ML20073P956
| ML20073P956 | |
| Person / Time | |
|---|---|
| Site: | Ohio State University |
| Issue date: | 05/13/1991 |
| From: | Myser R OHIO STATE UNIV., COLUMBUS, OH |
| To: | Michaels T NRC |
| References | |
| NUDOCS 9105240227 | |
| Download: ML20073P956 (12) | |
Text
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li Nuclear Rentor Laboratory 1244 Kinne.irN el Columbus. Oli 43212 1154
'i Phone 614 292-6"'$$
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4 LiNIVERSin May 13, 1991 Ted Michaels USNRC PDNP M.S. 11-B-20 Washington, D.C. 20$55
Dear Ted:
As we discussed 5/10/91, I have enclosed a copy of the Power increase Plan for the OSURR to go from 10 KW to 500 KW. _
We anticipate an outage of about one month to install the In-pool heat removal system, reload the core, and complete the necessary testing to return to 10 KW. This outage will begin about June 3, 1991.
All radiological monitorjng as indicated in the plan will be completed starting at 10 KW and including 50 KW, 100 iW. P90 FW nod 500 KW.
We do not anticipate operation above 100 KW until the lutter part of the summer. It is most likely the results of our power increase will be included in this year's annual report .
Please call me if you have any questions.
Sincerely.
-vi Richard D. Myser Associate Director RDM:al Enclosure l
910saOm ,10sm l ti\
gDR ADOCK 05000150 couege d Engineering PDR
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Power increase Plan i
pote and Personnni ,
The expected date when the outage to begin the power increase in May 31, 1991. The following Individunta shall be involved in the plan.
Richard D. Myser, Senior Reactor Operator, Associate Director of the Nuclear Reactor Laboratory. Supervised the unloading, ,
reloading, and reenlibration of the 11E0 core in 1981, 1984, and 1988, and the LEU core in 1988.
Joseptp W. Tainag!: Senior Reactor Operator, Senior Research Associate, Engineering, Supervised the recalibration of the LEU and llEU cores in 1988 and participated in the unloading, reloading, and recalibration in 1984. '
Joel M. Itatch, Senior Reactor Operator, Research Assistant, Engineering. Supervised control rod testing in 10f14 and 1988 reloading operations and participated in the unloading, reloading,- ,
and recalibrations in 1984 and 1988. l Michael J. Davis, Senior Reactor Operator, Student Research i Assistant. Participated in the unloading, reloading, and recalibration of the !!EU core la 108tl and the current I.EU core '
nIno in 1988.
Andrew Clark, P.E., is responsible for the design and installation of the heat removal system. The ex-pool part of the system is complete.
Instrumentation Figure 3.12 of the SAR (attached) shows the location of the core grid plate.
- start-up source, and current neutron detection instrumentation. For the core loading the following instrumentation shall be utilized.
- 3. Start-up Channel as described in Section 3.3.14 of the SAR l (Fission Chamber).
l l 4. Auxiliary Linear Power Level Monitoring Channel using a Compensated !;a Chamber (CIC) identical to those used in 1 and 2 above but placed in Beam Port #1.
- 5. Auxiliary (Fission Chamber) Pulse Channel to be placed in the Reactor Pool Thermal Column location 0-7.
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Lec -icnc of ins tru=en: and f:::i'.i:ies in and anund the OSt.TFJ. core.
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4 pre-installotton Acttv1tles Numerous items need to be completed before t's hoit removal system can be installed in the pool. First the pool water will be analyzed for radioactivity to assure releases to the annitary sewer system meet the requirements of 10CFR20. Then the core will be unloaded and the fuel placed in the storage pit. The control rods wil.1 be suspended from hangers along the N.E. corner o'f the pool. Lights and dry tubes will be moved to the East end. The CIF will be removed and stored in a shielded locatjun (BSF pool or BPf1 Areal. Then the reactor pool will be drained by siphon. After draining, the pool will be carefully surveyed. Entry will be by extension ladder. Previous exposures from the 1988 LEU fuel conversion outage were below the minimum detectable limit for film badges. The floor will then be mopped dry and the storage pit tealed. The walls will be cleaned, rinsed and dried if necessary to hu p remove (control) bacterial growth. In additjon, the NRL Staff committed to adding four additional film badges as environmentnl monitors about 100 feet away from t!- building in each direction. This was completed January 17, 1991.
!! eat Removal System Installat ton and Of her Dry pool Act ivit les The installation of the heat removal system will be directed by Mr. Andy Clark, P.E. 1le designed the system that has been reviewed and approve ' oy the ROC and NRC. The major components (decay tank, core shroud, and return line diffuser) have been fabricated by Allied Manufacturing. Installation should commence on or about June 3, 1991 and should require about two weeks.
During this same two week period the detector instrementation rack and/or housings holding the detectors may be modified to allow for appropriato adjustment for the new power level. Prior to this we plan to test (move) at least UIC al and CIC #1 to help determine if modifications are necessary.
Below are described the two proposed tests.
- 1. Level renctor power at 10 KW. Mark position of detector.
Move UIC *l-away from the core until it reads a factor of 60 less or 166.60 watts. This position should correspond to a trip at 600 KW which is the new T.S. limit. Stop movement before the detector la removed from the lower support bracket. Return detector to original position.
- 2. Level reactor power at 10 KW. Mark position of detector.
Move CIC #1 away from the core until it reads a factor of
-100 less than 10 KW which is 100 watts. This position should correspond to full range on the Log N Channel. The collars on the detector housings will limit movement of CIC
- 1. UIC *1.may need to be moved away from the core to allow CIC *1 to be_ fully extended. We may also want to check period indications with this configuration since this comes from the CIC al signal.
Once the heat removal system is in place the control rods need to be rejnstalled to check their alignment and the fit of the core rh.?oud cover plates. We anticipate using the mock control rod fuel element and the spare 3
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unirradiated control rod fuel element for the alignment checks. This will enable us to check two control rods at a time. The core shroud may also be heated to assure it does rat twist or otherwise move ouc of tolerance, t Control Rods Tests After loading the cont'of rods in the critical experiment, but before continuing with loadind, the following control rod tests shall be completed.
- 1. Magnet Release Times
- 2. Rod Insertion Times
- 3. Minimum magnet currents for holding, withdrawal, and pickup.
- 4. Hod drive times from lower limit to upper limit and upper to lower.
Critien) Experiment Utilizing where appropriate the five instrumentation channels a plot of 1/m versus the mass of_the fuel in the core can be extrapolated to 1/m = 0 to give the critical mass of the core. The following is the anticipated loading sequence in an attempt to maintain a symmetrical core. Actual loading shall be based on results f rom 1/m plots and may be more ,
conservative. Each fuel loading shall be less than or equal to 50% of-the difference between the amount of fuel previously loaded into the core and the critical mass predicted by the 1/m curves with the possible exception of the last few loadings of single elements. Please refer to the attached Figure 1 and Table 1 from our LEU Fuel Loading . It indicates the sequence used to load LEU *1 core.
.After the core 18 loaded, an approach to critical using 1/m vs rod height shall be completed. Following control rod calibrations. K can be recalculated using a new rod worth information.
The above core was measured to have approximately 1.5% K g
. Reactivity shimming can be accomplished by using fuel elements or inserting graphite
' elements if necessary. Data is rvallable to estimate the worth of various-grid locations. We anticipate loading additional fuel or graphite to obtain between.2.0-and 2.St-Kx ,.
Control Rod Calibration The fo))owing methods may be employed to complete control rod calibrations at low power estimated at a nominal 10w for rod drops; less than low for sub l- critical multiplication; and less than 1 kw for positive period measurements.
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Figt.re 1. LEU CORE I .
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TC A E C D E s
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Sa 5 3 lb. Sa OH-006 OHC-001 OH-007 Oiit.-002 i ep 011-008 Su
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,. , 5a N Puse Proposed LZ') Core for the OSURR With An Esticated E:: cess Reactivity of 1.5%3/K (X's Denote Filled Grid Plate Positions)
Loading Sequenca Grses Loading Sequence Grams 1 '
499.53 7 2697.64 2 1099.06 8 2897.49 3 1498,74- 9 3097.32 4 1898.43 10 3297,15 5 2297.97 11 3496.91 i 5a 2297.97 + Fillers 12 3696.73 6 2497.81 I
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TABLE 1 OSURR Core Loading Landing Positions Total U-235 Masa Number Loaded (grams) 0 source, CIF 0,00 1 2B, 2D, 4D, 4D 499.53 2 20, 3D, 3D 1099.00 3 10, 4C 1498.74
.4 3A -3E- -1898.43 5 1B, SC 2297.97 5a filler plugs 2207.97 6 50 2497.81 7 2E 2007.04-8 4A 2897.49 ,
9 2A 3007.32 10 4E 3207.15 11 1D 3490.91 12 5B 3600.73' ,
i 6
. l Rod Drop - Reactivity determination for full length of Shim Safety rods 1, 2, and 3 and for the minimum critical positions for the same rods.
Subcritical Multiplication - Reactivity determination for the lower portions of Shim Safety rods 1. 2, and 3.
Positive period Calibration of the entire length of the regulating rod and the upper portions of Shim Safety Rods 1, 2, and 3.
power Calibration and Temperature coefficient of Reactivity A determination of reactor power is to be completed by measuring the increase in reactor pool water temperature due to the operation of the ;
reactor. The-reactor will be operated at 10 kw as indicated on the Linear _
Level Monitoring Channel and reactor pool heat-up rate shall be monitored.
' Initial pool temperature shall be ambient and reactor operation shall allow I for a pool temperature rise of at least 5'F. From this heat-up rate the power is determined. A check of the power at 50 kw will also be made by measuring the heat up rate.
e Moderator Tempernture Coefficient of Reactivity To determine ti.e moderator temperature coefficient of reactivity the following measurements will be made:
- 1. Long (slow) period (100-300 sec.) starting at low power and continuing to = 10KW, making the following measurements:
- a. Reactivity measurement at low power from doubling time measurement.
- b. Heactivity measurenent at high power from doubling time measurement,
- c. Moderator temperature change above the core.
- 2. Short (fast) perjod (10-12 sec.) starting at low power and continuing to = 10 KW, making the same measurements.
Below is a description of how this method was used for the current LEU core.
The reactivity change for the transient of the long, slow period was 0.046%. It was 0.074% at the start and 0.028% at the finish.
The moderator temperature change was 4.2'C. It is assumed that the observed reactivity change of 0.046% was due to both moderator and fuel heating.
7
I The reactivity change for the aansient of the short, fast period was 0.02*.. It was 0.30% at the : tart and 0.28% at the finish.
There was no moderator temperature change measured during this short period. The reactivity change was therefore only due to fuel heating.
- 0.40% Moderator 4 Fuel Feedback
- - 0.2n4 Fuel Feedback only
- 0.20% Moderator Feedback only
- 0.20%/4.2*C = 0.10x10 ak/k/'C This meets Technical Specification 3.1.1(5) which required a negative moderator tgmperature coefficient of reactivity with an absolute value of at least 2x10 A k/k/'C.
Vold Coefficient of Renetivity l
The method used for measuring the moderator void coefficient of reactivity ,
at the OSURR requires plastic " void boxea" fabricated to fit into the i coolant flow channels in a reactor fuel element. One void box, with the internal volume air filled, is required. An exact duplicate box, with water-filled internala, is also required.
The critical reg rod position with the air-filled box inserted into a particular fuel element and the position with the water-filled box inserted are compared. The difference in critical rod height is due_to'the reactivity worth of the air void inside the box. This reactivity worth can be determined from the rod worth data compiled in the rod calibration experiment. This, in turn, gives the moderator void coefficient (dk/k/1".
void) when the air volume is rattoed to the total core floodable vol'ame.
Hndf at lon Monitoring for t he Power increase
- 1. Ar In previous correspondence with the NRC (attached) we have comaltted to extensive Ar-41 monitoring. This will begin during the power calibration at 10 KW after reloading the core. At the suggestion of ORS we will also monitor at 50 KW in addition to the a! ready planned 10, 100, 250, and l
500 KW levels.
- 2. Direct radiation monitoring will occur at each power levej monitored for Ar-41 production. At least the following locations shall be monitored:
- 1. pool top including over CIF and Dry Tubes
- 2. Beam Ports
- 3. Thermal Column Area
- 4. Rabbit Area i
- 5. Domineralizer Area i
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References Initial Calibration of Standard Core No. 1 of The Ohio State University Reactor January 1966, 1
i Recalibration of Standard Core No. 1 August 1975.
1981 Calibration Report for the OSURR.
Critical Experiment. Approved by the Advisory Committee on Reactor Operations April 13 1965.
Neutronic Scoping Calculations for OSURR Core Design with Standardized U 3S12 Fuel Plates. 1986.
Standard Core Catalog. OSU Reactor Core Fuel Reactivity Worth. !
1964. I 1
Fuel Load and Start-up planning Document for the LEU Core. 1988. l 1
Initial Loading and Testing of Low Enrichment Uranium Fuel in Tho Ohio State University Research Reactor. October 1989 I
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4 Testing to be completed per letter to NRC 2/28/90 jlen t i n t!, Vent ilot lon and Air Condi t lonint! System (HVAC)
As a part of the overall startup testing program for operation at 500 KW, a plan to determine Ar-41 concentrations in the restricted area during various operating conditions shall be implemented. The measurements will be made using a shielded volume air sampling detection system calibrated for Ar-41.
The system currently in use is calibrated annually and typically indicates ;
Mpc = 10 cps. Figure 0.1 on page 143 of the SAR indicates that the l equilibrium concentration for Ar-41 inside the building is reached in about i four hours. Measurements of air exiting the building at the exhaust fan will be made to verify when equilibrium has been reached. After this, other locations in the building shall be monitored. Furnace fans shall be running during all tests. The testing program to determine Ar-41 concentrations is described below.
Testo shall be completed at least for four power levels 10 KW, 100 KW, 250 )
KW and 500 KW. Three different operating conditions shall be investigated.
These are: Ar-41 production from the rouctor pool alone, Ar-41 production i from the reactor pool with the rabbit running continuously, and Ar-41 production from the reactor pool including a puff release from the rabbit.
A total of twelve locations may be sampled. These are:
1 Room 100 Reactor Bay Vent Fan
- 2. Room 100, Reactor Day Reactor pool Top Catwalk
- 3. Raum 100, Reuctor Bay BSP Pool Top Catwalk
- 4. Room 100, Reactor Bay Thermal Column Area
- 5. Room 100, Reactor Bay Deam port Area
- 6. Room 103, Office
- 7. Room 103A, Office
- 8. Room 100. Counting Room
- 9. Room 104, Office
- 10. Room 201, Conference Room
- 11. Room 205, Reactor Control Room
- 12. Room 209, Office.
It is anticipated that for the puff release, only locations 1, 7 and 11 will be sampled since the concentration of Ar-41 should peak and then decrease back ta equilibrium not allowing time to sample all locations.
After measurements are completed for each operating condition, the reactor shall be shut down. Additional measurements at all locations shall be completed after shutdown to determine the purge rate. This will facilitate a determination of how long to monitor after shutdown during normal operations.
10
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Testing to be coinpleted per let ter to NRC 2/28/90 IJenting, Ventilatton and Air Conditinning System (HVAC)
As a part of the overall startup testing program for operat ion at 500 KW, a plan to determine Ar-41 concentrations in the restricted area during various operating conditions shall be implemented. The measurements will be made using a shielded volume air sampling detection system calibrated for Ar-41.
The system currently in use is calibrated annually and typically indicates MpC e 10 cps. Figure 6.1 on page 143 of the SAR indicates that the equilibrium concentration for Ar-41 Inside the building la reached in about four hours. Measurementn of air exiting the building at the exhaust fan will be made to verify when equilibrium has been reuched. After thlH, other locatlons in the building shall be monitored. Furnace fans shall be running during all tests. The testing program to determine Ar-41 concentrations is described below.
Testa shall be completed at least for four power levels 10 KW, 100 KW, 250 KW and 500 KW, Three different operating conditionn shall be investigated.
These are: Ar-41 production from the reactor pool alone, Ar-41 production from the reactor pool with the rabbit running <ontinuously, and Ar-41 production f rom the reactor pool including a put f release from the rabbit.
A total of twelve locations may be sampled. These are:
1 Room 100 Reactor 11ay Vent Fan
- 2. Room 100, Reactor flay Reactor pool Top Catwalk
- 3. Room 100, Reactor liay llSP pool Top Catwalk
- 4. Room 100, Reactor Day Thermal Column Area 5, Room 100, Reactor Bay Beam port Area
- 6. Room 103, Office
- 7. Room 103A, Office
- 8. Room 109, Countlug Room
- 9. Room 104, Office
- 10. Room 201, Conference Room 1 11. Room 205, Reactor Control Room
- 12. Room 209, Office it is anticipated that for the puff release, only locations 1, 7, and 11 will be snmpled since the concentration of Ar-41 should peak and then decrease bark to equilibrium not allowing time to sample r.11 locations.
After measurements are ccmpleted for each operating condition, the reactor shall be shut down. Additional measurements at all locations shall be completed after shutdown to determine the purge rate. This will facilitate a determination of how long to monitor after shutdown during normal operations.
10
From these measurements we shall determine: 1. If any areas need to be posted as airborne radioactivity areas per 20.203(d); 2. what areas reach the highest concentrations of Ar-41, and whether these are typically occupied; and 3. how long we can operate at various powers and conditions while maintaining doses AhARA per 20.103(b)(2).
Table 2 shows the proposed plan for determination of Ar-41 concentrations during start up testing. After this testing la completed, regular monitoring during operations will be done using the effluent monitor.
Table 2. Proposed Plan for Determination of Ar 41 Concentrations Power Operatlig sample Sample
- f. eve l Conditinn flour Locatinn All (10,100,250,500 KW) Pool Top 1,2,3,4 Vent All Pool Top 4-5 All Shutdown @ 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> Pool Top 6,8 All All Pool Top 4 Rabbit 1,2,3,4 Yent All Pool Top 4 Rabbit 4-5 All Shutdown @ 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> Pool Top 4 Rabbit 6,8 All All Pool Top + Puff 1,2,3,4 Vent All Pool Top 4 Puff 4-5 1,7,11 Shutdown 4 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> Pool Top 4 Puff 6,8 All
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