ML20127A157
| ML20127A157 | |
| Person / Time | |
|---|---|
| Site: | University of Lowell |
| Issue date: | 06/10/1985 |
| From: | Perez P MASSACHUSETTS, UNIV. OF, LOWELL, MA (FORMERLY LOWELL |
| To: | Thomas C Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8506210038 | |
| Download: ML20127A157 (17) | |
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(617) 452 5000 June 10,1985 Mr. Cecil 0. Thomas, Chief ..
Standardization and Special Projects Branch Division of Licensing U.S. Nuclear Regulatory Comission Washington, D.C. 20555 RE: Responses to final questions for the safety evaluation -
University of Lowell License Renewal Docket No. 50-223
Dear Mr. Thomas:
Written responses to the final questions concerning the safety evaluation for the ULR are enclosed in this communiqul.
If you have any questions concerning these responses, please contact Mr. Thomas Wallace, Reactor Supervisor, or myself at (617) 452-5000, Ext. 2245.
Sincerely yours, alwk.
Pedro B. Perez Chief, Reactor Operations P8P:dm Enclosure (s) 8506
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FINAL QUESTIONS UNIVERSITY OF LOWELL .
1.Q. What is the effective delay neutron fraction?
1.A. 'The clean-cold core valde is 0.007 and due to the high enrichment of the fuel this value does not change considerably.
2.Q. What is the range of core inlet coolant temperature? What is the nominal value?
2.A. These values are given in section 9.1.2.1 of the updated SAR. The nomal operating range would be 80-90'F and the nominal value is 84*F.
3.Q. What is the range of core outlet coolant temperatures? What is the nominal value?
3.A. Operating history shows a range of 75-100'F; nominal 87'F.
4.Q. How was the limiting value of the withdrawal rate for the control blades determined?
4.A. The maximum rate of reactivity insertion by the shim safety rods which is allowed in Technical Specification 3.1.6, 0.025% Ak/k/sec, assures that the Safety Limit will not exceed during a startup accident due to a continuous linear reactivity insertion. Analysis in Paragraph 9.1.11 of the SAR shows that a maximum power of less than 1.4 MW would be reached assuming a continuous linear reactivity insertion rate of 0.035% Ak/k per second, which greater than the maximum allowed. l Together with the blade worth, the withdrawal rate is determined complying with the above limit on reactivity insertion.
5.Q. How much water does each' reactor pool contain?
5.A. Total volume is approximately 76,000 gallons, the bulk pool occupies
~45,964 gallons and the stall pool occupies ~30,000 gallons.
6.Q. What is the vertical position of the core in relation to the bottom and the surface of the pool?
6.A. Centerline of core to floor is 6'1"; centerline to pool surface is 25'1".
7.Q. What are the individual rod worths und what is the measured total 1 ,
m shutdown worth?
7.A. August 1984, surveillance on blade worth yielded the following total worths:
Blade 1: 1.98% Ak/k Blade 2: 2.81% Ak/k Blade 3: 2.74%Ak/k Blade 4: 3.72% Ak/k Reg. Rod.: 0.338% Ak/k The measured shutdown margin is ~9.913% Ak/k.
8.Q. Is the maximum excess reactivity of the reactor?
8.A. Technical Specification 3.1.2, limits excess reactivity to 4.7% Ak/k.
In practicy, we have not allowed loadings to exceed 4.5% Ak/k.
9.Q. Are the values of reactivity worth of experiments absolute values?
9.A. The limiting reactivity worth values of experiments are absolute values.
10.Q. Provide a list of signals that activate the rod withdrawal inhibit and rod withdrawal prohibit circuit.
10.A. The inhibits and prohibits are: .
a) Less than 3 cys-blade inhibit from the log Count Rate Meter -
(Start Up Channel) b) Start up counter detector traveling-Blade Inhibit ;
c) Less than 5% readings from safety channel meters-Blade Inhibit (Safety Channel) d) 110% reading on Safety Channels - Blade Inhibit (Safety Channel) e) 15 second reading on period meter - Regulating Rod Inhibit (from Log N Channel) ll.Q. What are ULR policies regarding doses to reactor-related personnel?
Are doses maintained below some level lower than the 10 CFR 20 limits, such as might be done in ALARA program? Is there an ALARA program at ULR7 ll.A. While the University does not operate a formalized ALARA program under the direction of a single responsible individual, the administration and staff do endorse and operate under the principles of ALARA. Under policies described in the University of Lowell Radiation Safety Guide quarterly doses are limited to the 10 CFR 20, 101 tabulated values.
Extensions to 3 rem per quarter are not allowed. The Radiation Safety Officer (RS0) oversees the routine operational health physics program and while a formalized ALARA program does not exist, the program is run 2
in keeping with the goals of maintaining doses as low as can be reasonably achieved. The Radiation Safety Officer attempts to maintain personnel whole body dose below 500 mrem per year.
All experimental use of the reactor facilities are reviewed by the RSO.
In addition, any experimental use which falls in a category which has not previously been reviewed by the Reactor Safety Subcommittee is reviewed by that committee. Experiments are reviewed from the point of view of overall safety with regard to facility personnel and experimenters.
Offsite releases of A-41, the only significantly produced airborne activity during routine operation are minimized by reducing A-41 production through the use of sealed hollow beam tube plugs inserted into the core end of the beam tubes when the facilities are not in use.
All liquid wastes are collected in holding tanks and activity is allowed to decay to achieve levels well below 10CFR20 release limits before dumping to the sanitary sewer.
12.Q. What is the sewer effluent dilution factor that is applied to the annual release rate (0.711 mci /yr) to meet 10CFR20 release limits?
12.A. Results of water metering at the University have shown a typical annual usage of 10llem3 of water per year. This water drains to the same line as does the sanitary sewer outlet from the Radiation Laboratory building from which the liquid waste is released. The 0.711 mci per year average release of 6 emitting radioactivity diluted by the average annual volume of 10 ll m3 e yields an average concentration of 7.1x10-9#Cicm-3 which is well below 10CFR20 release limit concentrations.
13.Q. What are the corrected values of average annual 41Ar release rates and the associated whole-body dose calculated for an individual outside the facility? What reference describes the expanding balloon cloud model used to calculate this dose?
13.A. The correct five year average value for the A-41 release rate is 0.34pCi sec-l. The whole body penetrating dose to an individual standing outside the reactor and immersed in an expanding cloud of the stack effluent containing A-41 emitted at this rate continuously for one year is 0.51 mrem. The reference describing the expanding balloon cloud model used to calculate this dose is " Evaluation of the Environmental Significance of the Projected 41Ar Release from the Lowell Technological Institute Reactor" by K.W. Skrable, G.E. Chabot, J. K11111ea, and H. Wedlick and published in llealth Physics, Vol . 22 pp. 49-56, January 1972.
14.Q. We need information on extreme wind conditions, hurricanes and tornadoes and notification system, if any.
3
14.A.
Tornadoes Affectina Middlesex County Within the Last 20 Years Date Cities Affected Damane Assessment (in Dollarsl 1960 Methuen 1972 Tyngsboro/Chelmsford $1,000,000 1974 Wakefield/Lynnfield 300,000 1974 Billerica/Tyngsboro ,
The tabulation of Mass tornadoes is not better organized or informative, but data on those storms are hard to come by. One has to recognize that a severe tornado can hit any particular area, with extremely high winds, perhaps 200 mph or more, and explosive change of air pressure. But even if the state has several a year with areas involved on the average of 5 square miles, which is too high, then the chance that a spot would be hit becomes once in over 1000 years. It !
would seem that the Thom [1] paper would be the best estimate of top winds to be expected in various expanses of time [2].
Hurricanes Within the Past 20 Years Within this time span the state of Massachusetts has had only one hurricane that was classified as " extreme". This storm occurred on September 12, 1960 and was named Donna.
Data exists for the past 335 years with an average of one storm in about 5 years. As only rare hurricanes affect all of Massachusetts, the threat for any one Bay State location will be lower, and the intervals longer, than the above statistics appear to indicate.
Threats of hurricane damage weaken as one goes inland.
The hurricane season in Massachusetts is from August into October [3]. ,
Notification System ,
Every morning prior to startup and again every 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of operations
- the National Weather Service is called for a forecast. If weather is predicted that could cause adverse conditions the reactor is either not started up or if in operation the reactor will be secured. Attachment
- 2 shows the extreme wind conditions at a 25 year and 50 year reoccurrence. These maps were taken from Reference 1.
15.Q. Section 10.5 is titled Review and Audit"; however there is no discussion in provisions for auditing reactor operations and performance. Please provide the information on the organization, number of professionals in the " auditing" group and procedures used at ULR for auditing the reactor facility operations, performances, records, etc.
4 t
l 15.A. The University is in the process of submitting a number of Technical Specification changes as a part of the licensing process. This question will be addressed by these Technien1 Specification changes.
16.Q. What is the composition of the Radiation Safety Committee?
16.A. The composition of the Reactor Safety Subcommitter is as follows: i Chairman: James P. Phelps, Professor of Nuclear Engineering
! Secretary: George E. Chabot, Radiation Safety Officer l Members: Cesare DeLizza, Electronics Engineer l
- Gunter Kegel, Professor of Physics
'Ihomas Wallace, Reactor Supervisor l The composition of the present Radiation Safety Committee is as follows:
l Chairman: Kenneth W. Skrable, Professor, Physics Department, Radiological Sciences Program Secretary: George E. Chabot, Jr., Radiation Safety Officer Members: Leon E. Beghian, Vice President of Academic Services and Director of Pinanski Center Paul Daigle, Campus Safety Officer Theresa Daigle, Purchasing Representative Research Foundation Nelson Eby, Professor, Earth Sciences ,
Michael Frechette, Professor, Clinical Lab. Sciences Gunter Kegel, Professor, Physics Department and Chairman of
- Accelerator Safety Subcommittee Mary Kloppenburg, Purchasing Representative, Business Office .
James Phelps, Professor, Nuclear Engineering and Chairman i of Reactor Safety Subcommittee Thomas Wallace, Reactor Supervisor l
17.Q. What are the limitations on fueled experiments?
l l 17.A. The University of Lowell proposes that fueled experiments be limited to j one gram of Uranium-235. This limitation would not apply to fission '
detectors or other " measurement devices". An analysis of the failure l of a fueled experiment containing 1 gram U-235 is attached.
(Attachment #1)
References
- 1. H.C.S. Thom, New Distribution of Extreme Winds in the United I States, Journal of the Structural Division. Proceedinas of l the American Society of Civil Enaineers, July 1968, pp. 1787.
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- 2. Personal Communication with R. 14utzenheiser, State Climatologist, New England Climate Service.
- 3. Hurricanes Affecting Massachusetts - A Brief History, R. lautzenheiser, State Climatologist, New England Climate Service.
Figures Needed:
Figure 4-10 will be fowarded within five (5) days.
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sr ATTACHMENT # 1 Impact of the Release of Fission Products from a One Gram Uranium-235 Fueled Experiment Introduction The meteorology description of atmospheric dilution of radioactive releases from the University of lowell reactor is complicated by the presence of buildings on the site and the fact that the stack is not high. In the event of a fission product release in association with failure of a fueled experiment airborne releases of activity may occur from the stack over a ahort time or may leak from the building over a longer time period if ventilation is shut down. A number of authors have investigated the effects of buildings on atmospheric releases. Work by Halitsky, Golden, Halpern, and Wu(1) and additional work by Dickson, Start, and Markee(2) has dealt with this area and has been reviewed in the noted documents.
. A significant concern in the event of releases of fission products from a fueled experiment would be the committed thyroid dose to individuals outside the reactor building. The following analysis uses the results presented in the above sited references to estimate concentrations or concentration-exposure time integrals from which doses are projected.
Proiected Radiciodine Doses from a One Gram U-235 Fueled Exseriment i For purposes of this analysis it is assumed that one gram of U-235 has been subjected to a thermal neutron flux of 1013c 1 for a time duration to produce saturation quantities of the isotopes I-131, 132, 133, 134 and 135. Immediately following irradiation it is assumed that the integrity l
of the experiment fails such that all the radioiodines are released from the experiment and leave the reactor building as a puff over a very short time duration. An individual standing outside the building and in the building i
wake is assumed to breathe the contaminated air during the duration of the 7
passage of the puff release.
From references (1) and (2) the concentration coefficient, Kc, is given by X
c = Q/AV where X is the actual concentration at any point, ,
Q is the release rate of activity of concern, AV is the wake volume flow in which A represents the effective cross sectional area of the weke, taken here as the maximum cross sectional reactor building area of about 595 m ,2 and V is the wind speed, taken as 1 me-l. ,
The above equation applies to a continuous release but may be applied to a short duration release in which total activity replaces Q and time integrated concentrations are used.
Reference (1) describes results from Halitsky et al that show the average value of Kc in the wake of a single building of dimensions very similar to the University of Lowell reactor building to be approximately unity, with a ,
maximum value of about 5. Reference (2) also gives results from Dickson, et al for a more complex site in which other buildings are present, perhaps more indicative of the lowell sites for this case Kc values are generally less than those from Halitsky. A value of Kc = 1 is used here.
Table I gives some of the physical data required.
Table 1 Thyroid Dose Curies Available Commitment per Iodine Fission Decay Constant, A at Time of Isotope Yield, y (day-1) Failure, N Ciinhaled)
Dc(rads /C1 131 0.029 0.086 11.7 1.48 x 106 132 0.043 7.14 17.3 5.35x104 133 0.065 0.795 26.1 4.00x105 134 0.080 19.0 32.1 2.50x104 135 0.064 2.47 25.7 1.24x105 The activity, N , of the ith radioiodine available at the time of failure is given by N = Fy g/3.7x10 where F is the U-235 fission rate (fa-1) and yi is the fission yield.
F is obtained from F= N,op4 where Nu is the number of U-235 atoms /g U-235 - 2.563x1021 op is the U-235 fission cross section, 580 barns, and 4 is the thermal neutron flux, 10 13 c e-2s -l.
The time integrated concentration in Ci-sec/m3 for the puff release of the ith radiofodine is given by C 1 ,t*
K c
C1 ,t " gy The total dose commitment to the thryoid, D.,1, fron inhalation of the ith radioiodine is then obtained:
D.,1 - Ci,t DeR, where R is the assumed breathing rate taken as 2.315x10-4m3s-1 The calculated values of D.,i are shown in Tabic 2.
9
Table 2 Iodine Isotope ,
Total Committed Thyroid Dose, IL.(rads) 131 6.74 132 0.36 133 4.06 134 0.31 135 1.24 Total 12.71 The above results are representative for a modest wind speed of 1 m ,
sec-1 and applies to adverse stable / meteorological condition in which appreciable mixing and dilution does not occur as a result of atmospheric instabilities.
The above analysis, as noted, assumes virtually instantaneous release of all the radioiodines with no credit taken for plateout or decay. If an experiment failure was followed by shutdown of the reactor ventilation such that the maximum building leak rate of 0.10 day-l prevailed, the resultant committed dose from the ith radioiodine outside the reactor would be given l by .I Di = Cg M,dt where Ci is the concentration of the i th radionuclide at time, t, post release and is given by KQ cg Ci= gy where Qi is the release rate of the ith radioiodine from the building.
Q3 is expressed by Qi = 0.10N c-k,t where ke = 0.10 day-1 + Ai ,
10
I and Qi is in Ci day-I*
Thus the committed thyroid dose, Di , over the exposure interval, T, beginning at the ting of failure is 0.10RR D ce1 -
Di- Avk O ~
'the value c* Di for T = 2 hrs and T = = are given in Table 3.
Table 3 Iodine Isotope Thyroid Committed Dose (rads) 2 hrs =
131 5.57x10-2 3.62 l 132 2.25x10-3 4,97x10-3 133 3.26x10-2 0.454 134 1.30x10-3 1.64x10-3 135 9.30x10-3 4.83x10-2 ,
Total 0.101 4.13 In reality coenitted doses would be less than those calculated above considering the fact that no credit has been taken for removal of radioiodines by any physical / chemical proceses. The analysis has also assumed an operating schedule which would result in saturation quantities of radioiodines in the experimental sample. Based on operating experience to date this is very unlikely particularly for the long lived and important I-131. In addition, if the release were protracted over any significant titue period traffic control over aieno close to the reactor would be effected to keep individuals removed from the,most likely affected areas. The thyroid calculated doses are well below the 10CFR100 limits and are also appreciably less than the EPA 11
d protective a'etion guide of 25 rem.
Projected Whole Body Gamma Ray Doses from a One Gram U-235 Fueled Experiment !
In case of failure of a fueled experiment resulting in releases of fission pro' duct activities the most significant external doses outside the
.~.
reactor building would be from release of the volatile fission products to the air.. Sectiod 9.2.1.2 of the updated Safety Analysis Report has reviewed the external dose consequences of the release of radioiodines and noble gases following damage to one-fuel plate of a fuel element. Complete escape of all the noble gases was assumed; the radioiodines contribute a relatively insignificant amount to the external penetrating dose. For an experiment s
involving I gram of U-235 exposed at a thermal flux of 1013cm-2sec-1 to achieve saturation activities of the noble gases, thequantitiesofsignificantfgpsionproductnoblegasesproducedwouldbe equal to 0.075 of those listed in Table 9.6 of the updated Safety Analysis Report and are given in Table 4.
Table 4 Radionuclides Saturation Activity (Curies) l Kr-85m 5.250 l Kr-87 10.05 Kr-88 14.55 Xe-131m 11.70 l
Xe-133m 26.10 Xe-133 26.10 Xe-135m 25.73 Xe-135 25.73 For a short duration puff type release of all the noble gases from the 12
1 i
reactor building the time integrated concentration in Ci-sec/m3 for the l
ith radionuclideo is as given earlier:
ci Ci ,t " AV where the parameters are as discussed earlier except that the equation refers to noble gases rather than iodines. Under the assumption of semiinfinite l
cloud exposure situation and using the (MPC)1 values given in Table 9.6 of l l
the updated safety analysis report, the external whole body penetrating dose, l Hext,1, from the i th radionuclide may be calculated from
-3 C 2.5x10 rem 1.t -7 f,t Hext.i " * = 6.94x10 3600 sec (MPC) (MPC)g Values of Hext.i are given in Table 5.
Table 5 Time Integrated Whole Body External Radionuclide Concentration, C1 ,t Penetrating Dose (rem), Hext.i Kr-85m 8.82x10-3 4.4x10-4 Kr-87 1.69x10-2 2.9x10-3 Kr-88 2.45x10-2 1.41x10-2 l
l Ke-131m 1.97x10-2 1,05x10-5 l
- o Ke-133m 4.39x10-2 7.08x10-5 Ie-133 4.39x10-2 9.51x10-4 Ie-135m 4.32x10-2 1.77x10-3 l
Ie-135 4.32x10-2 2.73x10-3 1
Total 2.30x10-2 The projected dose of 23 mrem is very low and would likely be even lower in view of the fact that the puff release would not yield a cloud of semiinfinite dimensions, and the lack of energy spatial equilibrium would result in notably lower doses than those calculated. If the ventilation were 13
shut down immediately following the experiment failure and leakage occurred at the rate of 0.10 per day from the reactor building, the above doses from noble gases would be reduced even further as a result of decay of the short lived noble gases.
References (1) Slade, David H., Ed., Meteorology and Atomic Energy 1968, pp. 230-250, Atomic Energy Commission /Div. of Technical Informations, July 1968.
(2) Randersson, Darryl, Ed., Atmospheric Science and Power Production, pp. 291-315, DOEA IC-27601, 1984.
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