ML20010J203
| ML20010J203 | |
| Person / Time | |
|---|---|
| Site: | Yankee Rowe |
| Issue date: | 09/17/1981 |
| From: | Heider L YANKEE ATOMIC ELECTRIC CO. |
| To: | Crutchfield D Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20010J204 | List: |
| References | |
| FYR-81-137, NUDOCS 8109290702 | |
| Download: ML20010J203 (12) | |
Text
Propond Ch:nga #158 Supplcment #4 YANKEE ATOMIC ELECTRIC COMPANY
' /*
T*
2.C.15.1
}
h 1671_ Worcest - Road, Framingham, Massachusetts 01701 FYR 81-137 ANKEEj t-8 4
f (A
September 17, 1981 United States Nuclear Regulatory Commission S"
i SEP2s 1981s q Washington,-D.C.
20555 2
tr o s Attention: Office of Nuclear Reactor Regulation dprosa (//
Mr. Dennis M. Crutchfield, Chief g
_g Operating Reactors Branch #5 g/
Division of Licensing ib References (a) License No. DPR-3 (Docket No. 50-29).
(b) Yankee Atomic Electric Company Letter WYR-78-61 to USNRC dated July 13, 1978, Proposed Change No. 158.
(c) Yankee Atomic Electric Company Letter WYR-80-95 to USNRC dated August 18, 1980, Proposed Change Nc,. 158, Rev. I to Supplement 3.
(d) USNRC Letter to Yankee Atomic Electric Company dated January 22, 1981.
Subject:
Crane Travel - Spent Fuel Pit, Spent Fuel Pit Water Level, and Design Features
Dear Sir:
l Pursuant to Section 50.59 of the Commission's Rules and Regulations, Yankee l
Atomic Electric Company hereby proposes the following modification to l
Appendix A of the Operating License. This letter is Supplement 4 to Reference (b).
PROPOSED CHANGE: Revise Section 4.9.7b to read:
b.
The spent fuel inspection stand, the temporary gate and the shielding panels shall be prevented from traveling over fuel assemblies in the spent fuel pit by administrative control; and Revise Section 3.9.11 to read:
3.9.11 At least 14 feet of water shall be maintained over the top of irradiated fuel assemblies seated in the storage racks.
Revise 5.6.3 to read:
5.6.3 The spent fuel pit is designed and shall be maintained with a storage capacity limited to no more than 721 fuel assemblies.
8109290702 y h 9 0
P b\\\\
To:
U. S. Nuclear Regulatory Commission September 17, 1981 Attention: Office of Nuclear Reactor Regulation Page 2 BASIS FOR CHANGE: The presently designed capacity of the spent' fuel pit is sufficient to provide storage until only 1986 without loss of. full. core discharge capability. The spent ftel inventory after the present outage is expected to be 225 elements. The proposed modifications summarized in Reference (b), provide the capability to increase on-site spent fuel storage by means of a double-tier arrangement of storage racks. The maximum storage caracity, as. outlined in Reference (b), will therefore increase to 721 elements.
REASON FOR CHANGE: As part of our overall effort to modify and upgrrde the spent fuel pool (see Reference (b)), this supplement describes.in detail the proposed modification to install additional racks to provide increased storage capacity. These racks will be installed on top of the existing racks, thereby providing a second tier for storage of fuel assemblies.
For this reason, the changes described above are necessary to (1) modify the minimum water level maintained over an irradiated fuel assembly seated in the storage rack, (2) modify the specified fuel assemblies to be stored in the spent fuel pool, and (3) to allow the new lx5 module racks to travel above the stored fuel in the first tier.
DESCRIPTION OF CHANGE: The basic elements of the double-tier storage rack system are:
a) the first tier storage racks, b) the second tier storage racks, and c) the second tier storage rack support framework.
1 Discussion on each of these elements is provided below.
The first tier will employ the existing, boral racks currently employed in the pit. These racks, consist of welded aluminum five-element modules which are bolted together to form racks holding from 30 to 45 fuel bundles.
The modules on the ends of the assemblies have special fittings for the legs which position the racks off the pool floor. Each rack is designed to be either free-standing or connected with other racks.
In the new application, the basic modular design will be retained. The first tier will consist of the existing racks plus additional racks formed in the same manner. The new five-element modules used in the first tier will be identical to those now in place except for minor changes in welding resulting from f abrication experience. There will be no change to the center-l to-center spacing, the poison material, or the cavity channels themselves.
The existing storage rack modules are currently free-standing on the floor
~
of the pit.
In the first tier of the new double-tier arrangement, these modules will be arranged in three, longitudinal rows. Adjacent rack modules in each row will be attached together; but those of different rows will i-
c To:
U. S. Nuclear Regulatory Commission September 17, 1981 Attention: Office of Nuclear Reactor Regulation Page 3 be separated by severe.1 inches. This gap leaves space for the second tier framework support columns which attach to the pit floor (see Figure 1).
To provid? lateral restraint for the racks, the rack feet will be placed in special supports which are anchored to the spent fuel pit floor. These supports will prevent impact against the second tier support columns.
The second tier of storage will also be made up of the basic, welded aluminum five-element modules. These modules provide the same center-to-center spacing, poison material and cavity channels as the existing racks. However, the five-element modules will not be bolted together to form rack assemblies.
Rather, each module will be inserted directly into a supporting framework.
This framework will consist of a network of north-south oriented beams supported by intermediate columns attached to the pit floor.
During erection of the liner (Reference (c)), spent fuel rack support fixtures were installed. The function of these fixtures are to transfer loads from the spent fuel racks through the liner and into the concrete structure. Part of the fixtures are designed to support the present spent fuel racks. The rest of the fixtures are to provide capability for increased spent fuel storage capacity.
The support fixtures that were installed for the present racks are pads anchored to the floor as shown in Figures RS-1 and RS-2.
The pads are machined with grooves and have a bolt hole in the center. Replacement leg assemblies, Figure RS-3, have been manufactured for the spent fuel racks which have a mating grooved surface. After a rack is positioned on the pads and the grooves engaged, bolts may be inserted through the legs using a long socket extension and tightened to provide a positive clamping force.
Likewise, the bolts may be disengaged to permit removal of the rack.
The anchors and replacement legs are designed, fabricated, installed and inspected to ASME Section III, Subsection NF, " Component Supports". Although seismic loadings will be addressed in conjunction with the Systematic Evaluation Plan, the anchors and legs have been designed for static lateral loads of 0.5g in two horizontal directions with a static 0.2g vertical load, based on a fully loaded rack.
The replacement legs are made of type 304 stainless steel and were inserted into the threaded holes in the spent fuel racks which are 6061-T6 aluminum.
Stainless steel, though high in the galvanic series compared to aluminum (more noble), does have the advantage of forming a high electrical resistance oxide (Cr2 3).
This film in combination with aluminum's high electrical 0
resistance oxide (Al 0 ) will tend to minimize the galvanic corrosion of 23 the aluminum. Also, test specimens consisting of altuinum to stainless couples have shown minimum signs of attack after pro 13 2ed exposure to the mildly aggressive conditions in the spent fuel pool.
i t
To:
U. S. Nuclear Regulatory Commission September 17, 1981 Attention: Office of Nuclear. Reactor Regulation Pade'4 1
The supports for the second tier racks consist of structural framework as i
shown on Figures RS-4 to RS-11.
These supports are completely separate from the present spent fuel racks. Connection to the pool walls is made i-by penetrating the liner as shown in Figures RS-8 to RS-10.
The double-weld seal is provided using backing and cover plates at. the penetration.
i j
The support structure is designed, fabricated, installed and inspected to ASME Section III, Subsection NF, " Component Supports". As with the anchors for the existing racks, the supportc are designed to carry static loads of 0.5g in two horizontal directions and 0.2g in the vertical direction.
Since putting fuel on the second tier will involve lif ting fuel assemblies higher than presently allowed,. provision has been made to place grating over the~ existing racks. The grating will be supported on the lower beams of the second tier support structure. The grating and support structure are designed to resist the impact of a fuel assemb?y dropped from 11 feet above the existing spent fuel racks. The analysis parmits plastic deformation but limits distortion to prevent contact with.the racks. This grating will be placed in the pool prior to placing any racks on the second
'ier.
SAFETY EVALUATION:
1 Criticality The criticality analysis for the proposed two-tier spent fuel rack design-has been performed. The procedure used is the procedure used to license the present fuel rack design. We will restate only the basic assumptions used in the analysis:
(1) Minimum water inside flux trap (2) Minimum B10 concentration inside the horal plate (3) Minimum boral thickness (4) No boron present in the spent fuel pit (5) A pool temperature of 680F l
(6) Fresh 4.5 w/o U-235 Zircaloy clad fuel (7) An ll-inch' assembly pitch (8) A 33-inch water gap between active fuel between-the first and second tier (9) Identical top and bottom racks of the present design-The result of this analysis is a very subcritical effective multiplicatien factor'of 0.782.
If a calculational uncertainty of 3.0% k/k is used,-the result is 0.801, well below the 0.95 acceptance criteria used by the NRC.
l The design has been evaluated using Yankee's standard methodology for
. calculating criticality for spent fuel storage. The design has sufficient margin to criticality to meet the present acceptance criteria and NRC
?
standards.
J 4
~,, ~.
,,.s L.
,.-,,..-..,,.,,,..--,.-n..,...erm.-~,._
To:
U. S. Nuclear Regulatory Commission September 17, 1981 Attentio.J Office of Nuclear Reactor Regulation Page 5 Water Level The head of water over the spent fuel is continuously monitored by level indication instrumentation (Low Level Alarms 251 & 226, High Level Alarms 250 & 225). Along with the unattended monitoring, the Operations Department physically checks the pool level every fcur hours.
When pool make-up is required, it is done manually by the operator. Make-up is accomplished by way of either, valve DW-V-665 (2-1/2" unmetered flow)
. valves DW-V-765 & DW-V-766 (1/2" metered flow).
A rate of evaporation calculation was done for the double-tier arrangement and concluded that at a pool temperature of 85 F, approximately 26 gal / day was liberated. We have calculated the evaporation rate at a pool temperature of 150 F at found it to be approximately 290 gal / day, assuming a sufficient volume of ventilation air to hold this water. The latter rate is well within the plant's ability to make-up.
Because the pool has continuous level monitoring along with periodic physical verification,v.e feel that maintaining the correct level of water over the spent fuel is insured.
Heat Remosal The Spent Fuel Pool Heat Exchanger (SFPHX) uses component cooling water which, in turn, uses service water as its heat sink. With the full pool loading of 721 assemblies, the tube inlet temperature of the SFPHX will be raised approximately 15%. This, in turn, raises the shell outlet temperature approximately 9% (from 78.3 F to 85.55 F).
The Component Cooling Water Heat Exchangers (CCWHX) have been designed for ar. Anlet temperature of 96 F.
Since the inlet temperature to the CCWHX will be below design limits, we conclude that the increased heat load to the SFPHX will not adversely affect plant operations.
Ventilation The ventilation system for the spent fuel building is rated t approximately 3100 cfm. This rate, combined with the ventilation of the Primary Auxiliary Building (PAB) at 10,900 CFM, passes through the filter bank in Mechanical Equipment Room #3 and to the atmosphere.
There would be no impact in heat released to the environment because:
1.
The temperature of the ventilation air for the spent fuel building would not increase appreciably (<10%) due to the standard flow rates.
To:
- 9. S. Nuclear Regulatory Commission September 17, 1981 Atteation: Office of Nuclear Reactor Regulation Page 6 l
l l
L 2.
A large percent of the energy available in the 150 F water will be used in converting water to vapor.
3.
The spent fuel pool ventilation will be mixed at a rate of over three to one with the PAB ventilation (PAB at ambient).
s Thermal-Hydraulic Analysis The thermal-hydraulic analysis indicates that there is adequate coolant flow within tLa spent fuel module. RETRAN-12B, a system of program modules used to analyze thermal-hydraulic transients, was used tc model the spent fuel pool. In the new mechanical design proposal, e second tier of spent fuel modules will be placed above the original level of spent fuel melules.
The thermal-hydraulic consequences of this change are:
1)
Increased axial power input to coolant 2)
Increased axial pressure losses to flow 3)
Decreased downcomer flow area.
To evaluate these changes, a RETRAN model was developed. The spent fuel pit model contains eight homogenized spent fuel module stacks. Each stack represents two spent fuel modules (Exxon fuel assemblies in place), and one grating separating the two modules. The eight stacks represent the longest span between the downcomer and the furthermost fuel module.
The Frictional Loss Coefficient (FLC) of the stack was calculated from the area weighted sum of its components. The FLC of the fuel assembly was calculated using d*.ts from the Exxon fuel report. This value was then adjusted to correspond with the spent fuel pit minimum flow conditions, which is defined as the flow rate that would exist if the fluid exiting the top of assembly were saturated liquid. Idel'chik's handbook was used to determine the FLC of the grating and Crane's handbook was used to determine the FLC of the spent fuel module wrapping.
The coolant reservoir model contains an arbitrarily large volume of water at constant temperature (150 F) and pressure ('.7.69 psia). The FLC of the reservoir was considered negligible.
The downcomer model contains the downcomer channel and the two support braces. The flow rate used to calculate skin friction contributions to the FLC was eight times the flow rate in one stack. The form loss factors of the FLC were taken from Crane's handbook.
The spent fuel pool floor model contains eight volumes corresponding to l
l
To:
U. S. Nuclear Regulatory Commission September 17, 1981 Attention: Office of Nuclear Reactor Regulation
' age 7 the eight stacks. Flow in these volumes is considered to be only between the fuel assembly nozzle and the pool floor. The only flow to the stacks is assumed to be through the fuel assembly nozzle.
The FLC of the pool floor was calculated from the area weighted sum of its components. The skin friction contributions to the FLC were calculated assuming the flow rate in all the stacks to be equal. The form loss factors of the FLC were taken from Crane's correlation for standard pipe tees.
The power level used in the spent fuel pool was one day decay. Assembly peaking factor was conservatively assumed to be 1.50 and 20% was added to the ANS Standard decay heat correlation.
To avo11 difficulties in a properly converged solution, the initial conditions in the spent fuel pool were zero flow rate and zero power level.
When the transient was initiated, the power level was ramped up to the level described previously. The transient was then continued until a steady-state solution converged, about 300 seconds.
0 The highest bundle outlet temperature calculated by RETRAN was 177 F, which is significantly less than the boiling point of 229 F.
To indicate the margin to boiling, all of the FLC values were doubled. The highest outlet temperature was 189 F.
The results of a conservative thermal-hydraulic analysis indicate that tnere is adequate coolant flow to preclude boiling in the double-tier spent fuel pit.
In addition to the examination of an assembly within a fuel module, cooling in the channel between modules was also investigated. Axial pressure loss to flow between modules was calculated to be about half of that inside a module. Combined with the fact that less heat will be deposited between modules than within the module verifies that cooling in the channel between modules will be better than that within modules and, hence, not a concern.
Fuel Handling As discrased in Revision 1 to Supplement #3 (Reference (c)), grating will be placed in the pool prior to placing any racks on the second tier. The necessity to place the grating in the rack support structure af ter completion of this installation is noneeded, since the existing administrative control does not allow lifting the fuel assemblies higher than presently allowed.
When lifting the fuel assemblies to the new height required for the second tier, the grating will be in place.
By administrative control, the fuel bundle will only be allowed to be lifted a maximum of 11' above the top of the first tier racks. As an added measure
~
To:
U. S. Nuclear Regulatory Commission September 17, 1981 Attention: Office of Nuclear Reactor Regulation Page 8 of assurance, a mechanical stop will be placed on the manipulators so as not to exceed the 11' height rest.? tion above the existir.g racks. During fuel transferring operations to the second tier, a total water depth of 36 feet will be required. This would result in five feet of water cover over the tops of the spent fuel. An evaluation was made to determine reasonable exposure levels with the ALARA concept in mind. The following restrictions will be imnosed because of ALARA considerations:
1.
Only fuel that has decayed 150 days or longer will be allowed to be placed in the second tier, and 2.
A minimum of five feet of water cover will be required over the spent fuel.
Considering these restrictions, an evaluation was made to determine increases in personnel exposure during fuel transfer operations. The results indicate increase in exposure of only 1.35 milliman-rem per assembly as compared to an already low level during fuel transferring operat;ons. Radiological measurements will be made though, at the time of the first shuffle, into the second tier to evaluate the margin of safety in the calculations as well as to determine if fuel of a shorter decay period than 150 days can be handled.
Even though the new 1x5 modules are heavier than a fuel bundle, the distribution of weight is such that it will impact no less than two sets of grating. With this as the case, and the shape of the rack less severe than a fuel bundle for an impact load; the worst drop accident has been evaluated based on previous submi ttals. This drop accident is that of a fuel bundle on to the grating.
As indicated in Revision 1 to Supplemen.., the liner in the fuel storage area was designed to provide a leaktight membrane. The intent of "leaktight membrene", as indicated in previous submittals, and typical of new construction, is to provide an added margin of assurance; the spent fuel pit will remain watertight during normal and upset conditions. Since the liner Jerves as only an upgrade item for the spent fuel pit, and lining the fuel transfar area was an additional upgrade, *' -- _aould be no re; iirement that the liner in the fuel transfer area remain watertight after a fuel bundle drop accident. The important point regarding this concern is that the spent fuel pit should remain watertight after such an accident.
An analysis was done on a fuel bundle drop of 40 feet. Assuming 37 feet of water drag, an impact velocity of 39.2 feet per second was calculated.
Using the modified NDRC formula, and the minimum contact area of the nose cone, the maximum penetration was equal to 2.88 inches. However, because of the nose cone geometry the penetration should be limited to 3/4 of an inch. Therefore, this indicates that the 36-inch thick floor slab will not be perforated, and the bundle drop will have no effect on the water
To:
U. S. Nuclear Regulatory Commission September 17, 1981 Attention: Office of Nuclear Reactor Regulation Page 9 retaining capacity of the pool concrete. However, since the liner is only 1/4 inch thick, it may be pierced by the prongs, before engaging the nose cone shell. Even if the liner is pierced, the integrity of the pool will be maintained. As an added measure of assurance, the fuel ransfer area could be isolated from the rest of the pool with the use of the temporary gates in the permanent gate support bracket. At that time, the fuel transfer area could then be dewatered and repatred.
Design Criteria The design, fabrication and installation criteria for the racks will be in conformance with the appropriate codes, standards, and specifications.
A static seismic analysis consisting of.5g horizontal in the two horizontal directions and.2g vertical was done on the second tier rack support structure due to lack of SEP resolution. Once the plant's response spectra is available, we will proceed with a more rigorous analysis and provide information concerning load combinations, design and analysis procedures, and the structural acceptance criteria.
The construction materials for the liner and the second tier rack support structure conforms to the requirements of ASME Code,Section III, Subsection NF.
The racks will be constructed from the same materials as the existing racks in the pool (aluminum). Through many years of experience, it has been shown that this material is compatible with the pool environment.
All the materials used are compatible with the storage pool environment and do nct contaminate the fuel assemblies or the water; and therefore, will maintain its material stability and mechanical integrity.
Radiological Impacts During normal operation, the spent fuel pool ion exchanger uses no more than one charge (20 cu ft) cf resin per year. During construction, wnsiderably more resin has been used due to an increase in chemical
. purities, not radionuclides. Expanding to double-tier storage will result in a somewhat higher input of radionuclides into the pool. This input has been analyzed using escape rate coefficients for fuel in a spent fuel pool.
The increase in concentration by calculation has been shown to be less than 10 percent of present values. For conservatism and to account for the uncertainties of crud product buildup and exchange with spent fuel pool water, the increase in solid waste from fuel pool expansion (resins and filters from spent fuel clean-up activities) is estimated to be less than 25% of normal values or approximatly 5 cu ft/ year.
Th-total gaseous Kr-85 released as measured from the plant vent stack for ti last two years was 7.6 curies in 1978 and 1.6 curies in 1979.
To:
C. S. Nuclear Regulatory Commission September 17, 1981 Attention: Office of Nuclear Reactor Regulation Page 10 The following table contains data on radionuclide concentrations in spent fuel pool water.
Jan '79 Sept '79 Syyt '78 Sept '75 H-3 4E-03 3E-03 4E-03 2E-03 Cs-134 2E-03 3E-04 1E-04 2E-05 Cs-137 2E-03 7E-04 3E-04 2E-05 Co-58, Co-60, Mn-54 1E-05
<1E-06 2E-05 2E-05 By measurement, the dose rate at the edge of the pool varies from 0.1-0.5 mR/hr. Since the exact source of this very low dose rate is not precisely known, no attempt will be made to model the dose equivalent rate from these radionuclides. The fuel in the pool, the pool water as well as local hot spots, all contribute. Also, the core is located about 100 feet from the spent fuci pool and could also be contributing a small umount.
Concerning airborne concentrations of radionuclides in the SFP area, noble gases have never been detected above SE-07 micro curies per cubic centimeter.
1-131 was detected during a fuel shipping operation in the fall of 1978 at 1E-12 to 8E-12 micro curies per cubic centimeter. Other than that, I-131 has never been detected in the spent fuel pool area above the MDA of 1E-13 micro curies per cubic centimeter. Particulate levels generally run on the order of SE-12 to SE-11 micro curies per cubic centimeter gross beta.
Since under normal conditions no detectable airborne concentrations of radionuclides er.ist in the spent fuel pool area at Yankee Rowe, it is expected that any increase due to expansion will be unmeasurable both in the SFP area and at the site boundary.
The spent fuel pool resin is treated the same as resin from primary coolant purification and refueling cleanup. During a recent change-out, 1.9 man-rem was accumulated during resin transfer, packaging and preparation for shipment. It is estimated that the curie content from the spent fuel pool resin accounted for about 10% of the total or about 0.2 man-rem. As stated earlier, expansion of the spent fuel storage capability is not expected to increase levels by more than 10%. Including a factor of 2.5 for conservatism and to account for uncertainties in the buildup of crud products, the estimated increase in man-rem from changing the SFP resin would be approximately 0.05 man-rem.
As stated previously, exposure levels in the Yankee Rowe SFP are very low.
Present levels along the edge of the pool are very low and would not justify cleanup action.
If, however, the bui.1 dup of crud products does become an exposure problem for personnel standing on the edge of the pool, underwater vacuum cleaning can be used effectively to remove this material. Adequate care must be taken to shield the filters used in such a device.
To:
U. S. Nuclear Regulatory Commission September 17, 1981 Attention: Office of Nuclear Reactor Regulation Page 11 The total man rem received by personnel occupying the fuel pool area is estimated to be less than 10 man-rem per year. The major point to be made is that under normal conditions, the spent fuel pool area at Yankee Rowe does not present a significant problem stea for radiological protection.
Control of this area and the maintaining of ALARA considerations in very
.ow radiation fields such as these, present no problems to a well qualified Health Physics staff. Therefore, estimates of dose rates and incremental dose rates and concentrations are limited in value. Certainly the experience to date would indicate that the fuel pool expansion will not significantly change the radiological conditions in the spent fuel pool and that such conditions are quite manageable utilizing current Health Physics procedures at Yankee Rowe.
The analysis for a fuel handling eccident onto the double-tier with 14 feet of water above the stored fuel was completed. The following table lists the decontamination factors, percent removal for iodine and the 2-hour site boundary thyroid dose and whole body gamma doses for various depths of water above the stored fuel.
2 Hour 2 iiour Depth of DF for Overall Percent Site Boundary Site Boundary Water Above Inorganic DF for Iodine Thyroid Dose Whole Body Dose Fuel (Feet)
Iodine Icdine Removal (120 Hrs Decay)(120 Hrs Decay) 23 133 100 99.0 42 REM 0.16 REM 14 58.5 51 98 83 REM 0.16 REM The DF of 133 for 23 feet of water above the fuel is the accepted DF for inorganic iodine as stated in Regulatory Guide 1.25.
The DF listed for 14-foot depth was derived in Calculation #YR-EDCR-78-12-C, Rev. 2.
The overall DF for iodine is calculated based on the assumed ratio of (99.75%)
inorganic to (0.25%) organic iodine species present in the gap activity.
The DF for the organic portion is assumed to be 1.
Note that changes in the amount of iodine released for the depths of water listed will not change reported whole body doses using 2 significant figures.
The Technical Specifications prohibit the use of a cask in the pool. This issue will be addressed in Supplement 5.
The determinations of the previous safety evaluation reports and final environmental statements have not changed significantly and the impacts are not significant as well.
The conclusion is that there is no increase in the probability of an accident
To:
U. S. Nuclear Regulatory Commission September 17, 1981 Attention: Office of Nuclear Reactor Regulation Page 12 (or equipment malfunction) or of an accident of a different type which has not been analyzed, and the margins of safety which have been defined in the bases of Technical Specifications have not been reduced. This Proposed Change has been reviewed by the Nuclear Safety Audit and Review Committee.
FEE DETF.RMINATION: This Proposed Change is a part of the overall project described in Reference (b) which was determined to be a Class IV amendment and for which a payment of $12,300 has been submitted.
7 SCHEDULE OF CHANGE: We solicit NRC approval of this Proposed Change before December 1. 1981.
We trust this information is acceptable to you; however, should you have any questions, please feel free to contact us.
Very truly yours, YANKEE ATOMIC ELECTRIC COMPANY Y Yd6 L. H. Heider Vice President COMMONWEALTH OF MASSACHUSETTS)
)se.
COUNTY OF WORCESTER
)
Then personally appeared before me, L. H. Heider, who, being duly sworn did state that he is Vice President of Yankee Atomic Electric Company, that he is duly authorized to execute and file the foregoing request in the name and on the behalf of Yankee Atomic Electric Company, and that the statements therein are true to the best of his knowledge and belief.
TM N A Robert H. Groce, Notary Public My Commission Expires September 14, 1984
- ,.'4 l(
i.>g%, f(
..s m e ;'
.4 y n
.