ML19241B725
| ML19241B725 | |
| Person / Time | |
|---|---|
| Site: | Zion File:ZionSolutions icon.png |
| Issue date: | 05/29/1979 |
| From: | Donohew J, Lantz E, Lobel R Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19241B710 | List: |
| References | |
| NUDOCS 7907230345 | |
| Download: ML19241B725 (12) | |
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UNITED STATES OF AMERICA NUCLEAR REGULATORY CCMMICSION BEFORE THE ATCMIC SAFETY AND IICENSING 90ARD In the Matter of
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COMMCNWEALTH EDISCN CCMPANY
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Docket Nos. 50-295
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50-304 (Zion Station, Units 1 and 2)
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NRC Staff Test.imony on Contention 2(g)
BY Richard M. Lobel Jack N. Donchew Edward Lantz Contention 2(g) is as folicws:
The Applicant's discussion of spent fuel boiling is in-adequate in that (1) there is no consideration given to the possibility that the pool might boil and (2) there is no discussion of possible damane to fuel cladding or of the consaquent release of radionuclides under such conditions; therefore, there is no assurance that public health and safety will not be endangered.
In addition, the heat removal cacacity of the Scent Fuel Pool Cooling Systens his not been shcwn to be adequate tc succort tne ex::anded occl capacity.
Eciling, as used in :nis contention, is ret clearly de#ined.
There are several Todes of boiling. At the conditions of tne spent fuel pool with loss cf cooling, tne ccolant surrcunding tne fuel rods would be in tne nucleate boiling Tcde. Nucleate boiling is a hignly efficient mode of neat trint f e r.
In the pressurized water reactor core, at full cower cceration, a small number of the fuel rods nor ally onerate in nucleate boiling.
In a hoiling water reactor core most of the fuel rods operate in nucleate boiling.
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. If the density of water bubbles were to increase at the surface of the fuel rod so that the mode of boiling changed from nucleate boiling to film boiling, the cladding temperature would increase significantly over that in nucleate boiling.
However, because of the low heat flux of a fual rod in a scent fuel pcol (with all its power coming only from decay heat), such a fuel rod would not undergo film boiling. Therefore, no damage is expected to fuel rods due to boiling of the spent fuel pool.
Cumulative spent fuel pool experience as recent as June 1978 has shown that "no ccmercial water reactor fuel has yet been observed to develop defects while stored in spent fuel pocis" at nomal spent fuel pool conditions.
(Reference 1).
Also, available evidence cited in Reference 1 shnws that a fuel rod which was already defected from operation in the reactor would not undergo further degradation. At the temceratures of fuel rods in a boiling spent fuel pool there should be no dissolving of the UO2 pellets if exposed to fuel pool water through a cladding defect. Observations at Karlsruhe, West Gemany showed no detectable dissolving of fuel cellets at nomal spent fuel :: col temceratures (Reference 1).
We would expect that the increased temceratures due to loss of spent fuel pool cooling would not change this result.
Cxication of Zircaloy claddi.q at boiling conditiens can be assumed to be negligible based cn data from Zircaloy 2 tubes excosed to treated Columbia 0
River water (less pure than spent fuel pool water) at a;croximately 9C C' Reference i:
A.5. Jonncon, Jr., " Behavior of Scent luclear Fuel in Water Pcol Storage", Battelle Pacific 'iorthwest Laboratories, BZL 2256 Sept.1977.
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. (Reference 1). Extrapolation of this data to 100 yeart yielded a conversion of Zircaloy to oxide of less than 0.1". (clad wall thickness).
fuel bundles to withstand An indication of the ability of Zircaloy clad UO2 boiling af ter reactor irradiation and subsequent spent fuel pool storage is the reirradiation of three fuel bundles of CANCU reactor fuel (Canadian reactor design). CANDU fuel rods are shorter than those used in the Zion reactor and have a thinner cladcing and a slightly larger diameter. However, the cladding material is Zircaloy 4 which is the same as the Zion fuel rods.
Af ter a first irradiatica to a low burnup, these bundles were placed in pool storage for 10, 9 and 5 years, respectively. The bundlas were then reirradiated for approximately one month at conditions apnroximately those in a CANDU reacto r.
The CANDU reactor operates at noninal inlet and outlet temperatures of 510 and 5940F, respectively. The reactor pressure drops from 1630 psia the 1449 psia across the ccre.
These conditions are much more severe than those which could occur in a boiling spent fuel pcol, The reactor conditions also would include a high energy neutron and garra ray flux which would r.ot be present in the scent fuel ; col.
The ceak oower in these fuel rcds was higher than that expected in the Zion peak power fuel rods and wculd therefore be at least an order of magnitude higher than the power o# a fuel rod in the scent fuel cool, Cne would there ore excect the stresses in the cladding to also be higher than those f
which could be imcosed on the cladding o' tne Zion fuel rods in a boiling scent fuel pool,
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4 Nevertheless, it is reported (Reference 1) that following this period of reirradiation, no defects were found in any of the CANCU fuel rods.
In conclusion, the spent fuel pool boiling node would be nucleate boiling.
PWR fuel rods are designed to operate in *.he reactor core in nucleate boiling at heat fluxes which are orders of magnitude higher than those which could occur in the spent fuel poo'.
Therefore, failures of fuel rods in the spent fuel cool due to boiling wculd not be expected.
Data also exists to show that a fuel rod defect would not be further degraded if boiling were to occur.
In regard to the possible release of radionuclides to the atmosphere it should be noted that the radioactivity of the additional spent fuel in the pool because of the pool modification would have decayed for several years.
The volatile radioactive nuclides in the defective failed fuel would.have, therefore, either decayed or diffused into the pool water.
The remaining radioactivity in tre spent fuel would then be non-volatile. For this activity, the leakage of activity from the fuel pin during cool boiling wculd not be significantly different from that at normal pcol operations.
Uncer normal concitions, experience indicates tnat taere is little radio-nuclite leakage from scent fuel stored in ccols af ter tne fuel has cooled for seser31 months.
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. The predominance of radionuclides in the spent fuel pool water appears to be radionuclides that were present in the reactor coolant system prior to refueling (which becomes mixed with water in the spent fuel pool during refueling operations) or crud dislodged from the surface of the spent fuel during transfer from the reactor core to the spent fuel pool.
Curing and after refueling, the spent fuel pool cleanup system reduces the radioactivity concentrations considerably.
It is theorized that most failed fuel contains small, pinhole-like perforations in the ' fuel cladding at the reactor operating conditions. A few weeks af ter refueling, the spent fuel cools in the spent fuel pool so that fuel clad temperature is relatively cool.
This substantial temperature reduction should reduce the rate of release of fission products from the fuel pellets and decrease the gas pressure in the fuel rod, thereby tending to retain the fission products within the fuel rod.
In addition, most of the gaseous fission products have short half-lives and decay to insignjficant level: within a few months.
Based on the operational reports submitted by the Licensees or discussions with the operators, there has not been any significant leakage of fission croducts from scent light water reactor fuel stored in the Morris Oceraticn (MO) (formerly vidwest Recovery Plant) at vorris, Illinois, or a Nuclear Fuel Services' (NFS) storage ; col ac West Valley, New York.
Spent fuel nas been stored in these two :cols wnicn, wnile it aas in a reactor, was determined to have significant leakage and was tnerefore removed fecm tne core. Af ter storage in the ensite scent fuel pool, tnis fuel was later snipted to either MO or NFS for extended O.7 n
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6-storage. Although the fuel exhibited significant leakage at reactor operating conditions, there was no significant leakags from this fuel in the offsite storage facility.
The conditions i'. the spent fuel during pool boiling which will affect the leakage of radioactivity from this additional spent fuel are not significantly different from the conditions in the pellet-cladding gap during normal 3001 operations.
dased on the experience discussed above for normal pcol conditions, we would not expect boiling of the pool to result in a significant increase, if any, in the leakage of activity frcm the additional spent fuel in the pool. Under normal pool conditions, any non-volatile radio-activity leaking from spent fuel into the pool water should remain in the pool water to be removed by the pool purification system. Urider conditicns of the pcol boiling, this radioactivity may be entrained in water droplets in the air above the pool.
These droplets will concense cut on surfaces in the fuel buildinc and a fraction of these droclets could ce entrained in the building ventilation air flow.
In the ventilation system, the droplets will concense cut on tne ducts cr Le #iitered ty tne filtration system.
The fil tration sysmen has pref 11:ers, HEPA filters and :harcoal #il'ers.
These filters will rerove the ::ater crcolets and the -adicacti,ity frcm the air until tne ceci cooling systen and purification systen is repaired or the het spent fuel is returned to the reactor vessel.
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Thus, it is our conclusion that fuel ;ool boiling, and its resultant effects on spent fuel stored therein, does not constitute a credible tnreat to public health and safety.
The following three sections of the NRC's Marcn 29,19/9 Safety Evaluatinn for the proposed modification audress the second part of thir, contention, which is on the adequacy of the heat removal capacity of tne heat remcval capacity of the spent fuel pool cooling systems:
2.2 Sn6nt Fuel Cooling The licensed thermal power for each unit of the Zion Station is 3,250 MWt.
The licensee olans to refuel both units 1 and 2 annually. This will require the replacement of acout 64 of tne 193 fual assemblies in each core every year.
Ihus normal refuelings will take place at 6-nonth intervais. To calculate the maximum heat load, the licensee assumed that it would take 10 days after the reactor was shut dcwn to comolete the transfer of both 1/3 of a core in a normal refuel-ing and full core in a full core offload. With these delay times, the licensee used the method given in American National Standard 5.1 to calculate 20.a x 106 Stu per hour as the 0
maximun heat load for an annual refueling and 35,0 x 10 Stu per hour as the maximum heat lead for a full core offload.
The spent fuel cool cooling systen as described in Chacter 9 of tre Final Safety Analysis Recort consists of two pumps and two heat e3 changers.
Each pumo is designed to pumo 2,330 gpn (1.15 x IOC pounds Der hcur), and each heat exchancer is dasigned to transfer 14.9 x lub 3:u cer hcur frce 120cp fue!
0001 water to 350F cccconent ccoling water wnich,is ficwinc tnrcugh tne heat excnanger at a rate of I
-9 x 100 ccunds oer ncur.
As shown in Chapters 6 and 9 of the Zion F1nal Safety Analysis Report, tnere are seismic Category I sources of makoup water fer tne scent fuel cool.
These are the re#ueling water storage tanks.
Inere is one of these stainless steel linec, reinforcea concrete, Class I structures for each units, and each one holds 389,0C0 gallons of water.
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. 2.2.1 Evaluation Using the method civan on pages 9.2.5-8 through 9.?.5-14 of the NRC Standard Review Plan dated November 24, 1975, with the uncertainty factor K equal to 0.1 for decay times longer tran 7
10 seconds, we calculated that the maximum peak heat load dt i.'.1g the 33rd refueling (the one t' hat fills the pool) could be 22.2 x 100 Btu per hour and that the maximum peak heat load for a full core offload that essentially fills the pool couid be 41.4 x 106 Stu per nour.
This full core offload was conservatively assumed to take place 6 months af ter the 30th refueling.
We also determined that the maximum incremental heat load that could be added by increasing the t,ur:ber of spent fuel assemblies in the poci from 868 to 2,112 is 5.4 x 106 Stu per hour. This is the difference in peak heat loads for full core affloads that essentially fill the present and the modified pools.
We calculated that with one pump operating with one heat exchanger, the spent fuel pool cooling system can maintain the fuel pool outlet water temperature below 1024 for the normal refueling.
In the highly unlikely event that both spent fuel pool cooling systems were to fail at the time when there was a peak heat load from a full core in the pool and the water was at its maximum temperature, we calculate that boiling could comrence in enut 7 he'.-s.
We also calculate that af ter boiling ccmences, the required water makeup rate will be less than 85 gallons per minute. We find that 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> will be sufficient time to establish an 85 gallon per minute makeup rate.
2.2.2 Conclusion We find that the preg.ent cooling capacity for the spent fuel pools at the Zion Nuclear Power Plant will be sufficient to handle the incremental heat load that will be added by the procosed modification. We also find that this incremenal neat load will not 3lter the safety consicerations of spent fuel :colirg from tnat which we previously reviewed and fcund to be acceptable.
These sections were prepared by Edward Lantz, and to the best of his knowlecge they are true and 20rrect.
In order to calculate the maximum heat load for this evalcation he assumed that a full care offload would take place six unths after the thirtieth refueling at the clant.
The reason for this is after fifteen years of oceration it is unlikely that a full core will be Y
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. discharged to the fuel pool right af ter a refueling. Also, in this regard, it should be noted that a core that has operated only six months af ter its last refueling will have fewer fission prodtcts and actinides in it than one that has operated a full year.
However, if we assume for extra conservatism that the core which is being offloaded ten days after a normal refueling dces have a full inventory of fission products, the 6
maxir,um possible heat load woul,d be 51 x 10 Stu/hr.
With both spent fuel p',ol cooling loops operating with this heat load in the pool the outlet water ten?ersture.' rom the spent fuel pool will be about 1250F.
If one of the cooling loops were to be stopped, i.e.,
if a single failure occurred, the outlet water tercerature would go up to about 17CoF.
If both of these cooling loops were to fcil at the time of this peak heat load, the rate of increate of the average fuel pool water temperature would be about 0
11 F/hr, Thus, there would be about eight he: s before boiling would comrence. After thi3 the maximum po.sible boil-off rate would be 105 gpm.
Since these numbers are not significintly different from those given in the 5efety Evaluation, the conclusion ' emains that tne present cooling capacit;,
is adequat; for the proposed mctification.
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PROFESSIONAL QUALIFICATI0MS 0F RICHARD M. LOSEL I am employed as a Reactor Engineer with the Division of Cperating Reacters, USNRC.
I graduated frca California State University at San Jose with a B.S. in Mechanical Engineering in 1966.
I ther. began work as a Mechanical c:yineer at Lawrence Livermore Laboratory, Livemore, California. At the same time I began work towards an M. S. degree in Mechanical Engineering at California State University at San Jose which I received in 1970. Since my masters degree I have taken an additional number of university course.s in nuclear and mechanical engineering.
In my present work at NRC I am responsible for reviewing reactor fuel reload applications and other safety matters concerning operating reactors.
My prime res;cnsibility is in the areas of nuclear fuel therral behavior and thermal hydraulic aspects of reactor behavior during steady state, anticipated transients and accidents.
Prior to my current assignment, I worked for three years in the Core Performance Branch where I was responsible for fuel ecd thermal performance including reviews of computer programs used by fuel vendors for predicting fuel conditions during steady state and transient conditions, fuel densification and analysis of fuel rods during a Loss-of-Ccolant Accident.
During the ceriod of 1966 to 1973 while I was emoloyed by Lawrence Liver-more Laboratory I was rest:onsible for the mechanical design of nuclear physics experiments.
I nave been a lecturer on nuclear fuel behavior at two University shcrt courses titled " Nuclear P0wer, Safety and tne Public" and " Nuclear Pcwer Reactor Safety Analysis."
PRCFESSIONAL CUALIFICATIONS CF EDWARD LANTZ An an Engineering Systems Analyst in tne Plant Sjs es Branch, I am resconsible for tecnnical reviews and evaluations of component and system designs and operating characteristics of licensed nuclear pcwer reacters.
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I have a Bachelor of Science degree in Engineering Physics from the Case Institute of Technology and a Masters of Science degree in Physics from Union College and a total of 28 years of professional experience, with over 20 years in the nuclear field. My excerience includes work on reactor transients and safeguards analysis, nuclear reactor analysis and design, research and development on nuclear reactor and reactor control concepts and investigations of their operational anc safety aspects.
I have held my present position with the Commission since December 1975.
My previous position, which I held for about two and one half years, was Project Manager in the Gas Cooled Reactors Branch, Division of Reactor Licensing, U.S. Nuclear Regulatory Comission, where I was responsible for the technical review, analysis, and the evaluation of the nuclear safety aspects of applications for construction and operation of nuclear power plants. For about ten years prior to that I was Head of the Nuclear Reactor Section in NASA. My section was responsible for the development and verification of nuclear reactor analysis computer programs, conceptual design engineering, and development engineering contracting. Prior to my employment with NASA, I was a nuclear engineer at the Knolls Atomic Power Laboratory for about six years, where I worked on the safeguards and nuclear design of the 53G reactors and the initial development of the nuclear design of the S5G reactors. Previour experience includes system engineering and electrical engineering with :ne General Electric Company and electronic develcpment engineering with tne Victoreen Instrument Company.
PROFESSIONAL CUALIFICATIONS OF JACK N. DCNCHEW, JR.
My name is Jack N. Donchew, Jr.
I an a Senior Nuclear Enginee, in the Enviror. mental Evaluation Branch in the Division of Coerating Reactors, U.5. Nuclear Regulatory Comission (NRC). My duties include the-review of rad-waste treat ~ent syst. ems and engineered safety feature ventilation systens for coerating reactors.
I received a Bachelor of Engineering Physics Degree frcm Cornell University in 1965, a Masters of Science Cegree in Nuclear Engineerinc frcm Vassachusetts Ins:'tute of Tecnncicgy in 1968, and a Occtor of Science Degree in Nuclear Engineeri g # rem VassacNusetts Institute c# Tectnoicgy in 1970.
I received my Professional Engineers License in Nuclear Engineering frcn tne Ccanonwealth of Denrsylvania in 1970 C l'
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Af ter graduation, I worked for Stone and Webster Engineering Corporation as an engineer in the Radiation Protection Group.
I was responsible for estimating source tenns, release rates and resulting doses for the Safety Analysis Report, Environmental Report and response to 1RC questions for boiling water nuclear reactors.
I was also responsible for shielding design for the reactor water cleanup system.
In February,1973, I became a Power Engineer in the Process Engineering Group, Stone and Webster Engineering Corporation.
I was lead engineer for the Shoreham Project and the equipment specialist for all nuclear plants for the containment iodine spray removal system, ventilation filter assemblies, and Boiling Water Reactor and Fressurized Water Reector gaseous wate treatment system.
In June 1975, I jointed the Nuclear Regulatory Commission as a senior nuclear engineer in the Effluent Treatment Systems Branch, Directorate of Licensing.
I was involved in rad-waste system licensing reviews of nuclear power plants.
I have conducted generic studies of the degradation of charcoal absorbers in ventilation filter assemblies.
In December 1975, I jointed the Environmental Evaluation Branch in the Division of Operating Reactors.
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