ML19210A393
| ML19210A393 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 05/24/1977 |
| From: | Herbein J METROPOLITAN EDISON CO. |
| To: | Reid R Office of Nuclear Reactor Regulation |
| References | |
| GQL-0703, GQL-703, NUDOCS 7910290619 | |
| Download: ML19210A393 (11) | |
Text
Nnr.,,,ptu 195 U.S. NUCLE AR nt t.UL AToHY CoMMISSloN DOCKLT NUMitr n
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FROM.
oATE Of C o CU *.t C N T Metropolitan Edison Con.pany 5/24/77 Mr. R. W. Reid Reading, Pa.
o,7c ncec,yc o J. G. Herbein 5/31/77 kETTEn CNoToR; ZED PROP INPUT FonM NUMOCR oF COPIES HECEIVCD R: GIN AL hNCLASSIFlE D CCoPY l
$l bN A oESCnirTioN ENCLoSU9C Consists of requested additional information regarding their tech spec change request No.
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PLANT NAME:
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Director cf iuclear Reacter Regulat y n..
Attn:
R. W. Reid, Chief Cperating Beacter Branch :To. h U. S. :Iuclear Regulatcry Cen=ission Washington, D.
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Dear Sir:
Three Mile Island :iuclear Station, Unit 1 ( 31I-1)
Decket IIc. 50-289 Operating License IIc. DFR-50 Attached please find cur response to your.2 quest for additional infer =aticn regarding cur Technical Specification Change Request 27o. LT to permit an increase in the storage capacity cf the 31I-1 spent fuel pccl. Feel free te call Mr.
J. M. Cajigas (Ext.16k) should you have any questiens regarding this natter.
Sincerely, lID f[,ff\\
J. G. Herbein g
Vice President
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Ms. Margaret Reilly
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FADER Il D Fulten Euilding Harrisburg, PA 17120 1469 DN 7715.3012:3
THREE MILE ISLAND NUCLEAR STATION UNIT 1 SPENT FUEL POOL MODIFICATION REQUEST FOR ADDITIONAL INFORMATION QUESTION #1:
Provide a detailed sum =ary of the stress margins due to the increased loading of the fuel pool walls and floor for the critical load combinations.
Include a discussion of the possibility of shear failures in the areas of contact of the rack supports with the floor and walls.
Compare nu=erically these results to those for the previous rack structure.
RESPONSE
Since the spent fuel pool was reanalyzed using the strength design method instead of the working stress design cethod, our answers are in the form of required sectica capacity versus existing section capacity instead of stress margins.
The loading combinations used are in accordance with U. S. NRC Standard Review Plan 3.8.4 as follows:
a.
U = 1.4D + 1.4F + 1.7L b.
U = 1.4D + 1.4F + 1.7L + 1.92 c.
U = 0.75 (1.4D + 1.4F + 1.7L + 1.7 T )
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U = 0.75 (1.4D + 1.4F + 1.7L + 1.7 To + 1.9E) e.
U = 1.2D + 1.9E + 1.2F f.
U = 0.9D + 1.4F g.
U = D + L + T ' + E' + F o
h.
U=D+L+T'+I+F o
a where:
U = section strength required to resist design loads based on the strength design methods described in ACI 318-71 D = dead loads including permanent equipment L = live loads including movable equipment F = bydrostatic loads To = loads generated by temperature with full capacity of pool cooling system operable T ' = loads generated by temperatures resulting from partial failure o
of pool cooling system E = loads due to OBE with maximum ground acceleration of 0.06g.
One hori: ental acceleration coeponent combines additively with the veritcal acceleration compor.ent.*
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Page 1 E' = lo'da due to SSE with maxi =um ground acceleration of 0.12g.
a One horizontal acceleration component combines additively with the vertical acccleration component.*
I,- loads due to hypothetical aircraft transmitted to the pool from exterior walls or roof by interconnecting members.
- Seismic forces consist of the su=mation of the following individual loads:
1.
Structural Seismic 2..
Fuel Rack Seismic Loading 3.
Hydrodynamic Loading The critical sections, loading cochinations required capacity, ex12 ting capacity and discussions of the fuel pool walls and floor are shown in Taba.e 1.
Our analysis of the concrete structure of the spent fuel pools A and B is a con-servative linear elastic analysis. The resultant forces and moments of all sections except local areas as shown in Table 1 are within the existirig capacity.
For critical localized sections 2, 3, 4 the amount of required section capacity in excess of the existing capacity is primarily caused by the te=perature effects.
As indicated in Table 1 discussion, the thermal stress is self-limiting and secondary in accordance with ASME Boiler and Pressure Vessel Code. ACI Standard 359-74,Section III, Division 2, 1975 ed., Section CC-3136.4, pp. 183-194. Hence, the excessive required sects.on capacities should not be a cause of concern.
The maximum estimated crack width of spent pool walls and floor above the bottom of the pool slab of the localized critical sections 2, 3, 4 is 0.0029 in., this compares favorably with the limiting crack width of 0.013 in. specified in Commentary on Building Code Requirements for Reinforced Concrete (ACI 318-71).
In comparison with the code permissible shear stresses (ACI 318-71) carried by the reinforced conc. rete, the nor=al shear and punching shear stresses are very small in the areas of contact of the proposed rack supports with the floor and walls. The forces due to the response of the new spent fuel storage racks are small, while the floor and wall thicknesses are relat.ively larger. Therefore, it is not possible to have shear failures in these areas.
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TABLE 1 CRITICAL SECTIONS LOADING REQUIRED & EXISTING DISCUSSIONS COMBINATIONS CAPACITY
- 1. Localized region, a, b, The required capacity The north-south is the bottom slab of Fool e, f slightly exceeds the weak direction of the two A at the south end existing capacity, way reinforced slab, and near the middle wall is slightly under capacity; extends 24'-0" north the structure will maintain from the niddle wall.
its integrity after local stress redistribution.
- 2. Localized region, c, d, The required capacity Since this wall is below South wall below g, h exceeds the existing the spent fuel pool and Elev. 305'-0", a capacity.
the thermal stress is self horizontal strip limiting and secondary, the of 5'-0" wide.
existing design is adequate.
Also, the required capacity at localized critical section is caused by the abrupt change of input temperature, and it will be reduced when refined tenperature gradients are applied.
- 3. Localized region, c, d, The required capacity Again, the large thermal North end of east g, h exceeds the existing stress is self li=iting and wall at Elev. 348'-0" capacity.
secondary, and with refined of Pool A.
modeling, the high coment would be reduced; the design is adequate.
4.
Localized region, c, d, The required capacity See 2 above.
West wall at the g, h exceeds the existing junction of the
- capacity, bottom slab and middle wall.
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QUESTION #2:
Provide the components of the stress value given in Table 5-2 for load combination "d" (as defined in Section 5.1.2) at grid beam location.
RESPONSE
The components of st tss at the worst grid beam location for load combination "d" are as follows:
Axial Stress (psi)
Bending Stress (psi)
Dead Load 270 2270 Thermal Load 40 3910 Impact Load 0
0 SSE 10760 6470 The contribution due to impact is zero because the loads due to fuel weight acting with the cans are higher than the impact loads due to fuel striking the cans.
QUESTION #3:
Provide justification for neglicting any a=plification of the seismic loads, transferred to the rack analyzed, due to the flexibility of the fuel cans in the adjacent racks.
RESPONSE
The fuel cans in adjacent racks were assumed to be infinitely flexible in that they provided no restraint between the upper and lower grids. Modeling in this manner tended to amplify the dynamic response since the significant mode shapes are governed by grid deflection rather than can deflection.
Some of this grid deflection could have been attenuated had the cans in surrounding racks been modeled.
QUESTION #4:
What has been the amount of solid wastes shipped from the plant in the last year?
RESPONSE
3 Between Janut.ry 1, 1976 and Dece=ber 31, 1976; 15,700 ft of solid waste was shipped from Three Mile Island Unit 1.
QUESTION #S:
On page 3-4 of your submittal of February 3,1977, you state that it is "i=possible" to predict the amount of waste generated from the precoat filter.
If the volu=e cannot be "upperbounded," there is no bcsis for you or us to reach a conclusion that the volume is negligible.
It is requested that you reevaluate the first pcragraph of page 3-4, discussing the projected frequency of operaticn of the filters, the basis for their replacement, the cubic feet of powdered resin used to precoat the filters and an estimate of the volume of solid waste presently attributable to the SFP operations.
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Page 5 RESPONSE; As stated in Section 3.3 of the Environmental Impact Evaluation of the TMI-l Fuel Rack Licensing Submittal, the "A" Spent Fuel Pool was purified by the RLVD System for a total of only 73 hours8.449074e-4 days <br />0.0203 hours <br />1.207011e-4 weeks <br />2.77765e-5 months <br /> during the period March 1976 to March 1977.
During this time, a full core of f load into "A" Pool occurred. The precoat filter used for this purification contains 1.33 ft of powdered resin per charge.
The criteria 3
for recharging the precoat filter is based on high differential pressure developing fuel across the filter.
Following the 73 hours8.449074e-4 days <br />0.0203 hours <br />1.207011e-4 weeks <br />2.77765e-5 months <br /> of operation of purifying the spent pool, the precoat filter was used for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> to purify the Borated Water Storage Tank. After use on this tank, the resin was discharged, then solidified.
Based on a full year of experience, the submittal stated that the a=ount of waste generated3 by the Spent Fuel Pool is negligible. To be extremely conservative, less than 10 ft of solid waste can be attributed to the SFP operations over a year's time.
QUESTION #6:
What has been the release of radioactive noble gases and tritium from the SFP Building in the last three years? What is the expected increase in the release of radioactive noble gases and tritium from the f acility due to the SFP odification?
RESPONSE
The SFP and Auxiliary Building Ventillation exhausts.are cocbined at TMI-1.
Continuous Within the monitoring of the combined exhc.ust began following plant startup.
sensitivity of the instru=entation, no tritium or radioactive nobic gases have ever been detecced, The Mini =um Detectable Activity (MDA) for radioactive noble gases is approximataly 5 s 10-8 uc/cc and tritium is approximately 1 x 10-8 pc/cc.
Similarly as discussed in Question #8, it is anticipated that the radionuclide concentrations in the fuel pool water will witness little change as a result of this modification.
Evaporation rates will be the same since the fuel pool te=peratures will remaic essentially unchanged from original calculations. Therefore, it is anticipated that the release tritium and radioactive noble gases will remain unchanged.
_ QUESTION #7:
What is the weight of any material (e.g., racks) that will be re=oved from the SFP due to the codification? What will be done with this caterial?
RESPONSE
The "B" Spent Fuel Fool presently is a dry, empty pool, There will be no material removed from this pool due to the fuel rack modification. The original fuel racks are stored in an open field next to Three Mile Island Nuclear Station Unit 1.
Since the racks, weighing approx 1=ately 15,000 pounds, are uncontaminated, they are scheduled to remain in storage until a use for the aluminum is found or disposed as ordinary scrap metal.
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Page 6 gUESTION #8:
Provide a di assion of increase in occupational man-rem exposure to personnel in the Spent Fuel Pool area from radionuclide concentration in the Spent Fual Fool due to the expansion of the capacity of the pool including the following:
(a) Identify the principC. radionuclides and their respective concentratiens in the spent fuel pool found by ga=ca isotopic analysis during all operations.
Identify the sample with respect to a specific operation (i.e., refueling, fuel handling, etc.).
(b) Provide an estimate of the man-rem exposure that will be received during removal of the old rac'<s and installation of new ones.
(c) Provide an estimate of the dose rates above the spent fuel pool from the concentrations of the radionuclides identified in (a) and the concomitant occupational exposure, in annual man-rem, due to all operations associated with fuel handling in the spent fuel pool area.
De5cribe the impact of the proposed modifications on these estimates.
Includ2 in your analysis the expected exposure from more frequent changing of the demineralizer resin and filter cartridges.
RESPONSE
(a) The TMI-l "A" Spent Fuel Pool has stored fuel for fourteen (14) months and has experienced two (2) yearly refuelings including a full core offload in 1976.
Based on various pool samples, the following table identifies the principal radionuclides and their respective concentrations:
Principal R.3dionuclide Concentration (uc/ml)
Operation Co 2.0x10-2)
S8 60 Co 1.8 x 104 l34 Ca 2.2 x 10-3 l36 Ca 1.6 x 10-4 Refueling Cal 37 2.0 x 10-3 131 1.7 x 10-3 1
Mr[4 1.2 x 10-4 S8 Co
- 1. 3 x 10-3 Co 4.0 x 10-5
} 4h 60 l34 Cs 4,1 x ig74 Midway Between Refuelings 137 5.0 x 10~4 CS M:E4 2.0 x 10-5
Pag'e 7-
RESPONSE
(Continued)
(b) There will be no man-rem exposure resslting from the removal of the old racks because the old racks were removed two years ago before plant startup. An estimated exposure of 0.15 man-rem is anticipated during the installation of the new fuel racks.
This estimate is based on actual fuel pool surveys and takes into accourt conservative rack installation requirements.
(c) Based on radia': ion ser eys during fourteen (14) months of fuel storage and two (2) refuelings, the following table provides typical dose rates:
Dose Rates (mR/hr)
Operation Annual Occupational Exposure (Mar.-Rem) 20 Refueling 11.5 0.4 Reactor Operating 1.5 The predominant contribution of radionuclide concentrations in the Spent Fuel Fool comes from the mixing of the pool water with primary coolant during refulcing. There-fore, it is anticipated that storing additional spent fuel will not significantly increase the radionuclide concentrations.
Consequently, the total annual man-rem exposure attributed to the fuel pool will be unchanged as a result of this rack modification.
It is not anticipated that the frequency of use of the fuel pool water purification system precoat filters will increase as a result of this modification; therefore, the man-rem exposurs from changing powdered resin will be unchanged.
QUESTION #9:
During the first resueling, 56 fuel asemblics were transferred into the SFP.
The submittal stated that during the current refueling, 48 fuel asse=blies will be replaced.
The submittal infers on page 5-3 that on the average, you plan to replace 52 fuel assemblies per year.
Based on your current fuel management plans, discuss the projected Jefueling schedcles, including the nu=ber of the fuel assemblies that will be transferred into the SFP at each refueling.
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RESPONSE
The phojected refueling schedule is as follows:
Year of Net Number of Fuel Asse=blies Total Nu=ber of Spent Fuel Refueling Discharged to the Spent Fuel Assemblies in the Spent Fuel Pools Fools 1976 56 56 1977 48 104 1978 52 156 1979 52 208 1980 52 260 1981 52 312 1982 53 365 1983 52 417 1984 52 469 1985 52
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521 1986 53 574 The refueling cycle continues in a re-occurring pattern of... 52, 52, 52, 53...
assemblies.
QUESTION #10:
The submittal (p. 5-3) states that the replace =ent cost of energy and capacity would be approximately $159 =illion per year.
Discuss whether reserves are such that replace-ment power for TMI-l would likely be available within the General Public Utilities Corporation System or from other utility syste=s af ter 1980.
If TMI-l were forced to shut down due to lack of storage space.for spent fuel, discuss the source and cost of replace =ent power of system reserves are not expected to be adequate without IMI-1.
If TMI-l were to be shut down, there still would be certain costs associated with the facility such as interest on investment, physical protecti;n, etc., apart from the costs associated with maintaining TMI-l in a " shutdown" condition.
RESPONSE
If TMI-l were forced to shut down in 1980, the Pennsylvania, New Jersey, Maryland (PJM)
Interconnection System reserves would be reduced by approximately 2 from values which range between 24 and 297. in the succeeding six (6) years. Though this will not precipitate PJM System breakdown, it will, however, lead to increased risk of conditions during which available capacity will be insufficient to meet load. PJM currently aims for a reliability index of one such occasion in ten (10) years and the planned reserves are generally able to meet this objective. A reduction of 2 percent will halve the reliability index to approximately one (1) occasion in five (5) years.
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RESPONSE
(Continued)
If TMI-l were shut down, CPU reserves in this period would be totally inadequate to provide reliable service without assistance from PJM. As presently planned, GPU reserves in the 1980's range from a low of 16.0 percent to a high of 27.5 percent with an average of about 22 percent.
Shutdown of TMI-l would reduce these values by approximately 12, at which level reliability would be so sericuly impaired that frequent involuntary load inters 2ptions would occur if GPU were obliged to depend solely upon its own resources. This evaluation assumes that no other nuclear units in the PJM are shut down for the sa=e or similar reasons.
If such shutdowns were to be considered, the state =ent of adequate PJM System reserves would have to be re-examined.
If TMI-l were to shut down in 1980, the cost per year to maintain TMI-l in this condition is approximately $75 million. This cost includes:
Physical Protection Requirements Routine Custodial Maintenance Necessary Decontamination Loss of Nuclear Fuel Investment Depreciation Federal and State Income Tax Franchise, Property, and Other Taxes Return on Investments QUESTION #11:
Discuss the number of spent fuel asse=blies that could be impacted in the proposed compact arrangement by the cask and associated lifting gear if the cask and lifting gear should tip and fall while in or near the spent fuel pool, RESPCNSE:
A cask drop analysis for Three Mile Island Nuclear Station Unit 1 was submitted to the Nuclear Regulatory Commission on February 14, 1976 by Metropolitan Edison Company letter CQL-0215.
In thzt analysis, specific consideration was given to the integrity of spent fuel asse=blies stored in the spent fuel pools "A & B) during handling of the cask.
It was de=onstrated that the cask transfer path will be limited so that the cask will be tipped in a direction away from the "B" spent fuel pool in the event of a cask drop.
Therefore, there are no assemblies that could be impacted by a dropped cask since the cask will not tip into the pool.
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