ML18142B985
| ML18142B985 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 01/30/1976 |
| From: | LeBoeuf, Lamb, Leiby & MacRae, Rochester Gas & Electric Corp |
| To: | Rusche B Office of Nuclear Reactor Regulation |
| References | |
| Download: ML18142B985 (110) | |
Text
r NRC DIST%i.UTION FOR PART 50 DOCKET Mi~%RIAL (TEMPORARY FORM)
R CONTROL NO:
PEPM.LEBOEBP,LANB,LELBY&NACRAE DATE OF DOC DATE R EC'D LTR TWX RPT OTHER Washington, DC Addressees 1-30-76 1-30-76 XXXX TO: ORIG CC OTHER, SENT NRC PPR Mr Rusche 3 signed SENT LOCAL PDR CLASS UNCLASS PROPINFO INPUT NO CYS REC'D DOCKET NO:
XXXXXXX 3 50-244 DESCRIPTION: ENCLOSURES:
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service...notarized..; Amdt to OL/Change to Tech Specs:~) Consistingc,"
w/attch certx. fr cate of . oi revision to tech specs with regard to 1-30-76....trana the following: modification to spent tuel pool storage.
racks....(40 cys encl rec'd)
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PLANT NAME. Ginna 'gp'i SAFETY FOR ACTION/INFC:RMATION 2-3-76 ehr 4SS It'NEO AD ASSXG'ARED BRANCH CHIEP
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LAW OFFICES OF LEBOEUF, LAMB,LEIBY 8L MACRAE 1757 N STREETi N. W, WASHINGTON, D. C. 20036 ARVIN E. UPTON l40 BROADWAY NEW YORK, N ~ Y. I0005 LEONARD M. TROSTEN WILLIAM O. DOVB EUGENE B THOMASi JR uary 30, 1976 WASHINOTON TELEPHONE HARRY H VOIGT
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Mr. Ben C. Acg) 5 Director CA Office of Nuclear Reactor Regulatio U.S. Nuclear Regulatory Commission Washington, D.C. ,20555 Re: Rochester Gas and Electric Corporation R. E. Ginna Nuclear Power Plant Unit No. 1 Docket No". '50-'24'4'
Dear Mr. Rusche:
As counsel for Rochester Gas and Electric Corporation, we hereby transmit three (3) signed originals and nineteen (19) copies of a document entitled "Applica-tion for Amendment to Operating License" together with forty (40) copies of a proposed change to Technical Speci-fication 3.8.1. This request for change in technical specifications is being submitted in connection with pro-posed modifications to the spent fuel pool storage racks for the Ginna plant. Attachment B to this application sets forth the safety evaluation for the proposed change in specification as well as a complete description of the proposed modifications.
In the opinion of RGEE's Nuclear Safety and Audit Review Board, the proposed modifications do not constitute an unreviewed safety question within the mean-ing of 10 C.F.R. g 50.59(a), as discussed more fully in Attachment B. Since, however, it has been the Commission's recent practice to review and, in effect, approve all modifications dealing with spent fuel storage pools, RGGE hereby requests approval of these modifications in ad-dition to approval of the proposed change in specifica-tions.
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Forty (40) copies of Attachment B are also enclosed.
A Certificate of Service showing service of these documents upon the persons listed therein is also enclosed.
Very truly yours, P7'iud LeBoeuf, Lamb, Leiby 6 MacRae Attorneys for Rochester Gas and Electric Corporation
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION In the Matter of )
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Rochester Gas and Electric Corporation ) Docket No. 50-244 (R. E. Ginna Nuclear Power Plant, )
Unit No. 1)
APPLICATION FOR AMENDMENT TO OPERATING LICENSE Pursuant to Section 50.90 of the regulations of the U.S. Nuclear Regulatory Commission (the "Commission" ),
Rochester Gas and Electric Corporation ("RGGE"), holder of Provisional Operating License No. DPR-18, hereby requests that Technical Specification 3.8.1 set forth in Appendix A to that license be amended. This request for a change in the technical specifications is submitted in view of proposed modifications to the spent fuel pool storage racks which will in-crease the storage capability of the pool.
The proposed technical specification change is set forth in Attachment A to this Application. A safety evaluation demonstrating that. the proposed change does not involve a signifi-cant change in the types or a significant. increase in the amounts of effluents or any change in the authorized power level is set forth in Attachment B. Attachment B also describes the proposed modifications to the spent fuel pool storage racks and supports
f the conclusion by Applicant's Nuclear Safety Audit and Review Board that the modifications to the facility as described in the facility ' Technical Supplement Accompanying Application for a Full-Term Operating License do not constitute an unreviewed safety question within the meaning of 10 C.F.R. rr 50.59(a) of the Commission's regulations.
WHEREFORE, Applicant respectfully requests that Appendix A to Provisional Operating License No. DPR-18 be amended in the form attached hereto as Attachment A.
Rochester Gas and Electric Corporation By L. D. W te, Jr.
Vice President Electric and ISteam Production Subscribed and sworn to before me this 2E day of January, 1976.
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ATTACHMENT A Add paragraph 3.8.1,g and h to Section 3.8.1.
3.8.l,g. The decay heat of the fuel stored in the spent fuel pit plus the fuel removed from the reactor for normal refueling will not exceed 5.3 x 10 BTU/hr.
3.8.1,h. If the full core is to be placed in the spent fuel pit, the decay heat of the fuel stored in the spent fuel pit plus the fuel removed from the reactor will not exceed 9.3 x 106 BTU/hr.
Add the following to the end of the Basis for Section 3.8.1.
During normal refueling the spent fuel pit temperature is limited to 120'F( ) . At this temperature the spent fuel pit heat exchanger will handle a heat load of 5.3 x 10 BTU/hr with a service water temperature of 80'F. To insure the spent fuel pit temperature will not be exceeded, a limit is placed on the system heat load.
During is limited to 1504F '.
full core diygParge the spent, fuel pit, temperature Under these conditions the system will handle 9.3 x 10 BTU/hr and the system heat load is correspondingly changed.
The decay heat will be calculated using Reference (4) plus 20%. The fuel assemblies will be assumed to have been irradiated at rated core power for the average burnup of the discharged fuel and decay time will be the average for the discharged fuel.
Add these References to the end of Section 3.8.
(3) FSAR Section 9.3.1 (4) ANS 5.1 (N18.6), October 1973
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Attachment B Spent Fuel Storage Rack Replacement
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Spent Fue torage Rack Replacement TABLE OF CONTENTS I. INTRODUCTION II'ESCRIPTION OF NEW DESIGN A. Spent Fuel Rack B. Rack Arrangement C. Rack Base III. EVALUATION OF NEW DESIGN A. ANSI N 18.2 1973 B. NRC Regulatory Guide 1.13 IV. INSTALLATION V. NUCLEAR ANALYSIS A. Methods of Analysis B. Evaluation of Reference Design C. Uncertainty Considerations D. Additional Considerations VI. THERMAL-HYDRAULICANALYSIS A. Heat Removal Requirements B. Service Water C. Analysis of Heat Removal Systems D. Cooling of Individual Assemblies VII. SEISMIC ANALYSIS VIXI. RADIOLOGICAL EVALUATION IX. ACCIDENT ANALYSES A. Fuel Handling Incident B. Shipping Cask Drop Accident C. Interruption of Spent Fuel Cooling X. REFERENCES
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<>> I INTRODUCTIGN The spent fuel pool at the Ginna Plant is described in Section 9.3 of the Final Safety Analysis Report (FSAR) and has a capacity for storing 210 fuel assemblies. There are currently 56 fuel assemblies in the pool, and it is anticipated that 36 more assemblies will be placed in the pool at the Spring 1976 refueling outage.
RGGE presently has a contract with Nuclear Fuel Services, Inc.
for fuel reprocessing. RGGE shipped the initial core loading of fuel, 121 fuel assemblies, to NFS in the Spring of 1973. However, in response to the present lack of storage space at NFS and to insure the capability for full core discharge following the refueling in March 1977, RGGE is planning to replace existing racks, in which fuel assemblies are stored with 21 inch center-to-center spacing, with new racks with a mean distance between centers of fuel of 12-1/2 inches. The number of spent, fuel storage positions will be increased to 595. This will allow RGGE to store all spent fuel assemblies from Ginna through 1985 and have the capability to unload all fuel from the reactor vessel.
The replacement of the spent fuel racks is planned for the Fall of 1976. The rack replacement will be performed while there is water and spent fuel in the pool. After approximately half of the existing racks are replaced, the spent fuel in the pool will be transferred to the new racks, and replacement of
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design criteria, including seismic capability for all racks containing fuel, will be satisfied at all steps in the rack-replacement procedure. The use of divers is anticipated to facilitate removal and installation operations.
Under the present fuel management plan, decay heat removal requirements will be less than the design capability of the Spent Fuel Pool Cooling System as described in the FSAR until after the refueling scheduled for 1981. The Spent Fuel Pool Cooling System is presently capable of removing 5.3 x 106 Btu/hr under Normal Refueling Conditions and 9.3 x 10 Btu/hr under Full Core Discharge Conditions.. Under the proposed change to the Technical Specification, the decay heat load from the fuel stored in the Spent Fuel Pool will be limited to these values until modifications can be made to increase the Spent Fuel Pool heat removal capability.
All design, analysis, and fabrication are being performed under direction of Wachter Associates, Inc. The nuclear analysis is being performed for Wachter Associates, Inc. by Pickard, Lowe, and Garrick, Inc. Installation will be performed under the technical guidance of Wachter Associates, Inc.
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II. DESCRIPTION OF NEW DESIGN A. S ent Fuel Rack The present spent fuel racks will be replaced by new spent fu'el racks that will increase the storage capacity'o 595 assemblies.
The new spent fuel rack is a modular design arranged in a checkerboard pattern. The inherent. strength of this rack design is its honeycomb box structure arrangement. (Every box in the module is solidly fastened to adjacent boxes, thus resulting in an extremely rigid structure.)
The rack assemblies are made up of a repeating array of square stainless steel boxes. Alternate boxes in the checkerboard pattern are designed to contain spent fuel assemblies. The remaining boxes vill contain pool water.
The stainless steel boxes are approximately l3-l/2 feet long, 8.25 inches square (on the inside) and 0.090 inch thick. The lower end of each box contains a horizontal plate, with a circular hole in the center, to hold the'pent fuel assembly.
There are three types of rack modules; Type A which contains 70 fuel assembly locations, Type B which contains 56 fuel assembly locations, and Type C vhich contains 49 fuel assembly locations. These units are structurally
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and differ only in
'dentical the number of boxes used to construct them. The empty weight of the Type A unit is approximately 20,700 pounds, the empty weight of the Type B unit is approximately 16,600 pounds, and the empty weight of the Type C unit is approximately 14,700 pounds.
B. Rack Arran ement The arrangement of the racks in the spent. fuel pool con-sists of seven Type A racks and one each of Type B and C, as shown in Figure 1. This arrangement provides space at the south end of the pool for the shipping cask and other items needed in fuel handling operations. The racks will occupy less space than is presently occupied by spent fuel racks.
C. Rack Base A stainless steel I-beam base will be installed in the pool for each rack module. These bases are provided with level-ing pads which are interconnected mechanically to each other N
and are laterally supported off the wall by means of large bearing pads. (See Figure 2.) The racks are solidly bolted to the rack bases. (See Figure 3.)
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4 III. EVALUATION OF NEW DESIGN Criteria for the design and performance of spent fuel storage systems are defined by ANSI Standard N 18.2 1973 and USNRC Regulatory Guide 1.13. The new spent fuel rack satisfies these criteria, as described below. Where compliance with a criterion is not affected by the modification, the criterion is not listed.
A. ANSI N 18.2 1973 5.7.4 Performance Criteria 5.7.4.1 "The design of spent fuel storage racks and transfer equipment shall be such that the effective multiplication factor will not exceed 0.95 with new fuel of the highest anticipated enrichment in place assuming flooding with pure water ... Credit may be taken for the inherent neutron absorbing effect of materials of construction or, if the requirements of 5.7.5.10 are met, for added nuclear poisons."
Assuming new fuel with an enrichment. of 3.5 w/o, the effective multiplication factor of the new rack design is less than 0.8871, including uncertainties. Ginna Technical Specification 5.3.l.c limits fuel enrichment to no more than 3.5 w/o of U-235. Credit is taken in the calculations for the inherent neutron absorbing effect of the-stainless steel used in the structure and for boxes that are utilized
as neutron poison. The method of calculation is presented in Section V. The requirements of Criterion 5.7.5.10 are satisfied as discussed subsequently.
5.7.4.2 "Fuel handling system facilities shall be designed to prevent damage to fuel assemblies while in storage or during transport from one location to another."
Each fuel assembly is stored in a stainless steel box, which physically separates that fuel assembly, from all other fuel assemblies. The stainless steel box will be strong enough to prevent damage to the contained fuel assembly in the unlikely event that another fuel assembly should be dropped 'anywhere on top of the spent fuel racks.
The new rack design contains no protuberances that could cause damage to a fuel assembly being lowered into or being lifted out of a storage position. Adequate lead-ins are provided at the top of the boxes.
- 5. 7. 4. 3 "The fuel storage pool capacity shall accommodate at, least one shipping cask and one complete c'ore in addition to the maximum number. of fuel assemblies normally stored in the pool.
Consideration should be given to potential for highly radioactive components which may require storage in the pool."
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The new rack design will provide the capability to store all spent fuel assemblies from approximately ll years of reactor operation and still retain the capability to accommodate one shipping cask and removal of the complete core from the reactor vessel. During refueling periods, and whenever the shipping cask is not in the pool, the cask area will be available to store radioactive components or to perform underwater inspection of or underwater mechani-cal operations on radioactive components. There is also space between the spent fuel racks and the walls of the pool which is available for longer-term storage of radioactive compo-nents.
5.7.4.6 "Suitable provisions shall be made in the design of the fuel storage pool cooling system to permit installation of instrumentation to monitor system performance."
The pressure and flow of service water through the Spent Fuel Pool heat exchanger and the temperature and pressure of Spent Fuel Pool water circulating through the Spent Fuel Pool heat exchanger are measured and indicated locally.
The Spent Fuel Pool water temperature is measured and a high temperature alarm is actuated in the control room if the Spent Fuel Pool water temperature exceeds 115'F. The Spent Fuel Pool water level is also measured and a High/Low alarm is actuated in the control room if the water level exceeds preset values.
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5.7.5 Mech cal Desi n Criteria 5.7.5.1 "The fuel storage pool and storage racks shall be designed to accommodate, within applicable code stress limits, normally imposed loads due to half the Design Basis Earthquake."
This criterion is satisfied as described in Section VII of this report.
5.7.5.2 "The fuel storage pool and storage racks shall be designed so that normally imposed loads plus loads imposed by the Design Basis Earthquake will not cause failure. Plastic deformation may take place but with a substantial margin to that which might result in failure. "
The fuel storage pool is founded on sound rock. The new spent fuel storage racks are capable of withstanding loads imposed by the Design Basis Earthquake without plastic deformation of the racks and without damage to spent. fuel assemblies. The bearing loads are sufficiently low to prevent damage to the stainless steel liner of the spent fuel pool and supporting concrete. The reinforced concrete structure of the pool is capable of transmitting these loads to the rock without plastic deformation of the pool structure.
5.7.5.3 "Lifting and transport equipment of the fuel handling system shall be designed to prevent dropping of fuel assemblies. Heavy loads shall not be carried over stored fuel assemblies.
The design shall prevent lifting a fuel shipping cask over fuel storage racks."
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I These provisions are included in the initial design. The outer envelope of the new spent fuel racks is entirely within the outer envelope of the existing spent fuel racks and, therefore, compliance with these criteria is not, affected.
5.7.5.4 "Fuel storage racks shall physically prevent placing more than one fuel assembly in a single storage location; specified minimum center-to-center distances between individual fuel assemblies shall be maintained to meet requirements of Section 5.7.4.1."
The new rack design permits only one fuel assembly to be inserted into a storage box. Minimum center-to-center spacings between fuel assemblies are maintained by the rack structure.
6 5.7.5.5 "Fuel storage rack design shall prevent geometric to environmental conditions character-changes due istic of this site. The design shall be stable against tipping with provisions to prevent unplanned movement of the fuel or 'the racks."
The geometry of the new rack design cannot be changed by seismic events, nor by other environmental conditions char- ~
acteristic of the site. The rack design is stable against tipping. Each stored fuel assembly is completely surrounded by a relatively close-fitting box.
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5.7.5.8 "The fuel storage pool and refueling canal shall have provisions, such as a watertight liner, to prevent leakage of pool water."
r A stainless steel liner is provided. The new spent fuel racks and their seismic supports are designed to limit local mechanical loadings on the pool liner to prevent damage to the liner. In addition, installation of the new racks does not require complete removal of the existing welded rack supports and the racks I
will not. be welded to the existing liner, thus. precluding possible damage to the liner during installation. The new racks are sup-ported as described in Section II-C of this report.
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minimum depth shall be determined by dose considerations at the top of the pool considering irradiated fuel or components stored in the pool or in transit and radioactive contaminants in the pool water." U I
The depth of water over the spent fuel is unchanged by the new rack design. A radiological evaluation of the new rack design is presented in Section VIII of this report.
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'I 5.7.5.10 "Fue torage racks using nuclea oisons additional to those inherent in the structural materials squall be designed and fabricated in a manner to prevent inadvertent removal of the additional poisons by mechanical or chemical action. Prior to installation of the additional nuclear poisons, the quantity and effectiveness of the additional poisons shall be verified.
Effectiveness of the additional poisons may be checked by isotopic analysis. Provisions shall be made to permit periodic inspection or verification or'oth, thereafter."
The proposed spent fuel rack design does not employ nuclear poisons in addition to those inherent in the structural materials.
5.7.5.13 "Provisions shall be made to accommodate the necessary heavy equipment loads in the'fuel storage 'pool without subjecting the pool liner to mechanical damage."
The bearing loads on the pool liner are low and will not cause mechanical damage to the liner. Bearing loads are described in Section VII of this report.
B. NRC REGULATORY GUIDE 1 ~ 13 "The spent fuel storage facility (including its structures and equipment'except as noted in Section 6 below) should be designed to Category I seismic requirements."
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The spent fuel pool is designed to Category I'eismic requirements as described in Section 5.1.2.4 of the FSAR.
The new spent fuel racks are also designed to Category I seismic requirements as described in Section VII of this report and as discussed in the responses to Criteria 5.7.5.1 and 5.7.5.2 of ANSI 18.2 1973.
"The spent fuel storage facility should have the following provisions with respect to the handling of heavy loads, including the refueling cask:
- a. Cranes capable of carrying heavy loads should be prevented, preferably by design rather than by interlocks, from moving into the vicinity of the pool, or
- b. The fuel pool should be designed to withstand, without leakage which could uncover the fuel, the impact of the heaviest load to be carried by the crane from the maximum height to which it can be lifted. If this latter approach is followed, design provisions should be made to prevent this crane, when carrying heavy loads, from moving in the vicinity of the stored fuel."
Section IX-B of this report, documents that these aspects of the facility are not. affected by replacement of the spent fuel racks because the outer envelope of the new spent fuel racks is within the outer envelope of the existing racks.
XNSTALLATiON The replacement of racks will be accomplished prior to the Spring of 1977 refueling outage. The rack replacement will be performed while water is in the poo l. After approximately half the existing racks are replaced, the
,92 spent fuel assemblies in the pool will be removed into the new racks, and the replacement of the remazn~ng racks will be completed. Applicable safety and design criteria will be satisfied in all steps of the rack replacement procedure. Xt x.s planne d to use divers to assist in the above operation.
NUCLEAR ANALYSIS A. Methods of Anal sis The LEOPARD( ) computer program was used to generate macro-scopic cross sections for'input to four energy group dif-fusion theory calculations which are performed with the PDQ-7 (2)'rogram. LEOPARD calculates the neutron energy spectrum over the entire energy range from thermal up to 10 Mev and determines averaged cross sections over appro-priate energy groups. The fundamental methods used in the LEOPARD program are those used in the MUFT(3) and SOFOCATE(
programs which were developed under the Naval Reactor Pro-gram and thus are well founded and extensively tested ana-lytic techniques. In addition, Westinghouse Electric Cor-poration, the developers of the original LEOPARD program, demonstrated the accuracy of these methods by extensive analysis of measured critical assemblies consisting of slightly enriched U02 fuel rods(
In addition, Pickard, Lowe and Garrick, Inc. (PLG) has made a number of improvements to the LEOPARD program to increase its accuracy for the calculation of reactivities in systems which contain significant amounts of plutonium mixed with UO2. PLG has tested the accuracy of these modifications by analyzing a series of U02 and PuO -U02 critical experiments.
These benchmarking analyses not only demonstrate the im-provements obtained for the analysis of Pu02-UO2 systems but also demonstrate that these modifications have not
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affected the accuracy of the PLG-modified LEOPARD program for calculations of slightly enriched UO2 systems.
The UO2 critical experiments chosen for benchmarking in-clude variations in H20/UO2 volume ratios, U-235 enrichments, pellet diameters and cladding materials. Although the LEOPARD model also accurately calculates the reactivity effects of soluble boron, these experiments have not been included in the benchmarking criticals since the spent fuel pool calculations do not involve soluble boron.
Neutron leakage was represented by using measured buckling input to infinite lattice LEOPARD calculations to represent the critical assembly. A summary of the LEOPARD results is shown in Table V-1 for the 27 measured criticals chosen as being directly applicable for benchmarking the model for spent fuel pool calculations. The average calculated keff is 0.9979 and the standard deviation from this average value is 0.0080 h k. Reference 5 raised questions concerning the accuracy of the measured bucklings reported for the experiments number 12 through 19. If these data are excluded, the average calculated keff for the remaining 19 experiments is 1.0006 with a standard deviation from this value of 0.0063 h k.
The PDQ series of programs have been extensively developed and tested over a period of 20 years and the current version, PDQ-7, is an accurate and reliable model for calculating the subcritical margin of the proposed spent fuel pool arrangement.
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As a specific demonstration of the accuracy of the calculational model used for the spent, fuel pool calculations, the, combined LEOPARD/PDQ-7 model has been used to calculate seven measured just-critical assemblies. The criticals are high neutron leakage systems with a large variation in U/H20 volume ratio
'nd include parameters in the same range as those applicable to the proposed spent fuel pool design. Experiments. including soluble boron are included in this demonstration since we are primarily interested in the. ability of PDQ-7 to calculate neu-tro'n leakage effects. The use of soluble boron allows changes in the neutron leakage of the assembly while maintaining a uniform lattice and thus allows a better test of the accuracy of the model.
These LEOPARD/PDQ-7 calculations, shown in Table V-2, result in a calculated'average keff of 0.9922 with a standard deviation about this value of 0.0014 h k. These results together with the I
previously discussed LEOPARD results demonstrate that the proposed LEOPARD/PDQ-7 calculational model can calculate the reactivity of the proposed spent fuel pool arrangement with an accuracy of better than + 0.01 I k.
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B. Evaluation of Reference Desi n The PDQ-7 program is used in the final predictions of the reactivity of the spent fuel storage pool. The calculations are performed in four energy groups and take into account, all the significant geometric details of the'uel bundles, fuel boxes, and major structural components.. The geometry used for most of the calculations is a basic cell representing one quarter of the area of a repeating array of two identical stainless steel boxes. The specific geometry and dimensions of this basic cell are shown in Figure V-l.
The calculational approach is to use the basic cell to calcu-late the reactivity of an infinite array of uniform spent fuel racks and to account for any deviations of the actual spent fuel rack array from this assumed infinite array as pertur-bations on the calculated reactivity of the basic cell. The effects of mechanical tolerances are also treated as pertur-bations on the calculated reactivity of the basic cell. The fuel bundles were assumed to be unirradiated with a U-235 enrichment, of 3.5 w/o which is higher than any anticipated reload enrichment for the Ginna core. Most. of the calcula-tions were performed at a uniform pool temperature of 80 F, but the reactivity effects of pool temperature are also taken J
into account as a perturbation on the basic cell calculations.
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The reference basic cell calculation is performed with the minimum dimension on all the stainless steel, boxes which results in a k = 0.8779. Other tolerances on the geometric array representing the racks are treated as perturbations on this reference basic cell calculation.
The stainless steel fuel and water boxes are nominally .090 inches thick with a tolerance of + .004 inches. Assuming a
'worst case in which all boxes were at the minimum thickness of
.086 inches the h of the basic cell is .8806. There-I fore, the maximum perturbation on the reactivity of the basic cell due to variations in the stainless steel box thickness is +.0027 a k.
With the fuel bundles located in their most reactive positions inside the stainless steel boxes, the k of the basic cell is .8807. Thus, the perturbation on the basic cell reactivity due to positioning uncertainties is + .0028 h k.
Most of the calculations with the basic cell geometry utilized a 50 x 25 two-dimensional array of mesh points. To test, the adequacy of this mesh description a calculation was run with a 100 x 50 mesh size and the resulting k was .8777. Thus the perturbation on the basic cell due to mesh spacing effects is .0002 g k.
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The k~
0 of the basic cell as a function of temperature is shown in Figure V-2. With a maximum pool temperature of 200'F under the worst possible conditions the k is 0.8838, which results in a perturbation due to temperature effects of
+0.0059 6 k. Although the overall steady state reactivity temperature coefficient of the spent fuel pool is positive, the temperature coefficient of the fuel bundles is negative.
The basic cell was also used to evaluate the reactivity effect of axial neutron leakage. Using an axial buckling based on a 142 inch active fuel length with a total reflector savings of 15 cm, the calculated k~ of the basic cell is .8759. Thus the reactivity pertubation due to axial neutron leakage is
.0020 5 k.
A summary of the perturbations to the basic cell reactivity calculation is shown in Table V-3. Thus the calculated reactivity of the spent fuel pool with 595 unirradiated bundles with 3.5 w/o U-235 is .8871 for a pool temperature of 200'F.
C. Uncertaint Considerations In Section V.A it was demonstrated that the uncertainty in the calculated k ff with the model utilized for criticality cal-culations is less than + 0.01 h k. It will now be demonstrated
~
that there are a number of conservatisms in the model's of the spent fuel pool such that these con- 'epresentation servatisms more than compensate for the uncertainty in the calculational model. Therefore, the effective mul'tiplication factors presented in Section V.B are conservative even when the effects of model uncertainties are included.
V-6
p e I
t I
~ 4 P'
The basic cell calculations of k apply to an infinite array of racks containing unirradiated fuel bundles with no burnable poisons and no net radial neutron leakage. The maximum reload batch size anticipated for the Ginna core is less than or equal to 40 bundles. Therefore even if the'ntire core were to be discharged shortly after the start of a fuel cycle, there would be at most 40 unirradiated fuel bundles in the spent fuel pool. In such a situation there would be significant
'll neutron leakage from the 40 unirradiated bundles to surrounding irradiated bundles or to empty fuel locations or to the water reflector. It is conservatively calculated that the resulting "
radial neutron leakage would reduce the calculated reactivity of the basic cell by .0102 I k.
The spacer grids utilized in the design of the Ginna fuel bundles contain inconel spacers which result in parasitic neutron absorption which is not included in the basic cell
,calculations. The spacer grids are calculated to reduce the k~ of the basic cell by .0086 6 k.
The inherent conservatisms in the analytical .model are such II as to reduce the calculated k~ of the basic cell by at least
.0188 h k. The reduction in k~ is nearly twice the possible increase in the k~ of the basic cell due to uncertainties in the analytical model. Therefore the multiplication factor of the spent fuel pool is ( .8871, the value reported in Section B above, and Criterion 5.7.4.1 in ANSI N 18.2-1973 is satisfied.
V-7
0 r
I'
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D. Additional Considerations These analyses take credit only for the inherent neutron absorbing properties of the type 304, stainless steel boxes which are the principal structural components of the spent fuel racks. Fe, Ni, Cr, and Mn account for 99% of the com-position of type 304 stainless steel and these are the only constituents which 'are considered to absorb neutron in these analyses. Other constituents, including impurities, will result in some small additional neutron absorption which will slightly increase the subcriticality of the rack.
The construction of the spent fuel racks is such that a dropped fuel bundle cannot under any conceivable circumstance pene-trate and occupy a position other than a normal fuel storage location. Therefore a dropped fuel bundle will end up in a final position that is somewhere between vertical and hori-zontal on top of the racks. The only positive effect of such a bundle on the reactivity of the rack would be by virtue of a reduction in axial neutron leakage from the rack. Since the calculations reported here show the total. axial neutron leak-age effect to be .0020 I k, a dropped fuel bundle would not 4
have any significant effect on the reported maximum possible reactivity of the spent fuel storage rack.
V-8
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I 'I TAt)t H V-1 Case'tefercnce I,atlice Critic:ll Calruhlted Number Nunlt<er 1'.nrichnlent (a loni .<I) 1120 Vo tun<a IV Fuel
~Denst t
,(6 icnl3)
Pellet DLamet<
(cm) r Cia<I Diameter (c nl)
Clad Tldcl n< ss (c ni)
Pitch (e nl) nl
-etr-Duckltnp. h 11 2.734 2.18 10. 18 0.'I620 0.8594 0.04085 1. 028'I 40 7r5 1.00FG<
11 2.'I34 2.93 10. 18 O.VG20 0.8594 O.ol085 1. 1049 53.23 1 0053
~
3 11 2.'I34 3.8G 10.18 0.7G20 0.8594 0.04085 1. 1938 G3.26 1.0013 12 2. 734 7.02 10.18 0.7620 0.8594 0.04085 1.4551 G5 64 1.0058 12 2. 734 8149 10.18 0.7620 0.8594 0.04085 1.5621 G0.07 1.0118 6 12 2.734 10.38 10.18 0.7620 0.8554 0.04085 1.G891 52.92 1.0072 2.50- 0'.8594 7 13 2. 734 10,18 0.7620 0.04085 1.0G1V O'I. 5 1:0008 8' 13 2. 734 4 10.18 0.7G20 Sr594 0.04085 1.2522 GB. 8 0.5987 5150 13 3. V45 10.37 0.7544 0.8600 0.040G 1.0G1V G8.3 1.0010 10 13 3. 745 4.51 10.37 0 7544 O.SGOO 0.040G 1.2522 95. 1 1.0025 11 14 3 745 4.51 10.3V 0.'7544 0.8600 0.040G 1.2522 95. GS 1.0005 12 15 4.0G9 2.55 9.46 1.12VS 1.2090 0.0406 1.5113 88. 0 0 5859 13 15 4.0G9 2.14 9.46 1.12VB 1.2090 0.0406 1.450 75.0 0.9830 14 ~
16 4.0G9 F 59 9.45 1.1268 1.2701 0.071G3 1.555 69. 25 0.9999 15 16 4.069 3.53 9.45 1. 12GB 1.2'I01 0.071G3 1.G84 85. 52 0 5958 16 16 4.069 8.02 9.45 1. 12GB 1. 2701 0.071G3 2.198 92.84 1.0040 17 1G 4.0GO 9.90 9.45 1. 12GB 1. 2701 0.07163 2.381 91. 'I9 0.9872 18 16 3.03V 2.64 9.28 1.1268 1. 2701 O.OV163 1.555 50. 75 0.9946 19 16 31037 8.16 9.28 1.12GB 1. 0.07163 2. 198 68. 81 0.9809 2701'.9931 20 8 0.714 1. GB 9.52 0.8570 0.0592 1.3208 108. 8 0.9912 21 8 0.714 2.1V 9.52 0.85'IO 0.9931 0.0592 1.4224 121 5 1.0029 22 8 0.714 4.70 9.52 0.85'IO 0.5531 0.0592 1.8GG9 155. G 0.9544 23 8 0.'I14 10.76 9.52 0.85VO 0.9931 0.0592 2.G41G 128.4 1.0008 24 5 0. 'I29 1.11 5.35 1.2827 1.4427 0.0800 1.752G 65.1 0.9502 25'G 5 0. 729 3.49 9.35 1.2827 1.4427 0.0800 2.4785 104.72 1 0055 9 0. 'l29 3.49 9.35 1.282'I 1.4427 0.0800 2 4'l85 75.5 0.9948 27 9 0.725 1.54 9.35 1.282V 1.442V 0.0800 1.5050 90.0 0. 98'IB These arc Pu02 in N<tur<<t UO2
+ Cases 1 throuGh 15 arc with sL<inless stell chid, Cases 20 throuGh 2V:<rc siren)toy clad.
TABLE V-2 WESTINGHOUSE UO2 CRITICAL EXPERIMENTS (References 6 and 7)
Boron H20/U02 Pitch eff
~Ex t ~(>m) (Volume) (In) (PDQ-7) 1.49 .600 .9905 2.42 .690 .9949 3 4. 35 .848 .9921 6.21 .976 .9918 306.0 1.49 .600 .9912 536.4 1.49 .600 .9925 727.7 1.49 .600 .9926
TABLE V-3 Reactivity Perturbations on the Reference Basic Cell Calculation Descri tion of Reactivit Perturbation Reactivit Effect, h k Mechanical tolerance spacing on stain- 0.00 less steel boxes Fuel position within stainless steel + .0028 boxes Mechanical tolerances on stainless + . 0027 steel box walls Mesh effects .0002 Temperature increase to 200'F + .0059 Axial neutron leakage .0020 Total perturbation on basic cell + .0092 reactivity calculation
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FIGURE V -1 NOTE: Boundary Condition at the Top of this Figure is 180 Rotational Symmetry RGE RACK REFERENCE DESIGN
- 0. 090 304 SS All Around CO 0, 233
- 8. 430
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0 FIGURE V-2 EFFECTIVE MULTIPLICATION VS.
TEMPERATURE I t I P
co" CO t
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,~+
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t CO C)
I t II
-t II I
j 150', 200
- . TEMPERAT URE F tI-!
VI . THERMAL HYDRAULIC ANALYSIS A. Heat Removal Re uirements The heat removal criteria of the Spent Fuel Pool Cooling System (SFPCS) are given in Section 9.3.1 of the Ginna FSAR and are that the system must be capable of maintaining the Spent, Fuel Pool (SFP) temperature less than or equal to 120'F during Normal Refueling operations and less than or equal to 150'F during Full Core Discharge situations.
Normal Refueling operations are conducted annually with nominally 40 fuel assemblies (one-third of the core) being removed from the core and placed in the SFP.
Full Core Discharge occurs when all the fuel in the reactor (121 fuel assemblies) is placed in the SFP. The full core will be discharged once every ten years to enable inspection of the lower reactor internals. Full core discharge may also occur'on other occasions when it is deemed necessary to remove the core to perform maintenance on the reactor lower internals or pressure vessel.
B. Service Water Tem erature The Spent Fuel Pool (SFP) heat exchanger transfers heat from the SFP water to the service water. The Service Water System I
is discussed in Section 9.6.2 of the Ginna FSAR.
0 The temperature of the service water go'ing into the SFP heat exchanger is a controlling factor in determining the heat transfer capability of the SFP cooling system. The service water temperature is the same as the intake (lake) water temperature except during the winter months when recirculation's used as necessary to maintain a water temperature of approximately 37'F.
Table VI-1 illustrates the monthly average of the daily minimum, average, and maximum intake water temperatures.
Table VI-2 presents lists of the minimum and maximum intake water temperatures that occur at any time during each month.
The intake water temperature has been recorded since December 1969. The data show the following:
the instantaneous daily maximum temperature has exceeded 80'F three times and then only by a maximum of two degrees.
- 2. the monthly average of the daily maximum tempera-tures has not exceeded 75'F.
- 3. the monthly average of the daily average tempera-
,tures has not exceeded 73'F.
The service water temperature to the inlet of the SFPCS heat exchanger can therefore be assumed to be 80'F or less.
VI-2
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C. Anal sis of Heat Removal S stem The SFPCS consists of a single loop containing a pump and heat exchanger. Water is drawn from the SFP by the SFP pump, forced through the heat exchange, and returned to the SFP. The heat exchanger is cooled by the Service Water. Approximately 10% of the water from the SFP bypasses the heat exchanger and is passed through a demineralizer and filter.
The design capabilities of the SFPCS were calculated for 120'F (maximum Normal Refueling temperature) and 150'F (maximum Full Core Discharge temperature). A service water flow of 700 gpm at 80'F was assumed with a SFP outlet flow of 610 gpm and with only 550 gpm flowing through the SFPCS heat exchanger. Under these conditions, the heat exchanger, with design fouling, will transfer 5.3 x 10 BTU/hr with a SFP outlet temperature of 120'F and 9.3 x 10 BTU/hr with a SFP outlet temperature of 150'F.
The impact of the propose'd modification on the heat load has been evaluated for Normal Refueling operation and the Full Core Discharge. Xn both cases the decay heat was calculated from the ANS 5.1, N18.6 standard plus 20% assuming finite irradiation. Table VX-3 illus-trates the results of these calculations using the following assumptions:
VI-3
- a. all fuel assemblies irradiated at rated core power for the entire design burnup of the fuel assemblies except for those assemblies presently in the SFP. These assemblies were assumed to be irradiated a rated core power for the actual average assembly burn~op.
- b. refueling takes place annually.
- c. one-third of the core (40 assemblies) is discharged annually except for the 1976 refueling when 36 assemblies will be discharged.
Full Core Discharge
- a. the emergency outage occurs one year after the last refueling and consists of a full core unload (121 assemblies) into the SFP in addition to the assemblies already in the pool from previous refuelings.
- b. a full core unload consists of 3 regions with burnups of one-third, two-thirds, and design assembly burnup.
As can be seen from Table VI-3 the heat load on the SFPCS decreases as the time between reactor shutdown and placement of the fuel in the SFP is increased. The SFPCS is capable of maintaining SFP temperature below 120'nd 150'F for several years without requiring unreasonable decay time.
VI-4
I g' f
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ll II I
During this period, modifications to the SFPCS will be considered that will increase the cooling capability of the system. Until these modif ications are complete the SFP heat load will be limited to 5.3 x 10 BTU/hr and 9.3 x 10 BTU/hr respectively.
The calculations summarized above are conservative for-the following reasons:
- a. No credit is taken for heat loss by evaporation from the pool surface. Heat is assumed to be removed only by the SFPCS heat exchanger.
- b. No credit is taken for the heat capacity of the SFP water. Transient calculations account-ing for this effect would allow greater instan-taneous heat loads without. exceeding required temperature limits.
- c. Measurements have shown that it is possible to have greater than design service water flow through the SFPCS heat exchanger. The cal-culations assume design flow; greater flow would result in larger heat transfer.
VZ-5
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- d. Refuelings are scheduled for March or April of each year when the lake water temperature is less than 45'F. Under these conditions, the SFPCS heat exchanger can be operated at higher heat fluxes without exceeding a SFP temperature of 120'F. Similar condi-tions would exist for a scheduled unload-ing of the core which took place during a normal refueling.
D. Coolin Anal sis of Individual Fuel Assemblies At present, water is returned to the SFP from the SFPCS heat exchanger through a discharge pipe entering the pool near the center of the south wall, Water enters the SFPCS through another pipe also located on the south wall.
To insure proper cooling of the fuel assemblies in the proposed spent fuel rack modification, the discharge pipe will be rerouted to run along the west wall of the SFP where it will discharge water in the wall-rack space just above the seismic supports. All of the fuel rack base I-beams have holes cut in them to provide 50%
free-flow area; this amounts to 7.38 ft alone in the I-beams facing the west wall. Another 1.64 ft is provided by the e
2" beam-to-floor gap (2" minimum, specified as 3"), although its presence is not crucial.
VX-6
The area inside the I-beams under the rack boxes is com-pletely open and free of obstruction except for the relatively minor effects of the jack screws and their supports. The bottoms of the fuel boxes and poison boxes are flush with others so that the only pressure losses inside the I-beams are flow-branching losses which are .
minor compared to passing through (and under) the I-beams.
Each fuel assembly's flow will depend upon its heat dissipation rate and the total pressure loss experienced by the base flow reaching its inlet (lower nozzle) location, which in turn depends slightly upon other fuel assembly heat rates and flows. Fuel assemblies having higher (than average) heat dissipations draw higher flow rates, but not enough to prevent a higher outlet temperature.
The flows from all the fuel assemblies mix above the fuel racks and move toward the south wall outlet.
The cooling of the individual assemblies has been analyzed assuming the worst case conditions. A full core discharge situation was assumed with a pool heat load of 9.3 x 106 BTU/hr, a pool outlet temperature of 150'F, and a pool flow of 610 gpm. The hottest fuel assemblies were assumed to be located near the east wall since the water reaching these fuel assemblies experiences the greatest pressure loss by having to pass through either 7 or 9 I-beams. The fuel
t t
I
assemblies o
having average heat dissipation were assumed to occupy racks in the center of the pool, where their cooling water passes through 3 I-beams.
The cooling analysis accounts for the pressure losses due to the I-beams and the 2-inch beam/floor gap. Flow branching ~ P's are negligible. Fuel assembly pressure losses are also accounted for and were found to be the dominant factor in comparison to the I-beam losses which were calculated on the basis of decreasing flow and velocity head due to branching.
By subtracting the pool-average fuel assembly h P and its associated I-beam losses, the hottest fuel assembly a P and its associated I-beam losses are calculated as a function of it its fuel assembly flow; increases when plotted against the ratio, hottest-fuel assembly-flow/pool-average-fuel assembly-flow. The driving differential for hottest-fuel assembly flow in comparison to pool-average-fuel 'assembly flow is the difference in the average water densities in the two fuel assemblies times the active fuel height, and decreases with increasing flow ratio. The intersection of these two curves defines the operating point of the hottest fuel assembly.
Results show the hottest fuel assembly with an outlet tem-perature of less than 155'F.
The maximum cladding temperature, accounting for the film- LT, is less than 160'F. Considering that the local saturation tem-perature is 242'F, the calculated temperatures for these worse case conditions are acceptable.
TABLE VI - I MONTHLY AVERAGE INTAKE IVATER TEMPERA~TURE oF VERSUS TIME OF YEAR GINNA STAT ON 1970 1971 1972 1973 1974 1975 MIN AVG MAX MIN AVG MAX MIN AVG MAX MIN AVG MAX MIN AVG MAX MIN AVG MAX
- 34. 3 32. 4 32. 8 33. 3 35. 0 36. 7 JAN 35. 0 32,9 33. 6 33. 7 35. 8 '7.7
- 35. 5 34. 3 35. I 34. 6 37. 9 39. 2
- 31. 4 32. 7 32. 0 33,1 32. 6 36.2 FEB 31. 8 33. 2 32. 1 ~ 33. 3 33. 0
'6. 9
- 32. 5 34. 1 32. 8 33. 8 33. 3 37. 9
- 32. 3 NR 32. 7 35,2 35. 0 36. 5
- 32. 8 NR 32. 9 35. 8 35. 1 36. 7
- 33. 2 33. 8 36. 7 36. 2 37. 0
- 40. 7 NR 35. 5 38. 8 40. 0 39. 1 APR 42. 1 NR 36. 0 39,3 41. 4 39. 6
- 43. 3 37. 0 40. 1 43. 5 40. 3
- 41. 8 42. 5 NR 42. 8 45. 2 49. 4
- 42. 8 43. 1 43. 4 46. 2 51. 1
- 43. 7 44. 2 NR 44. 0 47. 1 52. 2
- 55. 2 45. 9 48. 2 48. 9 51. 4 56. 1 JUN 56. 8 48. 1 50. 0 51. 7 53,6 59. 3
- 58. 3 50. 5 51. 7 54. 3 55. 8 62. 0
- 60. 9 62. 0 62. & &l. 5 61. 7 69. 9 JUL 62. 9 64. 3 64. 1 64. 6 64. 6 72. 7
- 64. 3 66. 6 65. 6 67. 5 66. 5 74. 6
- 69. 9 63. 4 60. 9 63. 1 66. 2 68. 6 AUG 71. 7 65. 7 64. 0 66. 4 68. 7 71. 0
- 74. 4 67. I 65. 9 69. 9 71. 0 73. 2
- 60. 7 57. 3 63. 1 65. 9 64. 4 57. 7 SEP 63. 5 60. 9 64. 3 66. 9 66. 4 59. 5
- 66. 0 63. 5 65. 3 68. 8 67. 5 61. 3
- 50. 9 52. 0 51. 9 55. 3 52. 4 54. 1 OCT 52. 4 54. 0 52. 4 56. 2 52. 8 55. 2
- 53. 3 55. 9 53. 0 57. 1 53. 8 56. 1
- 48. 2 45. 5 43. 1 45. 0 48. 1 50. 4 NOV 49. 0 46. 4 43. 3 45. 7 48 ~ 3 51. 1
- 49. 7 47. 3 43. 9 46. 6 48. 7 51. 7 39,3 37. 7 36. 2 40. 0 38. 7 DEC 40. 3 38. 8 36. 9 41. 1 39. 8
- 41. 6 39. 8 37. 5 42. 1 40. 9 Intake Structure Data:
Distance from Shore (ft): 3000 Water Depth (ft): 30 Average Water Withdrawal Depth (ft): 22, 5
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Table VI-2 MINIMUMAND MAXIMUMMONTHLY INTAKE WATER TEMPERATURE F GINNA STATION 1970 1971 1972 1973 1974 1975 Min Max Min Max Min Max Min Max Min Max Min Max JAN 30 39 32 43 32 40 32 39 33 53 35 42 FEB 31 35 32 37 32 37 32 37 32 36 34 43 31 35 NR NR 32 39 33 41 34 48 34 39 APR 39 45 NR NR 33 42 36 45 36 56 34 47 40 46 40 48 NR NR 40 48 42 52 43 64 JUN 45 64 38 60 42 55 41 63 43 63 43. 72 JUL 55 69 41 71 47 73 42 72 43 72 43 79
-AUG 58 82 41 72 46 69 40 76 43 75 43 79 SEP 41 75 40 73 45 68 49 78 45 78 43 68
'7 OCT 41 59 40 66 48 58 44 60 61 43 60 NOV 43 54 42 51 38 58 42 48 43 52 47 54 DEC 35 48 33 43 32 41 37 46 37 44
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Table VI-3 Total SFP Heat Load After Discharge Fuel is Placed in SFP Number of Discharge Fuel Assemblies Normal Refuelin Full Core Dischar e Year Dischar ed 10 da deca 15 da deca 25 da deca 30 da deca x 106 BTU x 10 BTU x 10 -BTU x 10 BTU Hr Hr Hr Hr 1974 12 1975 44 1976 36 4.50 3. 95 9. 37 8.74 1977 40 5.30 4.70 9. 74 9.11 1978 40 5.48 4.88 9.28 1979 40 5.66 5.06 9.46 1980 40 5.16 1981 40 5.23 1982 40 5.29
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I VII. SEISMIC ANALYSIS The new spent fuel racks for the Ginna Plant are designed for a maximum seismic event of 0.2g horizontal acceleration applied simultaneously with normal (1.0g) gravity plus or minus 0.2g vertical acceleration. The earthquake loads on the rack and base structures are calculated on the basis of the largest, fully-loaded spent. fuel rack. The direction of the horizontal seismic component is assumed to be in that worst-case direction which results in the maximum loads at any fuel rack corner joint. In addition, each fuel box (or cell) is designed to accommodate one fuel assembly with a rod control cluster assembly (RCCA) for a total design weight of 1,450 pounds.
The spent fuel racks are classified Seismic Category I in accordance with USNRC Regulatory Guide 1.29. They are designed for and will withstand the seismic loadings previously described. The honeycomb-like stainless steel structure of the rack not only provides a smooth, all-welded stainless steel box to protect the fuel assembly and preclude seismic damage, but also serves as a neutron absorber and will maintain the fuel in a non-critical (nuclear) array so long as the stainless steel box wall surrounds the fuel assembly.
The largest rack consists of 140 stainless steel boxes of which 70 are available for spent fuel storage and 70 are neutron flux traps. This 140 box rack is designated as a Type A rack. Two other rack geometries, the Type B containing 56 spent fuel
I
~ I C J ~ ~,< r e
n
assemblies and the'ype C containing 49 spent fuel assemblies, were necessary to accommodate the spent fuel pool dimensions and fuel storage restrictions. All calculations are based on the fully-loaded Type A fuel rack which serves as the worst case.
Each stainless steel box is securely fastened to its neighbors.
The resulting honeycomb-like structure is quite stiff; in fact, the rack is stiffer than the support base I-beams.
The I-beams serve as the load path to transfer seismic loads from the rack to the pool floor and walls, and also provide a redundant rack support structure. The stainless steel box configuration and thickness were selected on the basis of nuclear requirements as well as convenience in handling, shipping, stability and resistance to low frequency vibrations.
The loaded spent fuel rack (which includes water inertial effects and assumes that each fuel assembly contains an RCCA) and the base structure are capable of withstanding accident loads, including the Ginna OBE and DBE seismic requirements.
If the rack structure were subjected to the DBE (0.2g) load, the stresses of all applicable structural components would not exceed the following AXSC limitations:
a ~ Tension or compression aT < 0.60 ~y over a gross section <C. < 0.60 ay where ~y is the 0.2% yield strength of the stainless steel.
- b. Shear over gross section s 0.40 y c ~ Bending stresses tensile b 066 y and compressive
- d. Buckling stresses ~c 0 60 ~CR compression only where ~CR is the lowest load critical buckling stress.
- e. Tension or compression on ~ST 0. 75 ~y solid round or square bars; ~SC ~0.75 ~y also for bending stress of solid rectangular bars ~RB ~0.75 ~y about weaker axis Recognizing that yield stress ~y and elastic modulus E are functions of temperature, both properties were extracted from tables in Section III of the 1974 ASME Boiler and Pressure Vessel Code. The temperature at which these properties have been selected is assumed to be 200'F.
Since- the water in the spent fuel pool is not expected to reach 200'F, the values used for ~y and Ey are conservative.
Weld materials are generally considered to be identical to the base material since full-strength welds will be made in accordance with the AWS recommended sizes. In addition, all crucial structural welds were designed to and limited by the following stress values:
P L
1 II l
IV 4
Groove weld tensile stress ~y < 0.74 ~all go Groove weld shear stress ~WS< 0.60 o'all
- h. Fillet weld shear stress ~FS< 0.49 <all where all three stress limits and the allowable limit value v
11 were extracted from tables in Section VIII of the 1974 ASME Code. The shear stress limit for fillet welds was generally less than the shear stress over gross section limitation and, therefore, was conservative.
The results are based on the following:
- 1. The racks are made from Type 304 stainless steel which has a minimum yield strength (0. 2%) of 30,000 psi and a minimum tensile strength of 75,000 psi at room temperature. The values used in the stress and vibration analyses assume the pool temperature to be < 200'F which result in a yield strength of 25,000 psi and an elastic modulus of 27.7 million psi.
- 2. The trapped water for the horizontal motion occupies all the rack space at water box locations and 0.6 of the rack space at the spent fuel location.
- 3. The trapped water for the vertical motion occupies 0.3 of the rack space at the spent fuel locations only.
There is no vertical constraint at the water box location.
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4: No benefit is taken for the horizontal friction forces between the bottom of the rack leveling pads and the pool floor.
- 5. No benefit is taken for the damping effect of the water.
The actual stress values calculated and the results of the seismic vibration analysis are:
- 1. Lowest fundamental (1st mode cantilever vibration) horizontal natural frequency of fully-loaded Type A fuel rack = 36 Hz.
- 2. The spectral acceleration taken from the Seismic Response Spectra, 20% g, Fig. 5.1.2-8 of the FSAR which corresponds to 36 Hz = 0.2g.
- 3. Lowest fundamental (1st mode simply-supported beam vibration) vertical natural frequency of fully-loaded Type A fuel rack = 277 Hz.
- 4. The spectral acceleration taken from the Seismic Response Spectra, 20% g, Fig. 5.1.2-8 of the FSAR which corresponds to 277 Hz = 0.2g.
- 5. Horizontal (10 x 7) Seismic Weight = 165,289 lbs.
- 6. Vertical (10 x 7) Seismic Weight = 126,442 lbs.
- 7. Submerged dead weight = 121,520 lbs.
(neglecting buoyancy)
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- 8. Compressive Stress on N-S Seismic Restra'ints, 5 in. Schedule 80 pipe ec = 5,616 psi for 0.2g (SSE) c = 2,246 psi for 0.08g (OBE) 0.60 y = 15,000 psi 0.60 cr
= 2.47 x 106 psi
- 9. Compressive Stress on E-W Seismic Restraints, 8 in. Schedule 80 pipe
<c = 5,347 psi for 0.2g (SSE)
~c = 2,139 psi for 0.08g (OBE) 0.60 ~y = 15,000 psi 0.60 acr 6.04 x 10 psi
- 10. Bearing Stress on N-S Walls of Pool (ll" x ll" plate)
~c = 282 psi for 0.2g horizontally (SSE)
~c = 113 psi for 0.08g horizontally (OBE) ll. Bearing Stress on E-W Walls of Pool (12" x 19" plate)
~c = 299 psi for 0.2g horizontally (SSE)
~c = 120 psi for 0.08g horizontally (OBE)
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- 12. Compressive stress in East. Wall (Worst Case) of Pool due to Horizontal Seismic Loads (concrete compressive stresses) fc = 200 psi for 0.2g horizontally (SSE) fc = 81 psi for 0.08 horizontally (OBE)
Stress Limits of Concrete:
Gilbert Associates (4155 D-442-010) fc = 3000 psi minimum achieved after 28 days
- 2. 1963 ACI Building Code fc < 0.25 fc for 100% area of load application
< 0.375 fc for < 33% area of load application (0.25 fc = 750 psi; 0.375 fc = 1125 psi)
- 13. Bearing Stress on Pool Floor for 4 12"-diameter Leveling Pads under each 10 x 7 Rack
~B = 322 psi for 1.2g vertically (SSE)
~B = 290 psi for 1.08g vertically (OBE)
~B = 269 psi for 1.0g (deadweight) (Normal)
VII-7
P s 4 r
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- 14. Compressive Stress in Box Walls
~c = 780 psi for 0.2g horizontally, 1.2g vertically (SSE)
~c = 413 psi for 0.08g horizontally, 1.08g vertically (OBE)
~c = 169 psi for 0.0 horizontally, 1.0g vertically (Normal) 0.6 ~y = 15,000 psi 0.6 ~, = 7,105 psi VII-8
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VIXI~ RADIOLOGICAL EVALUATION A. Direct Radiation The principal source of radiation levels observed at the surface of the SFP water is due to the concentration of radionuclides within the pool water. This has been verified by calculations. The observed dose rate has been typically less than 5 mR/hr. The radionuclides are.
removed from the water by the SFP demineralizer with the need for changing the demineralizer resin determined by the pressure drop across the demineralizer. Increased fuel stor-age may result in an increased frequency of changing the demineralizer resin but is not expected to result in any increase in the radionuclide concentrations or in subsequent radiation levels at the surface of the water.
The top of the fuel assemblies stored in the spent fuel storage racks are at least 26 feet below the surface of the water. The'6 foot water shield reduces the direct radia-tion from the stored fuel assemblies to values which are negligible when compared to background.
In the original fuel racks, the sides of the fuel assemblies stored closest to the wall were approximately 12 inches from the concrete wall of the pool.'he new fuel racks will reduce this distance to 11.3 inches. The slight reduc-tion in distance and the closer fuel assembly spacing will result in a small increase in radiation levels outside the SFP. The resulting direct radiation levels outside the SFP wall will, however, remain below the design limits for the SFP wall.
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Therefore, for the reasons mentioned, the increased fuel storage will have essentially no impact on the radiation levels at the surface of the water or outside the SFP walls.
B. Airborne Radioactivit Increasing the storage capability of the SFP represents longer term storage of well cooled fuel. The additional spent fuel will have been stored for a year or more.
The escape of gaseous or volative fission products from any defective fuel is expected to be negligible since most of the iodines and xenons have decayed after 100 days cooling time. Kr-85 remains relatively constant because of its long half-life. The thermal driving forces required to cause Kr-85 to diffuse from the defective fuel are not present; therefore, Kr-85 is expected to remain in the old fuel. Also, there is no method for particulate fission products to become airborne. Therefore, increased fuel storage will have essentially no impact on concentrations of radioactivity in the air of the .auxil-iary building.
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IX. ACCIDENT ANALYSES A. Fuel Handlin Incident The extent of damage that might result to a fuel assembly during fuel handling is addressed in the FSAR Section 14.2.1. The new rack is inherently stronger because'f its box beam construction as compared to an open angle
'construction. Thus the above analyses are still valid.
B. Shi in Cask Dro Accident The proposed spent fuel rack modification is entirely within the outer envelope of the existing spent fuel racks. The storage capability is increased by decreasing the spacing of the stored fuel rather than by rearrangement of the pool rack configuration.
Due to the current shortage of offsite spent fuel storage space and spent fuel reprocessing capability, the spent fuel cask will not. be used for several years. An analysis of the "cask drop" accident will be submitted to the NRC prior to the use of a spent fuel cask.
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C.,Interru tion of S ent Fuel Coolin
,Nith the SFPCS in operation, the spent fuel heat load would be 5.3 x 10 Btu/hr with a 120'F pool temperature (Normal Refueling) and 9.3 x 10 Btu/hr with a 150'F pool temperature (Full Core Discharge) . The volume of water in the SFP is approximately 255,000 gallons. Complete interruption of cooling would, therefor'e, result in maximum heatup rates for the pool of 2.5'F and 4.4'F per hour, respectively. The time for the pool to reach 180'F would be 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> starting from an initial pool temperature I
of 120'F in the Normal Refueling case and 6.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> starting from an initial pool temperature of 150'F in the Full Core Discharge case. In the time available, equipment maintenance can be accomplished or backup cooling can be obtained.
The system is designed to facilitate the installation of a portable pump if the SFP pump should be lost. In the event of loss of the SFPCS heat exchanger, cooling for the SFP can be provided by using temporary connections to one of the component cooling heat exchangers.
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REFERENCES
- 1. R. F. Barry, "LEOPARD A Spectrum Dependent Non-Spatial Depletion Code for the IBM-7094", WCAP-3269, September, 1963.
- 2. W. R. Cadwell, "PDQ-7 Reference Manual", WAPD-TM-678, January, 1967.
- 3. H. Bohl, E. Gelbard, and G. Ryan, "MUFT-4 Fast Neutron Spectrum Code for the IBM-704", WAPD-TM-72, July, 1957.
- 4. H. Amster and R. Suarez, "The Calculation of Thermal Constants Averaged over a Wigner-Wilkins Flux Spectrum: Description of the SOFOCATE Code", WAPD-TM-39, January, 1957.
- 5. L. E. Strawbridge and R. F. Barry, "Criticality Calculations for Uniform Water-Moderated Lattices", Nuclear Science and Engineering, 23, 58, 1965.
- 6. "Large Closed-Cycle Water Reactor Research and Development Program Progress Report for the Period January 1 March 31, 1965", WCAP-3269-12.
- 7. "List of Equipment and Apparatus at WREC", Westinghouse Reactor Evaluation Center, February, 1967.
- 8. W. L. Orr, H. I. Sternberg, P. Deramaix, R. H. Chastain, L. Bindler, and A. J. Impink, "Saxton Plutonium Program, Nuclear Design of the Saxton Partial Plutonium Core", WCAP-3385-51, December, 1965. (Also EURAEC-1940)
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REFERENCES (continued)
- 9. R. D. Learner, W. L. Orr, R. L. Stover, E. G. Taylor, J. P. Tobin, and A. Bukmir, "Pu02-U02 Fueled Critical Experiments", WCAP-3726-1, July, 1967.'0.
A. F. Henry, "A Theoretical Method for Determining the Worth of Control Rods", WAPD-218, August, 1959.
- 11. P. W. Davison et al., "Yankee Critical Experiments Measurements on Lattices of Stainless Steel Clad Slightly Enriched Uranium Dioxide Fuel Rods in Light Water,"
YAEC-94, Westinghouse Atomic Power Division (1959).
- 12. V. E. Grob and P. W. Davison et al., "Multi-Region Reactor Lattice Studies - Results of Critical Experiments in Loose Lattices of UO Rods in H20," WCAP-1412, Westinghouse Atomic Power Division (1960).
- 13. W. J. Eich and W. P. Kovacik, "Reactivity and Neutron Flux Studies in Multiregion Loaded Cores," WCAP-1433, Westinghouse Atomic Power Division (1961).
- 14. W. J. Eich, Personal Communication (1963).
- 15. T. C. Engelder et al., "Measurement and Analysis of Uniform Lattices of Slightly Enriched UO2 Moderated by D20-H20 Mixtures,'AW-1273, the Babcock 6 Wilcox Company (1963) .
- 16. A. L. MacKinney and R. M. Ball, "Reactivity Measurements on Unperturbed, Slightly Enriched Uranium Dioxide Lattices,"
BAW-1199, the Babcock & Wilcox Company (1960). '
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