular gap between. spacers and tube wall, the damping- effect of the water on spacer motion can be substantially improved. Features required are: 0 Means for providing beering on the fuel end cap without interfering with flow distribution in subchannels of the fuel column.
0 Means for minimizing pressure drop at the outlet nozzle.
0 Adequate energy absorption characteristics (,as in C-1).
Reactor Piping Decontamination Fourteen-Tube Decontamination Test - PT-NR-91 . Results from radioisotopic analyses of carbon steel dummies before and after decontamination, and from analyses of samples of the decontaminating soliltion, have been obtained. Dummy analyses indicated thzt in the tubes given the optimum decontamination conditions, CO-60 (which accounted for 57 percent of the activity) was 99 percent removed. Fe-59, h-54, -and Ce-141, which together accounted for 30 percent of the activity, were 97 percent removed, and Sb-125 and ZrNb-95, which accounted for 10 percent of the activity, were 90 percent removed.
Analyses of the decontamination solution samples indiczted that CO-60, Cr-51 and Fe-59 accounted for about 90 percent of the activity taken into solution and Vn-54, ZrNb-95, and Ru-103 accounted for the remainder. Integrating the radioactivity in the solution samples from one tube over the 20-minute decontamination period, and extrapolating this result to. the entire reactor, provides a value of 180 to 200 curies expected to be removed during the full-reactor decontamination program, Full-Reactor Carbon Steel Decontamination - PT-NR-93 All material that will be in contact with the inhibited phosphoric acid decontaminating solution, Turco-4512-A, have been evaluated, except for 17-4 pH steel, 416 and 420 stainless steels, and Easy-Flo 45 braze alloy; these are being tested. No detrimental effects of exposure to the decon- taminating solution have been identified. A small-scale test of the effect of gamma radiation on the interaction between Turco-4512-A and Zircaloy oxide is also being performed. Results of these tests will be obtained before the reactor is decontaminated.
An engineering study of the proposed full-reactor decontamination was made to determine the potential risk to primary loop hardware. Grayloc couplings, V-11 and V-12 valves, and nozzle rolled joints were decontaminated with Turco 4512-A and disassembled for inspection. Examination of the component materials revealed no harmful effects from the decontamination . solution. It is felt that a relatively lair risk factor is associzted with decontamination, provided all dead legs, drains, vents and instrument lines art thoroughly flushed, and all cone-type valves are inspected to ensure I that no decontaninztion solution is trapped in the pockets formed by the valve body s... .
Two nozzle rolled joints were sectioned and micrographs taken, The section in question was the crevice fonhed by rolled joint.step and the process tube, Examination of the photos (taken at a magnification of 25OX) revealed no indications of intergranular corrosion in either the Zircaloy-2 or the carbon steel.
Cobalt Target Element Supports
It is planned, in the near future, to irradiate 20 cobalt target elements in a graphite cooling channel for isotope production. An improved support, consisting of a circular ring with three evenly. spaced ribs, was designed and a sufficient quantity fabricated, Work of attaching the supports to the target strips is complete, The graphite cooling tube mock-up in 1705-N is . being set up to run flow/pressure drop tests on the cobalt targets and their associated spacer train..
Five-Loop Reactor Operation
The capability of N Reactor to operate at the design conditions of 4000 liw and 800 MWe with one of the six cells out of service was discussed in the July report. This mode of operation has continued, although electrical power generation rates have dipped occasionally below 800 MWe because of WPPSS-BPA operating problems that were not reactor related,
Attention has now been focused on the. capability of the reactor to operate at a thermal power in excess of 4000 MW with five cells . .Three. basic limitations . must be considered in.determining maximum thermal power capa- bility in this mode of operat'ions
1, Fuel specific powers must be limited to satisfy emergency cooling and confinement criteria.to avoid excessive fuel heatup rates on interruption of cooling. i
2, Tube outlet temperatures must be limited to ensure adequate margins. for " single-tube" thermal hydraulic protection ,.
3, Equipment and instrumentation capabilities must not be over-t axed,
Fuel specific power limits for Mark I1 fuel permitted operation at 4800 lrlw, .These limits do not depend upon coolant flow rates; however, with the non- un,iform coolant flow distribution obtained with 5-cell operation, the reactor was purposely deflattened to match tube power with the tube coolant flow rateso At reduced flatten-ing efficiency, the total reactor power is decreased far a ' given specific fuel p6wer. In a practical sense, specific power limits would now probably permit reactor operation at some point above 4400 IN, but below 4800 MW.
:. . . . . - .. .'. i -. . .. I p
-n 4.: Tube outlet temperature limits are now approached r osely , However, by reducing the secondary system steam pressure, the reactor inlet teqera- ture would be reduced which would permit an increase in the reactor power. At 90 psig main steam header pressure (the minimum that can be attained without equipment alterations), it should be possible to operate at about 4400 MW with a small but reasonable margin to limits. This reduction in steam pressure would reduce the WPPSS output to about 540 file.
Confinement System Le &age
With the completion of lubricating oil piping vent modifications in the 109-Building cells, a path for air leakage from the Zone I confinement area to Zone I11 was formed. This leakage path is through the seals at each end of the primary pump shafts and the seals at the pump side of the extension shaft and finally out the oil system vents to Zone 111.
An analysis of the leakage through the primary pump lubrication system indicated that the leakage would be about 127 cfm from the six cells ,2t. 9. pressure differential (Zone I to Zone 111) of 6 inches wzter gaQge. This reFrescxSz =bout 5.8 percent of the total leakage from Zone I as-neesured durinz +,he last leaktightness test conducted in mid-1966.
In-Pile Boiling Test, V-11 Valve Modification
Equipment has been ordered to provide remote throttling capability to two 105 V-11 valves, Environmental testing of the components will be done at the test facility in the 189-D building, This work is aimed at determining the feasibility of remotely throttling the inlet valve for a planned Production Test on in-pile boiling.
In-Core Assemblies
Recently an approach to a combination assembly of fission chambers and self- powered detectors has been conceived. It is planned to combine a rhodium detector and a vanadium detector with a conventional fission ion chamber assembly, This assembly can be used to replace existing in-core fission chambers which are approaching end-of-life due to burnout. The advantage of this approach is that it will provide the required replacement of the existing in-core fission chambers with a chamber of a known type and per- formance, and wZll yield the experimental data from the self-powered rhodium and vanadium detectors ENGINEERING AND TECHNOLOGY -.FUELS AND TARGETS Mark IV Uranium Metal Quality
The circumferential cracking problem on the inner diameter of the outer billets has been discussed with feedsite personnel, It was 'generally agreed that the cracks are defect (gas pores) ingot related and are not quench caused, Quench cracks do not appear as circumferential cracks, but as transverse radius cracks, This analysis was confirmed by comparing photo- graphs of previous alloy work where cracks were definitely produced by quenching, It can be reasoned that cracking occurred on outer billets and not inner billets because of the higher primary extrusion reduction ratio on the latter, The higher reduction ratio would cause the metal to "pressure bond" together and close the void. Gas pores can be observed by autoradio- graphing the ingot, NLO will perform autoradiography on the next series of castings of this alloy to confirm the above &alysis, Mark IV End Weldin5
Double-pass welding on Mark IV fuels is perforning well compared to single- pass welding on the Mark I1 fuels, Available data indicate an inner fuel defective rate of 005 percent for 1427 ends, and an outer fuel defective rate of 1," percent, The improved performance is attributed to the lack of carbi.de inclusions in the weld zone, Surface porosity is virtually eliminated on the Mark IV closure, although a significant number of carbide inclusions still exist in the as-cast braze metalo
Mark I-C Rework Campaign
The Mark I-C rework campaign will be initiated as soon as.winged shoes become available, Initial delivery of winged shoes from an off-site vendor is scheduled for October 15, 1967; however, on-site fabrication of shoes should allow start-up in eerly September,
Outer diameter measurement' data on Mark I fuels indicate that many fuels will require 22-23 mil winged shoes to increase support circle diameters to 2.690-inch 2 0,005. Steel shoes have not been produced from material thicker than 18 mils, and initial tests indicate that it may be necessary to ,reshoe a number of fuels with shims under the shoes to develop the required support heights, To establish if the use of shims would be detrimental to shoe integrity a two charge-discharge tests of shimmed fuel were performed during the month, Results from these tests indicate that shimmed shoes are satisfactory, provided that at least a 15-mil shoe is used with the shim, Simulated Driver-Target (SDT "W" Spring Support Development Finished inventory and in-process storage.con$ains approximately 750 simu- lated driver-target (thick-walled) fuels. The support design used initially for the flow splitter was a buggy spring design which proved to be inadequate under reactor operating conditions. A Mark I1 type "W" spring should have sufficient strength and springback to accommodate .the flow splitter in the outer tube. Development work has been initiated to detail the design of this support which is to be followed by loop testing for final confirmation of performance.
Stress relaxation studies have been made. The SDT springs were initially compressed to 0.280-inch and releasedto 0.310-inch and fixed in the stress- relaxation fixture. Indications are that the Mark I1 spring converted to a SDT spring by increasing the loop height has as good or better spring proper- ties than the original Mark I1 spring. A column of fuels will be prepared . for flow loop testing with the SDT "W" spring. ADVANCED CONCE€TS AND PUNNING
ME'Cs AND OTHER PROPOSALS ME'C-10 - Centralized Reactor Operating Personnel Training Program This Manufacturing Facility Capability Study work order was closed out in . August. The funds provided for this study were underrun by about $4,100. MFC-12 - New Capabilities for Production Reactors
The AEC-RL has requested submission, by September 15, of a report summarizing. plans on a particular portion of MET-12. Preparation of this report is in progress.
Publication of a report by Battelle-Northwest summarizing the first year of activities for MFC-12 is anticipated in the near future. ME'C-l? - Studv of Amlication for Hanford-Produced Cobalt-60 No further contacts were made with firms engaged in using'cobalt-60.
OPERATIONS FiESEARCH CAGE Programming
The detailed allocation report for CAGE is now opera%ional and is being used for costing nondefense plutonium. Definition of the green sheet budget estimate report is complete with the exception of shutdown, standby, and startup costs, and programming can begin as soon as programmers are available. In most cases, data required for the green sheets can be derived directly from CAGE output. Mark IIA, which is an extension of the &rk I1 Hanford CAGE program, and which will permit flexible shutdown and startup of reactors, is essentially complete. Except for minor problems which are still to be ironed out, the program is now usable.
It is now indicated that the most difficult programming problems for CAGE bd 2 will be to provide the detailed display of fuel inventories at each of the various stations. in the pipeline, and to provide production in terms of year delivered rather than year generated. The rest of Mod 2 additions appear to be relatively straightforward. Mod 2 coding has been started in preparation for the September meeting at Fernald.
CAGE Applications
Pu/Oy Ratio Study
The study of the impact of advanced technology on the plutonium costs to oralloy value ratios was completed and documented (DUN-AOP-46). The study conal2ered the ratio associated Witk'eeversl 6ii'feyent techologies asci Trarying levels of reactors operating. Earlisr studies were generally liaited to oae react06 on current teehalogyo
Peacezul Plutonium Pricing $ .. All CAGE cases fer the peaceful plutociun pricirig st~dyhave been run on the technical pdrtion of CAGE. The cases remain to be rm OE the flmncial part of the. program. These are being held.up tereprarily because of a basic logic redefinition in the negkson cost &Location ecpaticrs, 'kis study is the first use of the eatire R-tchlancl CAGE srstem com2letely through the detailed allocstion format. Offsite fabricstion costs have been received from Oak Ridge for all the cases for this stxd~,and it is currently anticipated that tU.8 study will be cozpleted by the middie 02 Sep-kember. . Invectory Study
Six special cases were provided tc AECOP t~ germit 35em ~LG etxdy the inpact of 2lO &tal loadinqs on the diffxsion plats. One-page simnsries and detailed fuel dzta ws~etran,smitS;ed to AECOPv
Problem Listing A complete list of problems studied tc date wi%h CAGE &irk I1 is shown on pages E-4 and E-5,
ADVANCED PUNNING
Revised Producticn BB5ri.x
The revised prcduction matrix is I~GW nearly ccqlete. xatrix wiU, be much nore detailed than the previa-as one, s-r! %bst it -dllThis cover emicbments from natural up through oreUoy, an3 trill cover the isotoBes PU-238, 924-170, CO-60, Pc-2l.0, U-233, and triiXmO The entries in the matrix are now 'based on more soyhistieated treatinen?; thn pZ"ei%GUSlT and represent a firmer presentatior? of capabilit,y. Refhements Include accom-king fcr fud redesign at %he higher pover densities, better scccun+,ing lor product clecay dwbg the irradiations, and close.. definition cf ~rcduc",puri+,y. &spite these refinements, the production' capabilities match well earlier estLma.P;es except fcr PU-238; for this ma%erial there is a LO to 20 percen5 redue%+on ia stated capacity due to fuel redesign in the higher power densibj cases, and a reporting =ow in terms of atoms h-238 rather than in terms of total plutcnxiamo Plutonim-238 Production at anford
Closely coordinated studies have been in progress cluing the nonth with AECOP to provide for them the required cost and producticn permeters for the large- scale production of plutonium438 at Hmford. As of the and of the month, all t DUN-2873
information that was required had been transmitted. Representative production and cost figures are as follows:-.
1 PU-238 isotopic purity, $ 85 90 85 90 Np-237 processing rate , kg/yr 600 600 1200 1200
Pu-238 produced , kg/yr 112 224 155 .. _. .. 77 C .... . ~..:.
Long-Range Plan
Efforts are well along to incorporate N Reactor into the Company's long- . range plan, to provide further detail in the plan, and to recognize changes in market forecasts and environment since tlie previous plan was prepared. INDEX AND STATUS OF CAGE-MARK I1 STUDY PROBLEMS NFS Burnout i Number Submitted Tails Cost August 25 Reference Objective of Cases to OR Used bthod . Status Document :{ 1 Study the impact of ’ 124 Yes Yes Regular Complete DUN-AOP-19 advanced technology on RL Pu unit cost,
1 Sup. Assist in the analysis 136 No Yes Regular Complete of Problem 1
2 HQ reactor comparison 21 Yes Yes Regular DUN-2419 study
2A II 6 Yes Yes 5$ Disc. II I1 2A-1 2 NO No 2A-2 II 2 No NO I1 2B 3 Yes Yes It DUN-2460 II 2c 10 Yes 5 Yes DUN-24-96 5 No 2c-1 1 Yes No II 2D ,5 Ye s No II Complete Dm-2532 2E 5 NO No . It DUN-2886. . 11 (11 2l? 24 No ’ No DUN-2967 6 No DUN-3005 2G No It (11 3 Long-range plan 26 No Yes Regular Complete 4 HQ Pu/Oy study 14 Yes Yes 5$ Disc. Complete It 4A 5 Yes Yes II Complete ¶
5 Pu assay study 38 Yes Yes It Complete 6 FY-69 budget study 4 No Yes (2) Regular Complete DUN-2728
7 PU/OY aspects of 91 No Yes 5$ Disc. Complete DUN-AOP -20 Problem 1 DUN-AOP-46
. J
'I NFS Burnout Problem Number Submitted Tails cost August 25 Reference Number Ob.le ctive of Cases to OR Used &thod Status Document
8 Peaceful Pu pricing 60 To be determined. Federal In preparation. To be Register issued. and U03 credit schedule
Case B 10 No No 5% Disc. (1) Problem definition in progress, cases to be run.
Neptunium study 10 No N43) Regular Complete DUN-AOP-20 (To be issued) c=3 m Planning estimates See Footnote (4) PEA & PEB DUN-AOP-47 -e3 complete. (To be E-- PEC in issued) T . progress. Gdc3 HQ K level study 2 No NO 596 Disc. Complete DUN-2825 - -L AECOP inventory study 6 Yes No "(1) Complete DUN-AOP-48 RD r-mrj
With varying cascade tail assays. (1) (2) Mdified to eliminate tails rich in u-236 which were originally assumed for use in feed for E-68 charging.
(3) One case examined the effect of a 50-ton batch of Yankee tails scheduled through NFS in FY -68.
A number of cases were prepared manually and submitted informally to HQ and AECOP. (4) Additional cases are in preparation. It is expected that some of these cases will be run on CAGE as Case B type analyses. , ..I.
FUEL TECHNOLOGY The first of a series of U02-&3 cermet fuel pin capsules cooled by recirculating helium was discharged from a special test facility at KW Reactor. kradiation of this capsule was begun in April and the goal of 2000 hours irradiation had . been reached. In this experiment, the fuel temperature was controlled at 2000 to 2300 F by regulating the speed of the recirculating pump built into the outer end of the capsule. The pump and motor were salvaged and life performance tests will be run. A second recirculating helium capsule was charged and is operating at design conditions. These experiments are part of a NASA-Lewis Research Center program for the 'development of a nuclear space propulsion system.
The test assembly for the NAA-121-1 experiment was received from Atomics International. This capsule contains prototypic SNAP-8 fuel elements rather than sample specimens of candidate fuels, as have been irradiated in the previous SNAP-8experiments. Flow tests have been completed and the capsule is scheduled to be charged into a bottom front-to-rear test facility at KE Reactor in early October.
ISOTOPE PRODUCTION
Forty Be N2 target pieces were discharged from KW Reactor process tubes and shipped 2o ORNL for the recovery of C-14. The pieces had been irradiated for 615 operating days and have a calculated yield of 130 Cd of C-14.
ROUTINE IRRADIATIONS
The following routine irradiation services work was performed:
One hundred nineteen activation analysis samples were irradiated in the KE and KW Quickie facilities for Bat-telle-Northwest.
One pressurized uranium swelling capsule was discharged from a front-to- rear test facility at KW Reactor. .. . One creep rate measurement capsule, stainless steel specimen, was charged into a side General Purpose facility at KW Reactor.
One NASA U-m fuel development capsule was discharged and a new capsule charged into a KE Reactor General Purpose facility.
Two and one-half grams of strontium nitrate were discharged from a XE Reactor Snout facility and shipped to Battelle-Northwest for encapsulation. One capsule was sent to NASA at Cape Kennedy on August 23 and the second pair will be shipped Siptember 2.
4
c *. ... .+ y : c -1 Twenty-five,capsules containing KC1 were 'charged intb a figazine facility at Reactor. This material provides OM'with K-40 and ci-36 isotopes.
...... ADMINISTRATION GENERAL ..
.. ;. : EMPLOYEE RECRUI'l'MEIiT
The last two 1967 Technical-' Graduates joined. .Douglas United Nuclear in August , raising the total for the year to 28.
Company representatives attended the National Urban hague Convention. in' Portland , Oregon. A booth was set-up and staffed by DUN and other Hanford contractors ' for the purpose of advising attendees of this progect; and the types of employment opportunities available here.
SUPERVISORY .TRAINING
Pilot sessions of the Supervisory bknagement course were started with 24 participants. One group will mee% in 300 Area; a second group in 100-K Area. This eight-session course, prepared by the American'bknagement Association, covers various essential management functions: planning, organizing, communica- tions, etc. The pilot sessions will help determine if modifications are needed to the standard course before it is offered to all first-line supervisors.
AEC ANNUITlES
GE pension refund checks were received for employees who transferred to DUN and who were not eligible to vest their GE pensions. Packets were prepared and transmitted to these employees, providing them with'an opportunity to endorse their checks for the purchase of AEC annuities or to reject the annuities and cash-the checks. Approximately one-third of the checks were returned to DUN for transmittal to AEC for purchase of annuities. Since many of the employees eligible for AEC annuities are currently on vacation, it is anticipated that several weeks will elapse before all have made their decisions.
APPROVAL LETTERS At the close of the reporting period, the following approval requests have not been acted upon by AEC-RL: ATD Date of Transmittal Niunber to AEC-RL
AP-39 Pension Plan portion of letter January 12, 1966 entitled "Pension Plan, Salaried Savings Plan and Wage Savings Plan"
ATD- 90 DUN Participation in AEC Combined August 23J 1967 Add. #l Operations Planning Committee
. : . G-1 UNCIASSIF7ED ... $. _..* >J * UNCLASSIFED .DUN-2873
ATD .., ... Date of Transmittal . , _. .. Number Sub e c '- 3 t to.AEX-RL ATD-99 OffLSirte ,Colirses and Seminars Add. #1 Program ATD-U.7 Funding of Retirement Li'fe. Insurance: June 16, 1967 Premiums
ATD-122 Contingency Maintenance Job - Terminal.:. July 7, 1967 Add. #1 Site Work at Radiation Zones -.Abs-ndoned. Reactors. F and H..
ATD-127 N Reactor Safety July 14, 1967 ATD-128 ' Mo-12 Appendix "B" bdzfication August 1, 1967 Retiree' Group Term. Lif e Insurance :
ATD-129 Request for Revision of"0perating- August 14, 1967 Safety Limits. - N.Reactor
SAFETY.
bnth end safety statis.tics were: Disabling Injuries: August...... 0 CY to date ... e ...... 2 Days Since hst Disabling Injury ...... 66 Ihnhours Since Last Disabling. Injury ...... 717,800 Approval was received from AEC-RL for the planned Safety. Incentive Program. This enables the awarding of.prizes to individual employees on the basis of demonstrated Company safety achievement.
G-2 UNC.LASSIF'ED \
APPENDIX A
PROJECT STATUS SUMMARY - REACTOR FACILITIES
Percent Complete Number & Title Authorized Design Construction
B, C, & K Reactors Cleaning up exception items. C=!l CAI405 , Modifications for $429,000 100 100 Use of Bauxite - KE and KW Water Plants
* The high speed scanner was placed in CAI-174, High Speed Scan- $861,000 100 100 ning System for Temperature serGice in the reactor's No. 1 Safety Monitoring - KE Reactor Circuit on' August 28. * Exception work being completed. DCP-503, "C" and "D" Work $282 ,500 100 100 Plat form Safety Improve- ments - B Reactor
DCE-505, Boiler Control $240,000 65 100 Orders have been placed. for all engi- Improvements - 165-KE and neered equipment except the boiler Kw controls package. AEC-RL approved the "Record of Purchase" for a negotiated procurement of the control systems from Republic Operations, Beckman Instruments , Inc. , August 25, 1967.
DAP-507, Work Area Fog $144,000 100 90 Made tie-in to KW Reactor, which com- Spray Systems - lO5-B, C, pletes all tie-ins. Late delivery KE and KW of three PRV control 'valves is delaying completion of piping at B, KE and KW Reactors. An ATP is planned at C Reactor during the next scheduled outage. I PROJECT STATUS SUMMARY - REACTOR FACILITIES (contd)
I B, C, & K Reactors
DAP-509, Deactivation of $25,000 40 Started filter water tie-ins Hanford Production Reactor August 22.
DAP-510, Discharge Chute $190,000 10 Detail design was started June 15. Clearing Equipment - Design progress schedule was approved K Reactors by AEC-RL on July 20.
DAE-512, Replacement of $87,000 5 Comments were returned on HWS-7783, Turbine with Diesel Drive - "Procurement Speci fi cation for 18~~PU~P Standby Diesel Engine and Drive."
N Reactor
GCE-401, Upgrading Fire $150,000 100 * Design, furnish and install contract Protection - 100-N awarded on June 28. On August 23 AEC-RL issued a revision to the project proposal requesting extension of time from October 15 to May 31, 1968. Extension' is to permit %? additional time for detailed design by vendor, and to allow for inclusion of 1705-N. Building fire protection (yj coverage. p%"p GCP-404, Fuel Spacer 56 Purchase orders for the refuse dis- Disposal System and posal cask and container were placed Refuse Cask - 100-N on August 8 with delivery scheduled by November 1. ,
I i . PROJECT STATUS SUMMARY - REACTOR FACILITIES (contd)
Percent Complete Number & Title Authorized Design Construction Status
N Reactor
GCE-405, N Reactor $189,000 65 57 Sufficient work was completed to Temperature Monitoring permit upgrading of the Zone Tem- System Improvements perature Monitoring System prior to the increased power level demonstration in June. Balance of work is delayed pending receipt of permanent cable between Zone I and Zone 111; cable'bid opening is scheduled for September 7.
GCP-406, Safety Plat- $30,000 16 1 A project proposal, requesting forms and Accesses $300,000 (including $17,000 to be funded by WPPSS) was submitted to AEC-RL on 2-10-67. An interim authorization of $30,000 was issued by AEC-RL on August 1.
GCE-407, Fast Index $50,000 100 0 Detailed design work is complete and Rupture Monitoring System i'nstallation is schedulei2 to start
in October. SI
GCE-408, W, C, D Elevator $90,000 0 Design work started August 28. Safety
GCP-409, Fuel Handling $310,000 14 Detailed design continuing ,on Improvements schedule . Number & Title Aut ho ri zed
N Reactor
GAP-410, Decontamination $65,000 0 0 The design criteria and project Waste Loading Facilities proposal were revised to provide for rail car facilities at 100-N instead of truck loading facilities at 100-N and 200-E as originally scoped. Re-. vised criteria submitted to AEC-RL on .August 15 and the revised project proposal on August 25.
GCP-411, Effluent Control $i,670,000 7 0 Revised design criteria submitted to Program - 100-N WC-RL on August 15. GCP-412, Steam Generator ------Proj ect propos a1 submitted to AEC-RL Moni't ors on June 30 requesting $112,000.
I APPENDIX B
. . .. .<. - 2- Number clas S i ficat i oii Author -- Date
DUN-2 368 Secret ' SS Jones 7/25/67 Adequacy Review of AN Kugler the B-C Reactor Last- . .. Ditch. Cooling, Systems . ... I. ._ arter Completion of ' . . 5 .. , the PTA-014 Test
DUN-2456 Secret F'E Owen .7/31/67 Reactor .Effluent. I I .. -. An&-tic-al Data, .. *- . .. .. -- .. April, May, June, 1967 ,- ..'' CR Zook' '- DUN-2 467 Secret '- 7 I14 /6 7 PTA io. 079, . Severance and Wrap- .. . ping with Stainless Steel Shrouds Spe- cific KE and KW Reactor Ball 3X Entry Bellows
DUN-29 39 Unclassified PE Midkiff 8/7/67 Electrodeposition of Titanium on Uranium
DUN-2942 Secret RG Geier 8/7/67 Quarterly Report, Contamination Control Columbia River , April- June, 1967
DUN-AOP-44 Secret DH Curtiss 7/28/67 Pu-238 Production WM Harty Program at Hanford ' GF Owsley EE Voiland WK woods
DUN-AOP-46 Secret DH Bangerter 8/14/67 Hanford Pu/Oy Ratios
DUN-AOP-49 Secret WK woods 8/3/67 AECOP Task Assignment 85 (Second Letter)
DUN-AOP- 50 Secret WK Woods 8/17/67 AECOP' Task Assignment 85 (Third Letter)
DUN-AOP-51 Secret WK woods 8/25/67 AECOP Task Assignment 85 (Fourth Letter)
k , -. . . .-- r-. -. ...
p. SIGNIFICANT REPORTS ISSUED ( contd)
Number Class i fication .Author Date Title i HW-84238 A Unclas s fied KL Fowler 7/31/67 Steam. ..Generator Moisture' Carry-Over Detection Evaluation Report
RL-GEN-1065 A Unclassified DL ' purs ley 8/21/67 Final Report on Ball SUP 20 Safety System Functional Tests
RL-GEN-1617 D Confidential DL Renberger 8/2/67 N.React.or 4800 Mw Demonst rat ion Test. Summary Report
RL-GEN-1784 Unclassified DL Renberger 7/24/67 N-4 Test 5.7, Part I, Detailed Results from 800 M7e Load Rejection Test of December 9, 1966
. , .. .., .t ...... t;. a UNCLASSIFIED DUN-2873 .
APPENDIX C
EMJ?LOYMENT SUMMARY (as of 8/31/67)
Change Non- from Operations Division Exempt Exempt Total 7 /31/67 General 1 1 2 .o Manufacturing 222 Production Fuels 719 941 -24l -1 Subt ot a1 77 266 343 - 300 -m iX6- -25 Technic a1 Division
General 1 1 2 0 Research & Engineering 77 41 118 +1 Facilities Engineering 0 Advanced Concepts & Planning 54 18 72 10 2 12 +l Subtotal 142 --E *7iZ +2 N Project
General 1 1 2 0 N Reactor 61 N Fuels 213 274 +4 39 109 '-3 N Research & Engineering 148 48 20 0 N Praject Engineering 68 +1 Subtotal 56 73 - 20 5 7% 565 4-2 Finance & Administration
I General '1 1 2 0 Finance & Administration 150 +1 Employee & Community Relations 74 76 19 10 29 0 "Technical Graduat es'I2 30 0 0 Subtotal 30 - 124 87 211 +1 Company General
Management 2 +2 Summer Employees 6 8 -12 Subtotal 9 27 36 - 15 29 44- -10
Total - - 786 1 524 2 310 - 30 lRef1ects force reduction due to D Reactor deactivation.
2Employees in this interim classification are recorded in E&CR Section during their rotational training.
2' UNCLASSIFIED H-7 2. '' I) 1. FEATURE REPORT ,- .. .
POISON SPLINE SYSTEM FOR'SUPPLEMENTARY CONTROL ,. IN THE B, C AND K REACTORS ..
.I INTRODUCTION .._. - ." Should a designer of the original control systems for the Hanford pro- duction reactors return to observe current operation, he would be amazed to find the reactors functi.oning smoothly and 'efficiently at several times design power level , yet with 'loadings having reactivity parameters much more severe than contemplated origin'ally. . .
The capability for routinely and *flexibly meeting the greatly increased demands on both operating and safety-.control in the B, C and K reactors is provided by the poison spline, supplementary control system. &e purpose of this report is to describe this system, the background of its development, and some of the possibilities for extending its use which will hrther enhance the irradiation flexibility of these DUN-operated Hanford reactors.
POISON SPLINE SYSTEM DESCRIPTION .. .. The poison spline itself is- a' boron-cont&ining alhinum .tape (about 1/2" wide, 1/20" thick, and 30 feet long) which is inserted into the bottom cooling annulus of a ribbed process tube. Each spline has a poisoning effect somewhat greater than the reactivity equivalent of two 94 Metal enrichment co1umn.s. The system permits the insertion and/or withdrawal of these splines in any of several hundred process tubes, during op.eration as well as 'during outages. Thus a large'poten-tial is available for aug- menting the poison effect of the operating and safety control systems, not I only within their somewhat limited control configuration, but also outside the region in which the rod and ball columns are most effective.
A cross section of the process tube, the fuel element, and the water annulus, with a poison spline in the space between the tube ribs, is shown in Figure 1, appended. Also shown in Figure 1 is a longitudinal section view of the vinyl rubber sphincter seal through which the spline passes. This cap seal must permit passage of the small, limber spline, yet must preclude leakage of high pressure water from 'the process tube inlet nozzle. Such leakage can cause Panell-it trips.
The spline pattern at one of the K reactors , for which ribbed aluminum tubes were retained when most of the central tubes were replaced with smooth-bore Zircaloy, is shown in the appended Figure 2. Each of the 215 tubes in the pattern shown is fitted with the special spline cap and seal. The spline coiler mechanism, with which the spline is withdrawn onto a shielded windup reel for disposal, is shown in. Figure 3. v- Splines are inserted and withdrawn from the front face of the reactor by t
the regular operating crews. Since spline insertion and withdrawal ' introduces only a small reactivity increment (0.1 to 0.2 milli-k) over several minutes time, fine control is easily maintained by the control room operator through compensating moves of the normal horizontal control rod (HCR) system. The sequence and timing of spline insertion and with- drawal during reactor operation is- direct'ed by the Reactor Specialist , the man in charge of control room activities. The sequence and timing of spline insertion during a reactor outage is established to meet or exceed the process physicist's written instructions for fulfilling total control requirements.
Spline usage averages approximately 150 splines per startup at a K reactor and about 90 at a sm,aller reactor (B or C) . As many as 200 splines at a K reactor and 100 at B or C may be under irradiation'at a given time during an operating interval, as compensation for either a long-term fuel reactivity gain or an anticipated long-term fuel reactivity loss during an extended operating cycle. The splines cost about $12 each, and total spline costs for the. By Cy and two K reactors average about $130,000 per year. The . several hundred less expensive aluminum splines used each year for front-to- rear flux traverses are included in the above dollar total.
A reclaiming facility has been provided on-plant , whereby splines irradiated briefly during startups may be salvaged for reuse. Such splines are de- contaminated chemically, straightened, and then rerolled to assure the desired degree of surface smoothness and curvature. SplGes irradiated too' long to permit this recovery are disposed of by burial.
SPLINE SYSTEM DEVELOPMENT
It was apparent in the mid-1950s that the higher operating power levels expected to follow completion of a water plant expansion program for the small reactors would tax the capacity of existing operating and safety control systems. Development work therefore was conducted on two supplementary systems : (1) a poison column charging facility (PCCF) , and (2) a poison spline system. The PCCF was in operat,ional use at an earlier date, but the greater flexi- bility and neutron economy inherent in the spline concept, combined with its virtual freedom from excursion potential, provided strong incentives for continuing and completing the spline system development. This was done , and dramatic gains in operating efficiency were realized between 1959 and 1962 at all eight reactors through the use of.poison splines. The spline system is estimated to have increased production at least five percent during those years, at an increase in irradiation cost of only one dollar per incremental MWD produced.
Criteria called for the spline to be small in size, yet physically strong and an effective absorber of neutrons. It had to be compatible with the process channel environment , .and excessive pers.onne1 exposure rates during
? D
.. its removal had to be avoided. Thus the spline system as finally developed after a number of setbacks , has $he following- features: I Boron poison
Aluminum carrier .. Solid construction (sintered boron carbide in aluminum)
Remote removal by disposable windup reel (spline coiler)
Vinyl rubber sphincter seal in front noz,zle cap I
Use in only those tubes with adequate. rib height
Use of special upstream spacer elements to assure alignment during insertion.
Principal features discarded as a result of operational setbacks were the use of hollow splines packed with boron powder, and of a chopper-and-cask disposal method. Swelling w& encountered with the hollow splines, and the chopper system was plagued with jamming and associated high personnel radiation . exposure rates during breakdown and maintenance.
Spline utilization is now an operating routine at the B, C and K reactors , and the maintenance of insertion and ,withdrawal equipment is handled by plant forces.
PRESENT AND FUTURE BENEFITS FROM SPLINE USAGE General Cons i derat i ons
The original justification for the poison' spline system was primarily that I of improving startup efficiency (reducing .time to attain full power) by supplementing €he capacity of the operating control system. The horizontal control rods (HCRS) have sufficient capacity to handle the startup reactivity transients up to the point of reactivity turnaround only at very low power levels. Bootstrapping methods formerly used for attaining .full power level , such as a series of power raises and associated turnarounds, or the loading of temporary poison columns to be discharged in one or more short outages, accomplished return.to full power level only after loss of an effective day's full production in addition to the normal outage time. Early gains in startup efficiency: due to.spline usage were dramatic, as has been noted, and average production 1oss.per startup has been in the range of only one-fourth to one-third effective day during the last five years.
Other benefits from spline use have accrued which are just as important to current modes of operation as added startup control capacity, but whose quantitative measurement of value is less direct. These benefits, discussed in subsequent paragraphs, include the following:
8 Total Control - added capacity, geometric flexibility, and -reactor access.
e Flux Distribution Control - geometric and fine control flexibility for flux trimming and for oscillation damping.
o Loading Accommodations - compensation of long-term reactivity. transients and flux tailoring requirements of special loadings.
0 Axial Flux Traverse. Capabilikies.
e Control for .Unususl Situations - flexible geometry for recovering from emergency situations such as water leaks, or for. facilitating unusual. low-level experiments.
Spline Use for Total Control
The control capacity of the normal safety and operating control systems must be supplemented during certain outage and startup conditions in order to assure that subcriticality between the time of a postulated coolant, loss and the ensuing fuel meltdown would always be assured. Total control requirements increase with outage .length, and it is possible to add splines during the outage as require'd for the specific time the reactor has been down. Rear face access problems during the outage, such as from tae replacement or hot fuel hangup, would not prevent the charging of splines as required. The fact that the spline cap pattern can be made to include almost any process channel desired provides excellent geometric flexibility. If a control element is out of service, equivalent poison compensation may be made within the same region of the reactor. Where normal 'control area coverage is limited, a situation aggravated by radial enrichment, splines may be used to provide adequate control coverage.
It is somewhat fortuitous that the splines inserted for total control are generally needed also for operating control during startup. In the case of loadings whose reactivity gain subsequent to water loss would be . especially heavy in the reactor fringes, total control requirements may exceed startup control needs to some extent ; however, the fringe place- ment of splines may be planned in this instance such that total control requirements are met with minimum negative effect on .operating reactivity.
Spline Use for Flux Distribution Control
As the reactor power levels. and exposures have been increased, parameters related to flux oscillation tendencies have also increased. The term "xenon oscillations'' has been well publicized in the literature. A less publicized factor also of significance in the Hanford production reactor case is the positive moderator temperature coefficient of reactivity and its increasing effect with both specific power level and fuel exposure (plutonium buildup). "Trend control" methods developed at Hanford during I the early 1950s , for overcoming inherent oscillation tendencies through spatial rod control techniques, have been extended to incorporate the use of splines.
Operation under current combinations of specific power , flattening, and fuel exposure would simply not be possible without the routine use of the spline system or something equivalent. Splines supplement the normal HCR systems (nine rods at B Reactor, 15 rods at Cy and 20 rods each at KE and KW) such that these reactors are flexible competitors of the other Commis sion-owned production reactors whose basic control systems have several times as many control rods. The flexibility provided by the spline system was demonstrated in convincing fashion recently at the B Reactor when the manual control portion of the HCR hydraulic system failed for approximately two hours during power level ascension following turnaround, and excellent control of flux distribution and level were maintained largely by use of splines and variations in gas composition.
Supplementary control will be necessary also in any of the concepts considered for target space enhancement. Development plans include the procurement and testing of ribbed tubes of overbore size, of both aluminum and Zircaloy, to permit the use of poison splines in the modified reactor case.
Additional trim capabilities have been obtained with "grey" splines and "half" splines. As the names imply, the former contains a lower than normal poison density, whereas the latter consists of a poison-containing end attached to an aluminum-only spline. These two .concepts have been used experimentally for the most part, and the splines themselves have . been stored by the assigned physicist in order to preclude their in- advertent use for total control. Axial control, for which the "half" splines are designed, has been accomplished by the reactor operators usually with sequential partial insertions and incremental withdrawals of normal full length splines.
Spline Use for Loading Accommodations With plutonium-only production and low exposure discharge , long-term reactivity transient changes during a normal operating run may be accommodated easily within the capability of the HCR system. Long- term reactivity changes in this case refer to changes in fuel composition with exposure due to depletion and buildup of fissionable materials and fission products and to the effects of those composition changes on operating reactivity coefficients.
As other products are required, the driver and target combinations desired for optimum efficiency generally result in greater changes in reactivity with exposure than are experienced with natural uranium: In such cases poison splines can fulfill the role of a "burnable" poison because of their ease of removal. during operation. Conversely, splines may be used .. I- to compensate for a marked gain in reactivity during an operating run, as a is experienced with the depleted uranium targets used in the E-D program for
high-240 plutonium production. It is not unusual during extended .operating I runs , therefore , that in-reactor spline inventories may approach one' to one and one-half percent k in poison worth, with numbers of splines.on the order of 50 to 150, depending on reactor type and loading status. The preceding paragraph indicates the incentives for splines whose poison material would be converted to a useful isotope. Lithium appears to have most of the characteristics desired; i.e. , high cross section, compatibility with aluminum, marketable product, and no gamma radiation resulting from neutron capture. Lithium-bearing poison splines for trial use during ex- tended operating runs are being obtained under an offsite development contract.
The flexibility afforded by poison splines is particularly helpful in the planning and conduct of special small-scale block tests. Two examples under current irradiation are the "block-within-a-block" tests at KE and KW, the one to obtain preliminary 2.1 E-N conversion ratio GA~and the other to , demonstrate the neptunium-237 "quick-turnaround" concept for producing Pu-238. The first small block required the creation of a small reduced- flux zone to accommodate the 210 Metal fuel elements without excessive heat generation rates, whereas the Pu-238 test required the creation of a flux "plateau" above the normal level. Although both of these concepts consist basically of a combination. of "permanent" driver columns and target columns, the availability of splines permitted both tests to be accomplished with minimum need for stepwise loadings and the associated added costs for materials and time.
Axial Flux Traverse Capabilities
A front-to-rear traverse may be obtained after only a few minutes' irradiation of an aluminum spline in any available spline channel. The recording obtained automatically as the "transparent" aluininum spline is withdrawn onto the coiler reel gives an approximate plot immediately useful to an experienced observer. A mathematical fit -of the traverse data, corrected for withdrawal time and extrapolated to full column length, is obtained w&th a computer program quickly run from data keypunched from the original traverse.
Axial flux distribution knowledge useful for engineering and fuel as well as reactor physics studies has been advanced markedly by the spline system. Such information was available previously only from element-by-element radiation readings or from graphite temperature indications. The traverse data from discharged fuel elements are expensive in terms of special dis- charge and pickup time, and observations represent an integrated rather than an instantaneous flux distribution. Because of variations in the lattice and in the. instruments themselves , the relationship between graphite tem- perature and local flux could be considered as only an approximation.
...... _ & DUN-2873
Spline Control for Unusual Situations
Splines have facilitated efficient recovery from two situations which other- wise would have had severe 'adverse production effects. Recovery from both of these situations - one an internal water leak at F Reactor, and the other residual balls in DR Reactor following a 3X system trip - required operation (for water evaporation and boron burnout, respectively) and thus the enrichment of a small region of the reactor sufficient to override the normal operating control in that area. Splines provided the necessary reactivity compensation capability in both cases.
A final example of both the gross and the fine aspects of the control afforded by splines is illustrated by their use during the initial N Reactor startup. Their preinsertion into a large number of cross coolant channels (graphite cooling system), and subsequent withdrawal by . scattered array, simulated a uniform reactivity transient adequately to permit calibration of the N Reactor control and safety rod system in a variety of withdrawal sequences and configurations. An extension of this procedure to the operating N Reactor has been visualized should future N loadings have sufficient depletion transients to require poison to be removed (remotely) during operation. >.. UNCLASSIFIED . _-.-A.- DUN-2873
k
k 0
0c u
k 0 Q) n k 5 h0 R
I-8 ,W?CLAS SI FIEB 3 6- I c . L
H I I-'0
Figure 3. Spline Coiler on Reactor Front Face,