ML20148C003
| ML20148C003 | |
| Person / Time | |
|---|---|
| Site: | Zion File:ZionSolutions icon.png |
| Issue date: | 10/24/1978 |
| From: | Naughton W COMMONWEALTH EDISON CO. |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 7811010069 | |
| Download: ML20148C003 (69) | |
Text
. - - _ . . - . . - . - - -
o Camm:nw=lth Edis:n !
,nj V
One First National Plaza, Chicago. Ilknots Address Reply to: Post Office Box 767 f
Chicago lilinois 60690 October 24, 1978 ,
4 Director of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555 l
Subject:
Zion Station Units 1 and 2 Additional Information on Proposed
, Expansion of Spent Fuel Storage Capacity ;
NRC Docket Nos. 50-295 and 50-304 Reference (a) : September 26, 1978 letter from A. Schwencer j to Cordell Reed requesting additional I information on Zion Station's proposed spent fuel pool storage capacity expansion ,
1
Dear Sir:
Per Reference (a), the NRC Staff requested Commonwealth Edison Company to provide additional information in support of its request to expand the storage capacity of the Zion Unit 1 and 2 spent fuel pool. The attachment to this letter contains Commonwealth Edison's responses to the NRC Staff's request.
Please address any additional questions that you might have to this office.
one (1) signed original and thirty-nine (39) copies are provided for your use. ,
Very truly yours, William F. Naugh n Nuclear Licensing Administrator Pressurized Water Reactors attachment
( ccs' Susan N. Sekuler (Assistant Attorney General)
Richard J. ~ Coddard (NRC) ,
\
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g y . -e-- me- y r- , , -w --' '
. 4 Ccmmonwsalth Edisen NRC Docket Nos. 50-295/304 l
t I
I ATTACHMENT .
]
RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION ZION UNITS 1 AND 2 SPENT FUEL POOL STORAGE CAPACITY EXPANSION l
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NRC Docket Nos.
50-295 & 50-304 QUESTION NUMBER 1:
In view of the US DOE's announced intent to provide both interim and permanent spent fuel storage facilities and the probable availability of interim facili-ties in the 1983-85 time frame, what is your basis for selecting a storage capacity of 2112 assemblies to provide for storage of spent fuel to 19927 Include considerations of costs for interim storage and those for eventual permanent storage under the DOE's estimated cost breakdown.
1
RESPONSE
Based on the substantial uncertainty relating to the design, licensing and construction of government facilities, it seems prudent for a utility to plan for installed capacity beyond the announced date.
Further, once the government facility commences operation, there may be con--
straints on shipping to the facility in a timely manner, due to the backlog of spent fuel.
Finally, the design, licensing and construction of incremental additions to storage capacity involve costs which Comccawealth Edison can avoid by in-stalling the maximum feasible addition at this time.
01.1
NRC Docket Nos.
50-295 & 50-304 QUESTION NUMBER 2:
Have you investigated participation with other utilities tc construct an independent spent fuel storage facility? Is it likely that such a facility would be available prior to the time (approximately 1983) when an interim Federal facility is expected sn be available? Provide schedules and cost information for this option.
RESPONSE
We considered participation with other utilities to construct an independent spent fuel storage facility, but rejected it since Commonwealth Edison's needs are sufficient to justify the construction and operation of a facility which would service only the Edison system. l l
Further, a joint effort would seem to have no scheduling nor licensing advantages.
f 02.1
HRC Docket No3.
50-295 & 50-304 QUESTION NUMBER 3:
Have you corresponded with General Electric (Morris Operation) and NSF to confirm that no storage space is available for storage of spe 'uel from Zion Units 1 and 2?
RESPONSE
Both GE and NFS have confirmed that no storage space is available for spent I i
fuel from Zion.
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k 03.1
NRC Docket Nos.
50-295 & 50-304 QUESTION NUMBER 4:
Provide the additional cost to your customers for purchase of replacement power for each day that one and both Units of Zion Station do not operate.
Provide the basis for these figures. In addition, discuss the following:
RESPONSE
For each day that both Zion Units are inoperable, replacement power cost averages $400,000 (1978 dollars ) for the early 1980's.
If the Zion Units were inoperable during the early 1980's, not all of the replacement power could be obtained from the CECO system. Supplementary power will most likely be available on an emergency basis during this period.
The cost and availability of such replacement would, of course, vary widely with the season.
Costs, apart from the replacement power cost, associated with maintaining both the Units in a shutdown condition for a few years are estimated to be approxi-mately equal to the normal operating costs except for the fuel expense. This would be about 25 million dollars each year. Investment-related costs would total about 85 m!'. lion dollars per year for the Zion plant.
Q4.1
NRC Docket Nos.
50-295 & 50-304 QUESTION NUMBER 5:
Provide a breakdown of the cost of the modification.
)
RESPONSE: l The costs to complete the specified modification have been estimated as follows l
Engineering Costs $ 250,000 l 4 Fuel Storage Tubes 2,112,000 Fuel Rack Fabrication 2,000,000 Disposal of Existing Racks 400,000 Installation 350,000 l Total Plus 10% !
Contingency $5,623,000 l t
The above costs are estimates which include all project engineering, material and fabrication costs, disposal and installation costs. The disposal costs include crew labor to decontaminate the existing racks, handling costs, shipping crate construction, transportation and burial charges. As done in the 1976 modification, the existing racks will be decontaminated and shipped intact for burial. l l
Installation includes expenses for the fuel handling crew, divers and handling materials. Also included is the relocation of several fuel handling tools in the pool to storage locations which will not interfere with the new storage racks. The present value of the revenue requirements for this proj-ect will be approximately $9,900,000.
05.1
NRC Docket Nos. 50-295 and 50-304 Question Number 6:
Discuss the effect of energy conservation program within your service area as related to operation of Zion Units 1 and 2 and associated refueling schedules.
Response
Commonwealth Edison Company sponsors and participates in numerous energy conservation programs such as insulation l l
and improved thermal efficiency programs, heat pump testing I and certification program, commercial / industrial load management counseling, time-of-day rates for commercial /
industrial customers, residential time-of-dr.y rate experimental program, and advertising and public relations i programs which extensively utilize brochures and booklets on energy conservation. However, the resultant effect of these programs will be negligible on the operation of the Zion units and their associated refueling schedules. ;
This results primarily from the fact that Commonwealth Edison preferentially loads its system by incremental energy costs and the Zion units, along with Commonwealth Edison's other nuclear units, have the lowest incremental costs and are therefore usually base loaded. Thus, energy conservation in the Commonwealth Edison service
, area will more directly result in the less use of oil.
Q6.1
. NRC Docket Nos.
< 50-295 & 50-304 i
l QUESTION NUMBER 7:
2
" Discuss any plans to extend the time between refuelings" 1
RESPONSE
A
- Commonwealth Edison has performed a number of studies to investigate the feasibility of extending the time between refuelings at Zion Station. These studies have included
- 1) an 18 month cycle scheme, 2) Constant cycle length schemes between 12 and 18 months, and 3) a variable cycle length
- scheme. In each case the alternate cycle length could not be economically justified. Further, the number of assemblies discharged yearly into the spent fuel storage pool was higher i for the alternate schemes than for the 12 month cycle scheme employed at present.
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7 07.1 1
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NRC Docket Nos.
, 50-295 and 50-304 ]
Question Number 8:
. Discuss the commitment of material resources (e.g., stainless steel) required.to fabricate the replacement storage racks. I Will the existing racks be salvaged or used at another facility?
l Will the old racks be cut up for disposal, or disposed of "as-is"? )
Response
l A total of 602,766 pounds of stainless steel will be required to '
fabricate the replacement storage racks. ;
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We intend to dispose of the old racks "as-is". l l
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I 08.1
NRC Docket Nos.50-29f, and 50-304 QUESTION NUMBER 9:
Discuss the measures that you will take to ensure that the boral plates of the spent fuel racks will not deform during their expected service life due to the expansion of gas that may evolve from any water entrapped during the fabrication process or after placement into the spent fuel pool,
RESPONSE
The Boral neutron absorber plates are manufactured and assembled between the stainless steel tubes under the strictest quality con-trol procedures.
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09,1
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, NRC Docket )
Nos. 50-295 and 50-304 4 l
1 Question Number 10 Discuss the occupational exposure expected during this proposed SFP 1
modification. Address the expected dose rates, numbers of workers (including divers, if necessary) and occupancy times for each phase ,
1 of the operation. Include removal and disassembly (or crating) and disposal operations of the present spent fuel racks and installation of the new racks. Provide the estimated man-rem exposure. Identify and compare the measured and estimated man-rem exposure for your SFP modification approved in October 1976.
Response
The spent fuel rack modification is scheduled to be carried out over a period or thirty two working days. Twenty four old (i.e. present) racks will be removed at the rate of three per day and twenty four new racks will be installed at the rate of one per day.
A remote handling device will be employed to bolt and unbolt the racks. As a result divers will be needed, only to assist in aligning the new racks, The tank will require six workers working approximately six hours a day. The radiation exposure at the spent fuel pool water level should be 2 to 5 mrem /hr. Therefore, the average expected dose in 2.2 to 5.4 man-rem.
The health physics staff of Zion Station estimated that the average dose received by personnel during the 1976 rack modification was less than 3 man-rem. Specific dose records for the 1976 rack modification were not separately maintained. This was because the exposures were min imal , and recorded on the cumulative exposure histories.
The old racks shall be decontaminated and they shall be shipped intact in wooden crates to Barnwell, S.C. for burial.
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NRC Docket Nos.
50-295 & 50-304 l
QUESTION NUMBER 11:
Provide an estimate of the annual ran-rem expected from all operations in the 1
spent fuel pool area as presently configured and based on present concentra-tions of fission and corrosion products (e.g., 58 o, C 60Co) in the SFP water.
Discuss present levels of corrosion product (i.e., crud) build-up along the walls of the pool and the dose rates from this build-up. Describe the impact of the proposed modification (e.g., additional fuel elements and crud) en I these estimates, including the effect of more frequent changing of the de-mineralizer resin and filter cartridge.
RESPONSE
Under current operations, seven fuel handlers receive about 4 man-rem per year from all operations in the spent fuel pool area. The major part of this j exposure is caused by activation products (e.g., 58Co, 60Co) dispersed into the water during refueling. The dose is therefore relatively independent of the amount of spent fuel in the pool. l l
l Relative to corrosion products (i.e., crud) build-up along the walls of the l SFP, the depth of the pool water is more than adequate to provide shielding to eliminate any dose rate concern.
The proposed modification will have no impact on fuel handlers exposure nor corrosion product build-up since these factors are mainly dependent en the l l
frequency and length of the refueling operations when the SFP is directly l connected to the reactor cavity and activation products are dispersed into 1
the water.
1 Ol) 1
NRC Locket Nos.
50-295 & 50-304 QUESTION NUMBER 12:
Describe the method that will be used to dispose of the present racks (i.e.,
crating intact racks or cutting and drumming). If the racks are to be cut and drummed, show that the exposure received by this disposal method, as com-a pared to crating the intact racks for disposal, will provide as low as is reasonably achievable ( ALARA) exposure to personnel. Your response should include a description of the disposal method used and resultant exposures from the 1976-1977 SFP modification.
RESPONSE
The last set of spent fuel racks were shipped intact in wooden crates lined with plastic to Barnwell, S.C. for burial. No cutting or drumming was per-formed. Crating in one piece keeps exposure to personnel as low as reasonably achievable ( ALARA). The present racks will be disposed of in a similar way.
Fuel handling and radiation protection personnel cannot associate any appreciable dose rate with the decontamination and shipping of the spent fuel racks in 1976.
012.1
NRC Docket Nos.
. 50-295 & 50-304 QUESTION NUMBER 13:
Discuss in detail the following:
- a. - the leakage of water from the pool and the pool leak collection system.
- b. - the adequacy of the spent fuel pool purification system (SFPPS) to maintain low concentrations of radioactivity in the pool after the modification with no changes to the SFPPS, and
- c. - the ability of the SFPPS to otherwise clean up the pool water.
Q13.1 l.
13 A.) The spent fuel pool has six drain lines which will direct any water leaking between the fuel pool liner and support structure to the radwaste system. Leakage may be directly observed at six sight glasses located in the 542' pipe tunnel in the Auxiliary Building. Presently, no signif-icant leakage has been observed, as the glasses show only a slow drip rate. No increase in pool leakage is anticipated as a result of the rack modification. In addition, the pool water level may be directly observed by the roving Auxiliary Building Operator, in addition to the level alarms provided in the control room.
B + C) The spent fuel pool clean-up system, as described in the i 1
FSAR, consists primarily of two demineralizers, two filters and a skimming pump and filter. In addition to )
maintaining water quality in the fuel pool, the system I is designed to also clean water used to fill the reactor cavities of both units.
The fuel stored in the pool contributes very little to the level of suspended solids or " crud" in the pool.
Almost all of the crud in the spent fuel pool results from the solids released from the reactor vessel during refueling operations, not from the spent fuel assemblies in the pool. Experience has shown the cleanup system to be adequate. The demineralizer resins and filters are l l
changed as needed. To date,for a total of five refuelings, l
013.2 I
t each demineralizer bed has-been changed once, resulting
- in a total of sixty cubic _ feet of spent resin being 2
discharged to radwaste. Filters are changed infrequently also, so that the impact on the total radioactive waste discharge of Zion Station is minimal.
Leaking fuel has not been detected at Zion Station. Fuel performance has been excellent, and:there is no reason to expect leaking fuel to be a problem in the future.
Should fuel leakage occur, migration of the fuel into the water is limited. Equilibrium pressure is established between the leaking fuel rod and the hydraulic pressure, thus there is no' driving force to extract the solid fuel from the fuel assemblies.
Thus increasing the quantity of stored fuel will not adversify impact the clean-up system, nor the radwaste system.
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Q13.3 4
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- NRC Docket Nos.
50-295 & 50-304 QUESTION 1RTMBER 14:
Discuss the relationships between and the effects of the fuel assembly move-ments during your 1976-1977 SFP modification on the amount of crud in the pool water and on the radiation levels in the vicinity of the pool. Your response should include measured radioactivity concentrations.
RESPONSE
There was no noticeable change in the radiation levels around the spent fuel pool area during the 1976-1977 modifications. The average dose rate a few feet above the spent fuel pool during the period of the modification (July 14 to November 9,1976) was approximately 1 mR/ hour.
The effect of the 1976 SFP modification on the amount of crud in the pool water was negligible, as most crud is deposited from the opening of the re-actor vessel during refuelings. The operation of the clean-up system removes a substantial portion of any crud making the assessment of any noticeable effect difficult. Demineralizer resins in the clean-up system have required only one change, and filter changes are also infrequent.
Results of s1mpling of I-131 concentration in the pool water for the modifi-cation period appear below:
Time Period (1976) Av. I-131' Specific Concentration, uCi/ml June 3.84 X 10-12 July 2.37 X 10-12 August 1.03 X 10-11 September (fuel movements) 5.34 X 10-12 October 5.35 X 10-12 Novedber 1.09 X 10-12 December 2.37 X 10-12 It is evident that the modification under discussion caused no increase in the I-131 specific concentration of the SFP water.
Q14.1
NRC Docket Nos. 50-295 and 50-304
. Question Number 15:
Supplement your April 26, 1978 submittal by addressing the impact of the proposed SFP modification on the following.
- a. Radioactive gaseous effluents from the pool
- b. Radioactive liquid effluents from the plant, including leakage of the water from the pool
- c. Radioactive solid wastes from the plant, including the frequency of replacing the SFP demineralizer resin and filters.
Response
- a. The amount of radioactivity in the spent fuel storage racks will differ only slightly from that in the current design. One year after shut-down the amount of radioactivity in spent fuel is less than 1% of its ,
l value at shutdown due to radioactive decay. After cooling for two I
more years, the radioactivity is reduced to about 0.3% of its initial value. Ten years after shutdown, it decreases further to slightly over 0.1%. Thus, the great majority (say> 85%) of the radioactivity will come from fuel just discharged. Therefore, only a very slight change in the amount of radioactivity, and the corresponding dose rates at the pool: surface, will be a result of expanding the spent fuel pool storage capacity. Since gaseous effluents like Kr-85 and Xe-133 con-tribute only a fraction of the total radioactivity, the impact of the modification on radioactive gaseous effluents will be negligible.
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- b. As stated above, most radioactivity comes from recently discharged fuel.l Thus, liquid effluents from the plant could contain only negligibly more radioactivity as a result of the spent fuel pool modification.
Nothing in the proposed SFP. increases the probability of leakage of I water from the pool as described in the FSAR.
Q15.1
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C. The amount of additional solid radioactive waste generated by the proposed modification will be insignificant. Evidence indicates that, the frequency of filter and demineralizer resin bed change outs are dependent upon activation products in the SFP water. These are present during the refueling operation when the reactor cavity is connected with the SFP. Therefore, the solid radioactive waste generated is mainly dependent upon the number of refueling operations and not on the number of spent fuel assemblies in the SFP 1
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l 015.2
- ' NRC Docket Nos.
50-295 and 50-304 QUESTION NUMBER 16:,
Pruvide the estimated volume of contaminated material (e.g., spent fuel racks, seismic restraints) expected to be removed from the spent fuel pools during the modffication and shfpped from.th_e. plant to a licensed burial site.
RESPONSE
1 The esi:imated volume of contaminated material in the existing spent fuel racks and auxiliary equipment is as follows:
3 Volume of Racks in present configuration = 17,166 ft Volume of Steel in racks = 586 ft3 Q16.1
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NRC Docket Nos.
50-295 & 50-304 ,
QUESTION NUMBER 17:
Provide a list of typical loads representing the range and type of loads that you would intend to carry near or over the spent fuel pool. Provide the weight and dimensions of each load. Discuss the load transfer path, including whether the load must be carried over the pool, the maximum height at which it could be carried and the expected height during transfer. Provide a descrip-tion of any written procedures instructing crane operators about loads to be carried near the pool. Provide, the number of spent fuel assemblies that could be damaged by dropping and/or tipping each typical load carried over the pool.
RESPONSE
The fuel building crane is rated at 125 tons, and it is equipped with inter-locks which prevent it from going over the spent fuel pool area. The inter-locks can be defeated only with keys held by the fuel handling foreman and the shift engineer. The transfer path is from the area to the east of the pool, along the north wall, and over to the western area of the spent fuel pool (new fuel vault and unloading area).
The most typical loads moved near the pool and their approximate weights and dimensions appear below:
Load ' Weight Dimensions Single Fuel Assembly 1600 lb 8".X 8" X 13' Westinghouse Fuel Container (full) 6700 lb 3' i f A 14' Movable Shielding Blocks 20 tons 6' X 6' X 15' Waste Drums 500 lb 2' X 2' X 4'
, Q17.1
NRC Docket Nos.
50-295 & 50-304 QUESTION NEMBER 18:
Discuss the instrumentation to indicate the SFP water level and water temperature. Include the capability of the instrumentation to alarm and the location of the alarms, i
l RESPONSE: l l
l The spent fuel pool water level and temperature are continuously 1 monitored. Alarm annunciators are provided in the control room.
The parameters and their respective setpoints are shown below: i Spent Fuel Pool Parameter Setpoint Normal High water level 615' 7" 615' 3" Low water level 614' 4" High water temperature 1250F 60 F In addition local indications of the water level and temperature are provided by the pool side, their values are monitored shiftly by the operators.
Q18.1
NRC Docket Nos.
50-295 & 50-304 QUESTION NUMBER 19:
Propose a Technical Specification which prohibits carrying loads greater than the weight of a fuel assembly over spent fuel in the storage pool; or justify why such a specification is not needed to limit the potential consequences of accidents involvitig dropping heavy loads, other than casks, onto spent fuel to conseqt.ences already evaluated for the design basis fuel handling accident.
RESPONSE
Heavy loads and cask drop analyses have been submitted to the NRC Staff via the following letters:
G.J. P11m1 to Mr. Robert A. Purple, Chief Operating Reactors -
Branch 1, dated April 8, 1976.
R.L. Bolger to Mr. Albert Schwencer, Chief Operating Reactors -
Branch 1, dated September 14, 1976.
D.E. O'Brien to Nb. Albert Schwencer, dated August 9, 1977 D.E. O'Brien to Ux. Albert Schwencer, dated March 3, 1978.
W.F. Naughton to Director of Nuclear Reactor Regulation dated July 13, 1978 The last of these letters was in response to an NRC Staff request for additional information on control of heavy loads near spent fuel. In this response, Commonwealth Edison indicated that no heavy loads are required to be moved over the fuel pool, with the exception of the spent fuel cack. Heavy load and cask drop accidents have been analyzed per the Ol9.1
first four references above, In addition, administrative control precludes the movement of heavy loads over the spent fuel pool during refueling; therefore, no procedures exist for moving heavy loads.
Any required movements of heavy loads will be evaluated on a case by case basis'and a special procedure for the move will then be written and be subject to the onsite review process.
Q19.2
NRC Docket Nos.
50-295 & 50-304 QUESTION NUMBER 20:
"To preclude any unreviewed increase, or increased uncertainty, in the calculated value of the neutron multiplication factor, which could raise the actual keff in the fuel pool above 0.95 without being detected, a limit on the fuel loading is required.
This limit should be imposed in the Technical Specifications either by specifying the numbers of grams or uranium-235 per axial centimeter of the fuel assemblies that were used for the calculations in the Licensing Report forwarded as Attachment 1 to your April 13, 1978 submittal and then making both this Licensing Report and the NRC's Safety Evaluation of this report the basis for the keff in the Technical Specifications, or by directly limiting future fuel loadings in assemblies that are placed in these high density racks to the maximum number of grams of uranium-235 per axial centimeter of assembly that was used in these calculations. Propose a Technical Specification using one of these alternatives."
RESPONSE
The proposed limit of a maximum fuel enrichment of 3.1 w/o U-235' in our April 13, 1978 submittal serves to limit the future fuel loadings that are placed in the high density racks.
Q20.1
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. I This enrichment limit is equivalent to a loading of 39.4 grams i l
of U-235 per axial centimeter of the fuel assembly lengths based on the analysis input assumptions contained in Table'3.3-4 of the Nuclear Services Corporation Licensing
~ Report. (The typographical error in the table should be
! corrected so that the pellet density reads "94.95"). l 1
l Incorporated in the use of the 3.1 w/o limit is the 0.01 1
w/o enrichment uncertainty and the conservatively high j i
- pellet density of 94.95 percent of Theoretical.
l The enrichment limit is utilized in lieu of a gram per l
centimeter limit to allow operators to more easily relate to the limit.
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Q20.2 i
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1:RC Docket Nos.
50-295 and 50-304 ZION REPLY TO NRC QUESTION NO. 21 Part a:
In order to verify the accuracy of the CHEETAH-XSDRN-CITATION calculations when applied to poisoned spent fuel rack neutronic analysis, two benchmark calcula-tions have been performed. In the first benchmark calculation, experimental results from the Plutonium Recycle Critical Facility were used (reference 1).
In the second benchmark calculation, critical experiments performed at Oak Ridge National Laboratory were used (reference 2).
For the first benchmark calculation, the fuel design values used are summarized in Table 1. Figure 1 shows the boral blade design used in the experiment and Figure 2 shows the experimental fuel assembly configuration. The calculational procedure is as follows:
- 1) Fuel rod cell ' calculation for fuel region using CHEETAH.
- 2) Fuel rod cell calculation,123 groups collapsed to 27 groups 4
using XSDRN.
- 3) 1D supercell calculation, 27 groups collapsed to 4 groups using XSDRN. Dimensions of the supercell are shown in Figure 3.
- 4) 2D CITATION criticality calculation, cross section sets for non-fuel region obtained from XSDRN for fuel region obtained from CHEETAH. The assembly configuration used in the calcu-lation is shown in Figure 4.
For the second benchmark calculation, the fuel design values used are summar-ized in Table 2. Figure 5 shows the experimental fuel assembly configuration.
The calculational procedure is as follows:
E O21.1 l
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- 1) Fuel- rod cell calculation for fuel region using CHEETAH.
- 2) . Fuel- rod cell calculation,123 groups collapsed to 27 groups using XSDRN.
i 3) '1D supercell calculation, 27 groups collapsed to 4 groups
, using XSDRN. Dimensions of the supercell are shown in j Figure 6.
4
- 4) 3D CITATION criticality calculation, cross section sets for j non-fuel region obtained from XSDRN, for fuel regien obtained i from CHEETAH. The assembly configuration used in the calcu-
- lation is shown_in Figure 7. ,
l The results of the two benchmark calculations are shown in Table 3. l i
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021.2 i
TABLE 1 Basic fuel pin assembly array 9x9 Fuel enrichment 2.35 0.03%
Fuel rod pitch 0.75" Fuel density 9.2 gm/ce, 84% T.D.
Pellet 0.D. 0.44" Clad 0.D. 0.50" Clad thickness 0.03" Clad material Aluminum Active fuel length 36" Buckling 0.00089 cm2 Boral blade core density 2.484 0.002 gm/cc
- /o of BgC ,
34.9 0.2
- /o of A1 65.1 0.2 l
021.3 l
l
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TABLE 2 Total. number of fuel rods 324 Fuel enrichment 4.95 w/o U-235 Fuel rod pitch- 2.05 cm Fuel rod diameter 0.762 cm Active fuel length 30 cm ,
Buckling 0.006 cm 2 Fuel material Unclad Uranium nietal Fuel rod mass 264 gm Fuel rod volume 13.681 c.c.
Fuel rod density 19.297 gm/c.c. 1 Depleted Uranium block:
Mass 640 kg Length 60.6,425 cm Width 21.9075 cm Height 25.87625 cm Composition 99.795 w/o U-238 0.185 w/o U-235 0.02 w/o trace elements 021.4
TABLE 3 CHEETAH.-XSDRN-CASE EXPERIMENTAL CITATION KEN 0 REFERENCE i
Single boral blade 1.0018 1.0174 - 1 in array of 9 x 9 2.35 w/o UO 2 fuel -
Bare U rods with 1.000 1.01321 1.01748 2 depleted U block i and boral sheet, j water reflected 1
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W 4
021.5
4 i
r l
i
, u e
- s O.041" -
AL WALL y
f j ' B4C 0.168"- 5.25" a
0.041" -
- AL WALL i
-U i
1 l- .
i = 5,25 " . =
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1 i
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e Figure 1. Boral Blade Design for First Benchmark Calculations .
' 021.6
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, 00000000000000000 000@O00 00000008000000000< )OO@OOOO 2
00000000000000000< >O000000 000000000000000000060000000 eeessessessecee9E d49909000 OOOOOOOOe@@@@@888@c >O9980000 000000000@O00000@O( 90000000 000000000@O0000@O0<@O000000 000000000@O000@O00e@O000000 000000000@O00@O0000@O000000 000000009@O0@O0000000000000 00000000000@O00000000000000
. 0000000093@O000000000000000 000000000@O0000000000000000 seseessesseseesecessessesse O000000@8000000000000000000 00000@O8000000000e0000000 0000@O000000000008000'0000 000@O00000000000000000000 0000000000000000000000000 0000000000000000000000000 0000000800000000000000000 0000000000000
~. W _
BASIC 9 X 9 WITH BORAL BLADE; UO -2.35= 235g 2
EXPERIMENTAL POWER DISTRIBUTION:
POSITION NO. RELATIVE POWER POS'ITION NO. RELATIVE POW'ER
! :M !* 1:58 7 1.080 30 1.144 8 1.720 31 1.527 9 1.989 32 1.698 10 1.414 33 1.471 11 1.287 34 1.377 12 1.226 35 1.236 13 1.141 36 1.146 14 1.049 37 .891 15 .886 38 .637 16 .666 39 .478 17 .346 40 .997 18 .200 41 .788 19 .295 42 .51 5 20 .31 6 43 .369 21 .287 44 .260 95 1.716 45 .31 5 26 1 .61 4 46 .370 27 1.633 47 .460 Figure 2.
Q21.7
g Al H0 BORAL R-4 R
3 R =
FUEL 3
R) = 9.67303 CM R = 10.03129 CM 2 !
R3 = 10.09005 m R = 10.30341 CM 4
Figure 3. 1-0 Supercell Configuration for First Benchmark Calculations 021.8
e - _ a ------
h
- 1.905
-1.905 -1.905 -1.905 -1.905
_ 25 _.
11.43 _ , ._ 10.795 6.35 , ,_4.445 , 6.958 ,
,_25_,
'll' ' llI ~l lI i ~l l 'l WATER BORAL BLADE l
l 4
1 FUEL l
}
(~' .
l 1 1 k
-meme - l L i L J ALL DIMENSIONS IN CENTIMETERS Figure 4.
021.9
0000000000000000.00' '
000000000000000000 0000Q0000000000000
. 000000000000000000 000000000000000000 000000000000000000 000000000000000000 000000000000000000' OO,0000QOOOOOOOOOOO 000000000000000000 000000,000000000000 BORAL SHEET .
000000000000000000 (0.25INTHICK) l 0000000,000000'00000 000000000000000000 0000000000'00000000 .
0000000000'00000000 )
O00000000000000000 0000000000000000,00 l
FUEL PINS AND DEPLETED URANIUM: SEE TABLE 2 BORAL SHEET : 0.041 IN. CLAD OF 1100 Al 3
bR ENiT = 0.0949 LB/IN l CORE IS 38.9 W/0 B C, REMAINING IS 1100 Al 4
LENGTH = 18.549 IN.
HEIGHT = 10.1875 IN. (CENTERED)
)
REFLECTOR: 16.6 CM OF WATEP. AB0VE CORE, 15.24 CM IN ALL I OTHER DIRECTIONS l l
Figure 5. ORNL Critical, Run 105, With 324 Rods, Depleted Uranium and Boral Sheet 021.10
l
-BORAL AL , -AL HO HO HO 2 , ,2 2 HO FUEL FUEL DEPLETED HO 2 ll " 2 l l lI II I I O 15.24 29.59a n 31.64 54.19 54.817 76.7245 91.9645 l 30.2975 - - 30.9325 ALL DIMENSIONS IN CENTIMETERS D
e 1
Figure 6 1-0 SUPERCELL CONFIGURATION FOR SECOND BENCHMARK CALCULATIONS 021.11
-6.76402 CM 6.76402 CM- j 5.10723 CM 5.10723 CM-
_15.24 CM_
36.9 CM' -
, _15.24 CM_-
JI l 4
15.24 CM i
if i J
o 1
l DEPLETED URANIUM 21.9075 CM ;
1 lI
_ -0.627 CM n
WATER FUEL 22.55 CM i
BORAL SHEET i 0.429 CM 11 I f I f f f f / f I f f J f f f I f f f f f f f f f f f J f I / / 1 / / A 0.104 0.706 FUEL 14.35 CM u
I 15.24 CM y
Y h
1
~X 0
Figure 7.
021.12 l
Part b:
- 1) The definition of the " grey" boundary condition is the same as the " black" boundary condition except that the boundary constant (D/A) has a value smaller than transport-theory predicted value of 0.4692 for the totally black zone.
- 2) Use of the " grey" boundary constant is necessary in diffusion theory be-cause Fick's law on which the diffusion theory is based, is not valid in a strong absorbing medium.
- 3) The formula used in our calculation is from ANL-5800(3)
Constant =
0 ,
1 1-2E3 (Ia t)-
A 2 1+2Eg (I,t)_
where 1
I a= macroscopic neutron absorption cross section for the " grey" l zone (cm-l),
t = thickness of the " grey zone (cm).
E3 and Eg are exponential integrals of the 3rd and 4th orders, respectively.
This formula is an approximation. As I,t - -
D 3
--0.500 which is greater than the theoretical limit ~of 0.4592.
0 Whenever this formula predicts a value of greater than 0.4692, we replace it by 0.4692.
- 4) In finite-difference diffusion code CITATION I4) , the internal black (or !
" grey") boundary condition is used just the same way as an extrapolated boundary condition in that the flux shape within the finite-difference l
element is extended into the " grey" zone. For a more detailed descrip-tion, see Sections 701 and 702 in CITATION manual (Appendix A).
021.13
l
. l
. i i
Part ci
- 1) Our calculation corresponds to a boron-ten areal density of 0.0197 gm/cm2 , l l
- 2) The k,gf vs. boron-ten areal density is given in the following j i
table (for fuel enrichment of 3.2 W/o) '
1 l
B 10 Density (gm/cm ) keff j
0.010 0.94804 l 1
0.0197 0.92629 -
0.020 0.92585 0.030 0.91314 J
0.040 0.90443 and also plotted in Fig. 8 Part d:
- 1) The estimated maximum value of 0.7% for the transport theory correction is based on the worst case value determined in the diffusion to transport theory comparison documented in Reference 5 (extract attached as Appendix B) .
- 2) We would like to emphasize that our criticality experiment benchmark shows that our method of calculation is conservative (calculational bias = 0.0048) and includes the effect of using diffusion theory inatead of transport theory. We did not take credit for this negative factor.
021.14
,,_m - .,- y -
i
, 0.95 ,
)
t 0.94 0.93
\
- K,ff 0.92 1
i 0.91 0.90 0.89 0.00 0.01 0.02 0.03 0.04 B-10 AREAL DENSITY (JM/CM )
Figure 8. Effect of Boron Ten Density of K,ff 021.15
REFERENCES
- l. " Experimental and Calculated Results for UO and 2
UO2 -PuO 2 Fueled H 0-moderated Loadings", by L. D.
2 Williams and J. B. Edgar, BNWL-1379, August 1970.
- 2. " criticality calculations on Maine Yankee Spent Fuel )
Racks Containing Boron", by D. J. Denver, B. F. Momsen, I and E. E. Pilat, YAEC-1090, November, 1975.
- 3. " Reactor Physics Constant," ANL-5800, 2nd Edition, USAEC, July 1963.
1
- 4. T. B. Fowler, D. R. Vondy, and G. W. Cunningham, " Nuclear Reactor Core Analysis Code: CITATION:. ORNG-TM-2496, Revision 2, July 1971.
- 5. Wisconsin Electric Power Company, Publication March 18, 1978, Docket Nos. 50-266 and 50-301.
l 1
i Q21.16
t 4
- . - = - .
APPENDIX A CITATION MANUAL SECTIONS 701 AND 702 l
l 021.17
) Section 701: The Discrete-Energy Orcup Diffusion Equations Described here are the equations which are used to obtain a finite .
difference . type of neutron balance over discrete elements of volume. A l basic equation expressing the diffusion approximation to neutron trans-I port at some location r and energy E is
~
r, E r,E * ( a,r,E + s,r,E r, E X (vI)
=
(I r,E'~E #
k r,E,dE ,
(701-1)
E' The continuous energy spectrum is divided into discrete energy groups, i
a buckling term is allowed when appropriate, the source distribution function, x, is assumed to have no spatial dependence, and a simplification is made in the transport tem,
)
-D r, g V8 4 r, g + (I a,r,g +5IL s,r,g-n +D r, g B,* g ) 4 r,g n
X (VI) l
( s , r, n- g + k )d r, n .
(701-2) n l
Where 7* = The Laplacian geometric operator, 2 2 2
{ + + in slab geometry, cm',
, Dx* By* B z*
- '6 rg =
The neutron flux at location r and.in energy
} group g, n/sec-em*,
I =
a , r, g The macroscopic cross section for absorption, normally weighted over a representative flux enerEy spectrum, em' ,
021.18 4
I l
L 1
- j. *
) I s,r,g~n
- The' macroscopic crous section for scattering of neutrons from energy group c to energy Group n a (a set of these makes up a scattering kernel),cm' ,
- D = The diffusion coefficient, normally one-third of the rg recriprocal of the transport cross section, cm, i B' = The buckling term to account for the effect of the 2g Laplacian operator (leaka6e) in a dimension not
~
treated explicity, em *,
VI f g
= The macroccopic production cross.section (v is the number of neutrons produced by a fission and I f is the cross section for fission), em' ,
j X g
= The distribution function for source neutrons I (normally X = 1.0),
j g g ..
j k, = The effective multiplication factor, ratio of rate-I of production of neutrons to rate of loss of neutrons l from all causes, an unknown to be detennined. i fg When criticality is to be effected by altering nuclide concentraticns, [
i} normally by a direct iterative process, the k, is fixed at some specified 1
! value (near unity) and contributions to the cross sections frem the stored
} nuclide concentrations must be included in the, equations. Eq. 701-2 is ;
then altered to the form '
-@D r, g 8
4 r, g + (I a,r,g. +
, r, g
+ I s , r, g--n
+D r, g 3Ag) 6 r,g n -l 1
X -
4
=
I s,r,n~g + k, 3 (vI f, r, n + hvI p, r, n)_4 r, n , (701-3) l'
- where h is the . relative nuclide density, the eigenvalue of the problem, and I rg and v! are absorption and production cross sections as-p,r,n 4
acciated with the search nuclides'and a unit value of h.' Contributions from the search nuclides to the scattering kernel and to the diffusien aThe in-group term rg is excluded from the calculation.
4
.Q21.19
l
} coefficient have not been shown here, but these may be included. For '
the usual 1/v cearch, E b, r,g values are r cipr c 1 velocity rer t.he material and temperature of interest, so that h is w in the approximation j that in the acymptotic sense, the flux is given by d r, g *
- l Fixed Source Perhaps the major application of fixed source calculations has been in playing computation games. However, the fixed source probles arises in certain situations as reactor start-up with an inserted source, ,
or analysis of an experiment located outside of the reactor core with a fine-mesh description and coupling to the core with a leakage cource.
The fixed source problem is simply Eq. 701-2 with he set to unity and .
the cource added to the right hand side. The equation which applies is l-i ,
k -D v* 6 r, g +- (Ia,r g '
s,r,g~n +D r, g ~B"az) r,6 g I -
T6 s
o n >
l I
( 3,r, n g * *g("A)f,r,n Dr, n 0 r, g , (701h) n where Sr ,g is the fixed source. As is frequently useful, the fixed source may be separable in space and energy, in which case i r, g =X g r S #
j and data transfer from disk is then minimized.
l 1
- 1 l
\
END OF SECTION Q21.20
I l Ucetion 702: Finite Difference Representation of the Laplacian Operator The Laplacinn operator is to be reprm:ented in a finite-difference frem.
First, the finite-difference mesh will be examined. Consider a traverse in i space direction r. A region is traversed between r1 and r2 boundaries cr mot.erial interfaces. Input data specifies the number of mesh points to be located between ry and r and the spacing a = r ~# r ss the region. ;
l 2 2 1 The equations will be treated as if the problem were three-dimensional; l
]
for fewer dianensions, the extent in the undefined dimensions is considered infinite and the associated undefined dimensions cancel out of the final equations.
- The layout of mesh points is described in Table 702-1. The seven- l
- point diffusion approximation to neutron transport is formulated belev.
Figure 702-1 presents a three-dimensional sk' etch showing the flux location at mesh point (1,j,m) and the surrounding six flux locations in slab
. gecmetry (x-y-z). The finite-difference volume about mesh point (1,j,m) I is (x t-xg ) (y -yg ) (Z,-Zg) where these are locations of the surfaces of the finite-difference element. (In hexagonal geometry, the fomulation involves nine mesh points, seven in the hexagonal plane, see Figure 201-3).
Neutron leakage from (1,j,m) to (1,j,m-1), L(zm-1), through the front face of arc'a (xg -x3 ,3) (y -y ) is approximated as follows. Le t Cy a the unknown flux at the d.nterface. Leakage out is given by approximating the Clux slope at the surface by the average within the element (between the
! central point and the surface), '
j (Y3-Y3 1) ( x i-Xt_1)
~
b(*m- 1} i., j , m 1,j,m 1 3-2 (3 3,1) 2.
I b*here D 1 , is the diffusion constant at (1,j,m). Similiarly fer inward Icakage frem ths adjacent finite element, I
i _ _
(Y j yj _1) bt - XI-1}
~
m-1) = hij,m-1ji,j,m-1 ~
1_ ~*
( _"m-1 m-2) _
t 021.21
[ , .- ..... ,
.._ ._ ~.
Table 709-3 ,
Layout of Meshpoint:
System S3nb Cylinder Cylinder Sphere Hexagonal Triagonal Picture X y G g R X y X y _
Z Z 3 R Z ' E 60* 6c-
~
Geometry X-Y-Z 0-R-Z R-Z R li-Z T-E Specified regicn dimensions A A 'A A r x E A A X 0 z y x A Ag A A z z r Specified internal f O mesh points- J,,I ,M J ,I ,M 0 r' z r #x ' y' z x' x'"z lo 3
Volume of region AAA (r -r ) na (r2 -r ) (r-rf) 2 A,A,A A,A,A Volume about each mesh point = Volume of region + number of intemal points Mesh' point locations are at finite-difference centroids b'
W
ORNL-0WG 69-5569
. Y O
Z )
- i ,j +1,m
- i ,j ,m /
- 1-1,3,!m j *i+1.j ,m gf - - X \
FINITE-DIFFERENCE ELEMENT Y
j-1 / T Z m-l HAVING UNIFORM NUCLEAR PROPERTIES
- i,j,m-1 Xg_3 X
"*i,j-1,m V
FACE Z m-1 Figure 9. The Seven-Point Finite Difference Mesh Q21.23 l
Eliminating Cyfrom the equations gives 1
4 _ _
i 2(y3 -y,,1)_(x 1 - x 1,1)
- ~# 1,j,m-1 m-1}" (702-1) z, - zm-1 , *m-1 ~* m-2
- 1,J,m D
1,j,m . i,j,m-1 _
Gince the -term which multiplies the flux difference is simply sane constant, Eq. 702-1 reduces to the form 1
L(*m-1) = C 1,j,m,m _4i ,j,m - 4 1,j,m-1_ . '(702-2) I It may.be noted that within a region having uniform nucles. properties and uniform mesh spacing,
~X Cg =
(Yj '- Y j_1) (X i i_1)D 1,j,m
, internal, slab geometry.
(2 e ~Z m-1} l
\.
3 The leakage frcm the whole element is given by e
L(sm) + M sm-1 *
- i}~ *i-1) ' b(YJ)
- L(Yj.1) i,J,m _ 1,j,m,m+1 ^ 1,j,m.m-1 -C 1,J,m,1-1 + Cg,J,,,g,y .
i
+C 1, j , m, J + 1 + C1, j , m, j -1_J -C 1,j,m,m+1 4 1,j,m+1 s
-C 6 -C 4g g,3,,,,,1 g,g,,,1 1,j,m,1-1 i+1,j,m - C1,3,,,g,1 ,1,3,,
~
1, j , m, j + 1 1,j+1,m -C 1,j,m,j-1 1,j-1,m (7 02 -3)
For a zero gradient boundary condition, the associated C con-g,3,,,,,1 stant is cet to :ero. For an extrapolated boundary eendition, external or internal black boundary, the flux slope within the finite-difference element is extended., The boundary condition to be satisfied at the element surface is t
t 021.24 e
r em a
1 l , = C, . (702 h)
Where C, is a specified constant.
Let 43be the internal flux, 4, be thi. boundary flux and h be the 4
distance to the boundary from the internal point.
A linear approximation of the flux within the element gives
~
B4 i s g= , or b .,
B4 'D
-[Ds g , = d (4 - 4,) - C, 1
s
~
l Representing the normal area by An, the boundary leakage from one face of an element volume is given by L
s,n =-D t A. n b" a ,
(7024)
C D s t which gives the required constant for Eq. 702-3 of course the external leakage is considered lost from the system, but leakage into an internal black absorber is accounted for as an absorpticn in the region.
For curvilinear geometries, the surface areas of the finite-difference 1
element faces must be used which lead to somewhat more involved equations than above. For exac:ple, in r-z-0 geometry the surfaces are indicated in Figure 702-2.
The repeating boundary condition causes flux values at opposite ends of a row to be coupled so that the row is a closed loop. For 90* rotational symmetry, coupling is from the right-hand edge colw:n to the bottom edge rew.
For 180* rotational syrreetry, the right-hand edge column couples wi,th itself inverted.
l Q21.25 L' .. . - - - -
i 1
P ORNL OWG 69-5590 l
l 82 1/2(e2-ei)(r22 _ 7,2) !
I l
r2(e2 - el)(zz - z1)
I #
f
/ / s A
,/ }
z z / "2
/ (rg - r )(ziz ~2)1 A
I l
l '
rj(e2 ~'1)(12 ~2)1 i
NOTE THAT IN R-Z GEOMETRY, WITH SYMMETRY IN e, 0 2 ~ '1 = 2r; e IS IN RADIANS.
Figure 10. Surfaces on a Finite-Difference Element in R-Z-e Geometry.
,021.26 l
9 8
l I
1 l
l l
APPENDIX B i 1
WISCONSIN ELECTRIC POWER COMPANY PUBLICATION MARCH 28, 1975 DOCKET NUMBERS 50-266 AND 50-301 l
l l
l l 02.1.27 l
l
,/[bg
'gb l L Wisconsin Electnc wecomw 231 WEST MICHICAH, MILWAUKEE. WISCONSIN 53201 March ' 28, 197 5 .
9 Mr. Benard C. Rus'che, Director Office of Nuclear Reactor Regulation U. S. NUCLEAP. REGULATORY CCFMISSION Washington, D. C.- 20555
Dear Mr.'Rusche:
DOCKET NOS. 50-266 AND 50-301 SPENT FUEL STORAGE RACK REPLACEMENT POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2
- In accordance with Section'50.59 of 10 CFR 50, Wisconsin Electric Power Company and Wisconsin Michigan Power Company (Licensees in the above-named dockets) request an amendment to Facility Operating Licenses DPR-24 and DPR-27 to incorporate the attached changes to the Point Beach Nuclear Plant Unit Nos. 1 and 2 Technical Specifications and Final Facility Description and Safety Analysis Report (FFDSAR). These changes will result from Licensee's modification to the spent fue. pool storage racks ,
which will increase the storage capability from the original 208 to 399 irradiated fuel assemblies. ]
1 The modification involves removal of four south pool storage' racks, relocation of two south pool racks to the north pool, and installation of six new racks in the south pool with twice the s torage capacity. Appendix A identifies the need for this modification, describes the new storage rack physical arrangement, specifies the design criteria employed for the required analyses, and presents the results of these. analyses. .
Analyses are presented for the following:
- 1. Spent fuel pool criticality,
- 2. Rack structural design (seismic analysi's).
- 3. Pool structural design.
- 4. Decay heat removal capability.
- 5. Radiation and radjological considerations. '
021.28
' U, Q
I
'C. DESIGtt CRITERIA AMD AflALYSES l
l C.1 Criticality Considerations In order to put twice the fuel assemblies in the same area the center-to-center spacing had to be reduced by the factor 1/ 4. Reducing the space between assemblies increases the K 9ff of the fuel array in the pool, thus requiring an evaluation of this effect on the criticality I analysis of the spent fuel pool. A center-to-center spacing of 15.5 l inches was selected because prior experience indicated that this would be acceptable from a criticality standpoint without requiring special poison curtains and would provide twice the storage capacity at only a slicht increase in 'ack size.
C.I.1 Criticality Analysis Conclusion l 1. The 15.5 center-to-center spacing provides ample criticality safety margin to a Keff limit of 0.90. The Keff limit was estimated to he less than 0.95 but the calculations have shown that the present Keff limit of 0.90 can be met.
- 2. A high degree of confidence can be placed in the diffusion theory results based on the excellent agreement shown with independent transport-theory results and comparison between calculations and experimental results.
C.l.2 Method of Analysis and Results The reference set of calculations were carried out using the '
t standard ARK-PDQ' sequence. ARK was used to generate cross sections for each of the spatial regions represented in the GQcalculations(seeFigureC.1-1). The fuel region consisted
- of 3.5 w/o U-235 and 95" theoretical density unirradiated fuel rods. The reference moderator temperature was 120*F for all 4
regions. The only structural materials used in the analysis were the 304 SS angles at the corners of the fuel storage module.
These angles were hotogenized in the inner reflector recion.
The outer reflector region consisted of pure water. Zero ppm .
boron was assumed throughout. One mesh per fuel rod was used in the fuel region in the P00 calculation. Nro current boundary 4
conditions were used throughout to simulate ' infinite array.
Geome,tric axie.1 buckling with total reflector savings of 15 cm.
was used.
Figure C.1-2 shows the mechanical tolerances used in calculating the closest center-to-center assembly seoaration.
F,igure C.1-3 shows the P00 calculated Keff as a function of center-to-center spacing. Included in this figure are transport cal-culated Keff values for three different center-to-center assembly separations. Table C.1-1 summarizes the calculated K e
from both diffusio.n and transport theory calculations.ff values Note that Q21.29
\
'., i
on a unit cell basis, the two' calculations agree very well, . !
indicating a substantial ennsistency in the basic cross- !
section sets used in the two calculatinns. The additional different spatial neutron scattering treatments in codes.
and in Figure C.1-3 include cofrections to the orig to account P0Q for the following differences with respect to the calculations:
End 3) T
-0.2%anEga=170*F.
- 1) the absence of grids, 2) zero axial buckling listed above. 0.2%ap, respectively, for each of the differencesT i tion is shown on Table C.1-1.The effect of using SS in the diffusion calc Figure C.1-4 shows the variation of K the pool. eff with water density in in the ARK runs of all -This curve was obtained by varying the water d cross-sections in - v s, and inputting the resulting as the water tem' the plotted Keff values.
(upperlimit Keff decreases
' eases in the range of 68*F to 212*F Figure C ' .perature following a loss of SFP cooling).
of apr- optimum moderation occurs at a water density i 0."' :c and result :
o
. than the new s in a Keff value equal to I fuel storage rack requirement tion shows that new fuel of an enrichment e stored in the new spent fuel racks even modera tion.
At Poiht Beach, the spent fuel condition of optimum moderation..ed with borated wate
/
. , anal uncertainties of the criticality analyses can be assessed 101 by comparing the results of standard calculations with critical experiments.
difference between the calculations and experimental bias in the computation, was 0.1%aK. , or resul 4K.for level. calculational variations becomesonfidence Since no suitable, directly comparable, experimental 1.42%data 4K a were availsble for validating the accuracy of the diffusion theory calculations duplicating the P0Q (P00),
cases. transport theory calculations using 00T we As is shown on Table C.1-1 and Figure C.1-3 the diffusion transport theory calculations verify the accuracy of th calculations.
assigned to this comparisen of calculational Hiethods.An avera A worstacase assuming K pff criticality consideration is given in Table C 1-1.
calculational uncertainties (1.42%) and the use of the absorbing characteristics of the structural members (1.05 error The resulting tolerance. allowable Keff is 0.8758 which includes dimensiona .
By referring to Figure C.1-3 the minimum center-to-center spacing can be no less than 13.3 inches if a K 0.875 n)ust be maintained. eff limit of The minimum center-to-center spacing 9
021.30
,m s +n'
- between adjacent racks is designed to be no less than 14.5. inches, ,
thus there exists an additional margin of 2.4% aK. The minimum i spacing within a rack is 14.892" as shown in Figure C.1-2 and it is not as limiting as the rack-to-rack spacing of 14.5".
C.l.3 Other Criticality Considerations The transportation of heavy loads above the south pool racks is not pennitted. The same restrictions on the transportation of i heavy loads that are employed now will continue to be used after i this modification is made. Therefore, the spacing limits ' dis-cussed above are valid limits to be considered in the criticality J analysis. l No additional limitations on the handling of fuel assemblies in l l
the pool are needed, because the consequences of dropping an '
assembly on the top of the new racks is no worse than dropping an assembly on top of the old racks. The consequences are no worse because the additional storage locations make the new racks stronger in the vertical direction, and the distance the fuel assembly would )
fall is limited to about 3 feet in each case due to crane travel 1 limits. The design of the pool and administrative controls do l
not permit storage of fuel assemblies in any location other than l those l'ocations provided by the storage racks or transfer devices.
The structural design of the spent fuel storage racks is based n maintaining their structural integrity during a seismic event. l Refer to Section C.3 for a discussion of this subject. The same result is assumed for the cask drop accident, i.e., no change in j rack geometry. Refer to Section C.4 for a discussion of the cask drop considerations. In all cases, these events are not considered in the criticality evaluation be.cause rack geometry and storage I l
spacing is not altered as a result of the event.
.e I e
Q21.31
l l
TABLE C.1-1 COMPARISON OF KEFF VALUES FOR DIFFUS10ft ATID TRAt1 SPORT CALCULATIONS l
C-to-C T-D Spacing (in) Diffusion Transport T (%)
12.5 0.9069 0.9133 0.7 l l
15.5 0.8387 0.8441 0.6 20.0 0.8208 0.8171 -0.5 Unit Cell 1.4041 1.4031 -0.1 EFFECT OF STAINLESS STEEL ON K EFF C-to-C Keff
.Keff Spacing (in) With SS Without SS aK 15.5 0.8387 0.8746 0.0359 6
WORST CASE CRITICALITY CONSIDERATI0fl Safety Limit 0.9000 Calculational Uncertainties .0142 Structural Absorbing Uncertainty .0100 0.8758 Keff at 14.5" C-to-C spacing -0.8518 0.0240 Additional margin 2.4%aK I
t Q21.32
c 3.885" : C 1.576" = t 2.290" =
--> e-- . 55 5"
,.! ' ,/
I
/
j/ '/
i -
X '
FUEL REGION ,-
X X /
/
///
/ i
/ i
/ ,/ ,/ / / /
,// -
' ,- /,' ,
, l '
'/ /
HOMOGEllIZED SS AND WATER REGION /
/
///,/// / / /
WATER REGION GUIDE TUBES Figure 11,14 X 14 Fuel Rack 2-0 Geometry 021.33 e
-4
1.10 1.05 -
1 l
A TRANSPORT RESULTS CORRECTED
., TO DIFFUSION CONDITIONS 1.00 i 0.95 l
. \ ;
\
a i 0.90
\ :
. i
! l l l 0.85 ! l\ ! i j i
0 1
j 'l 0.80 i i 11 12 13 14 15 16 17 18 19 20 CENTER TO CENTER SPACING (INCHES)
Figure 12. Effective Multiplication Factor as a cunction Of Center to Center Spacing.
021.34 j
NRC Docket Nos.
50-295 and 50-304 QUESTION NUMBER 22:
On page 3-9 it is stated that a minimum areal density of 0.02 grams of boron-ten per square centimeter of Boral plate was used in the I calculations. How will this be verified for all of the plate area?
What is the nominal concentration of baron-ten in the Boral plates?
l
RESPONSE
The manufacture of Boral and fabrication into plates is controlled by the Brooks and Perkins Quality Assurance Program, which includes detailed procedures for the inspection and verification of baron-ten
]
loading in each Boral plate. The inspection plan for these activities .
will include the following:
- 1. Documented laboratory analysis for chemical boron content and isotopic baron-ten content of each lot of boron carbide powder.
- 2. Inspector's verification of the weighing and mixing of bc on carbide and aluminum powder into a batch, according to the l production plat., and the assembly of this batch into one u,'
several ingots, all identified and traceable to the batch.
- 3. Documented laboratory analysis of a selected sample of batch mixes to verify boron carbide content. l
- 4. The rolling of the ingot into a sheet, and the subsequent blanking of a sheet into two or three plates, each identified and traceable to the ingot and batch.
Q22.1
- i
!' )
I
- RESPONSE (Continued)
- ,
- 5. Visual inspection of the perimeter of each plate to verify that l the core extends to the edge (the aluminum edge filler has been
- completely sheared.away), and a check of plate thickness at several points.
1 6. Documented laboratory analysis, according to a sampling plan, ,
i of the boron carbide content of coupons cut from each end of each.
plate.
- 7. Documented neutron transmission tests over the surface of a selected sample of plates to verify the uniformity of boron-ten loading across the entire plate area.
4 The detailed sampling plans will be established prior to manufacturing, based on the specific production lot sizes to be used, in order to
- establish a 95 per cent confidence that the minimum areal density of 0.02 grams of boron-ten per square centimeter is present over the entire area 'of each plate.
1 It is estimated that the nominal level of baron-ten areal density will be 0.0220 0.0012 grams per square centimeter, to provide an assured minimum level of 0.02. The exact range will be established in the detailed production plans.
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022.2 l
NRC Docket h l
50-295 and-50-304 QUESTION NUMBER 23:
Provide a description of the onsite test you intend to perform to verify, within 95 per cent confidence limits, that a sufficient number of Boral plates in the installed racks will contain the ;
0 95. j required boron content to maintain the keff 3 RESPONSE: )
As discussed in the response to question 22, the Brooks and Perkins Quality Assurance Program which will be used to control the manufacture of Boral and the assembly of Boral plates into poison tubes will verify and document that all of the plate area in every poison tube I contains at least the minimum areal density of boron-ten. The -
fabrication Quality Assurance Program which will be used to control the assembly of poison tubes into a complete fuel rack will verify and document the traceability of accepted poison tubes into specific locations in the completed racks.
In summary, the presence of the required baron content will be .
verified by establishing a 95 per cent confidence that all the Boral plate contains a minimum boron-ten content, and by verifying l and documenting the traceability of individual plates to specific poison tubes, and then to specific rack locations. Onsite verification will not be required.
023.1 9
NRC Docket Nos.
50-295 and 50-304 QUESTION NUMBER 24:
With the consideration of all of mechanical and fabrication tolerances included, what will the minimum storage lattice pitch be, which will be verified by QA inspection records?
RESPONSE
The fabrication QA inspection records will verify the dimensional location of the fuel assembly storage tubes as follows:
In one horizontal dimension, the racks are assembled from clusters of two or three storage tubes, welded into a row of two or three tubes. Within these clusters, the corresponding inner walls a-e inspected to be within ! 0.06" of true pitch spacing (10.35"); that is, the inner wall of the second tube is 10.35 ! 0.06" from the I corresponding wall of the first tube; the inner wall of the third tube is 20.70 ! 0.06" from the corresponding wall of the first tube.
These measurements are made at both the top and bottom of each tube.
When these clusters are assembled into a row across the rack, the inner wall of the first tube in the cluster is located and inspected to be within t 0.06" of true pitch spacing for that tube measured at the top of the tube. Thus, for the second and subsequent clusters in a row, the corresponding inner walls of each tube (at the top) are within ! 0.06" plus 10.06" = I 0.12" of true pitch position.
In the second horizontal direction, the clusters are assembled side-by-
- side, and inspected to be within ! 0.06" of true pitch position (10.35"),
l measured from the inner wall of the tubes in the first row, at the top of the tubes.
Co n ti nued . . . . . .
024.1
. O RESPONSE (Continued):
In summary, at the top of each tube, the corresponding inner walls are located within not more than I 0.12" in one direction and within ! 0.06" in the othar direction of true pitch position. These i dimensions will be verified by QA inspection records.
This fabrication inspection is conservative with respect to Condition I 3 discussed on page 3-15 and illustrated in Figure 3.3-3B on page 3-23 of the Zion licensing report, which assumes both eccentric positioning j of the fuel assembly in the storage tube, and eccentric positioning of the storage tube in the lattice (offset 0.15" from true pitch position in both directions).
s Q24.2 l
NRC Docket Nos.
l 50-295 & 50-304 1 4
I QUESTION NUMBER 25: I 1
[ The calculated peak spent fuel pool outlet water temperatures i given in Figures 3.6-4 through 3.6-6 are lower than would be
- expected from the cooling system data given on Table 9 5-2 l
of the FSAR. What is the cause of these lower temperatures?
l
RESPONSE
The calculated spent fuel pool outlet temperatures in the licensing report differ from values given in FSAR Table 9.5-2.
' The FSAR table is based on the upper limits of the pool I design. The value of 120aF was selected as design criteria by the operating departments to be used in sizing the heat exchangers. Actual pool temperature is much less, usually 1
70-80aF. Thus, the values in the FSAR are selected limits, l 1
not tlie actual operating values.
i l The data generated by Nuclear Services Corporation in the license report, is computed based on the heat load as it is generated in the pool. The results are the pool i
operating temperatures as a function of fuel loading and heat removal capacity. Thus apparent discrepancy is the result of comparing the upper limits quoted in the FSAR, to the pool temperatures as generated in the report.
025.1
./
i NRC Docket Nos.
. 50-295 and 50-304 5-QUESTION NUMBER 26:
Figure 3.6-5 is entitled " Zion, Normal Refueling i HX" and noted
. to be Case 8-1, but Case B-1 is stated to be full core discharge.
Please clarify this apparent discrepancy.
4 i
RESPONSE
Figure 3.6-5 should be entitled, " Zion, Full Core Discharge 1 HX",
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l 1
i i
1 d
4 4
i 4
4 026.1
a e
, . o' NRC Docket Nos.
50-295 & 50-304 QUESTION NUMBER 27:
f Are the temperatures given in FS AR Table 9.5-2 under the title
" Spent fuel pit heat exchanger" for one heat exchanger in oper-ation with one pump? If not, what configuration of the system are these temperatures for?
RESPONSE
The valves in FSAR Table 9.5-2 for the heat exchanger are given for each unit based on single train operations. This assumes full component cooling flow (3000 gpm) and one spent fuel pit pump (2300 gpm) for each heat exchanger.
027.1
.