Text

Proposed Changes to Tech Specs Re Spent Fuel Storage Facility Mod
	ML19257D713
Person / Time
Site:	Prairie Island
Issue date:	01/31/1980
From:	; NORTHERN STATES POWER CO.
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ML19257D711	List: ML19257D710; ML19257D712; ML19257D713;
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	NUDOCS 8002050517
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[k PRAIRIEISLAND NUCLEAR GENERATING PLANT Red Wing, Minnesota UNITS 1 AND 2 I

/

l MlHNE APOLI5 e ST. PAUL

"* L 0l'? %

i k SPENT FUEL STORAGE LICENSE AMENDMENT REQUEST DATED JANTIARY 31, 1980 I877 041,

m. NORTHERN STATES POWER COMPANY 7@ MINNE APOUS. MINNESOTA 3 0 (l 2 0 5 0 '/"I

License Amendment Request dated January 31, 1980 Table of Contents Exhibit Subject Page A. 1. Specification 3.8.B.1 (Refueling and Fuel Handling) 1 Proposed Change, Reason for Change, Safety 1 Evaluation

2. Specifications 5.3.A.2, 5.6.A, 5.6.B (Design Features) 2 Proposed Change 2 Reason for Change 2 Safety Evaluation 3 1.0 Int roduc t ion 4 2.0 Nuclear Safety Evaluation 8 3.0 Thermal-Hydraulic Evaluation 10 4.0 Mechanical, Structural and Material Evaluation 12 5.0 Radiological Evaluation 15 6.0 Nonradiological Impact Evaluation 23

,.0 Accident Evaluation 24 8.v Procedural Impact Evaluation 25 9.0 Summary 26 10.0 References 27 11.0 Tables 28 12.0 Figures 30 B. Revised Technical Specification Pages ---

TS3.8-2 TSS.3-1 TSS.6-1 C. Nuclear Services Corporation Report QUAD-1-79-509 ---

" Licensing Report for Prairie Island Nuclear Generat ing Plant Unita 1 and 2 Spent Fuel Storage Modification" D. Nuclear Services Corporation Report QUAD-1-79-558 ---

" Fuel Pool Building Structural Evaluation for Prairie Island Nuclear Generat ing Plant Units 1 and 2 Spent Fuel Storage Modification" E. Letter, L 0 Mayer (NSP) to D K Davis (NRC), dated ---

April 14, 1977

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EXHIBIT A Prairie Island Nuclear Generating Plant License Amendment Request dated January 31, 1980 PROPOSED CHANGES TO THE TECHNICAL SPECIFICATIONS APPENDIX A 0F OPERATING LICENSES DPR-42 & DPR-60 Pursuant to 10CFR50.59, the holders of Operating License DPR-42 and DPR-60 hereby propose the following changes to Appendix A, Technical Specifications:

1. Specification 3.8.B.1 Refueline and Fuel Handling PROPOSED CHANGE Change the footnote at the bottom of the page to read -

"For the purpose of completing the fuel storage pool modifications, the movement and placement of loads shall be in accordance with the ins tallation procedure approved by the plant on-s ite review commit tee ,"

REASON FOR CHANGE The current footnote referring to hearing testimony and licensing sub-mittals will not be applicable to the forthcoming modification.

SAFETY EVALUATION The plant Operations Committee (on-s ite review commi t tee) is composed of senior plant management personnel. These personnel were res pons ible for review / approval of the procedures used for the modifications performed in 1977. Currently the Operations Committee has 8 of the 10 members originally responsible for review / approval of procedures used in the last rack modification. Thus, extensive experience and familiarity with the appropriate procedures and precaut ions exist to ensure no safety hazard exists for the procedures used in the modification.

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2. Section 5 De s i gn Features PROPOSED CHANGE A. Specification 5.3.A.2, Reactor Core Ch a nga to read:

2. The average enrichment of the reload core is a nominal 2.90 weight percent of U-235. The h ighest uranium - 235 loading is a nominal 39 grams of U-235 per axial centimeter of fuel assembly (ave rage ) .

B. Specification 5.6.A Fuel Handling Criticality Consideration In the second paragraph, change the third sentence to read:

"In addition, fuel in the storage pool shall have a U-235 loading of < 39.0 grams of U-235 per axial centimeter of fuel assembly (average)".

Delete the following from the last sentence in Specification 5.6.A -

"except for initial naw fuel s torage".

C. Specification 5.6.B, Fuel Handling Spent Fuel Storage Change the first paragraph, first sentence to read:

"The spent fuel storage f acility is a two compartment pool that may contain up to 1582 storage locations for spent fuel as semb lies". >

REASON FOR CHANGE Specifications 5.3.A.2 and 5.6.A are related in that spent fuel criticality cons ide rat ions establish the upper bound on tne U-235 loading. From a conse rva t ive standpoint, the reactor core should not have a higher loading than the spent fuel pool requirement since one should always assume that full core discharge of slightly used fuel (which would be the highest enrichment) might be required.

Northern States Power Company has used or intends to use fuel produced by Westinghouse Electric Corporation and Exxon Nuclear Company. The fuel supplied by these two vendors has dif ferent clad thicknesses. In ';P orde r t o ach ieve the same optimum fuel burnup, these suppliers use 'f' different enrichments that would correspond to 39 gm U-235/ axial em '3 fuel ascembly length, the loading assumed in the safety analyses. ,

ts-Th e reference in Specification 5.6.A to initial new fuel storage is cr) no longer appropriate. . - - -

Specificat ion 5.6.B must be ch anged since Northern St at es Power Company intends to increase the available spent fuel storage locations from the cur re nt 687 to 1582. This change sbould provide adequate storage capacity until either an AFR (away from reactor) storage facility or reprocessing facility is built that can provide a location to ship the Prairie Island spent fuel assemblies.

SAFETY EVALUATION On April 14, 19 78, the NRC Staf f distributed the document "0T Pos it ion for Review and Acceptance of Spent Fuel Storage and Handling Applications" (Re f e re nce 1). This safety evaluation provides the informat ion recommended by that document. Reference will be made to Exhibits C and D, Nuclear Services Corporation documents specifically prepared to address selected areas of this amendment submittal.

This safety evaluation addresses the following areas -

1. Nuclear Evaluation

2. Thermal Hydraulic Evaluat ion

3. Mechanical, Material, and Structural Evaluation

4. Environmental Ef fects Evaluation (Cost Benefit, Radiological, Accident) i

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1.0 INTRODUCTION

The Prairie Island Nuclear Generating Plant ( PI NG P) fuel storage facility was previously described in sections 9.3 and 9.5 of the Final Safety Analysis Report (FSAR). The fuel storage facility has two spent fuel storage pools as shown in Figures 1 and 9, which have a combined storage capacity for 687 spent fuel assemblies.

Since Spring 1976, forty spent fuel assemblies have been transferred annually from each reactor to the spent fuel pool for interim storage.

There are currently 320 assemblies stored in Pool 2. At this rate, Pool 2 will be filled during the 1983 Unit 2 Spring refueling outage (as shown in Table 1). Thus spent fuel would have to be placed in Pool 1.

The full core-off-load capability will be lost after the 1983 Unit I refueling outage. The cur rent storage capacity limit will be met after the Fall 1984 refueling outage.

An Away From Reactor storage facility or reprocessing facility is not expected to be operational by the time when the PINGP spent fuel storage pools would be full. For this reason, Northern States Power (NSP) plans to increase the storage capacity of the current fuel storage f acility from 687 to 1582 storage locations.

1.1 Storage Needs/Contractural Arrangements Northern States Power Company had a fuel reprocessing contract with Nuc lear Fuel Services of West Valley, NY since December 31, 1965.

This contract initially covered the Pathfinder plant, but was amended on May 7, 19 70 to include the Prairie Island and Monticello plant s.

On Septemaer 20, 19 76, NFS not i fied NSP that the company was withdrawing from the nuclear fuel reprocessing business. Since no other facilities were available and there was inadequate storage space, a modification was applied for on November 24, 1976 and authorized on August 16, 1977. This original modification expanded the storage capacity to the current 687 storage spaces.

The first Prairie Island reactor refueling occurred in March 1976.

After the initial cycle, each Prairie Island unit has operated at ve ry high availability, result ing in one refueling per unit per year.

Thus, 80 fuel assemblies per year have been added to the spent fuel pool resulting in the current number of 320 spent fuel assemb lies . In addition to spent fuel, there are 1 Rod Control Cluster Assembly (RCCA), 2 Source assemblies, 100 - Burnable Poison Rod Assemblies (BPRA's), 8 Part Length Rod Control Cluster Assemblies (PLRCCA's), and 11 Thimble Plugs stored in the storage pool. These components are inserted into spent assemblies to reduce the storage space consumption.

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1.2 Storage Modification Plans There are three 7x7, six 7x8, thre< Sx8, and one 3x4 storage racks presently in the pools. Northern States Power Company plans on reracking the small pool (Pool 1) in February 1981 with the nine racks (as shown in Figure 2), which would have a storage capacity for 462 assemblies.

All spent fuel assemblies (400 expected) would then be transferred to the small pool . So also would the spent assemblies from the 1981 Unit I refueling outage. The modification is being scheduled for early 1981 because enough time must be available to procure materials, fabricate new racks, remove the existing racks and install all of the new racks before the Unit 2 1982 refueling. With that outage, 482 spent fuel assemblies would have been discharged which is greater than the new Pool I capacity. After the modification, all spent fuel assemblies would be trans ferred to Pool 2. Th e four 7x7 racks on the southeast corner of Pool 1 would be removed if a cask is brought into the SFP facility for fuel shipment. Th is wou l d re s u l t in a storage configuration as shown in Figure 3. These four racks would only be used if a full core of f-lead is required. Thus the result ing storage capacity for normal refuelings would be 1386 spaces.

1.3 Construction Costs The estimated installed cost of the proposed modifications to the Prairie Island f acility is $5,230,000 (1980 dollars). Key costs are broken down as follows:

Material $2,930,000 Labor 830,000 Permits and Licenses 340,000 Escalation, AFC, A&G, Engineering 1,130,000 Construct ion Supervision, AE for both installation of the new racks and removal of the old racks.

Considerat ion has also been given in the estimate to escalation and depreciation costs.

1.4 Alternatives to Increasing Storage Capacity NSP has considered the alternatives -

(1) Ship fuel to a fuel reprocessing facility (2) Ship fuel to an Away-From-Reactor ( AFR) storage facility (3) Ship fuel to another reactor site (4) Shut down the reactors Spent fuel is currently not being reprocessed in the United States on a commercial basis. None of the three commercial reprocessing facilities in the United States is currently operating. The General Elect ric Morris, Illinois aand Nuclear Fuel Service West Valley, N.Y. facilities are both in a decommissioned status. The Allied General Nuclear Service (AGNS) facility at Barnwell, South Carolina has not been licensed to operate.

-S-

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The Morris facility owner is primarily using the storage space for GE-owned fuel (which had been leased to utilites) or for fuel which GE had previously contracted to reprocess. GE as a matter of policy will not store spent fuel unless they had previously committed to reprocess that fuel. There is ,o such commitment for Prairie Island.

The West Valley facility has capacity for about 260 FfrHM, with approxi-mately 170 FfrilM presentiv stored in the pool. Although the storage pool at West Valley is not full, of ficials have indicated they are not accepting additional spent fuel for storage, even from those reactor facilities with which it had reprocessing contracts, e.g. Northern States Power Company (as discussed previously).

The Allied General Nuclear Service facility has not been licensed to operate and cannot be depended upon for receipt of spent fuel due to the indefinite deferral of licensing proceedings for the plant.

Department of Energy (DOE) plans for AFR storage have not been finalized nor has funding been received. Thus it is unlikely that DOE's 1983 target will be met.

The indefinite deferral of reprocessing has left potentially all commercial nuclear reacter operators with limited storage capacity and no place to ship spent 'uel. Other utility neclear plant operators are unwilling to risk prematu.e shutdown of thei r own plant because of storage of off-site fuel generated by another facil'ty.

Shipping the Prairie Island spent fuel for storage in the Monticello spent fuel pool has been considered. The fuel asse-blies for the two plants dif fer significant ly. Because of this, the fuel handling and storage equipment for fuel assemblies from the two plant s are not c om p at ib le . Storage of Prairie Island fuel at Mont icello has been rejected for the following reasons:

1. A substantial modification of the Mont icello fuel handling and storage equipment would be re q u i red .

2. The storage space at Monticello is needed to ensure continued operation of the Mont icello plant.

3. Storage of Prairie Island fuel at Monticello would result in increased handling and shipping of spent fuel.

Thus the alternat ives that involve fuel shipment from the Prairie Island Plant are not viable at this time nor in the immediate future when the PINGP spent fuel pools would be full.

NSP aoes not consider shuttic; down PINGP a viable option for the follcwing reasons.

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Operation of the Prairie Island Plant could be in jeopardy following loss of full core reserve in 1983. Even if NSP were successful in avoiding full core discharges in 1983 and 1984, both units at Prairie Island will have to be shut down by 1985 if additional spent fuel storage is not provided. Assuming knowledge of an impending long term shutdown occurs in early 1980, a new generating facility to replace Prairie Island could not be placed in operation until 1989 or 1990.

The ef fects of a shutdown were examined for the year 1985. Without Prairie Island, NSP's remaining generating capacity, including oil generating f acilities, would barely equal the ant icipated peak demand.

In addition to decreased reliability of its power supply system, NSP's electric production expenses for the twelve month period examined are projected to increase by approximately $160 million if Prairie Island is not available. Most of this expense would be caused by the increased utilization of NSP's coal-fired and oil-fired generating f acilities which would have to supply much of the energy that would have been produced by Prairie Island. The projected increase in oil consumption would Le approximately 40 million gallons, wh ile coal consumption would be projected to increase by over 3 million tons in the initial twelve month period. Increased purchases from other utilities to offset the deficit caused by the shutdown of Prairie Island also contribute to the increased electric production expenses.

Beyond 1985, the electric production expenses will continue to in-crease due to escalation of fuel prices and other operating costs, until re p l a ceme nt generating capacity is in service. Expected load growth in 1986 will cause additional increase in electric product ion expenses since there will not be a corresponding generating capacity addition in that year. Existing generat ing capacity additions scheduled for 1987 and beyond will keep pace with load growth, but will not re place the Prairie Island capacity.

1.5 Material Resource Ef fects For the new spent fuel storage racks the material resource commitment would be:

Material Weight Stainless steel 273 tons Silicone polymer 32,300 lbs Boron carbide 31,900 lbs not including usage attributable to manufacturing tolerances. Th is usage does not constitute a significant resource commitment since the materials are not in short supply.

The silicone polymer and boron carbide are used in the form of Boraflex, previously authorized by the NRC for use at the Point Beach.and Nine Mile Point facilities.

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2.0 NUCLEAR SAFETY EVALUATION Exhibit C, Section 3.3, describes the analyses conducted to verify that Keff < 0.95 under a variety of conditions. Included are (1) normal storage and (2) postulated accident conditions. Sect ions and tables described below refer to Exh ibit C, unless otherwise noted.

2.1 Design Criteria Section 3.3.1 describes the design criteria (fuel design, Keff limit, standard review plan compliance, and temperature and fuel position cons ide rat ions ) .

2.2 Analytical Methods Section 3.3.2 describes the analytical method used to evaluate un-cert ainty considerations including: (1) t rans port correction, (2) method bias, (3) fuel location ef fects, (4)i5t r Re tube pitch tolerance effect, (5) methods bias uncertainty, (6) B tolerance ef fect , (7) storage tube dimens ional e f fect . This section discusses the calcula-tional model and computer code used to determine Kef f, as well as the comparison of code calculations with actual experimental data.

2.3 Normal Storage Case Results Section 3.3.3 ontains the analytical results for the normal s torage case (including ef fects of uncertainties). Tables 3.3-4 and 3.3-5 show the ef fects of storage tube pitch and pool water temperature on Ke f f, res pect ive ly.

2.4 Postulated Accidents Section 3.3.4 describes the evaluation of the postulated accidents specified by Reference 1 Sect ion III 1.2. As noted in sect ion 9 of the FS AR, the spent fuel pool structure is designed as a Class I structure chat meets seismic and tornado criteria in Appendix B of the FS AR.

2.5 Conclusicts Section 3.3.5 provides the conclusions that proposed design meets the specified criteria of Keff < 0.95 for the postulated conditions.

Thus the Section 3.3 description addresses the section III 1.1-1.4 requirements of Reference 1.

2.6 Acceptance Criteria for Criticality Since the new spent fuel racks will contain baron as the neutron absorbing medium, on-site verification tests will be conducted to ensure within 95% confidence limits that the neu*ron absorber is present in the racks.

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A separate check of each tube will be conducted. Recalibration will be performed on a periodic basis to assure adequate equipment calibration.

A device having a neutron source and thermal neutron detectors would be lowered ir.to the storage tubes. This device would monitor the ef fect iveness of the borated sheets at absorbing the neutrons.

Sample coupons will be removed every five years and tested to confirm that the aeutron absorber condition has not changed.

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3.0 THERMAL-HYDRAULIC EVALUAfION 3.1 Decay Heat Calculations The calculations for dotarmining the amount of thermal energy that would be removed by the spent fuel pool cooling system were conducted in accordance with Branch Technica1 Position APCSB9-2 entitled " Residual Decay Energy for Light Water Reactors for Long Term Cooling". These calculations are summarized in Sect ion 3.5.2 of the Exh ibi t C report.

3.2 Analytical Results Sect ion 3.5 of the Exh ibit C report addresses th e following - (1)

Fuel Assembly Cooling Analysis (2) Pool Cooling Analyses. These analyses assume that no fue' assemblies are reme,ved from the core until 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months have elapse after reactor shutdown (in accordance with Technical Specificat ion 3.8. A.7) . For both the normal refueling

=d f ull cute di::ch arga racee, the complete transfer of fuel assemblies is assumed to be completed by 150 hours0.00174 days 0.0417 hours 2.480159e-4 weeks 5.7075e-5 months after reactor shutdown. This assumption is conservative. Based on the eight refuelings conducted at Prairie Island to date, at least 150 hours0.00174 days 0.0417 hours 2.480159e-4 weeks 5.7075e-5 months after shutdown is ren uired f or normal re f ue lings .

3.3 System Description

The spent fuel cooling system is described in Section 9.3 of the FSAR.

Figura 4 che m a sketch of the existing system. The capacity of the SFP cooling pumps has b.!en increased to 2200 gpm.

Haat ic t rano le t t ed t rom t he s pent fuel cooling system fluid to the component cooling system in the SFP heat exch ange rs . Either unit's component cooling system may be used to remove the heat from the spent fuel cooling system. If an event occurs wh ich requires safety injection, component cooling will continue to be supplied to the SFP heat exchangers. The operator would transfer the spent fuel cooling load to the alternate unit 's component cooling system in the eve nt of a LOCA to reduce the heat loads on the component cooling system of the af fected unit .

This modification does not affect the largest heat Icad (due to full core discharge) which is nominally 58% of the total spent fuel cooling heat load at design conditions. The increased heat load, due to those spent fuel assemblies stored for longer periods of time, can easily be accommodated by the component cooling system design. Compared to the maximum g heat rejection rate from the plant to the cooling tower water, 8.37x10 BTU /hr, (Reference 4), the increased heat rejected would be insignificant.

3.4 System Indications The water in the spent fuel pools serves a two-fold funct ion - (1) provides adequate radiation shielding and (2) acts as a cooling medium to remove decay heat from the spent fuel assemblies. Three systems have,been provided to alert the operator to the existence of unusual conditions (in addition to those described in Section 5). These .

are the SPP level and temperature detection systems and the SFP leak detection system as described below.

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Spent Fuel Pool Level Level detectors are provided for both pools. These sensors input to provide high and low leve. alarms for each pool on each unit 's control board alarm panel (8 total). Since the high level alarm is set at 753' 8" (nom i nal), and the low level alarm is set at 751' 6" (nominal), the operators will be alerted promptly so that correc t ive action can be taken to eliminate the cause of the alarm.

Spent Fuel Pool Temperature Temperature detectors are provided for both spent fuel pools. A h igh t emperature alarm, nominally set at 130F, is provided for each pool on each unit 's control room alarm panel (4 total) .

Spent Fuel Pool Leak Detection The spaces between the spent fuel pools' stainless steel liners and the concrete support structures are divided into twelve (12) sections.

Any leakage into these areas is routed via a se arate pipe to a common open sight drain trough and then to the Waste Disposal systems.

3.5 Operational Monitoring The auxiliary buildir.g operator monitors SFP level, radiat ion, temperature, and leak detection systems periodically as a routine shift responsibility. Anomalous conditions are reported to the control room operators and shift supervisors. In addition, the control room operator reviews SFP radiation monitor levels as a normal shift responsi-bility. Unusual level and temperature are alarmed on both units' cont rol board panels, as described in section 3.4.

3.6 Conclusions The analytical results demonstrate that the existing systems will satisfactorily handle the additional heat load. The FSAR design basis (normal refueling conditions) SFP temperature was 120F. The Exhibit C Sect ion 3.5 analysis demonstrates that with the larger SFF heat exchanger and one SFP cooling pump in service that the maximum t emperature expected would be 124F. With two pumns and heat exchangers in service this temperature would be lower. The rate of evaporation and thus the need for makeup water is not expected to be changed by the proposed modifications. (Reference 3) 1877 053

4.0 MECllANICAL, STRUCTURAL, AND MATERIAL EVALUATION 4.1 Description of the Spent Fuel Pool and Racks The SFP storage facility and systems were described in the FSAR Sections 9.3 and 9.5. Figures 5-9 of th is report show the location of the spent fuel storage facility in relation to other plant structures.

No changes in the plant structures are planned. Exh ib i t C Figure 3.1-1 shows the typical 7x8 rack structure. Exh ib i t C figure 3.3-1 s' ows the normal 2 nit cell configuration. The neutron absorber shown wxtl be in the form of Boraflex, previously authorized by the NRC for use at the Point Beach facility. (Reference 2) 4.2 Structural and Mechanical Analvses Exhibit C, Section 3.4 describes the mechanical and structural analysis for the new spent fuel racks.

Exhibit D contains the structural evaluation for the spent fuel storage pool facility. That report addresses the following areas:

1. Loads, load combinations, and acceptance criteria

2. Method of analysis and comr utat ion of design loads

3. Evaluation results and cor clusions.

4.3 Small Pool Protective Cover A report describing the structural evaluation for the small pool prot ect ive cover was provided by letter, L 0 Mayer (NSP) to Don K Davis, Acting Chief ORB #2 (NRC) dated April 14, 1977. This report provided a system description and discussion of loadings, allowable stresses and design procedure, and design result s. Th i s report is included as Exh ib i t E.

4.4 Material Evaluation The stainless steel sheath provides support for and protect ion of the Boraflex. Gases that might be generated by irradiation of the Boraflex are vent ed to prevent bulging and swelling of the stainless steel shrouds.

No evidence has been determined that indicates any significant deterior-ationo{gBoraflex through a cumulat ive irradiat ion in an excess of lx10 rads gamma occurs to effect the suitability of Boraflex as a neutrou shielding material Testing to date of Boraflex in a neutror. and gamma field indicates that Boraflex is suitable for the proposed use.

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All permanent structural material exposed to the spent fuel pool environment that is used in the fabrication of the spent fuel storage racks is 304 stainless steel with the exception of the leg adjusting bolt s wh ich are 17-4Ph stainless steel. These materials were chosen for compatibility with the spent fuel pool water.

At the design operating temperature of 120 F, there is no deterioration or corrosion of stainless steel in this environment. There is also no corrosion problem at temperatures up to and including pool boiling.

All other structural components in the spent fuel pool system, such as the pool liner, cooling system pipe connections, etc., are made of stainless steel.

The high density spent fuel storage racks to be used at PINGP will ut ilize Boraflex sheets within an inner and outer stainless steel clad. The cells will be vented to prevent bulging and swelling of the structural steel. The stainless steel clad is Type 304 meet ing the requirements of ASTM A 240 The Bora flex ggeets will be demonstrat ed, at a g5% confidence level, to have a minimum B content of 0.04 gn/cm of sheet surface area. The Boraflex is designed to operate in a 2100 ppm nominal boric acid concentrat {on, normal pool t amperatura 80F-140F and tot al radiation exposure of 10

Rads (gamma).

In summary, the pool liner, rack lattice structure, and cell extericrs are all stainless steel, which has demonstrated good corrosion resistance in PWR spent fuel pool environments. The design, material nelect ion, and the NDE program provide a high degree of assurance that the integrity of the fixed absorber material will be maintained. Th e material used in the new spent fuel storage racks is similar to present components and does not affect or alter previous evaluations.

4.5 Neutron Absorber Verification Programs Close control and verification of the material properties utilized in the manufacture of the Boreflex is assured through the manufacturer's Quality Assurance Program and is documented on appropriate material certification reports. Prior to inserting the Boraflex sheets into the finished cell configuration, each sheet is identified in order to allow traceability to the end product. Records are generated for each cell ident ifying the plates installed in that cell by serial number, thereby providing positive assurance that the required plates .are in place.

During rack fabrication, addit ional care is exercised to prevent damage to the stainless steel cladding of the fuel storage cells.

Packaging and shipping will be done in accordance with approved procedure to minimize the possibility of degrading the quality of the racks during transit. A thorough receipt inspection at PINGP is performed to assure no damage has occurred.

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Documer.tation is maintained on all testing and surveillance performed on the fuel storege cells as well as material certification reports on all materials used in the construction of the cell.

4.6 In-Pool Surveillance Program Surveillance specimens are provided to allow for surveillance over the lifetime of the fuel storage racks. The purpose of these specimens is to provide assurance that no unexpected corrosion is occurring which could compromise the integrity of the Baraflex. The surve 11ance specimens are in the form of removable stainless steel clad Boraflex eheets, which are typical of the fuel storage cells. These speci-mens can be removed and examined.

4.7 Conclusions The fuel pool structures are structurally adequate to withstand additional loads that would result from the proposed modification in ahich the present racks will be replaced by high density racks.

The Borafic: material will provide the required neutron absorption and based on test data to date will perform satisfactorily in the intended environment. The stainless steel used in the modification will provide the required structural integrity.

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5.0 RADIOLOGICAL EVALUATION We have evaluated the radiological impact of this modification.

This section addresses considerations of

1. Radioactive Waste

2. Radiation Exposure and describes the systems used to reduce the likelihood of spent fuel radionuclides reaching the environment and the systeme used to monitor radionuclides and radiation levels associated with the spent fuel storage.

5.1 Impact on Wastes 5.1.1 Solid Waste The thirteen (13) existing spent fuel racks will be removed from the spent fuel pool and disposed as low level radwaste at licensed disposal sites. Based on prior experience., the crevices in the racks do not make it practical to deconte*inate the structures to levels low enough to recycle the staintess steel. The total solidragwaste, generated by this modification, is expected to be 15050 ft , most of which is the old spent fuel racks. The radio-act ivity content has been conservatively estimated to be less than 270 Ci. The total estimated weight of the racks and associated structural waste is 310,000 lbs.

With ag estimated radioact ivity of less than 40 Ci, approximately 132 ft of solid r. dioactive waste are produced by operation of the spent fuel facility annually. There is not e gect(d to be a significant increase due to this modification. The solid waste generated by the spent fuel pool cleanup system represents approxi-tuately 2% of the totajvolumeshippedfromtheplant. In 19 78, approximately 6870 ft of solid radioactive waste was shipped f rom all sources at the facility.

Typically the frequencies for replacing cleanup system filtration com po ne nt s a re -

SFP Filter 1/yr SFP Skimmer Filtet 5/yr SFP Demineraliz<r 2/yr 5.1.2 Liquid Waste The only radioactive liquids result ing from the construction act ivity are the evaporator distillates from processing the mop water and SFP resin sluicing. Th is is estimated to be less than 3000 gallons, which af ter further processing through a demineralizer, will have a radioact ivity content of less than 12 uCi and less than 1.25 Ci tritium.

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There should not be an increase in the normal liquid wastes due to th is modification since the spent fuel cooling system operates as a closed system with removal of soluble ionic and insoluble particulates by the cleanup system. Water originnt ing f rom cleanup of SFP floors and sluicing of SFP resins are estimated to be less than 1000 gallons per year. The associated radioact ivity content is less than 4 uCi and less than 0.4 Ci tritium.

Existing liquid ef fluent technical speci fical ions provide assurance that liquid releasas are enveloped by the radioact ive liquid waste design basis.

5.1.3 Gaseous Wastes No radioactive gases are expect ed to be generat ed wh ile performing th is modi ficat ion that are ove r and above those associated with normal use of the SFP.

Gaseous vaste from the SFP during operation is primarily t rit ium evaporated from the pool surface and released through the normal SFP ventilation system. In 1978, the tot al plant gaseous tritium release was 143.4 Ci, mos t of which can be assumed to come from the SFP.

The only significant noble gas isotope remaining in the spent fuel is Kr-85 (Reference 7). Th is modificat ion is not expected to have any significant on-site or off-site consequences due to Kr-85

( re fe re nce s 2, 3). There have not been detectable releases of Kr 85 from the SFP to date.

On occasion, a small quant ity of I-131 has been introduced into the SFP water from makeup water and subsequently released to the a t mos phe re . In the first aix mont hs of 19 79, this amounted to 1.143 mci, approximately one pe rce nt of the plant design object ive for 1-131 releases.

5.2 Radiation Exposure 5.2.1 _ Mod i f ic a t io n in the 19 77 rerack, the man-rem exposure associated with poolside work, removal and disposal of the old racks, and installation of the new racks was 4.54 man-rem. For this re rack , the overall radiat ion expoaure associated with removal of the old racks, movement of spent fuel current ly in the SFP, inst allat ion of the new racks, and pre-paration of the old racks for shipment is conservatively estimated to be less than 40 man rem. Based on experience with these work tasks, it

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is anticipated that the actual accumulated dose will be significantly less than the estimate. This occupational exposure will not be incurred on a continuing basis. The exposure is small compared with the total plant occupational exposure over the useful life of the modification (approximately 200 man-rem annually).

5.2.3 Annual Exposure Based on experience to date, the expected annual man-rem expasure due to all operations associated with spent fuel pool area related activities, is as follows:

Total Operation Man-hours Exposure (Man-Rem)

Fuel Handling 600 4.20 Equipment Checkout & Maintenance 140 0.54 System Maintenance 60 1.20 Radwaste handling 16 0.80 Cleanup 100 0.20 Total 6.94 5.2.3 Internal Exposure Operations in the spent fuel pool area have not contributed any me asur ab le internal doses to plant personnel and are not expected to be a significant source of internal doses in the future.

Periodically the health physics staf f conducts whole body counting of personnel who may be exposed to airborne radioactivity. Intake of airborne radionuclides is measured by this process.

5.3 Measurements 5.3.1 Radiat ion Levels The measured approximate radiation levels near the spent fuel pool are as follows:

Location Dose Rate (mrem /hr)

On SFP Crane above pool area 10.0 At handrail around pool 2.0 Other areas 1.0 60 The major radionuclides contributing to this exposure are Co & Co .

_1,_ 1877 059

The radiation levels experienced at the surf ace of the spent fuel pool are due primarily to the radioact ive contaminents in the pool water. Measurements conducted at the Morris storage facility for PWR fuel, similar to the Prairie Island fuel, are shown in Figure 10.

These show that for water levels more than eight feet above the fuel there is no significant change in radiat ion level due to the fuel.

The normal surface level at Prairie Island is 25 feet above the fuel; thus there would have to be a significant level change before any change is expected in the radiation level.

5.3.2 Airborne Radioactivity Samples from the SFP area usually only contain tritium as the detect able nuc lide. Weekly and monthly integrated radioactivity samples are taken from the SFP ventilation system. For 19 79, the following radioactivity releases were measured:

Sample Nuclide Total Activity, Ci Frequency IHf33 * "" 7 Xe 135 35.3 Weekly X

f31 *

Y I 1.16 E-3 Weekly Particulate 0 Weekly To our knowledge, there has not been detectable Kr in the SFP air. These findings are consistent with those presented in Reference 7, where the NRC reported that experience at the NFS West Valley reprocessing plant showed that almost all of the Krypton was retained in the fuel until its dissolution during the reprocess-ing operation.

5.3.3 Measured SFP Liquid Activity The following is a summary of SFP liquid activities (in microcuries per milliliter) during different periods of operation:

Before After Recent Radionuclide Refueling Refueling Data Cs 4.08E-04 *

Cs 4.01E-04 *

Sb 4.09E-04 5.04E-03 5.IlE-04 Co 2.25E-03 2.16E-02 5.32E-04 Co 3.38E-03 4.05E-03 1.98E-03 Na 4.18E-05 1.13E-04

Co

7.76E-05

Sb

3.12E-04

Notes 1 Data taken 27 March 79 prior to Unit 1 Reload 4 Refueling

2. Data taken 1 May 79 af ter the Unit i Reload 4 Refueling

3. Data taken 29 August 79

Not detected Sample act iviti*es are determined using a Ge-Li detector with mult ichannel analyzer (4096 channels) .

_18 i877 060

5.4 Radioactivity Control Systems Three systems provided at the Prairie Island Nuclear Generating Plant reduce the likelihood of spent fuel radionuclides reaching the environ-ment:

(1) Refueling cavity cleanup system (2) Spent fuel pool cleanup system (3) Spent fuel pool vent ilation system 5.4.1 Refueling Cavity Cleanup System During refueling operations, the re fueling cavity cleanup system (Figures 11 and 12) is used to remove suspended solids. This system substant ially aids ir. prevent ing crud buildup in the spent fuel pool so that there currently is no noticeable buildup of crud along the sides of the spent fuel pool.

Th is system consists of a single pump which takes a suction from low in the refueling cavity and discharges through filters (in a parallel or series parallel mode) back to a higher elevation in the refueling cavity.

The filters are changed whenever radioactivity levels reach 3 R/hr or the filter dp exceeds 25 psid, conditions which indicate possible red uc t ion in filter efficiency.

5.4.2 Spent Fuel Pool Cleanup System The spent fuel pool cleanup system is a bypass loop around the spent fuel pool heat exchangers consisting of a demineralizer, several filters , and associated piping, valves and fitt ings. A skimmer loop augments the cleanup loop and consists of a skimmer in-take, strainer, pumps, filter, piping, valves, and fitt ings. The cleanup and skimmer loops provide turbidity and soluble radioactivity control via the filters and demineralizer, re s pe c t ive ly . Th e systems are shewn in Figures 13 and 14.

Two spent fuel pits have combined volume of 355,617 gallons. One pit cont ains approximately 2-1/ 2 t imes the water volume capacity of the other pit. Approximately 60 gpm of pool water is diverted, when necessary, through the cleanup loop. One spent fuel pool water volume can therefore be processed in less than 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months .

The spent fuel pool filter is a cartridge type filter designed to retain 98% of particles of sizes down to as small as 5 microns.

The filter is designed for a maximum differential pressure of 5 psi at rated flow. Local pressure indicators are located upstream and downstream of the filter to monitor loop pressure.

Two parallel pre-filters, normally operated singly, located up-stream of the demineralizer, are designed to reduce the effect of insoluble part icles on the demineralizer operation.

-'9~

. l877 06F ,

The demineralizer is a deep bed H-Oli type containing 20 cubic feet of res ins. Spent resins are flushed to a resin storage tank of the waste disposal system. Demineralizer resins are changed when pool water samples indicate reduced decontamination ef fectiveness.

The spent fuel pool water is sampled to determine radioactivity levels. The purification loop is placed in service whenever a significant concentration of radioactive materials exist. The purification loop is also placed in service during any refueling ope ra t ions .

Radioact ivity and turbidity in the fuel pool wcter are caused by the release of crud buildup on spent fuel assemblies. Operating experience has shown that the greatest quantity of crud is released when loose deposits are dislodged during movement of spent fuel assemblies. Once the handling of spent fuel assemblies is finished the addition of materials and radioactivity to the pool water is greatly diminished.

Failed fuel, if present, may add some fission product contamination to the spent fuel pool. This contaminat ion is proport ional to the amount of failed fuel present. This addition of contamination is much smaller than the crud added to the pool by refueling operations.

Consequently, a change in the capacity of the fuel pool does not overburden the cleanup system which has been sized to remove the impurities resulting from refueling operations.

No changes or equipment addition to the cleanup loop are necessary for the augmented storage facility. As no change in refueling frequency is anticipated, the frequency of operation of the cleanup loop is not expected to change.

The redesign of tqe SFP racks increases only the storage capacity of the pool and not the frequency or the amount of the core to be replaced for each fuel cycle. Thus, the amount of corrosion product nuclides released into the pool during any year will be about the same regardless of the length of time or number of assemblies stored in the pool.

5.4.3 Spent Fuel Pool Ventilation System The normal and special spent fuel vent ilation systems are described in Section 9.6 of the FSAR. The fuel handling accident, (described in Section 14.2.1 of the FSAR), which forms the design basis for the special ventilation system, is not af fected by the expansion of the fuel pool. The normal vent ilation system has no safeguards functions and is designed to isolate upon detection of high radiation in the fuel pool area.

1877 062

5.5 Radiation Monitors The radiation monitors used at Prairie Island are described in Section 11 of the FSAR. The following is a summary of those process and area radiation monitors appropriate to this modification.

5.5.1 Process Radiation Monitors Radiation process monitors (R-25 and R-31) are installed in the SFP normal ventilation exhauat ducting. Upon sensing high radionuclide conce nt ra t ions in the air leaving the spent fuel pool area, the SFP normal ventilation system (Figure 15) shuts down and SFP special vent ilation system starts which discharges through the spent fuel special and inservice purge PAC filters to the shield building ex-haust stack (Figure 15).

5.5.2 Area Radiation Monitors A criticality radiation monitor (R 28) is installed on the spent fuel /new fuel storage facility operating deck. This monitor would alarm in the unlikely event of criticality in the area.

Six additional radiation area monitors are installed in close proximity to the spent fuel storage facility that could alert the operations staff that abnormal radiation conditions exist. All of these radiation monitors, alarm locally and in the control room, so th at correct ive action may be taken promptly. These monitors are R5, R8, R29, R32, R33, and R34. RS, the spent fuel storage area monitor, is mounted on the operating deck of the storage facility.

R8 is mounted in the waste gas valve gallery area below the storage facility.

R29, 32, 33, and 34 are mounted in the shipping and receiving area and the three floors of the radwaste building. G-M and scintillation detectors provide diversity in gamma detection.

If any of these area or process radiation monitors alarm, there is an individual qualified in radiat ion protection procedures on site to direct the appropriate corrective actions.

All of these alarms provide defense-in-depth to a design which assures th at Keff < 0.95 even if pure water and new fuel are used in the pool.

5.5.3 Continuous Air Monitor (CAM)

A portable CAM unit is normally located in the area near the SFP vent ilation system to detect airborne part iculates and iodine. Th is unit is normally monitored by the Auxiliary Building operator during routine shift equipment checks.

1877 063

. s

5.6 Radiation Protection Practices The PINGP radiat ion protect ion program is formulated to maintain radiation exposures ALARA ( As Low As Reasonably Achievable). Work in the controlled area, of which the SFP is a part, is governed by radiation wo rk pe rmi t s . The radiat ion work pe rmit (RWP) is used to ide nt ify prot ec t ive req uireme nt s , such as Anti-C clothing, respiratory equipment, dosimetry, special instructions, etc., for work in the controlled area. The radiation protect it; specialist prepares th e RWP for jobs performed in the controlled area and attemots to assure that the radiation exposures will be ALARA. In addition, all plant personnel are recommended to look at their work at the plant in order to maintain exposures ALARA.

Utilizing theae principles has helped and will continue-to help in ensuring that the man-rem exposures associated with this modification are ALARA.

5.7 Technical Specifications The PINGP technical Specifications, Section 3.9, impose limits on releases of radioactive effluents. These specifications include releases associated with the operation of the SFP storage facility and provide additional assurance that there will not be any signifi-cant impact on of fsite radiat ion exposures as a result of this modification.

5.8 Conclusions Sections 5.1-5.7 describe the review of the radiological impact of th is modificat ion. As a result, we conclude that there will not be a significant impact on radiation exposures either onsite or offsite.

l877 o64

6.0 Nonradiological Impact Eveleation 6.1 Nonradiological Effluents There will be no change in the chemical or biocidal effects from the plant as a result of the proposed modifications.

Thermal impact of this modification was addressed in Section 3 of this safety evaluation.

6.2 Impact on the Community The new racks will be f abricated of fsite and shipped to the plant .

No environmental impact on the community is expected to result during or af ter completion of this modification.

9 1877 00

7.0 Accident Evaluation 7.1 Eleavy Loads Analysis The existing technical specifications preclude handling of any heavy loads over or in either spent fuel pool when fuel is stored in that pol . The consequences of fuel handling accidents are unchanged from those presented in the FSAR. For the purpose of completing the modification, the footnote added at the bottom of page TS.3.8-2 allows movement and placement of loads described in the installation procedures for this modification as described in Exhibit C. The small pool covers are designed to sustain a heavy load drop (see Exhibit E) without af fecting the fuel in the small pool.

7.2 Fuel flandling Accidents The FSAR Section 14.2.1 addresses fuel handling accidents. The conclusions presented in that report are unchanged by this modifica-tion. Fuel assembly drop accident s are also addressed in Section 3.3.4 of Exhibit C.

7.3 Cask Drop Accidents The FSAR sect ion 9.5 provides a description of the cask drop accident.

Cask drop is also described in Sect ion 3.3.4 of the Exh ibit C report.

7.4 Conclusions The existing technical specifications and analyses provide assurance that this modification will not adversely affect the public health and safety.

1877 066

8.0 Procedural Impact Evaluation The fuel rack installation sequence and a summary procedure are described in section 3.7 of the Exhibit C report. The fuel handling procedures as previously reported in the FSAR Se.ction 9.5 will not need to be changed as a result of this modification.

As noted in the separate safety evaluation for the proposed change in Technical Specification 3.8, the plant Operations Committee will review the specific fuel rack installation procedures for this modification.

1877 067

9.0 Summary Reference 1 described the information required for the NRC to conduct a review of license amendment requests involving spent fuel storage facility modifications. This report has addressed the two general areas of staff review - (1) Safety Evaluation Report, (2)

Environmental Impact Appraisal.

Evaluation of the nuclear, thermal-hydraulic , mechanical, material, structural, and environmental aspects of this modification that are presented in this report provide assurance that there will be no significant effect on the public health and safety. This report addresses Sections III through V of Enclosure 1 to Reference 1.

This spent fuel storage capacity increase license amendment request must be approved on a timely basis (September 1980) to assure continued operation of the Prairie Island Nuclear Generating Plant.

O 1877 068

10.0 References

1. Letter, B K Grimes (NRC) to All Power Reactor Licensees, dated April 14, 1978. Enclosure 1 "0T Posit ion for Review and Acceptance of Spent Fuel Storage and Handling Applications".

2. Letter and Enclosures, A Schwencer (NRC) to S Burs tein (WEP) authorizing amendments 35 (DPR-24) and 41 (DPR-27), dated April 4, 1979.

3. Letter and Enclosures, A Schwencar (NRC) to J Dolan (I & MEC) authorizing amendments 32 (DPR-58) and 13 (DPR-74) dated October 16, 1979.

4. Letter and Enclosures, K Goller (NRC) to L 0 Mayer (NSP) authorizing amendments 22 (DPR-42) and 16 (DPR-60), dated August 16, 1977.

5. Battelle Pacific Northwestern Laboratory Report PNL-3065, Commentary on Spent Fuel Storage at Morris Operation, K J Eger and G E Zima, July 1979.

6. Letter, L 0 Mayer (NSP) to Victor Stello (NRC), dated November 24, 1976 Exh ibit A.

7. Generic Environmental Impact Statement on Handling and Storage of Spent Light Watec Power Reactor Fuel, NUREG 0575, August 1979.

i877 069 11 , Tables Table Title 1 Prairie Island Spent Fuel Storage Historical Data and Forecast O

-2e- e 877 0'l0

TABLE 1 .

PUIRIE ISLAND SPENT FUEL STORAGE HISTORICAL CATA ANO FORECAST PRESENT SPENT FUEL STORAGE 4PUNGEMENT TOTAL TOTAL NO.(0I

CF TOTAL (6)

REFUELING UNIT II UNIT I TO REMAINING SPACES CATE ASS EMSLIES 4SSEMBLIES DATE SP4CES OCCUPIEC 1980 40 40 360 327 52 1981 40 41(I) 44f(2) 245 64 1982 41(I) 40 522(3)(4) 165 76 1983 40 40 602 85 87 Full Core Storage for 121 Assemblies is no Longer Available 1984 40 40 682 5 99 PROPOSED SPENT FUEL STORAGE MODIFICATION (to be complete in late 1981) 1982 40 40 522 857 38 1983 40 40 602 777 44 1984 40 40 682 697 49 1985 40 40 762 617 55 1986 40 40 842 537 61 1987 40 40 922 457 67 1988 40 40 1002 377 73 1989 40 40 1082(5) 297 78 1990 40 40 1162 217 84 1991 10 40 1242 137 90 1992 40 40 1322 57 96 1993 40(7) 40 1402 100 NOTES (1) Exxon delivers 201 fuel assemblies for each unit ch their 5-year contract-(2) Maximum contemplated capacity of Pool I with high density spent fuel storage racks - 462 asse-blies. The modificatien can be made during Sursrer and Fall 1981 with all fuel stored in Pool 1.

(3) Pool 2 capacity - 555 asseelies.

(4) The next refueling discharge will require storage of spent fuel in Pool 1.

Presently our license dces not allow handling heavy objects above a pool storing irradiated fuel. Pool 1 is in the crane movement pathway and also is required for shipping spent fuel.

(5) Pool 2 mcdified capacity - 1120 assenblies.

Crane modification required to move heavy loads over Pool 1 after next refueling. (See Note 4)

(6) Pool 1 and 2 total modified capacity - 1582 assemblies. However, four modules must be remved from Pool 1 for spent fuel shipping cask handling. Therefore ,

only 1379 storage locations are available for normal storage.

(7) After the first refueling in 1993, nor .al storage capacity will be exhausted.

1877 071 .

12.0 Figures Figure Title 1 PINGP Fuel Storage Facility 2 Interim Storage Configuration (early 1981) 3 Final Storage Configuration 4 Spent Fuel Cooling System 5 General Arrangement - Fuel Handling & Ventilation Fan Room-East 6 General Arrangement - Operating Floor-East 7 General Arrangement - Sections B-B & C-C 8 General Arrangement - Sect ion D-D 9 General Arrangement - Spent Fuel & New Fuel Storage 10 Gamma Exposure Rates above Spent Fuel 11 Unit 1 Refueling Cavity Cleanup System 12 Unit 2 Refueling Cavity Cleanup System 13 Spent Fuel Skimmer System 14 Spent Fuel Cleanup System 15 Normal and Special SFP Vent ilation 9

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(~ . m w) This sketch is an FIGURE 14 enla rgement of Figure 4 SPENT FUEL POOL CLEANUP SYSTEM F877 092

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EXHIBIT B License Amendment Request dated January 31, 1980 Docket Nos. 50-282 License Nos. DPR-42 50-306 DPR-60 Exhibit B consists of revised pages of the Prairie Island Nuc let.r Gene rat ing Plant Technical Specifications, Appendix A, as listed below: Pages (TS-) 3.8-2 5.3-1 5.6-1 0 ' 1877 094,

TS.3.8-2 REV

6. Direct communication between the control room and the operating floor of the containment shall be available whenever changes in core ge ome t ry a re taking place.

7. No movement of irradiated fuel in the reactor shall be made until the reactor has been subcritical for at least 100 hours.

8. The radiation monitors which initiate isolation of the Con-tainment Purge System shall be tested and verified to be operable immediately prior a refueling operation.

B. During fuel handling operations, the following conditions shall be satisfied:

1. No heavy loads will be transported over or placed in either part of the spent fuel pool when irradiated fuel is stored in that part.*

2. Prior to spent fuct hand 1'ng in the auxiliary building, tests shall be made to determine the i erability of the spent fuel pool special ventilation system inicuding the radiation monitors in the normal vent ilation system that actuate the special system and isolate the normal systems.

3. Prior to fuel handling operations, fuel-handling cranes shall be load-tested for operability of limit switches, interlocks, and alarms.

4. When the spent fuel cask contains one or more fuel assemblies, it will not be suspended more than 30 feet above any surface until the fuel has decayed more than 90 days.

C. If any of the specified conditions in 3.8.A or 3.8.B above are not met, refueling or fuel-handling operat ions shall cease. Work shall be initiated to correct the violated conditions so that the specifications are met, and no operations which may increase the reactivity of the core shall be performed.

For the purpose of completing the fuel storage pool modification, the movement and placement of loads shall be in accordance with the installation procedures approved by the plant on-site review committee.

4-

                                                                      \        e

TS.S.3-1 REV 5.3 REACTOR A. Reactor Core

1. The reactor core contains approximately 48 metric tons of uranium in the form of slightly enriched uranium dioxide pellets. The pellets are encapsulated in Zircaloy-4 tubing to form fuel rods. The reactor core is made up of 121 fuel assemblies. Each fuel assembly contains 179 fuel rods.

2. The average enrichment of the reload core is a nominal 2.90 weight per cent of U-235. The highes t Uranium-235 loading is a nominal 39 grams of U-235 per axial centimeter of fuel assembly (average).

3. In the reactor core, there are 29 full-length RCC assemblies th at contain a 142-ingig length of silver-indium-cadmium alloy clad with s t ain le s s steel B. Reactor Coolant System

1. The design of the gyctor coolant system complies with all applicable code requirements.

2. All high pressure piping, components of the reactor coolant system and their supporting structures are designed to Class I requirement s, and have been designed to withstand:

) a. The design seismic ground acceleration, 0.06g, acting in the horizontal and 0.04g acting in the vertical planes simultaneously, with stressed maintained within code allowable working s tresses.

b. The maximum potent ial seismic ground acceleration, 0.12g, acting in the horizontal and 0.08g acting in the vertical planes simultaneously with no loss of function.

3. The nominal liquid volume of the reactor coolant system, at rated operating conditions, is 6100 cubic feet.

C. Protection Systems The protection systems for the reactor and engineered safety features are designed to applicable codes, including IEEE-279, dated 1968. The design includes a reactor trip for a high negative rate of g nge of neutron flux as measured by the excore nuclear instrument s. The system is intendedtogipthe reactor upon abnormal dropping of more than one control rod If only one control rod is dropped, the core can be operated at full power for a short time, as permitted by Specification 3.10. References (1) FSAR, Section 3.2.3 (3) FSAR, Table 4.1-9 (2) FSAR, Sections 3.2.1 and 3.2.3 (4) FSAR, Section 7 1877 096 ,

TS.S.6-1 REV 5.6. FUEL HANDLING A. Criticality Consideration The new and spent fuel pit structures are designed to withstand the anticipated earthquake loadings as Class I (seismic) structures. The spent fuel pit has a stainless steel liner to ensure against loss of water. (1) The new and spent fuel storage racks are designed so that it is impossible to insert assemolies in other than the prescribed locations. The fuel is stored vertically in an array with the center-te-center distance between assemblies sufficient to assure k gg 1 0.95 even if unborated water were used to fill the pit. In addition, fuel in the storage pool shall have a U-235 loading of axial centimeter of fuel assembly (average)139.0 grams of U-235 per The spent fuel storage pit is filled with borated water at a con-cent rat ion to match that used in the reactor cavity and refueling canal during refueling operations or whenever there is fuel in the pit. B. Spent Fuel Stora2e The spent fuel storage f acility is a two-compartment pool that may contain up to 1582 storage locations for spent fuel as semb lies . The pool is enclosed with a reinforced concrete building having 12- to 18-inch thick walls and roof (1) The pool and pool enclosure are Class 1 (seismic) structures that afford protection against loss of integrity from postulated tornado missiles. The s torage compartment s and the fuel trans fer canal are connected by fuel transfer slots that can be closed off with pneuma-tically sealed gates. The bottoms of the slots are above the tops of the active fuel in the fuel assemblies which will be stored vertically in specially constructed racks. O 1877 097

EXHIBIT C PRAIRIE ISLAND NUCLEAR GENERATING PLANT License Amendment Request dated January 31, 1980 Docket Nos. 50-282 License Nos. DPR-42 50-306 DPR-60 Exh ibi t C consists of the Nuclear Services Corporati.on document: QUAD-1-79-509

      " Licensing Report for Prairie Island Nuclear Generat ing Plant Units 1 and 2 Spent Fuel Storage Modification" 1877 098,

QUAD-1-79-509 JOB NO. NOR-0174 LICENSING REPORT FOR PRAIRIE ISLAND NUCLEAR GENERATING PLANT UNITS 1 AND 2 SPENT FUEL STORAGE MODIFICATI0h' NSP PROJECT NUMBER E-78YO75 Prepared For: NORTHERN STATES POWER COMPANY Minneapolis, Minnesota By: NUCLEAR SERVICES CORPORATION A Division of Quadrex Corporation 1700 Dell Avenue Car.pbell, California 95008 }8(( Qgg 2 f 0- g ? ( f2e/B0 l h SNh < h h /2f29/79 0 Qd$$ l!hb (&S /ol/9l79 REV. PROJECT QA PROJECT ISSUING DATE OF NO. ENGINEER REVIEWER ENGINEER ENGINEER MANAGER APPROVAL

QUAD-1-79-509 NUCLEAR SERVICES CORPORATION A Div,5 f CN OF QURDREX CO APO A ATION TABLE OF CONTENTS PAGE

1.0 INTRODUCTION

1-1 2.0 PURPOSE l-1 3.0

SUMMARY

OF DESIGN MODIFICATIONS & ANALYSES 3-1 3.1 General Description 3-1 3.2 Mechanical Design 3-1 3.3 Nuclear Analysis 3-5 3.4 Structural Analysis 3-22 3.5 Thermal Analysis 3-54 3.6 Radiochemical Anaylsis 3-70 3.7 Fuel Rack Installation 3-73

4.0 CONCLUSION

S 4-1

5.0 REFERENCES

5-1 O 1877 100 i

QUAD-1-79-509 DUCLCRR SERVICES CORPORATION A Civ'siON OF WORDREE CO APO A ATION

1.0 INTRODUCTION

This report is occassioned by a proposed modification to the Spent Fuel Storage System for the Prairie Island Nuclear Generating Plant Units 1 and 2. The proposed .nodification will provide 15S2 spent fuel storage spaces for Unit 1 and Unit 2 combined. Of this 1582 storage spaces, 1386 spaces will be available for normal storage of fuel and 121 can be used for a full core discharge. The remaining 75 spaces will be used only during installation of the new racks. This layout is described more completely in Section 3.1. The two units of the Prairie Island Plant share two interconnected fuel pools for spent fuci storage. The proposed modification applies to both of those fuel pools. 2.0 PURPOSE The purpose of this report is to examine from a technical standpoint those aspects of the proposed modification which may have a bearing on the storage of spent fuel in the Prairie Island spent fuel pools. The nuclear, structural and thermal aspects of the proposed modification have been examined and those aspects of the modification are presented in the following sections of this report. i877 . 101 1 -1

QUAD-1-79-509 RUCLEAR SERVICES CORPORATION Q.CO.A O!w1$'ON 08(lRDREX A AO A ATION 3.0

SUMMARY

OF DESIGN MODIFICATIONS AND ANALYSES 3.1 General Description The Prairie Island Spent Fuel Storage Facility consists of two storage pools. The first is a small fuel storage pool (pool 1) which is used for fuel storage and for loading of fuel into the shipping cask. The other pool (pool 2) is a larger pool which is used only for fuel ;torage. The arrangement of these two pools is shown in Figure 3.1-1. In order to use a spent fuel shipping cask in pool 1, it will be necessary to remove the four spent fuel racks located in the southeast corner of that pool. Therefore, only the five remaining racks in pool l can be used for normal fuel storage. This results in the availability of 266 normal storage spaces in pool 1. The racks in the southeast corner of pool I can be used for a full core discharge, since it is not necessary to use a shipping cask during a full core discharge. The spent fuel pool structure and supports have been analyzed and found to be acceptable for the additional load imposed by the increased fuel storage capacity. A description of this analysis and the results are presented in Report Q-1-79-558. 3.2 Mechanical Desian Figure 3.1-1 shows that two sizes of spent fuel racks will be used; a 7 x 7 space rack and a 7 x 8 space rack. The 7 x 8 rack is shown in Figure 3.2-1. In this design, upper and lower grids are used to interconnect the storage tubes. These grids also ensure proper location of the storage tubes on 9.5 inch pitch in both directions. The upper and lower grids are tied together by vertical and diagonal members. These members will transmit seismic and handling loads from the tubes to the rack base. 3-' i877 102,

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FLOOR LEG CONNECTION RACK LEG 1877 104 FIGURE 3.2-1: 7X8 SPENT FUEL RACK 3-3

QUAD-1-79-599 DUCLEAR SERYlCES CORPORATIOR b b5bbEX CO APO A ATION & The rack base is composed of heavy box beams connected at the four corners to box section legs with aC ustable feet. These adjustable feet will provide adjustment during installation to ensure that the storage tubes are vertical. The box beams of the base are elevated above the pool floor to allow flow of cooling water below the rack and up into the storage tubes. Reactivity control is provided by the 9.5 inch storage tuue pitch and by the storage tube material. Each storage tube consists of three components: an inner stainless steel tube, a layer of neutron absorbing material, and an outer skin of stainless steel. The inner tube is of adequate length to extend from below the bottom of the fuel assembly to above the top of a stored fuel assembly with the control cluster in place. Two support bars are welded into the bottom of this inner tube, and the stored fuel rests on these support bars. The layer of neutron absorber is located on the four outer surfaces of each inner tube. The neutron absorber is in the form of solid sheets of material provided by Brand Industrial Services Company. This type of material has been previously licensed by the USNRC for use in spent fuel racks. The material is composed of a silicon polymer base material with sufficient boron in the form of boron carbide to result in an area density of 0.04 grams / square centimeter of boron-10. The neutron absorber extends the full axial length of the active fuel region. Quality assurance procedures for the neutron absorber fabrication will ensu e to a 95% confidence level that the boron-10 area density is a minimum of 0.04 g/cm _2 The outer skin is a thin sheet of stainless steel whdch covers the neutron absorber and holds the absorber in place. This outar skin will completely enclose the neutron absorber, except that some 3-4 1 877 , 105

QUAD-1-79-509 DUCl.CRR SERVICES CORPORATIOD A Div'liCN OF QUADREX CO ADO A ATION regions at the joint between of the auf.er skin and the inner tube will not be welded. This will result in a venting of the neutron absorbing region and will prevent pressure buildup in this region which might result from gas generatioa or hydrostatic pressure buildup. With the exception of the above described neutron absorber and the screws in the ajustable feet, all material used for rack construction will be type 304 stainless steel which meets the appropriate ASTM or ASME material specification. The adjustable foot screws will be 17-4 Ph stainless steel in accordance with ASTM A564. These stainless steel materials are the same materials used in the fuel racks presently in use at Prairie Island. Each rack sits on the pool floor liner. There are no bolted or welded connections between the rack and pool. Vertical loads are transmitted in bearing from the rack feet directly to the floor. Horizontal loads are transmitted from the feet to the floor in shear by friction only. There are no connections between adjacent racks, nor are there any supports to the fuel pool walls. Spacer beams are located as shown in Figure 3.1-1 to preclude the insertion of a fuel assembly between racks. 3.3 Nuclear Analysis A nuclear analysis has been performed for the proposed spent fuel rack to determine the k eff for the rack while storing the fuel. 3.3.1 Design Criteria The accepte.nce criteria established for this spent fuel storage rack is as follows: 3.s 1877 106 9

QUAD-1-79-509 DUCLEAR SERVICES CORPORATIOD

                                 & Lev.5 CN Of QURDREX CO A AQ A ATION                                     $

TABLE 3.3-1 Fuel Design Parameters Fuel Type Westinghouse Exxon Future Rod Array 14 x 14 14 x 14 14 x 14 No. of Fuel Rods 179 179 179 No. of Water Holes 17 17 17 n n ,, Rod Pitcn 0.556 0.556 0.556 Pellet 0.0. 0.3659" 0.3565" 0.3444" Clad 0.D. 0.422" 0.424" 0.400" Clad Thickness 0.0243" 0.0300" 0.0243" Clad Material Zircaloy 4 Zircaloy 4 Zircaloy 4 Pellet Densi ty, i T.D. 94 94 94 U L ading 39 g/cm 39 g/cm 39 g/cm 235 Nominal Active Fuel Length 144" 144" 144" 3-6 1877 107,

QUAD-1-79-509 DUCT. ERR SERVICES CORPORATIOD

                                     & DivlSICN C7
                                   /IADREX CO A AO A /h TION When immersed in ciean, unborated water and completely filled with new fuel as described in Table 3.3-1, the calculated k        f r this eff rack shall be 1 0.95. This shall include all probable uncertainties and anticipated pool water temperatures.

This spent fuel rack is designed in accordance with USNRC Standard Review Plan 9.1.2. Revision 2. Included in the conditions to be analyzed are: pool water temperatures from 40 F to 212 F, eccentric position of fuel in the proper location, fuel tube and rack assembly tolerances, and single fuel assemblies outside the storage racks. 3.3.2 Analysis Method The value of k eff is determined as follows: 2 k eff 1 k0+Ak)+ak2 + (ak3 + ak4 b Ak 5 + ak6 + ak7) where k 0

         = n minal calculated k (2-D diffusion theory) ak) = transport correction ak2 = method bias ak 3
         =    uel locadon effect ak = storage tube pitch tolerance effect 4

akg = uncertainty in methods bias (95% confidence level) ak6 = neutron absorber boron-10 tolerance effect ak7 = storage tube dimensional effect Verification of k eff is obtained using a two dimensional (X-Y) diffusion theory computer code calculation. Fuel neutron cross sections are developed for a four group energy range using the 3-7 1877 108,

QUAD-1-79-509 OUCL.CRh SERVICES CORPORRTI0n A Dival:0N OF QUROREX CO APOR ATION CHEETAH code (Ref.1) which is an adaptation of the LEOPARD-CINDER code. XSDRil (Ref. 2), which is a one dimensional discrete ordinates spectral averaging code, was used for two calculations:

1. Fuel rod cell,123 group; collapsed to 27 groups.

2. Rack supercell,1-D (cylindrical) approximation, to collapse from 27 to 4 groups.

Cross secticn sets for all non-fuel regions were obtained from XSDRN. The output of CHEETAH and XSDRN are used for a diffusion theory calculation using the CITATION code (Ref. 3) to establish the k eff' In order to verify the accuracy of the CHEETAH-XSDRN-CITATION calculation, a comparison was made with two critical experiments as shown in Table 3.3-2. The actual fuel used in the critical experi-ments was placed into CHEETAH-XSDRN-CITATIOC calculation with the results shown for comparison. These results may be compared with the measured k and with the value of k eff calculated using the eff KEN 0 code (Ref. 4). As a further check the keff f r the proposed fuel rack has been calculated with the KEN 0 code. The data provided in Table 3.3-2 show that the CHEETAH-XSDRN-CITATION analysis performed for the proposed racks provides a conservative result for k eff' The methods bias and uncertainty in the methods presented in the following section are based on the comparison to KENO for the proposed design. Based on the values of k eff calculated with CHEETAH-XSDRN-CITATION and KEN 0 for the proposed design, the transport theory correction factor is conservatively taken as zero. The following assumptions were used in the calculations: 3-8 O 1877 109

QUAD-1-79-509 NUCLEAR SERVICES CORPORATION A O'weliCN 08 bADREX C O APO A ATION TABLE 3.3-2 CHEETAH-XSDRN-CASE EXPERIMENTAL CITATION KENO II) Single BORAL blade in array of 9x9 2.35 1.01680 w/o U02 fuel 1.0018 1.0174 1 00550 (2) Bare U rods with depleted U block and BORAL sneet, water reflected 1.000 1.01321 1.01748

                                                                               + 00634 Prairie Island rack                                                        0.85377 with Westinghouse fuel         ---

0.89998 1 00624 (1) Critical experiment data from BNWL-1379 (2) Critical experiment data from YAEC-1090, Run 105 1877 ' 110 3-9

QUAD-1-79-509 RUCLEAR SERVICES CORPORATION g . o ...

                           %(URDREX o.

CO A AC A ATION

a. Pool is filled with fuel at the highest enrichment stored in an infinite array.

b. The water in the fuel storage pool is clean and unborated.

c. The pool temperature is 400F (4.40C).

d. No credit is taken for the fuel assembly support structure.

e. fieutron absorption in fuel assembly grids is excluded.

f. Axial neutron leakage is included.

g. lio credit is taken for U 234 and U 6

in the fuel.

h. The neutron absorber boron-10 area density is 0.04 g/cm2 ,

The analysis is based on nominal stainless steel thicknesses, because the Boron-10 content is taken at the minimum 95% confidence level value and the Boron-10 is the dominant neutron absorber. The above assumptions are considered as a conservative base for the calculations. 3.3.3 Analysis Results The nominal value of k eff nd the applicable uncertainties are pre-sented in Table 3.3-3 for each of the fuel assemblies to be stored. The results of the keff calculations for potential storage and handling conditions are listed as follows: K ggg CONDITION Fuel Type Westinghouse Exxon Future

1. tiormal positioning .89998 .90210 .93414 in the spent fuel array See Figure 3.3-1 3-10 1877 111

QUAD-1-79-509 DUCLERR SERVICES CORPORATI0n A OlviSiON OF

                              '[.1 7aD Q 3 X CO APOM ATION TABLE 3.3-3 Nuclear Analysis Results Term                        Fuel Type                       Method W           Exxon                 Future k

0 0.89998 0.90210 0.93414 Calculated for cell ak j 0 0 0 Comparison with KENO ak 2

        <0             <0                    <0        KEN 0 and critical experi-ments ak 3
        <0             <0                    <0        Fuel moved to corner ak 4

0.00059 0.00059 0.00065 .060" displacement in X and Y ak 5 0 0 0 KEN 0 and critical experi-ments 95% confidence level ak 6 0 0 0 Anaiysis is based on minimum Bio content with 95% confidence ak 7

        .00808         .00812                 .00801   Analysis based on worst case tube dimensions k      .90808         .91024                 .94218 eff                                                   i877 112 where:

2 2 k eff =k0+ak)+ak2 + (ak3 + ak4 + ak5 + ak6 +ak7) 3-11

QUAD-1-79-509 DUCLEAR SERVICES CORPORATION

                           . o.vis,o. o, QdADAEX CO APO A ATION TABLE 3.3-3     (Continued)

NOTE: From the data presented in Table 3.5-2, it can be seen that the combination of methods bias and uncertainty results in a negative value. These correction factors are here conserva-tively assumed to equal zero. O 3-12 l077 ,

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QUAD-1-79-509 NUCLEAR SERYlCES CORPORATIOR A DIV SeCN OF

                           %f41ADREX CO APO A ATION

2. Eccentric posi- .89062 .89270 .92457 tioning in the spent fuel storage array with fuel at corners of the channels See Figure 3.3.2

3. Normal positioning .90057 .90269 .93479 in the spent fuel storage array with the channel offset 0.06" (.152 cm) in both X and Y See Figure 3.3-3

4. One extra fuel .90742 .90958 .94056 assembly at side of rack See Figure 3.3-4 Sensitivity studies were performed to evaluate the influence on eff f storage tube pitch and pool water temperature.

k The effect of change in pitch on k w s calculated, and the eff results are presented in Table 3.3-4. The reference rack unit cell was analyzed at several temperatures in order to determine the temperature effect on k The fuel eff. pellet, clad, moderator, and rack pitch were assumed to expand no rmally. The results are shown in Table 3.3-5. The numbers quoted in this report. are the higher values for the lowest temper-ature (40 F). 1877 114, 3-13

QUAD-1-79-509 DUCl. ERR SERVICES CORPORATIOD

                                  . A D6viss0N 08 VdRDREX CO APO A ATION                          $

TABLE 3.3-4 Effect of Storage Tube Pitch Variation at 40 F Pitch (inches) eff W_ Exxon Future 9.44 x 9.44 0.90837 0.91053 0.94257 9.50 x 9.50 (base case) 0.89998 0.90210 0.93414 9.56 x 9.56 0.89189 0.89397 0.92601 O O 3-14 1877 115

QUAD-1-79-509 DUCLEAR SERVICES CORPORRTIOD

                                      & DavtSION OF QZIADREX CO APO A ATION TABLE 3.3-5 Effect of Fuel Pool Water Temperatures Temperature ( F)                                    eff W_

Exxon Future 40 (base case) 0.89998 0.90210 0.93414 130 0.89000 0.89207 0.92339 212 0.87772 0.87974 0.90896 3-15

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QUAD-1-79-509 DUCLEAR SERVICES CORPORRTIOn A DivaliCN 05 MADREX CO RPO A ATION 3.3.4 Accident Conditions Potential accident conditions considered in the design of these spent fuel racks include: dropping of a fuel assembly on top of a rack or between a rack and the pool wall; effect of an earthquake on the relative position of fuel racks and fuel storage tubes; loss of all cooling systems. The drop of a fuel shipping cask and tornado effects are not concerns in this modification for the reasons described below. The effect of a dropped fuel assembly in the most reactive position next to a rack is shown as condition 4 in Section 3.3.3. The drop of a fuel assembly onto the rack was considered. The fuel rack is designed to prevent any plastic deformation in the fuel region for these loads. Therefore, the reactivity effect of these abnormal loading conditions is insignificant. Considera-tion was given to a fuel assembly lying on top of a rack. Due to the greater than 18 inch separation of the fuel assembly on top of the rack from active fuel in the rack, the fuel on top of the rack is isolated and k eff is not affected. The fuel racks have been designed as seismic Class 1 equipment and, as such, can withstand the plant SSE with no damage to the structure which maintains Keff <0.95. Although the racks may move during an earthquake, the dimensions of the fuel rack structure preclude reduction of the fuel storage spacing between adjacent racks to less than the nominal analyzed dimension. As shown in Table 3.;-S the fuel rack has been analyzed for the full potential water temDerature range. It can be seen that K eff actually decreases as temperature increases. Therefore, loss of the pooi cooling systems will not result in Keff >0.95. 3-16 O 1877 117

QUAD-1-79-509 GUCt. ERR SERVICES CORPORRTIOR A Division OF WUROREX CO A PO A ATION The drop of a fuel shipping cask or other heavy equipment into pool 2 is not a concern because the building crane is physically restricted and cannot be moved over pool 2. During installation a temporary crane will be used to move fuel racks. However, all fuel will be stored in pool I while equipment is being lifted in pool 2. For pool 1 a cover was designed and fabricated during the previous fuel storage modification. At any time that pool 1 is used for fuel storage, this cover will be installed following completion of fuel handling. The use of this cover has been reviewed under the application for the previous fuel storage modification. When fuel is stored in pool 1, the cover will be used and no fuel shipping cask will be inserted into pool 1 until further licensing action is completed. Because of the above, no consideration in the nuclear analysis has been given to the drop of a fuel shipping cask into the pool. The building in which the fuel pools are housed has been designed for the maximum anticipated site tornado. Therefore, no tornado effect on the fuel stcrage equipment is anticipated. 3.3.5 Conclusion The proposed design meets the specified criteria of keff " 0.95 for all conditions included in the postulated accidents. Therefore, storage of the fuel described in Table 3.3-1 presents no safety problems with respect to the nuclear aspects of the design. In order to ensure that the fuel rack will contain the analyzed amounts of neutron absorber, a fabrication quality assurance program will confirm and document that each sheet of absorber is in location. Inspection and analysis of the absorber material will confirm to a 95% confidence level that the boror-10 content is equal to or above the level assumed in this analysis. 3-17 )b

O 9.88296 FUEL 0.61994 INNER H2 0 - --- _ _ . _ _ __ . _ _ . _ 0 27940 0.22860 SS 304 0.31750 ABSORBER HO2 0 09144 SS 304 0 92456 0 UTER H2O ALL DIMENSIONS IN CM ALL B0UNDARIES ARE REFLECTIVE Figure 3.3-1: Normal Unit Cell Configuration O 3-18 1877 119

9 0.92456 0 UTER H2 0 0.09144 _ _ _ _ _ SS 304 0.31750 H2O ABSORBER HO 2 1 0.22860 SS 304

                                                                           --~ ------

1.23988 INNER H2O ja 20.447

;            19.76592                                     FUEL 0.22860 0.31750                   H02                             HO 2

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c3 ALL BOUNDARIES ARE REFLECTIVE ALL DIMENSIONS IN CM Figure 3.3-2: Fuel Offset

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QUAD-1-79-5D9 DUCLCAR SCRVICCS CORPORATI0n A DeviSION 08

                         ~ I_lA D R E X CO APO A ATION As a further backup each storage tube will be inspected after installation in the pool to ensure that all neutron absorber sheets are in position.

3.4 Structural Analysis 3.4.1 Rack Structural Description The proposed high-density spent fuel racks are free-standing type, i.e., these are placed on the pool floor without any floor attach-ment or lateral wall support. Two different sizes of spent fuel racks have been used in the proposed pool arrangement. These are designated as 7 x 8 and 7 x 7 size racks. Configurations for these two rack sizes are similar. Each rack consists of a frame assembly, an upper grid, a lower grid, a base assembly, four leg assemblies, and the tubes supported on the lower grid and base assembly and welded to the top grid. The structural components of the rack assembly are shown in Figure 3.2-1 and listed in Table 3.4-1. 3.4.2 Loads, Load Combinations, and Evaluation Criteria 3.4.2.1 Loads The following loads are considered in the design and evaluation of the proposed racks in their installed condition in the pool: D = Deadweight: Weight of the rack assembly including the spent fuel bundles. B = Buoyance: The effects of the buoyancy on the submerged racks and fuel bundles. O 3-22 1877 123

DUCLEAR SERVICES CORPORATIOD Q(_A OlviS40lu 08 lADREX C O A PO R ATIO N TABLE 3.4-1

MEMBER SIZES FOR RACK ASSEMBLY PLANE MEMBER SIZE COMP i ENT OUTER PLANE E&S 10"X3/4" INNER PLANE UPPER GRID A&3 8,, X1/ 2, INNER PLANE B,C,D,2&4 6"X1/2" PLANE LOWER GRID A,B,C,D,E 4"X1/2" AND 2,3,4,5 VERTICAL ALL 4 CORNER L BEAM CORNERS 8"X8"X1/2" PLANE A&3 4" X1/ 2" CROSS BRACES OUTER PLANE 4" X 3/ 4" E&5 OUTE LANE MIDDLE STRAP 4"X3/4" BASE ASSEMBLY PLANE 5 10 1/2" X 4" CROSS (BOX) BEAMS PLANE A,C,E 6" X 6" X 3/4" RACK LEG ALL 4 CORNERS 9" X 9" X 3/4" FLOOR LEG ALL 4 CORNERS 2 3/4" 0 SOLID CONNECTION ROUND BAR (1) See Figure 3.2-1; (2) See Figure 3.4-7 l~877 , 124 3-23

QUAD-1-79-509 DUCLERR SERVICES CORPORATIOD

                                , . owo~ e, URDREX CO APO A ATION E = Operating Basis Earthquake (0BE): Loads from the two horizontal and the vertical component of the OBE using pool floor CBE vibratory motion from Reference 14. The mass of the water inside the rack and hydrodynamic mass effects of the water surrounding the racks are also considered.

E' = Safe Shutdown Earthquake (SSE): SSE loads were conservatively assumed to be twice the OBE loads. M = Fuel Bundle Drop: Loads resulting from the accidental drop of a spent fuel bundle from a height of 18 inches. Q = Thermal Gradient: Loads resulting from thermal gradient due to a single " hot" spent fuel bundle being placed in a rack cavity with the adjacent cavities empty. Tg , T, = Pool Water Temperature: Loads resulting from the increase in fuel pool water temperature during normal operation (T ) and during accident condition (T ). However, for free-st nding a racks with no floor attachment, stresses due to T g and T are a insignificant and are not considered. U = Grapple Load: Loads that might occur if a fuel assembly or a fuel handling grapple were to jam accidentally in the racks during removal. The design vertical and horizontal loads for this condition are considered to be 7,000 and 3,500 pounds, respectively. However, the resultant of the simultaneously applied forces is taken to be 7,000 pounds. 3.4.2.2 Load Combinations and Evaluation Criteria In accordance with USNRC Regulatory Guide 1.29 (Reference 5) and ANSI Standard N210 (Reference 11), the proposed racks were classified as Seismic Category I and Safety Class 2 structures. Structural adequacy of the racks was verified for the applicable loading com-binations and stress allowables listed in USNRC Stancard Review Plan Section 3.8.4 (Reference 6). Basic allowable stress (S) values were taken from Table I-7.2 of ASME Boiler and Pressure Vessel Code Section III (Reference 7). Elastic working stress 3-24 i877 125

QUAD-1-79-509 IlUCLERR SERVICES CORPORRT10fi A DIVISION OF QUADREX CO A AD A ATION method of analysis and AISC (Reference 17) method of evaluation were used for all loads except for impact loads, in which case an elasto-plastic method of analysis was used to evaluate the consequences. Table 3.4-2 shows the loading combinations and the respective structural evaluation criteria for which the proposed spent fuel racks were to be evaluated. The basic acceptance criteria in the evaluation of the spent fuel rack is that the rack configuration shall always maintain the nuclear criticality coefficient Keff less than 0.95. For the proposed racks this was ensured by using the stress limits set by USNRC in Standard Review Plan Section 3.8.4, and shown in Table 3.4-2. Since the proposed racks are free-standing and are not tied to the pool floor, these may or may not slide during a seismic event and so may or may not impact on each other. This would depend on the seismic intensity to which the racks may be subjected, and the coefficient of friction between the rack and the pool floor. For evaluating the consequence of such impact as well as the consequence of accidental drop of a fuel bundle on the rack, the criticality criteria was translated to the following equivalent structural criteria: The resulting deformation state shall be such that the structure, which maintains the fuel spacing in the active fuel region, remains within the elastic limit. Free-standing racks have the potential for overturning during severe seismic events. To ensure the stability of these racks, an acceptable factor of safety needs to be established as a minimum. Following USNRC Standard Review Plan 3.8.5 (Reference 12), a factor of safety of 1.5 for OBE and 1.1 for SSE has been used. 3.4.2.3 Seismic Input Structural evaluation of the proposed racks was based on the . seismic data provided in Reference 14 in which the spectra cor-3-25

QUAD-1-79-509 DUCLERR SERVICES CORPORRT100 A Divisf 07W OF QUADREX CO APO A ATION TABLE 3.4-2 LOADS, LOAD COMSINATIONS AND STRUCTURAL ACCEPTANCE CRITERIA (4) Load Combination U) Allowable Stress (2)

1. D+B S

2. D+B+E() S

3. D+B+Q l.55

4. D+b+Q+E 1.55

5. D + B + E'( ) 1.6S

6. D + B + E' + Q l.65

7. D+B+U (4)

8. D+B+M (4)

Notes: 1. These are applicabie loading combinations generally in accordance with USNRC Standard Review Plan 3.8.4.

2. These allowable st.' esses are in accordance with USNRC Standard Review Plan 3.8.4. The 'S' value for stainless steel shall be from the applicable appendix of ASME Boiler and Pressure Vessel Code, Section III, Division 1, 1977.

3. Factors of safety against overturning shall not be less than 1.5 for OBE and 1.1 for SSE (USNRC Standard Review Plan 3.8.5).

4. The general acceptance criterion for all the loading conditions is that the final configuration of the rack array shall maintain a Keff < 0.95. This criticality criteria is translated to the following structural criteria:

The resulting deformation state shall be such that the structure which maintains the fuel spacing in the active fuel region remains within elastic limit. 3-26 - 1877.127

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QUAD-1-79-509 DUCLERR SERVICES CORPORRTIOD a DivisiCN OF QURDREX C O AP O A ATIO N responding to Mass Point 25 represents those for the spent fuel pool floor. Figure 3.4-1 and 3.4-2 show the horizontal and vertical OBE response spectra at the pool floor. The spectrum values were multiplied by appropriate scaling factors to account for eccentricities (Reference 14). The final values representing the design OBE response spectra are listed in Tables 3.4-3 and 3.4-4. For nonlinear sliding analysis, SSE time history of floor horizontal seismic motion was used as input. Such a time history was generated from SSE response spectrum using NSIC's proprietary computer program NSCTH (Reference 15). The SSE response spectrum for horizontal motion was developed by multiplying the OBE design response spectrum values shown in Table 3.4-3 by 2. This is shown in Figure 3.4-3. The generated acceleration time-history is shown in Figure 3.4-4. Figure 3.4-5 shows its compatibility to the target floor response spectrum in the frequency range of interest. 3.4.3 Method of Analysis Four major types of analyses were performed on the proposed racks tc evaluate their structural adequacy. Since the spent fuel racks would rest freely on the pool floor, it was necessary to determine the maximum horizontal movement and velocity of racks relative to the pool floor when subjected to the vibratory motion of the most severe postulated earthquake, i.e., SSE. This was computed by performing a nonlinear sliding analysis. To ensure that stresses in the spent fuel racks, when subjected to different combinations of loads, are within the allowable stress limits, elastic finite element stress analyses were performed for D, B, Q, U and E loadings. For a free-standing rack without any floor attachment, the stresses resulting from pool water temperature (T g and T,) are negligible. Hence, no analysis for Tg and Taloadings was necessary. Since, SSE response spectrum input was not available; no separate analysis was performed for SSE loading condition. Conservatively, SSE 3- S

1877 130

DUCLERR SERVICES CORPORRTIOR Q(.&DiV8S'08WOf.lA CO A PO R ATION DREX g TABLE 3.4-3 RESPONSE SPECTRUM FOR OBE HORIZONTAL MOTION (1% DAMPING) Period Frequency Acceleration (sec) (cps) (g's)

 .025                   40.00                  .073
 .050                   20.00                  .083
 .075                   13.33                  .094
 .100                   10.00                  .1 04
 .150                    6.67                  .130
 .200                    5.00                  .190
 .275                    3.64                  .595
 .325                    3.08                  .822
 .375                    2.67                 1.227
 .425                    2.35                  .599
 .600                    1.67                  .308
 .800                    1.25                  .164 1.000                    1.00                  .140 1.500                     .67                  .092 2.000                     .50                  .043 9

3-30 .

NUCLEAR SERVICES CORPORATIOR A DIVA $ ION OF QUADAEX CO APOR ATION TABLE 3.4-4 RESPONSE SPECTRUM FOR OBE VERTICAL MOTION (1% DAMPING) Period Frequency Acceleration (sec) (cps) (g's) 0.025 40.00 0.061 0.050 20.00 0.065 0.100 10.00 0.095 0.150 6.67 0.138 0.200 5.00 0.210 0.238 4.20 0.635 0.250 4.00 0.635 0.400 2.50 0.340

  .500                    2.00                   .190
  .600                    1.67                   .120
  .800                    1.25                   .068 1.000                    1.00                   .060 1.500                      .67                  .042 2.000                      .50                  .025 1877 132 3- 31

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QUAD-1-79-509 RUCLCRR SERVICES CORPORRTIOD

                                & OlVill0N OF Q-UADREX CO APO AATION stresses were taken as two times OBE stresses. Stability analyses of racks were performed to determine the factor of safety against overturning due to the action of SSE. Lastly, the consequences of accidental dropping of a fuel bundle (M) and the impact between two racks during an SSE event were evaluated.

The configurations of 7 x 8 and 7 x 7 racks are almost identical and their sizes are also not too different. Hence, it was judged that their structural behavior would be very much similar, and both rack sizes can be qualified structurally by analyzing only the critical rack size in detail. From past experience the 7 x 8 rack was judged to be the most critical (see Section 3.4.5). Section 3.4.4 describes the various structural analyses performed on 7 x 8 size racks. Evaluation of the 7 x 7 size rack has been performed approximately, by conservatively extrapolating the results of the 7 x 8 size rack. This is described in Section 3.4.5. 3.4.4 Analysis of 7 x 8 Rack 3.4.4.1 Sliding Analysis The objective of this analysis was to determine the maximum displace-ment and velocity of the rack relative to the pool floor under the action of SSE vibratory motion. Upper bounds of these values were determined for a very conservative combination of various parameters affecting the movement. The conservative use of these parameters in the nonlinear sliding analysis is described below:

1. The coefficient of friction between the rack and the pool liner (i.e., between stainless steel and stainless steel in a wet condition) was assumed to be 0.2. Judging from the test results reported by Professor Rabinowicz of the Massachusetts Institute of Technology (Reference 8), this value is very 3-35 1877 136

QUAD-1-79-509 NUCLEAR SERVICES CORPORATIOR Q~UADREX

                                  .m,o~ o, COR AD A ATION conservative. Results of his 199 tests under simulated condition had shown a mean value of 0.503, upper limit of 0.753 (mean plus two times the standard deviation), and a lower limit of 0.253 (mean minus two times the standard deviation).

2. The effect of vertical components of SSE was considered conservatively by assuming a constant upward acceleration on the rack equal to the peak vertical SSE acceleration. The frequency of the rack in the vertical direction was computed to be more than 21 cps for which the peak floor acceleration is 0.139 Howevcc, a conservative value of 0.169 was applied to the rack in the upward direction thereby reducing the frictional resistance against sliding.

3. In computing the mass properties of the mathematical model, the water inside the rack was considerea as added mass. The hydrodynamic mass effect of the surrounding water was considered conservatively by adding equivalent virtual mass computed in accordance with Reference 10. Due to the participation of larger hydrodynamic mass, inertial force and so the sliding displacement and velocity parallel to the shorter side of the 7 x 8 size rack would be more than those for the other rack size. Hence, sliaing displacement and velocity parallel to the shorter side of the 7 x 8 rack were computed for the SSE motion. These values were used for computing the energy of impact between the racks.

4. For the sliding analysis, even though the hydredynamic mass was considered in modeling the inertial properties of the rack, the associated damping was ignored, and only structural damping (2 percent) was used. For a structure vibrating in a water medium, tne use of 2% damping is judged to be conservat ve.

i 0 3-?6 lb(( }3[

QUAD-1-79-509 00CLERR SERVICES CORPORPTI0li Q.CO(_lA AAPODivision D R E X 08 A ATION Non-linear time-history analyses using the computer code ANSYS (Reference 9) were performed for the fully-loaded rack and for the empty rack. The mathematical model used in this analysis is shown in Figure 3.4-6. Equivalent stiffness properties of the members were obtained by matching the structural behavior of the stick model with that of a detailed finite element model of the actual rack (see Section 3.4.4.2). Input time-history is shown in Figure 3.4-4. A damping value equivalent to 2% modal damping was used in the analysis. Direct integration method of analysis was employed using time-step size of 0.01 second. 3.4.4.2 Dead Load, Buoyance, and Seismic Analyses 3.4.4.2.1 Mathematical Model For the purpose of dead load (D), buoyance (B) and OBE seismic (E) analyses, the 7 x 8 rack was represented as an assemblage of finite elements. Figure 3.4-7 shows the finite element model used in the analysis. All the stress analyses on this model were performed using the computer program STARDYNE (Reference 16). 3.4.4.2.2 Stress Analysis for Deadload and Buonnce Loads Stresses resulting from deadload and buoyance were evaluated simultaneously using the finite element mathematical model described in Section 3.4.4.2.1. These loads included the following:

1. Buoyant weight of the rack body.

2. Buoyant weight of the fuel bundles.

3-37 1877 138

RUCLCAR SERVICES CORPORATIOR A 01wt5 04 OF W WLIADREX CO APOR ATION A

                               ,             O g                 O 8

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                           &l FRICTION ELEMENT y                   7 b               l   t h

FIGURE 3.4-6: MATHEMATICAL MODEL FOR NONLINEAR SLIDING ANALYSIS O 1877 139' 3-38

DUCLEAR SERVICES CORPORATION hk To PRRRIE ISLRND 7XS SPENT FUEL STORRGE RRCK, ce N f -/ y'D e ,, s

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1877 140 " FIGURE 3.4-7: FINITE ELEMENT MODEL FOR 7 X 8 RAC 3-39

QUAD-1-79-509 NUCLEAR SERVICES CORPORATION Q(_A 08vil'04 08lADAEX CO RPO R ATION W 3.4.4.2.3 Seismic Response and Stress Analysis Modal analysis of the 7 x 8 loaded rack was performed using the finite element mathematical model described in Section 3.4.4.2.1. Eighty-six dynamic modcs of vibration were extracted. In perfonning tbase analyses, the mass of the water inside the rack wi.s considered as added mass. The hydrodynamic mass effect of the water in between the racks was included in the dynamic mathematical model. This was computed in accordance with Reference 10. Once the eigenvalues and the eig.nvectors were extracted, the stresses due to three OBE components were computed by the response spectrum analysis method. For the horizontal and vertical components, OBE horizontal and vertical response spectra (Tables 3.4-3 and 3.4-4) were used. In these analyses it was assumed that the friction coefficient is such that the rack :iould not slide and tilt. But ouring a postulated seismic event, the racks, having no floor attachment nor lateral supports, may or may not tilt depending on the level of seismic excitation. In the worst possible case, each rack may be supported on one or two legs for a very short duration in a dynamic state of equilibrium. Since the racks are free to slide, the probability for the rack to be tilting on one leg is judged to be very small. However, for the ourpose of stress evaluation, the 7 x 8 rack was also analyzed for this extreme condition (i.e., when supported on one leg) in addition to the normal storage condition (i.e., when resting on four legs). For the convenience of reference, these two conditions will be termed as: Configuration I. Rack supported on four legs Configuration II: Rack supported on one leg

     '                        ~4 1877 141

DUCI.CRR SERVICES CORPORATION a DW15:04 Of QUADREX CO APO A ATION To simulate Configuration I, the lower nodes of the members representing the four corner legs were restrained from translation. Seismic re_ponse and stress analyses described in the previous paragraphs assumed rack Configuration I. Stress analysis performed using Configuratica II is described in the next paragraph, During a postulated seismic event, when the rack is supported on one leg for a very short duration, the equilibrium is maintained by inertia loads. For this condition, the external loads acting on the rack, which constitute dead load and seismic loadings were simulated by trial and error method to determine the condition which causes incipient tilting motion of the rack on one leg. The rack structure was then analyzed for the resulting loadings. In general, these loadings caused stresses higher than those obtained from Configuration I, and are considered as upper bound loadings. 3.4.4.3 Thermal Stress Analysis Thermal stresses due to loads resulting from temperature gradient due to a single hot spent fuel bundle being placed in a rack cell with adjacent cavities empty (Q) are self-relieving, and so can be neglected in accordance with USNRC Standard Review Plan 3.8.4 (Reference 6). However, as a conservative approach these stresses were considered in the present analysis. These stresses were evaluated for a temperature differential as shown in Figure 3.4-8 and using a the finite element model of the rack. 3-41 s.

DUCLERR SERVICES CORPORRTIOn Q(_A DIVIlick 08lADREX CO AGO A ATION $ ELEVATION A TOP 0F THE FilEL TUBES TOP OF ACTIVE FUEL - O BOTTOM 0F ACTIVE FUEL

                                                         >  AT(*F) 0              35 FIGURE 3.4-8 THERMAL GRADIEtiT O
                                                         !877 !45 3-42

QUAD-1-79-509 DUCLEAR SERVICES CORPORATIOD 10iveS104 08 QUADREX CO A POR ATIO N 3.4.4.4 Grapple Load Analysis The forces that might occur if a fuel assembly were to jam in the racks during removal are conservatively estimated at: 70C0 lbs. in the vertical direction and 3500 lbs. in the horizontal direction; however, the resultant of the simultaneously applied forces not to exceed 7000 pounds. Stresses due to these acciderital U loads were computed by an elastic analysis. 3.4.4.5 Analysis of Impact Between Two Adjacent Racks In the event of an SSE, the racks may potentially slide and impact on each other. In this analysis the potential damage to the rack due to this postulated impact is evaluated. The following acceptability criteria was used: Any structural part of the rack which is necessary to maintain Keff < 0.95, should remain elastic. Thus, any local plastic deformation caused by the above postulated impact must be limited to that portion of the rack which is not required to maintain fuel assembly spacing. The rack construction is such that the base assembly is projected outward from the rack storage tubes and frame. Thus, in the event of an impact, the two adjacent base assemblies will impact on each other. A portion of the impact energy may be absorbed by the local deformation of the base assembly. The remaining energy, if any, will be spent by a combination of the following: a) Bottom of the racks will rebound moving away from each other. 1877 144 3-43

QUAD-1-79-509 AUCLEAR SERVICES CORPORATION A DiviliO4 08 Q/lADAEX CO APO A ATION b) The tops of the racks may tip towards cach other due to inertia, in which case, a part of the impact energy will be spent in tilting the rack. c) If there is any more of the energy left, the racks may impact on each other at the top corners, in which case the remaining energy will be spent in deforming the top corners of the racks. The above-mentioned phenomena were analyzed to determine the following: a) The energy spent in deforming the base assembly structure. b) The energy spent in tilting one rack to a position in which its top corner would just touch the top corner of the identically-tilted adjacent rack. The energy of impact between two racks was assumed to be equal to the kinetic energy of the racks. The latter was computed using the peak sliding velocity of the rack determined from the nonlinear sliding analysis described in Section 3.4.4.1. The kinetic energy was computed conservatively using the mass value of the heaviest rack (i.e., 7 x 8 size rack). The energy spent in tilting the rack to a position in which its top corner would just touch the top corner of the adjacent rack was computed as the work done in this process. In this computation it was conservatively assumed that the bottom base assembly of the adjacent racks are still in contact while the top of the racks are tilting toward each other. The energy required to tilt the rack was deducted from the initial impact energy to determine the energy 9 3-44 \0 ,

QUAD-1-79-509 RUCLERR SERVICES CORPORRT100 a Devv510m 08 QfIADAEX C O AP O A ATIO N with which the top of the racks may impact each other. Thus, any energy absorbed in the elastic impact of the base assembly and, more significantly, in the rebounding of the bottom of the racks are neglected. 3.4.4.6 Analysis of the Impact Resulting From the Drop of a Fuel Bundle This section describes the analysis performed to determine the consequences of an accidental drop of a fuel bundle onto the rack from a height of 18 inches above the top of the rack. The objective of this analysis was to ensure that, in the accidental event of dropping a fuel bundle on the proposed rack at any location, the deformed configuration of the rack would still maintain the criticality coefficient keff < 0.95. This criticality criteria was translated to the following equivalent structural criteria: The resulting deformation state shall be such that the structure which , maintains the fuel spacing in the active fuel region remains within the elastic limit. Using energy balance methods, an elasto-plastic analysis of the rack was performed to determine the maximum length of the rack that might be stressed beyond elastic (yield) limit in the event of a postulated 18 inches drop of a fuel bundle at the most critical location on the rack. No credit was taken to account for the fluid energy dissipation during the drop phase, nor was any credit taken for the energy absorbed due to the crushing of the fuel bundle. 3-45 1877 146

QUAD-1-79-509 DUCT. ERR SERVICES CORPORRTIOD Q(.A Devill0*W 08.lADREX CO APO A ATION 3.4.4.7 0verturning Analysis To ensure that the proposed racks are laterally stable against overturning during an SSE, stability analysis has been performed very conservatively using an energy method which assumes the following phenomenca of sliding and overturning: At the instant the rack attains the maximum sliding velocity, it has the maximum kinetic energy. At this instant, it is assumed that all the kinetic energy of the rack is suddenly transformed to cause overturning of the rack. During this process when the rack is getting tilted, no sliding motion is assumed. Since the rack can potentially slide during the time required to tilt the rack to an unstable position, and thereby can dissipate a significant part of the energy in overcoming the friction force, this assumption is very conservative. Other sources of conservatism in the stability analyses are as follows: a) To compute the kinetic energy of the horizontal motion that contributes to the overturning potential, the sliding velocity of the rack is computed using the minimum coefficient of friction (see Section 3.4.4.1), which produced the largest velocity. On the other hand, this energy was assumed to be suddenly transformed to cause overturning, which is equivalent to an assumption that the translatory motion of the rack is stopped suddenly because of infinite friction coefficient. Thus, the overturning potential computed by this method is likely to be an upper bound value, i.e., the method is very conservative. b) Viscous drag force of water during sliding and overturning has not been considered. 3-46 1877 147

QUAD-1-79-509 DUCT. ERR SERV!CES CORPORATI0n

                                    & Devi5tDN OF Q'JUADREX CO APO A ATION c)    Effect of vertical earthquake is considered conservatively by assuming that a constant vertical acceleration equal to tt.e vertical floor acceleration acts constantly upward, thereby increasing the oterturning potential. Since, it is unlikely that the vertical component of the earthquake motion will peak at the same time as the horizontal motior., the above assumption is conservative.

Using the above-listed conservative assumptions, the factor of safety against overturning was computed as the ratio of the potential energy of the rack at the incipient overturning position (i.e., when itr center of gravity is directly above the tip over edge) to the maximum kinetic energy of the sliding rack. 3.4.5 Analysis of 7 x 7 Rack From past exper;ence it has been observed that, of all the loads listed in Section 3.4.2.1, seismic loads are by far the most dominant loads. It has also been observed that the rack which has the least base width and highest ratio of plan length to plan width, has higher seismic stresses. From these two considerations, the 7 x 8 rack was considered to be more critical than the 7 x 7 rack. So, the 7 x 8 rack was analyzed in detail (Section 3.4.4). Stresses in the 7 x 7 rack are conservatively evaluated in this section extrapolating the stresses from the 8 x 7 rack, when necessary, tionlinear sliding a ilysis was performed for the 7 x 8 rack. The sliding movement is proportional to the horizontal inertia force resulting from the horizontal virtual mass which is equal to the actual mass plus the added hydrodynamic water mass. The force resisting the sliding rrevement is proportional to the actual mass only. The higher is the ratio of the virtual mass to the actual mass, the higher is the potential for sliding. Since, the 7 x 8 3-47 lh(( .

                                                                    }40'

+ 0

QUAD-1-79-509 DUCLEAR SERVICES CORPORATIOD a CiviliO4 0F QUADREX CO APO A ATION rack has higher plan length to plan width ratio, it has larger virtual mass to actual mass ratio, and so has higher sliding and overturning potential. Thus, the sliding and overturning potentials calculated for the 7 x 8 racks are upper bound values for the 7 x 7 racks. Also, since the mass of the 7 x 8 rack was used to compute the energy of impact between two racks, the factor of safety against having unacceptable damage potential during SSE event computed in Section 3.4.4.5 is also applicable to 7 x 7 size racks. Stresses due to dead load plus buoyancy loads in 7 x 7 rack will be less than those in 7 x 8 rack since 7 x 8 rack is bigger and heavier. Stress analyses due to Q and U loads (Sections 3.4.4.3 and 3.4.4.4) and the analysis of the impact resulting from the drop of a fuel bundle (Section 3.4.4.6) were performed on local models which are applicable to both 7 x 8 and 7 x 7 racks. Thus, it is concluded that the stresses computed for the 7 x 8 rack are upper bound values for the stresses in 7 x 7 racks. 3.4.6 Analysis Results and Design Evaluations Results of the analyses described in the previous sections showed that the proposed spent fuel racks, when subjected to various possible load combinations shown in Table 3.4-2, would meet the indicated structural acceptance criteria, and would have adequate factors of safety against any deformation state for which K can eff be greater than the permissible value of 0.95. Results of the upper bound sliding analysis for the loaded rack showed that the maximum sliding was 0.47 inch with a maximum sliding velocity of 4.43 inches /sec. For impact and overturning evaluation, the peak sliding velocity value was conservatively 3-48 9 1877 i49

QUAD-1-79-509 NUCLEAR SERVICES CORPORRTIOR A Devl5lCN CF Q~(352[351EE); CO A AQ A ATION rounded to 4.5 in/sec. For the empty rack the maximum sliding was 1.30 inch with a maximum sliding velocity of 5.6 inches /sec. Although the sliding velocity is higher for the empty rack, the total kinetic energy of the loaded rack is greater due to the added mass of the fuel. Therefore, the impact analysis was bast. on the loaded rack. Results of the eigenvalue analysis for the 7 x 8 rack are summarized in Table 3.4-5. Results of stress analyses are summarized in Tables 3.4-6 and 3.4-7. The values presented were calculated for the 7 x 8 rack and are upper bound values for the 7 x 7 rack. Comparison of the computed stresses for various load combinations with the corresponding allowable stresses shows that the stresses are well within the allowable limits. Results of stability analysis of the fully-loaded and the empty racks, when subjected to SSE motion, are summarized in Table 3.4-8. The resulting factors of safety against possible overturning are more than the minimum required value of 1.1. Since the minimum required factor of safety against overturning during OBE event is 1.5, while the computed minimum factor of safety for SSE loading is 25, no overturning analysis was needed for OBE loads. Elasto-plastic analysis of the rack for the accidental fuel bundle drop showed that the maximum length of the rack which might be stressed beyond the elastic limit is 3.17 inches, whereas the available length of the racks above the active fuel length is about 17.4 inches. 1877 150 3-49 -

DUCLERR SERVICES CORPORRT100 A DriviliON 08

                       'UADREX CO A AQ A ATION TABLE 3.4-5 RESULTS OF FREQUENCY ANALYSIS OF 7X8 RACK liodal Participation Factors Mode (I) Frequency Horizontal No.          (cps)

Short Dir. Long Dir. Vertical 1 4. 71 2 2.115 ** 0.06 2 5.738 0.01 6.64 0.22 3 5.829 0.03 5.74 0.23 4 6.789 1.23 0.25 ** 5 8.234 0.35 1.21 0.01 6 10.252 1.39 ** ** 7 10.253 1.28 ** ** 8 10.509 ** 1.14 ** 9 10.510 ** 1.08 ** 10 10.527 ** 1.07 ** 11 21.519 ** ** 1.44 12 33.292 0.09 0.31 1.23 13 39.781 ** 0.03 1.32 (1) Only the significant modes are listed

  **   Modal participation factor less than one-thousandth 3-50

RUCLSAR SERVICES CORPORATION A civiSions OF

                                        % I U R D R I 'T CO APO AATION TABLE 3.4-6 STRESS EVALUATION FOR CRITICAL ELEMENTS CF THE SPENT FUEL RACK CONFIGURATION I Rack (1).                 Critical                     Comouted     Allowable Component                    Load                      Stress (2)  Stress Index Combination                     Index Upcer Grid                       D+S+Q+E'                        .717        1.000 Lower Grid                       D+B+E                          .555         1.000 vertical Corner                  0+B+Q+E'                       .197         1.000 L 8eam Cross Brace                      D+8+Q+E'                       .399         1.000 Middle Strap                     D+B+E'                         .609         1.000 absorter Tube                    0+B+U                          .974         1.000 Base Assembly 5eam               0+B+E                          .434         1.000 Rack Leg                        0+9+Q+E'                        .280         1.000 Floor Leg Connection            0+B+Q+E'                        .348         1.000 NOTE: (1) See Figure 3.2-1 x2 (2) Stress Index =      +
                            "a    "bx2     "bx3 Where, f, = computed max, axial stress f       computed max. bending stress h2 = about the major axis.

f = computed max. benaing stress bx3 about the minor axis. F, = allowable ccmcressive stress Fbx2 = allcwable bending stress about the major axis. Fbx3 = allowable bending stress about the minor axis. 1877 152 3-51

DUCLERR SERVICES CORPORATION n . o,,,,,,, o, W

                                   %IUADREX CO A AO A ATlDN TABLE 3.4-7 STRESS EVALUATION FOR CRITICAL ELEMENTS OF THE SPENT FUEL RACK CONFIGURATION II Rack                          Computed                 Allowaole Com onent (1)                      Stress (2)               Stress Index Index i

l Uccer Grid ,979 1.000

Lower Grid .765 1.000 O

i

                                              .678                      1.000 lVerticalCornerLBeam Cross Brace                                  .561                      1.000 i

Micole Strap .326 1.000 i Absorcer Tube ! .161 1.000 Base Asseetly Beam .966 ! 1.000 l i Rack Le9 .502 1.000 Floor Leg Connection .834 1,000 Note: (1) See Figure 3.2-1

          '2} See Tacle 3.4-5 footnotes for explanation of syrcols.

9 1877 153 3-52

DUCT. ERR SERVICES CORPORATI0n A DiviS40% 08 M(lADREX CO A AD A ATION TABLE 3.4-8 RESULTS OF OVERTUP,NING ANALYSES Parameters Computed (1) Fully Loaded Empty Rack Rack KE (Kip-inch) 4.64 3.54 PE (Kip-inch) 390.85 87.59 AV (inch) 5.76 5.76 F.S 84 25 (1) KE = Maximum Vsinetic Energy of the Sliding Rack Due to SSE Motion PE = Potential Energy when the Rack Mass is Lifted Through 4 V AV = Vertical Displacement of the Center of Gravity of the Rack necessary to change the rack equilibrium from a stable state to critical one F.S = Factor of Safety against Overturning @ l877 154 3-53

QUAD-1-79-509 nUCLERR SERVICES CORPORRT100 Q(.& Devilf 04 Of.lADAEX CO APO A ATION W The possibility of two adjacent racks impacting on each other was evaluated using the method described in Section 3.4.4.5. Results of the analysis show that: a) When two adjacent racks impact on each other at the maximum sliding velocity of 4.5 inches /sec the postulated plastic deformation of the base assembly is very localized and insignificant. b) Even if the energy absorbed in deforming the base assembly is ignored, the kinetic energy due to seismic motion is less than the potential energy required to tilt the racks to a position where the tops of the racks can impact on each other. These, together with the f act that the available length of the racks above the active fuel length is about 17.4 inches, show that the racks are safe against postulated impact between each other during seismic events. 3.5 Thermal-Hydraulic Analysis The decay heat generated by the spent fuel stored in the proposed racks is disipated into the fuel pool water. This water is then cooled by the spent fuel pool cooling system. In order to ensure adequate cooling of the fuel and of the fuel pool water two themal analyses have been performed. Th? first considers cooling of ar.

"dividual fuel assembly by natural circulation of water through the fuel storage tubes. The second considers overall cooling of the spent fuel pool by the Spent Fuel Pool Cooling System (SFPCS).

3.5.1 Fuel Assembly Cooling Analysis 3-54 1877 155_

QUAD-1-79-509 DL' CLEAR SERVICES CORPORATIOR Q(.A DivtSsON 08.IADREX CO A AQ A ATION 3.5.1.1 Analysis Method This analysis is performed to ensure that each fuel assembly receives adequate cooling. All cooling of the stored fuel is assumed to be the result of natural circulation caused by the heat generation of the fuel in the storage racks. Nuclear Services Corporation has developed the computer code CIRCUS for analysis of natural circulation in spent fuel assemblies. The code treats a series of fuel assemblies fed by a single down-comer, through a series of inlet areas. Viscous losses through the downcomer, inlet and fuel channels are balanced with buoyant forces developed by power generation in the fuel channels. The upper pool is assumed to be maintained at a constant temperature. Outlet temperatures are computed based on an energy balance. An iterative procedure is used to balance the forces. Natural circulation cooling in the spent fuel pool was modeled as shown in Figure 3.5-1. A peak power fuel assembly was assumed to be stored in the center of.the pool (position 11) at the end of a row of average power fuel assemblies (fuel assemblies whose decay heat is based on the average fuel assembly power in the core). Flow to this row of fuel assemblies was assumed to follow a path which takes the cooling water from the upper pool, down the side of the pool (between the fuel racks and the pool wall) and under the fuel storage racks. This model gives an upper bound for outlet temperatures, since flow from other directions is neglected. The flow of water into the corner storage tubes of each rack is more restricted than to the other tubes, since water must flow through the leg assemblies to reach the corner tubes. Because of this restriction, a check was made for the above described model with the peak power fuel assembly stored in position 8 and average power assemblies in the ten o'Ser positions. 3-55 1877 156

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QUAD-1-79-509 DUCLCAR SERVICES CORPORRTIOR 4 Olvil'O4 Of QUADREX CO APO A ATION Using the models described above, the temperature increase of the water as it passes through the storage tube containing a spent fuel assembly was calculated for a time af ter reactor shut down of 100 hours. 3.5.1.2 Analysis Results For the base case with the peak power assembly in position ll, the temperature increase of the water flowing through the storage tube was determined to be 35 F with a corresponding flow rate of 7485 pounds / hour. The case with the peak assembly in position 8 produced almost identical results of 35.2 F temperature increase and 7454 pounds / hour flow rate. In both cases the temperature increase for positions with the average power assemblies was less than that described above. The numbers calculated above are based on a bulk pool temperature of 150 F. However, the bulk pool temperature has a very minor effect on the temperature increase in the fuel storage tube due to water density variation. 3.5.2 Pool Cooling Analysis 3.5.2.1 Analysis Method The heat generation and pool temperature were calculated using the Nuclear Services Corporation computer code P00LHT. P00LHT calculates fuel decay heat based on Branch Technical Position APCSB 9-2. P00LHT performs an analysis of fuel pool temperature as a function of heat input from spent fuel, heat rejection through the pool 3-57 *$77 I50

Q'JAD-1 509 DUCl. ERR SERVICES CORPORATI0n

                             ...mo, QUADREX CO APD A ATION cooling systems, pool water mass and time. The heat re.iection rate in the heat exchangers is calculated based on heat exchanger inlet temperatures, heat transfer coefficient, effective heat transfer surface area, and primary and secondary water flow rates. Finally the time-dependent pool temperature is calculated by an energy balance on the spent fuel pool water.

For this analysis it has been assumed that the first fuel assembly is removed from the reactor and placed into the fuel pool 100 hours after reactor shutdown. The total amount of fuel to be discharged is assumed to be transferred to the fuel pool within 150 hours af ter reactor shutdown. This same assumption applies to both normal refueling and the full core discharge. Fuel loading occurs annually for each unit and will be staggered between the two units so that about one-third of a core will be unloaded approximately every six months. Under normal conditions, this will continue until 1362 locations are filled. As stated earlier a total of 1582 storage positions will be provided. However,121 positions will be reserved for a full core discharge and 75 positions will be used only during installation of racks in pool 2. With 1362 positions filled from normal refueling the remaining 24 positions are potentially available for r.ormal fuel storage. There are no plans to use these positions, but they are maintained for storage required by nexpected fuel replacements. The increased heat generation due to these 24 assemblies is negligible. In confonnance with the analysis performed for the previous fuel storage modification at Prairie Island the following definitions a'pply for potential pool cooling conditions:

                                                              \s11 \S9 3-58

QUAD-1-79-509 DUCLERR SERVICES CORPORRTIOD A Divil6DN OF @ WUADREX CO APO A ATION Normal - Spent fuel pools contain 1362 spent fuel assemblies, leaving room for a full core discharge. One pump and the larger heat exchanger are assuned to be in operation. Abnormal - Spent fuel pools filled with 1483 fuel assemblies, including a recently removed full core. Both pumps and both neat exchangers are assumed to be in operation. Faulted - Less of operation of a pump from the spent fuel pool cooling system when the spent fuel pools are filled with 1483 fuel assemblies, including a recently removed full core. One pump and the larger heat exchanger are assumed to be in operation. Accident - Loss of all external spent fuel pool cooling when the spent fuel pools are filled with 1483 fuel assemblies, including a recently removed full core. The " Normal" condition assumes that a routine refueling has just taken place, and that the discharged one-third core results in 1362 spent fuel assemblies stored in the pool. The POOLHT code is then used to determine the heat generation and fuel pool temperature as a function of time after reactor shutdown. The " Abnormal" condition assumes that after normal operation which results in 1362 spent fuel assemblies in the pool, a full core discharge of 121 assemblies occurs. It is assumed that this full core discharge occurs from one unit 30 days after refueling of the other unit. The assumed refueling data used in these analyses is shown in Table 3.5-1. The cooling system data used is shown in Table 3.5-2. 3-59 igJ7 140

TABLE 3.5-1 REFUELING DATA Normal Refueling Full Core Discharge Batch Number of Total Assenblies Cooling Irradiation Cooling Irradiation Number Assemblies In Pool Time Time Time Time 1 40 40 6022.5 876 6052.5 876 2 40 80 5840 876 5870 876 3 40 120 5657.5 876 5687.5 876 4 40 160 5475 876 5505 876 5 40 200 5292.5 876 5322.5 876 6 40 240 5110 876 5140 876 7 40 280 4927.5 876 4957.5 876 8 40 320 4745 876 4775 876 Y2 9 40 360 4562.5 876 4592.5 876 $ 10 40 400 4380 876 4410 876 11 41 441 4197.5 876 4227.5 876 12 41 482 4015 876 4045 876 13 40 522 3832.5 876 3862.5 876 14 40 562 3650 876 3680 876 15 40 602 3467.5 876 3497.5 876 16 40 642 3285 876 3315 876 17 40 682 3102.5 876 3132.5 876 18 40 722 2920 876 2950 876 19 40 762 2737.5 876 2767.5 876

    ---  20       40           802             2555               876      2585              876

((] 21 40 842 2372.5 876 2402.5 876

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   ,    TABLE 3.5.1, REFUELING DATA, Cont'd.

Normal Refueling Full Core Discharge 8atch Number of Total Assemblies Cooling Irradiation Coolir.g Irradiation Number Assemblies In Pool Time Time Time Time

      .22               40                882         2190              876      2220               876 23              40                922         2007.5            876      2037.5             876 24              40                962         1825              876      1855               876 25              40               1002         1642.5            876      1672.5             876 26              40              1042          1460              876      1490               876 27              40              1082          1277.5            876      1307.5             876 28              40              1122          1095              876      1125               876 29              40               1162          912.5            876       942.5             876 30              40              1202           730              876       760               876 y      31              40               1242          547.5            876       577.5             876 23      32              40              1282           365              876       395               876 33              40              1322           182.5            876       212.5             876 34              40              1362            0               876        30               876 35              41              1403            -                -

0 754 36 40 1443 - - 0 462 37 40 1483 - - 0 170 Co N N __. Note: All times are given in days. ch N.

DUCl.CRR SERVICES CORPORRT100

                                 & DevilsDN 08 VdADREX CO APO A ATION                                    $

TABLE 3.5-2 Spent Fuel Pool Cooling System Data A. Heat Exchanger 1 Type Four-pass shell and U-tube 0 Design Heat Transfer 7.89 x 10 Btu /hr Shell Tubes Design Temperature 200 F 200 F Design Flow Rate 1810 gpm 1,415 gpm Design Inlet Temperature 95 F 120 F Design Outlet Temperature 104 F 109.6 F Fluid Circulated CC wtr SFP wtr B. Heat Exchanger 2 Type Two-pass shell and U-tube 6 Design Heat Transfer 7.89 x 10 Btu /hr Shell Tubes Design Temperature 200 F 200oF Design Flow Rate 1810 gpm 1,415 gpm Design Inlet Temperature 95 F 140 F Design Outlet Temperature 104oF 129oF Fluid Circulated CC wtr SFP wtr Note: The data shown above is the original heat exchanger design data based on the original tuba side flow rate of 1,415 gpm. In 1977 the SFPCS pumps were modified to int ease the tube side flow rate to 2,200 gpm. The fuel pool calculations are based on this increased flow rate. 3-62 g7 }63

QUAD-1-79-509 DUCLERR SERVICES CORPORRTIOR A DfveliO4 08 QUADREX CO APO A ATION 3.5.2.2 Analysis Results Figures 3.5-2 and 3.5-3 show the heat generation and pool water temperature as a function of time for the normal condition. The maximum pool temperature for this condition is 124 F. Figures 3.5-4 and 3.5-5 show the heat generation and pool water temperature as a function of time for the abnormal condition. The maximum pool temperature for this condition is 141"F. Figure 3.5-6 and 3.5-7 show the heat generation and pool water temperature as a function of time for the faulted condition. The maximum pool temperature for this condition is 164 F. This calculation is based on the conservative assumption that only one pump is in operation from the time of reactor shutdown. In actual plant operation, the full core discharge would be delayed if the SFPCS was not in full operation. Therefore, the temperature curve shown in Figure 3.5-6 is an upper bound. For the accident condition the pool water temperature would rise until boiling at the pool surface began. For this condition the report submitted for the racks presently in service showed that the maximum fuel clad surface temperature for this condition would be 252 F. This is well below the normal fuel operating temperature. Based on the assumption that all cooling is lost at the peak heat generation point for the abnormal condition, 2.9 hours would elapse before initiation of boiling in pool 1. The maximum evaporation rate would occur at this time and would equal 44.7 gpm. This assumes ~..i 1 is isolated and the recently discharged full core is located in pool 1. If complete mixing of water in the two pools is assumed, 9.9 hours would elapse before initiation of boiling. 3-63 1877 164

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+ fiORMAL C0tIDITION P0OL TEMPERATURE 1877 166 3-65

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0.00 l'0.00 2'O.00 3'O.00 tiO.00 5'O.00 6'O.00 DAYS AFTER SHUTDOWN rIGURE 3.5-7: PRAIRIE ISLAND 1877 170 FAULTED CONDITION . POOL TEMPERATURE 3-69

QUAD-1-79-509 DUCLERR SERVICES CORPORATIOD [kb3EX C O A P O A ATIO N $ Makeup to the spent fuel pit cooling system can be provided from

                                                                                  ' " ' ~ ~

the following systems: a) Chemical Volume and Control System - provides borated water (at concentration selected by the operator) to the discharge side of the SFP heat exchanger b) Demineralized Water - provides demineralized water to the discharge side of the DFP demineralizer c) Reactor Makeup Water - provides demineralized water to the discharge side of the SFP heat exchanger d) Refueling Water Storage Tank e) CVCS Holdup Tanks via the CVCS Holdup Tank Recirculation Pump f) Fire Protection System 3.6 Radiochemical Analysis The purpose of this analysis is to determine the dose rate at the pcol surface as a result of the concentration of fission products in the fuel pool water. 3.6.1 Analysi, Method Based on the refueling schedule and fuel leak rate, the radioactivity content of the fuel pool water in curies per unit volume is calculated. From this value the surface dose rate is determined. To solve this problem, Nuclear Services Corporation has developed the P00LRAD computer program. The model on which the code is based 3-7 1877 17i ,

QUAD-1-79-509 DUCLERR SERVICES CORPORRTIOR

                                  & Dewf 510N OF QUADREX CO A AD A ATION treats the fission product concentration in the pool as a dynamic equilibrium between the pool cleanup system and the diffusion of fission products from the fuel matrix. A closed form solution for the leakage rate from the fuel was developed based on time dependent diffusion from a sphere.

Initial concentrations of 30 fission product isotopes are computed for the time of fuel discharge from the reactor in each of several fuel batches. The equilibrium concentration is then computed for each isotope on a batch-by-batch basis for each of several input time points. Total curies per cubic centimeter and pool surface dose rate are then computed at each time. The results are obtained for 6, 10, 20, 40 and 100 days after shutdown. The isotopic data were obtained from the ORNL-4628 "0RIGEN- The ORNL Isotope Generation and Depletion Code". The cleanup efficiency for the different isotopes were obtained from NUREG-0016. It is assumed in this analysis that one percent of the fuel leaks. The diffusion coefficient for Cesium is taken as 10 19 sq. cm/sec. This is the same as the value for Krypton-85 below 800 C and is considered to be a conservative value. Data published for reactor primary systems at power would lead to the use of a lower diffusion coefficient for Cesium and, therefore, lower calculated dose rates. 3.6.2 Analysis Results The results of this analysis indicate that Cs will be the main fission product contaminant in the spent fuel storage pool. Table 3.6-1 presents the calculated water radioactivity content and surface dose rate for 1% fuel leakage. The results are well within acceptable limits. 3-71 lb77 172

QUAD-1-79-509 RUCLEAR SERVILES CORPORATIOD A Devel10N OF QdADREX C O A P O R A TIO N TABLE 3.6-1 FUEL POOL WATER RADI0 ACTIVITY CONTENT AND SURFACE DOSE RATE FOR 1% FUEL LEAKAGE Radioactivity Surface Dose Days From Content Rate Shutdown (curies /cc) (mr/hr) 6 4.05 x 10 10 0.44 10 2.69 x 10 10 0.29 20 1.51 x 10 10 0.16 40 1.00 x 10 10 0.11 100 8.07 x 10 11 0.09 0 3-72 1877 173

QUAD-1-79-509 DUCLEAR SERVICES CORPORATION A DIVISION D8 Q'CO/JADREX APO A ATION 3.7 Fuel Rack Installation The spent fuel racks will be installed in the two spent fuel pools as shown in Figure 3.1-1. Th installation will be done according to the following sequence:

1. The spent fuel (442 assemblies) will be stored in pool 2.

2. The portable 15 ton crane will be installed on the fuel handling bridge rails.

3. Using the portable crane and the main crane the present racks will be removed from pool 1.

4. Some of the existing anchor bolts will be cut off flush with the pool liner in order to provide clearance for the rack feet.

This work will be done in accordance with approved procedures and proven methods in order to avoid damage to the pool liner.

5. The new racks will be installed in pool 1.

6. All spent fuel will be transferred from pool 2 to pool 1.

7. The protective cover will be installed on pool 1.

8. The present racks will be removed from pool 2.

9. The new racks will be installed in pool 2.

10. Spent fuel will be returned to pool 2.

This procedure will not require the movement of heavy equipment above stored fuel with the exception of movement of equipment above pool 1 when the protective cover is in place. The protective cover has been designed to withstand the drop of a spent fuel rack. 1877 174 3-73

QUAD-1-79-509 DUCLERR SERVICES CORPORATIOD

                                     & DIVillON OF Q7lRDREX CO APO A ATION

4.0 CONCLUSION

S The proposed modification to the Prairie Island Nuclear Generating Plant spent fuel storage as described herein has been evaluated in the following areas:

1) Fuel Rack Mechanical Design

2) Nuclear effects

3) Structural and seismic effects, including fuel handling accidents

4) Thermohydraulic effects

5) Radiochemical effects

6) Fuel rack installation The results of the above evaluations have shown that there are no unresolved safety problems associated with the proposed modification.

O 4-, 1877 175

QUAD-1-79-509 DUCLEAR SERVICES CORPORATION 4 DeVISION 08 QUADREX CO A PO A ATIO N

5.0 REFERENCES

1. CHEETAH - A unit cell homogenization and spectrum generation (LEOPARD-CINDER) program with a fuel depletion option. Output in PDQ-7 format.

2. XSDRN - A discrete ordinates, space dependent spectral code for the generation of multi-group cross sections in the fast, resonance, and thermalization regions. Output in CITATION, MORSE, and DOT, format.

3. CITATION - (Revision 2, Supplement 3) 1, 2, or 3 dimensional diffusion theory, multigroup, finite difference code designed to solve depletion problems with or without elaborate refueling treatment for multi-cycle analysis. First order perturbation theory edits available.

4. " KENO IV - An improved Monte Carlo Criticality Program", ORNL-4938, Oak Ridge National Laboratory (Nov.1975), L.M. Petrie and N.F. Cross.

5. " Seismic Design Classification", USNRC Regulatory Guide 1.29, Revision 2, February 1976.

6. "Other Category I Structures", USNRC Standard Review Plan Section 3.8.4.

7. "ASME Boiler and Pressure Vessel Code", Section III, Division 1, 1977 Edition. ,

8. " Friction Coefficients of Water Lubricated Stainless Steel for a Spent Fuel Rack Facility", by E. Rabinowicz, Professor, Massachusetts Institute of Technology, November 5, 1976.

~ 1877 176 5-1

QUAD-1-79-509 RUCI. ERR SERVICES CORPORRTIOD

                                   . o,m o. o, 2.lADREX
                               ~

CO A AD A ATION

9. "ANSYS - An Engineering Analysis Program", Swanson Analysis Systems, Inc., Elizabeth, Pennsylvania, 1975.

10. " Table of Hydrodynamic Mass Factors for Translation Motion", by K. T. Patton, ASME Publication No. 65-WA/UNT-2.

11. " Design Objectives for Light Water Reactor Spent Fuel Storage Facilities at Nuclear Power Stations", ANSI Standard N210-1976, published by American Nuclear Society, Hinsdale, Illinois, April 12, 1976.

12. " Foundations", USNRC Standard Review Plan Section 3.8.5.

13. " Combination e' Modes and Spatial Components in Seismic Response Analysis", USNRt, Regulatory Guide 1.92, December 1974.

14. " Revised Earthquake Analysis for Prairie Island Nuclear Generating Plant", by John A. Blume & Associate Engineers, February 16, 1971.

15. NSCTH - A Computer Program to Generate Spectrum Compatible Time-History Motion, Nuclear Services International Corporation, 1977.

16. STARDYNE - A Static and Dynamic Structural Analysis Systems, Mechanics Research Institute, Los Angeles, California.

17. " Specification for the Design, Fabrication & E"ection of Structural Steel for the Building", American Institute of Steel Construction, February 1969.

O 1877 177 5-2

EXHIBIT D PRAIRIE ISLAND NUCLEAR GENERATING PLANT License Amendment Request dated January 31, 1980 Docket Nos. 50-282 License Nos. DPR-42 50-306 DPR-60 Exh ibit D consists of the Nuclear Services Corporation document: QUAD-1-79-558

   " Fuel Pool Building Structural Evaluation for Prairie Island Nuclear Generating Plant Units 1 and 2 Spent Fuel Storage Modification" 1877 178

QUAD-1-79-558 JOB NO. NOR-0174 FUEL POOL BUILDING STRUCTURAL EVALUATION FOR PRAIRIE ISLAND NUCLEAR GENERATING PLANT UNITS 1 AND 2 SPENT FUEL STORAGE MODIFICATION NSP PROJECT NUMBER E-78YO75 Prepared for NORTHERN STATES POWER COMPANY Minneapolis, Minnesota By: NUCLEAR SERVICES CORPORATION A Division of Quadrex Corporation 1700 Dell Avenue Campbell, California 95008 1877 174 0 h h ', 'd }lQ ,, L. Y l0 ' ' ll * ? ~19 REV. PROJECT QA PROJECT ISSUING DATE OF N0. ENGINEER REVIEWER ENGINEER ENGINEER MANAGER APPROVAL

DUCI. ERR SERVICES CORPORATI0n A Ceuts*CN CF JLIADAEX CO A AC A ATION CAMPBELL CALIFORNIA 95008 TABLE OF CONTENTS Page

1.0 INTRODUCTION

1-1

2.0 DESCRIPTION

OF SPENT FUEL P0OL STRUCTURES 2-1 3.0 LOADS, LOAD COMBINATIONS & ACCEPTANCE CRITERIA 3-1 3.1 Loads 3-1 3.2 Load Combinations & Acceptance Criteria 3-2 4.0 METHOD OF ANALYSIS AND COMPlfTATION OF DESIGN LOADS 4-1 4.1 Spent Fuel Pool Floor 4-1 4.2 Spent Fuel Pool Walls Above Pool Floor 4-2 4.3 Shear Walls cnri Colurine Supporting the Pool Floor 4-2 5.0 EVALUATION RESULTS & CONCLUSIONS 5-1 5.1 Evaluation of Pool Floor 5-1 5.2 Evaluation of Spent Fuel Pool Canci Wall 5-1 5.3 Evaluation of Shear Walls & Cot .s Supporting the 5-1 Pool Floor 5.3.1 Shear Walls 5-1 5.3.2 Evaluations of Columns M b

                                                 - 10band Mb -7 b   ~

5.4 Evaluation Results and Conclusions 5-2

6.0 REFERENCES

6-1 i 1877 180

AUCLEAR SERVICES CORPORATION e . . . --,

                                 %l/_lADAEX CO APC'1 ATION

1.0 INTRODUCTION

This report presents the structural analyses and evaluations performed by Nuclear Services Corporation to determine the structural adequacy of the spent fuel pool building structures of Prairie Island Nuclear Generating Plant (Units 1 & 2) to support the additional loads that would result from the proposed fuel pool storage modifications. The configurations of the fuel paol structures and the arrangement of the proposed high density racks in the pool are described in Section 2.0. Section 3.0 lists the loads and load combinations for which the pool structures were evaluated. This section also describes the acceptance criteria used to determine the structural adequacy. Methods of analysis and load computations are presented in Section 4.0. The evaluation of section capabilities to withstand the computed loads has been summarized in Section 5.0. Section 6.0 lists the references used. g 1877 181 0542

DUCLEAR SERVICES CORPORATION A OlvisiO4 0F Ql.lADREX CO A AQ AATION

2.0 DESCRIPTION

OF SPENT FUEL POOL STRUCTURES The fuel storage area at Prairie Island Nuclear Generating Plant is located between the two reactor buildings, and consists of a new fuel pit, two pools for storing spent fuel, and a canal for transfer of fuel elements. Figures 2-1 and 2-2 show plans of the fuel storage area at elevations 715 ft. and 693 ft. A typical cross section through the pool is shown in Figure 2-3. The two spent fuel storage pools are designated as Pool No.1 and Pool No. 2. Pool No. 1, the smaller of the two pools, has inside plan dimensions of 18'-11" x 18'-3". Pool No. 2 has inside plan dimensions of 18'-11" x 43'-5" Normal water depth for both pools is about 40 feet. Pool No.1 is presently designated as cask loading and unloading area. However, in the proposed modification, as shown in Figure 2-4, this pool would store nine racks. Pool No. 2 will hold 21 racks. These racks are free standing and are not anchored to the floor or the walls (Refer-ence 1). For the purpose of evaluation, the fuel pool structures were divided into three groups: (1) the spent fuel pool floor supporting the racks, (2) the pool walls above the pool floor, and (3) the shear walls and columns M b

             - 10band M b -7 b below the pool floor.

g 1877 182 0542

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                                                                                                    \877 \8-3 FIGURE 2-3 TYPICAL CROSS-SECTION THROUGH SPENT FUEL POOL 2-4

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                                     & OsWis.ON OF QLIADREX CO A AO R ATIO N 3.0 LOADS, LOAD COMBINATIONS & ACCEPTANCE CRITERIA 3.1 Loads Fuel pool structures were evaluated for the following loads:

a) Deadweight (D) -- Weight of pool structure components including the buoyant weight of the proposed spent fuel racks. b) Hydrostatic Load (H) -- Loads on the pool flocr and walls exerted by the water in the pool. c) Operating Basis Earthquake (E) -- Loads from horizontal and vertical components of the Operating Basis Earthquake (0BE) using the appli-cable vibratory motion from Reference 2. d) Safe Shutdown Earthquake (E') -- These loads were assumed to be twice that of OBE loads. e) Thermal Load (To, Ta) -- Loads resulting frorr the increase in fuel pool water temperature during normal operatian (To) and during accident condition (Ta). For the modified pool arrangement shown in Figure 2-4, the buoyant deadweight of the loaded racks was computed as 2.0 kips per square foot (ksf). However, for the present evaluation this was conservatively increased to 2.4 ksf. Response spectra values for OBE horizontal and vertical motion corresponding to 1% damping are listed in Tables 3-1 and 3-2 and are developed from Reference 2. To determine the thermal loads, normal operating temperature of the pool water was assumed to be 120 F; water temperature during accident condition was assumed to be 212 F. 3-1 @ 0542 1877 187

DUCLEAR SERVICES CORPORAT100 A DsVsSiOAl Of Q(JADREX compO A ATION 3.2 Load Combinations and Acceptance Criteria The spent fuel pool structures were evaluated as Seismic Category I structures in accordance with USNRC Regulatory Guide 1.29 (Reference 3). Their structural adequacy was verified in accordance with USNRC Standard Review Plan Section 3.8.4 (Reference 4), using section strength per American Concrete Institute (ACI) Building Code 318-77 (Reference 5). Table 3-3 lists the pertinent loading combinations for which the pool structures were evaluated. It also lists the respective permissible loading limits. O 0542 32 0 1877 . 188

RUCLEAR SERVICES CORPORATIOR

                        -     ....s.o~o,
                        - COGRCREE A AC A A TION TABLE 3-1 RESPONSE SPECTRUM FOR OBE HORIZONTAL MOTION (1% DAMPING)

Period Frequency Acceleration (sec) (cps) (g's)

    .025                   40.00                   .073
    .050                   20.00                  .083
    .075                   13.33                  .094
    .100                   10.00                  .104
    .150                     6.67                 .130
   .200                      5.00                 .190
   .275                      3.64                 .595
   .325                      3.08                 .822
   .375                      2.67                1.227
   .425                      2.35                 .599
   .600                      1.67                 .308
   .800                      1.25                 .164 1.000                      1.00                 .140 1.500                        .67                .092 2.000                        .50                .043 1877 189

DUCI. ERR SERVICES CORPORATIOR

                         , & Oswi$ SON Cf URCREI:

canpanarian TABLE 3-2 RESPONSE SPECTRUM FOR OBE VERTICAL MOTION (1% DAMPING) Period Frequency Acceleration (sec) (cps) (g's)

  .025                   40.00                 .061
  .050                   20.00                 .065
  .100                   10.00                 .095
  .150                      6.67               .138
  .200                      5.00               .210 .
  .238                      4.20               .635
  .250                      4.00               .635
  .400                      2.50               .340
  .500                      2.00               .190
  .600                      1.67               .120
  .800                      1.25               .068 1.000                      1.00               .060 1.500                        .67              .042 2.000                        .50              .025 3-4 1877 190

NUCLEAR SERVICES CORPORATION A OtviSsCRs QF QUADREX CO A AQ A ATION TABLE 3-3 LOAD COMBINATION AND EVALUATION CRITERIA Loading Design Pertinent Load Combination Used (1) Loading (2) Condition Method Limit Service Working D+H S Stress D+H+E S Ul tima te 1.4 D + 1.4 H U Strength 1.4 D + 1.4 H + 1.9 E U (1.4 D + 1.4 H + 1.7 Tg + 1.9 E) 0.75 U Factored Ul timate D+H+T +0 E' V Strength 0+H+T a U D+H+T +a 1.25 E U D+H+T +a E' U (1) For definitions of symbols, see Section 3.1. (2) S is the required section strength based on the working stress design method and the allowable stresses defined in ACI-318-77. U is the section strength required to resist design loads based on the strength design methods described in ACI-318-77. 1877 191 3-5

DUCLEAR SERVICES CORPORATION

                                       & DivisiO4 0F WijRDREX CO A AQ A ATION 4.0 METHOD OF ANALYSIS AND COMPUTATION OF DESIGN LOADS This section describes the analyses and the computations of design loads for the components of the fuel pool structure listed in Section 2.0.

4.1 Spent Fuel Pool Floor The spent fuel pool floor consists of a 5'-11" thick reinforced concrete slab supported as shown in Figure 2-3. This was analyzed for dead loads and hydrostatic loads using a finite element model shown in Figure 4-1. For evaluating the pool floor, this model is conservative since it assumes one-way slab action. Also, the support provided by the 4'-9" thick canal wall (supported between two north-south walls and acting as a deep beam) was conservatively ignored. Instead, the dead weight of the canal wall was assumed to be supported by the floor. The analysis was performed using the computer program STARDYNE (Reference 7). To determine the worst loading, two load cases were considered, one with the transfer canal empty and the other with the canal full of water. To determine the seismic loads on the pool floor due to the vertical component of the earthquake motion, the out-of-plane fundamental fre-quency of the pool floor was computed using the method outlined in Reference 8. An upper bound value of the amplification factor in the vertical direction was obtained by computing the upper bound and lower bound frequencies of the pool floor slab and selecting the largest amplification factor in that frequency range. The OBE response in the vertical direction at 1% damping, thus computed, was 0.0639, but a conservative value of 0.089 was used in the evaluation. For SSE vertical motion, a value of 0.16g was used. The seismic loads on the pool floor resulting from the horizontal seismic motion of the pool water were computed using methods outlined in Reference 9. These loads were combined with those due to vertical seismic motion by the square-root-of-the-sum-of-the-square (SRSS) method. 4-1 . Os42 1877 192

RUCl. ERR SERVICES CORPORATIOD n CivissGN 09 Q(JADREX CO ApO A ATION Thermal loads in the pool floor slab due to T gand T were a computed using methods outlined in Reference 10 and assuming a cracked concrete section. Dead load, hydrostatic loads, seismic loads and thermal loads were combined in accordance with Table 3-3 to determine the design factored forces, moments, and shears. The resulting design values for the critical load combinations are listed in Table 4-1. 4.2 Spent Fuel Pool Walls Above Pool Floor The proposed high density spent fuel racks are free standing and would not have any lateral wall supports, and so would not impose any trans-verse loads on the pool walls above the ficor. Hence, with the proposed modification, the design loads for these walls would be smaller than those with the existing racks, since the existing racks are designed to transmit lateral seismic and thermal loads to the pool walls. Thus, no new evaluation of the pool walls above the floor is necassary. To be conservative, the above justification was not used in the case of the 4'-9" thick wall separating the pools and the transfer canal (see Figures 2-1 and 2-3). Even though, as a conservative approach, the spent fuel pool floor was evaluated assuming that it supports this canal wall, the latter was evaluated assuming that it carries half of the lead from the spent fuel pool floor as a deep beam supported between the two north-south cross-walls. The resulting design moments and shears for the critical load combinations are listed in Table 4-2. 4.3 Shear Walls and Columns Supporting the Pool Flcor The shear walls and columns supporting the pool floor are shown in Figure 2-2. Due to the increased deadweight of the racks and fuel bundles on the spent fuel pool floor, dead load stresses and seismic overturning stresses (compressive and shear) on these walls and columns 4-2 g 0542

                     ^                                                  1877 ~

m

NUCLEAR SERVICES CORPORATI0n a Divisione CF WLIADAEX CO APO A ATiON will be larger than those predicted during the earlier densification study (Reference 6). The difference between the existing rack loads on the pool floor (1.7 ksf) and the proposed rack loads with the high density racks (2.0 ksf) is only 0.3 ksf. This additional load is about five percent of the total floor load resulting from dead weight of the slab, existing racks, and canal wall and the hydrostatic loads. However, to be conservative, the shear walls and columns supporting the pool floor were evaluted for an additional floor load of 0.7 ksf instead of 0.3 ksf. The dead load compressive stresses and seismic bending and shear stresses thus computed were added to those computed earlier in Reference 6. These are listed in Tables 4-3 and 4-4. Columns bM - 10band M b -7 b support the canal wall separating the transfer canal and the pools. As was stated in Section 4.2, to be conservative, this wall was assumed to support the pool floor in deep beam action; hence the loads computed for the supporting columns are also very conservative. The design loads on these columns were determined by adding the reaction loads from the canal wall (acting as a beam) to the other existing loads. The latter leads were obtained from Reference 6. The combined loads are shown in Table 4-5. The computed additional loads, shown in this table, include the seismic effect on the additional rack weight on the pool floor. 4-3

~~

0542 - 1877 194

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O O FIGURE 4-1 FINITE ELEMENT MODEL USED FOR POOL FLOOR EVALUATION 4-4 1877 195

DUCT. ERR SERVICES CORPORATION O(_& DiWiStans OFlADREX CO APO A ATION TABLE 4-1 MAXIMUM POOL FLOOR LOADS FOR CRITICAL LOAD COMBINATI0ES Consideration Canal Critical Load Moment Axial F'rce Shear l Condition Combination (K-f t/ f t) (k/ft1 (k/ft) 1.4 D + 1.4 H -350 47.3 ---- 1.4 0 + 1.4 H + 1.9 E -498 59.5 ---- Full D+H+T a 773 33.8 ---- D+H+T +a E' 855 46.6 ---- Bending 1.4 0 + 1.4 H -448 65.8 ---- 1.4 D + 1.4 H + 1.9 E -600 78.9 ---- Empty D+H+T a 865 47 ---- L+H+T +a E' 1022 60.8 ---- 1.4 0 + 1.4 H -112 ---- 75.7 Full 1.4 0 + 1.4 H + 1.9 E -224 ---- 86.0 Shear ! 1.4 D + 1.4 H -210 ---- 68.6 Empty 78.3 1.4 D + 1.4 H + 1.9 E -324 ---- 4-5 1877j96

TABLE 4-2 MAXIMUM BENDING MOMENTS AND SHEARS IN THE CAftAL WA. '_ Location , Critical Load Consideration Load At Critical At Support Combination Section Bending Moment 17,200(1) -34,117 (k-ft) 1.40 + 1.4H + 1.9E Shear Moment -29,012(2) -34,117 (k-ft) Shear 2,170(2) 3,226 (kips) i NOTES: (1) At the center of span (2) At the gate 4-6 1877 197

TABLE 4-3 SHEAR STRESSES IN THE WALLS SUPPORTING THE POOL FLOOR Shear Shear Additional Shear Total Shear Wall (1) Existing Stress (Psi) Stress (Psi) Stress (Psi) OBE SSE OBE SSE OBE SSE 1 29.2 58.4 2.0 3.9 31.2 62.3 2 48.0 95.9 3.6 7.2 51.6 103.1 3 34.4 68.7 3.8 7.6 38.2 76.3 4 34.4 68.7 3.8 7.6 38.2 76.3 5 3.2 6.3 3.5 6.9 6.7 13.2 North 18.9 39.7 2.5 5.0 21.4 44.7 Wall South 24.6 49.1 3.1 6.1 27.7 55.2 Wall (1)See Figure 2-2 4-7 )Q}} k00

TABLE 4-4 COMPRESSIVE STRESSES Ill CRITICAL SHEAR WALLS SUPPORTIflG THE P00L FLOOR (1) Existing Comp. Additional Comp. Total Comp. Shear Wall (2) Stress (Psi) Stress (Psi) Stress (Psi) 1 670 53 723 2 670 53 723 5 184 54 238 South 618 19 637 Wall 110TE: (1) For D + H + E' Loadings (2) See Figure 2-2 1877 199

TABLE 4-5 LOADS Ott COLUM.'lS M - 10 AND M b b b

                                                       -7 b Column           Loads with (2)              Loads with flo .        Existing Racks (kips)       Proposed Racks (kips)

M b

         -7 b                1773                       2195 2972                        3549 Mb - 10b NOTE:       (1) Includes SSE Loads (2) From Ref. 6 4-9

@ 1877 200

RUCl. EAR SERVICES CORPORATION Q(_A Divistosu OFlADAEX CO A AQ A ATION 5.0 EVALUATION RESULTS & CONCLUSIONS 5.1 Evaluation of Pool Floor The pool floor slab was evaluated from two considerations: (1) combined bending and axial loads, and (2) shear. Like most reinforced concrete floor slabs in nuclear power plants, the pool slab is under-reinforced, i.e., its bending strength is limited by the amount of tensile reinforcement. The tensile steel cross-sectional area requirements for combined bending and tension loads listed in Table 4-1 were computed using the strength method. These requirements are compared to the available steel area in Table 5-1. Shear capacities of the pool slab were also calculated for load combin-ations and loads listed in Table 4-1, and compared to the computed factored shear loads. This comparison is shown in Table 5-2. 5.2 Evaluation of Spent Fuel Pool Canal Wall The canal wall separating the pools and the transfer canal was analyzed as a beam spanning between Columns Mb - 10b and M b -7.b For evaluating the bending capacity, the minimum depth of this beam (at the two gates) was used. Shear capacities were computed at the face of the supporting columns as well as at the location where the depth is minimum due to the cut-out for the gates. In both cases, the ultimate strength method was used to compute the capacities. Table 5-3 compares the available moment and shear capacities with the computed moments and shears. 5.3 Evaluation of Shear Walls and Columns Supporting the Pool Floor 5.3.1 Shear Walls Compressive, shear and bending stresses due to additional rack 0542 , 1877 201

DUCT. ERR SERVICES CORPORATION O(_A Dive $104 CFlADAEX cCapO AADON weight including their seismic effect on the shear walls supporting tne pool floor were listed in Tables 4-3 and 4-4. The combined shear stresses for the critical load combinations are shown and compared to the allowable shear stresses in Table 5-4. The al-lowable shear stresses were computed using the ultimate strength method. Combined compressive stresses due to bending and axial compression for the critical load combinations are shown in Table 5-5. Allowable compressive stresses, computed using the working stress method of design, are also shown in this table. 5.3.2 Evaluations of Columns M - 10 and M -7 3 Load carrying capacities of Columns M b

                                                - 10band M b ~7 b were obtained from Reference 6. These capacities were computed for service load conditions using the working stress method. These are listed in Table 5-6 and compared to the computed loads shown in Table 4-5. Even though the computed loads include the effect of SSE on the additional rack weight, for simplicity and conservatism, these were compared with the capacities computed for service load conditions using working stress method.

5.4 Evaluation Results and Conclusions Results of evaluation show tnat fuel pool structures are structurally adequate to withstand the additional loads that would result from the proposed modification in which the present racks will be replaced by the high density racks described in Reference 1. A comparison between the safe load carrying capacity of the structural components and the computed loads for various critical load combinations as shown in Tables 5-1 through 5-6 shows the calculated margin of safety. The actual margin of safety, in general, is more than depicted in these tables. The reasons are listed below: 5-2 0542 1877 202

DUCLERR SERVICES CORPORATION a OlviliO4 08 WLIADREX CO A&JO A ATION a) Due to age-hardening effect, the actual strength of concrete is more than the design rated (28-day) strength. This difference can be as high as 20 percent. b) Even though the floor load from the proposed high density racks was computed to be 2.0 ksf, the evaluation presented was performed using a rack load of 2.4 ksf. c) The canal wall separating the pools from the transfer canal was evaluated assuming that it supports the pool floor in a deep beam action. Thus, the design loads for this beam-wall, as well as for Columns bM - 10b and Mb~7b which support it, are conservative. d) For evaluating the pool floor, the canal wall was assumed to be resting on it. Hence, the design loads for which the pool floor was evaluated are conservative. e) For Columns M b

                     - 10 b nd Mb - 7b , for simplicity, capacities computed for service loads by the working stress design method were compared to loads which included the effects of SSE. Since SSE loads are included only in factored load combinations for which higher allowables are permitted, the above comparison was conservative.

5-3 1877 203 - 0542

TABLE 5-1 EVALUATION OF BETIDING CAPACITY OF POOL FLOOR Canal Critical Load Required} teel Condition Combination Available} Area (in teel

                                           )         Area (in )

1.4D + 1.4H 7.62 2.46 Full 1.40 + 1.4H + 1.9E 7.62 3.38 0+H+T a

                      + E'          7.62               5.1 Empty     D+H+T     a
                      + E'          7.62               6.2 9

s.4 g 1877~204

TABLE 5-2 EVALUATION OF SHEAR CAPACITY OF POOL FLOOR Canal Critical Load Safe Shear Computed Shear Condition Combination Capacity (kip /ft) Load (kip /ft.) Full 1.4D + 1.4H 300.5 75.7 1.4D + 1.4H + 1.9E 300.5 86.0 Empty 1.40 + 1.4H 273.1 68.6 1.4D + 1.4H + 1.9E 195.3 78.3 5-5 1877 205

TABLE 5-3 EVALUATION OF SPEf4T FUEL POOL CAf1AL WALL Critical Load Load Safe Capacity Computed Load Combination Type At Critical At Support At Critical At Support Section Section 1.40 + 1.4H + 1.9E Moment 31700(I) -52,700 17200(1) -34117 (k-ft) Shear 2653(2) 6,471 2170(2) 3226 (kip) fl0TE: (1) At tnc center of span (2) At the gate O 5-6 1877 206

TABLE 5-4 EVALUATION OF SHEAR CAPACITIES FOR WALLS SUPPORTING THE POOL FLOOR Shear Computed Shear Allowable Shear Wall (y) Stress (psi) Stress (psi)(2) 1.4D + 1.4H + 1.9E D+H+E' 1 59.3 62.3 126.5 2 94.2 103.1 126.5 3 72.6 76.3 126.5 4 72.6 76.3 126.5 5 12.7 13.2 126.5 North Wall 40.7 44.7 126.5 South Wall 52.6 55.2 126.5 NOTES: (1) See Figure 2-2 (2) Contribution from steel reinforcement not included 5-7 1877 207.

TABLE 5-5 EVALUATI0tl 0F COMPRESSIVE STRESSES FOR WALLS SUPPORTIrlG THE POOL FLOOR Critical Load Shear Computed Comp. Allowable Comp $) Combination Wall ll) Stress (Psi) Stress (Psi) ( 1 723 1800 0 + H + E' 2 723 1800 5 238 1800 South Wall 637 1800 fl0TES: (1) See Figure 2-2 (2) Allowable Stress computed by Working Stress Method 5-8 1877 208

TABLE 5-6 EVALUATION OF COLUMNS b M -b10 ANDb M ~ b Column Computed Allcwable Loads (2) No. Loads (Kip)(1) (Kips) M - 76 2195 2424 b M - 10 3549 3636 b b NOTES: (1) Includes SSE Loads (2) Allowable Loads are from Ref. 6 and are computed by Working Stress Method 5-9 ' @ 1877 209

DUCLEAR SERVICES CORPORATIOD A Olvissom Of

                                 % ( (_l R D R E X CO APC A ATION

6.0 REFERENCES

1. " Licensing Report for Prairie Island Nuclear Generating Plant, Units 1 and 2, Spent Fuel Storage Modification," Report No. QUAD-1 509.

2. " Prairie Island Nuclear Generating Plant Earthquake Analysis,"

Revised February 16, 1971, John A. Blume & Associates.

3. " Seismic Design Classification," USNRC Regulatory Guide 1.29, Revision 2, February 1976.

4. "Other Category I Structures," USNRC Standard Review Plan, Section 3.8.4.

5. " Building Code Requirements for Reinforced Concrete (ACI-318-77),

American Concrete Institute,1977.

6. " Safety Evaluation Report for Auxiliary Building Spent Fuel Storage Pools and Their Supporting Structures Affected by Alterations in Fuel Storage Capability, Revision 2," by Flour Pioneer, Inc. ,

Prairie Island Huclear Generating Plant, Units 1 & 2, Project No. 21-7450 C.0001 (E75GA01), Activity No. 000-232, June 21, 1977.

7. "STARDYNE - A Static and Dynamic Structural Analysis Program,"

Mechanics Rasearch Institute, Los Angeles, California.

8. " Introduction to Structural Dynamics," by J. M. Biggs, McGraw &

Hill Book Company,1964.

9. "kJclear Reactors and Earthquakes," TID-70?4, U.S. Atomic Energy Commission, August 1963. *

10. American Concrete Institute Standard 307-69, 1969.

1877 210 6-1 0542

DUCLEAR SERVICES CORPORATION

                                     & 04Wila04 05 QLIADAEX CO ApQ A ATION

11. Pioneer Service & Engineering Company, Project No. .216197, Drawing Nos: NF-39213E, 38303-8L, 38303-15K, 38303-lQ, and 38303-4K.

O 1877 211 6-2 0542

EXHIBIT E PRAIRIE ISLAND NUCLEAR GENERATING PLANT License Amendment Request dated January 31, 1980 Docket Nos. 50-282 License Nos. DPR-42 50-306 DPR-60 Exh ibit E consists of the following: Letter, L 0 Mayer (NSP) to Don K Davis (NRC), dated April 14, 1977. 1877 212 0

MSE2 NORTHERN STATES POWER COMPANY MI N N E A PO Ll e, M IN N E SOTA 55401 April 14, 1977 Mr Don K Davis, Acting Chief Telecopied to the NRC Operating Reactors Branch #2 4/14/77 Division of Operating Reactors U S Nuclear Regulatory Commission Washington, DC 20555

Dear Mr Davis:

PRAIRIE ISLAND NUCLEAR GENERATING PIRIT Docket No. 50-282 License No. DPR-42 50-306 DPR-60 Modification of Spent Fuel Storage Pool On April 11, 1977 we were requested by your staff to provide a structural evaluation of the protective cover which is to be used during the in-stallation of the modified spent fuel storage racks in the Prairie Island Spent Fuel Pool. The requested information is contained in 'r he Attachment to this letter. Full-sized drawings of the cover were provided to Mr M Gror.enhuis earlier and, due to the size of the drawings, are not included with this submittal. If additional copies of these drawings are necessary, please contact us. Yours very truly, /s/ L 0 Fhyer L 0 Mayer, PE Fbnager of Nuclear Support Services LOM/LLT/ak cc J G Keppler G Charnoff MPCA Attn: J W Ferman 1877 213

STRUCTURAL EVALUATION FOR PROTECTIVE COVER OF AUXILIARY BUILDING SPENT FUEL STORAGE POOL #1 I System

Description:

In order to protect the small spent fuel pool (spent fuci pool #1) against the accidental drop of a spent fuel rack onto the stored spent fuci due to the proposed new spent fuel storage arrangement which provides the additional storage capacity, a protective cover is provided. The cover is designed in three separate pieces for easy handling. Each piece weighs approximately 1.85 tons. The capacity of the gantry crane presently being used is 3 tons. The cover is made of 3/16" stainless steel AST11 A167 Type 304 plate welded to a frame of built up wide-flange beans and structural tecs using structural steel ASTM A588 Grade A (COR Ten B). The compressibic material pads (silicone rubber SE-551 - AFG Specification 3335A) are used underneath the end supports to absorb a part of the energy generated from an acci-dentally dropped rack. The cover is designed to cicar the crane, which has a cicarance of Sh" (as measured in field) above the top of the pool walls. The cicarance between the cover and the crane is designed to be % inch. The cover clears the temperature and water icvel indicator. However, the interference with light fixtures, fuel handling test racks and electrical receptacles requires relocation of these items. II Loadings: The weight and size of the spent fuel rack used in the design is as specified in our License Amendment Request dated November 24, 1976. No other external loads are assumed to act during the time the racks are being installed. When transferring the rack over the small pool the height of the bottom of the rack in the non-tilted position is limited to a maximum of 6" above the top of the protective cover. The spent fuel racks shall be put down on the protective cover at a vertical speed not exceeding 2 feet per minute. The racks can be unloaded any place on the cover, maintaining a minimum clearance of 5" between the nearest edge of the rack's leg and outside edge of the cover, i877 214.

The following loading cases are investigated: CASE 1 - Sudden application of a load due to manipulation of the rack in the non-tilted position (a normal operation case) . CASE 2 - Sudden application of load due to crane failure, dropping the rack from a height of 6" in the non-tilted position. CASE 3 - Sudden application of load due to crane failure, dropping the rack in a titled position from a height of 3". III Allowable stresses and design procedure: Concrete: ACI 318-71 Bailding Code requirements for reinforced concrete. Steel: For Case 1 - Elastic design. Allowable stresses as per AISC, 1969, 7th Edition. For Cases 2 & 3 Inelastic design. The procedure of Williamson and Alvy is used to determine the equivalent static loads caused by the inpact of dropped racks. The allowable stresses are as per Part 2 of AISC, 1969. Only the wide-flange sections are designed to carry the load due to an accidental drop of a rack; not the stainless steel plate. In the event the plate is punctured, following an accidental drop, the ductile nature of stainless will prevent the generation of secondary missiles. IV Results of the design: Protective Cover The designed protective cover is shown on Prairie Island structural drawing NF-38303-29 (three copics of this drawing were provided separately). Two wide-flange sections W14x26 are used for each piece of cover supporting the inverted IE4x3.25 members (or alternate plates 5/16' :5") space at 6". The 3/16" stainless steci plate covers are approximately 6'-6" in width. The stresses and deficctions under the investigated loadings are as follows: For Loading Case #1: lbximum bending stress in the wide-flange section is 23.3 KSI against an allowable of 33 KSI. Maxicati bending stress in the FE sections is 27.3 KSI (30 KSI for plates) against an allowable of 30 KSI. Fbximum shear stress in the wide-flange section is 3.5 KSI against an allowable of 20.0 KSI. Maximum bending stress in the 3/16" cover plate is 14.3 FSI against an allowable of 24 KSI. The maximum deflect. ion is 3/4". 5

For Loading Case #2: The ductility ratio 3.3 is obtained against an allowable of 10. For Loading Case #3: When the load is applied at the middle of the span:

a. Maximum deficction is 11.2".

b. Ductility ratio is 8.28 against an allowabic of 10.

When the load is applied near the support, the whole energy is absorbed by the 1" thick compressible material pads.

a. The equivalent static load is 243 KIPS and deflection of compressibic material pad is 13/16".

b. Bending stress is 48.72 KSI <,.F = 50 KSI y

c. Shear Stress is 9.5 KSI 4 F V all " *

d. Bearing on concrete F = 3.4 KSI .t F = 3.57 KSI p 77

c. The capacity of the weld is 400K against the shear force of 318.8 KIPS.

Lifting Rin ' Each lifting ring is designed to carry twice the entire weight of the protective cover (2 x 3.8 = 7.6 KIPS) allowed for 1007. impact. The shear stress in the ring so obtained is 6.8 KSI against 20 KSI allowabic. Existine Pool Structure The existing spent fuel poel concrete structure was reviewed for the new load condition by reviewing Structural Drawings NF-38303-1 through 41, Auxiliary Building - Concrete Fuel Pool (latest revisions) and project files of the original calculations files. The accidental drop of a rack is assumed not to coincide with the other loading conditions which are controlling the supporting structure design, 4,e , carthquake, accident t emp e ra tu re , tornado, etc. Hence, the supe cting s'ructures are adequate to carry the additional loads imposed b- che system u.'er consideration. 1877 216}}

ML19257D713

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1.0 INTRODUCTION

3.3 System Description

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1.0 INTRODUCTION

SUMMARY

4.0 CONCLUSION

5.0 REFERENCES

1.0 INTRODUCTION

SUMMARY

4.0 CONCLUSION

5.0 REFERENCES

1.0 INTRODUCTION

2.0 DESCRIPTION

6.0 REFERENCES

1.0 INTRODUCTION

2.0 DESCRIPTION

6.0 REFERENCES

Dear Mr Davis:

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