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Staff Exhibit S-1,consisting of Undated Safety Evaluation Accepting Licensee Applications for Reracking of Spent Fuel Pools as Related to Amend 8 to License DPR-80 & Amend 6 to License DPR-82.Technical Rept Re Spent Fuel Racks Encl
	ML20237K380
Person / Time
Site:	Diablo Canyon
Issue date:	06/17/1987
From:	; Office of Nuclear Reactor Regulation
To:	;
References
	OLA-S-001, OLA-S-1, NUDOCS 8709040320
	Download: ML20237K380 (138)
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                                                                                .p nacn Safety Evaluation By The Office of Nuclear Reactor Regulation Relating to the Reracking of the Spent Fuel Pools At the Diablo Canyon Nuclear Power Plant, Units 1 and 2 As Related to Amendment No. 8 to Unit 1 Facility Operating Licence No. DPR-80 And Amendment No. 6 to Unit 2 Facility Operating License No. DPR-82, Pacific Gas and Electric Company Docket Nos. 50-275 and 50-323
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t b TABLE OF CONTENTS PAGE 1

1. INTRODUCTION 1

1.1 Licensee Submittal and Staff Review 1.2 License Amendment Request -2 1.3 Sumary Description of Reracking 2 3

2. CRITICALITY CONSIDERATIONS 6 j

3. MATERIAL COMPATIBILITY AND CHEMICAL STABILITY 9

4. STRUCTURAL DESIGN 4.1 Description of Spent Fuel Pool and Racks 9 4.2 Applicable Codes, Standards and Specifications 10 4 4.3 Loads and Load Combinations 10 4.4 Design and Analysis of Racks 11 4.5 Design and Analysis of Pool Structure 13 4.6 Fuel Handling Accident Analysis 14 4.7 Conclusions 14

5. INSTALLATION OF RACKS AND LOAD HANDLING 14

6. SPENT FUEL POOL COOLING SYSTEM 15

7. SPENT FUEL POOL CLEANUP SYSTEM 16

8. RADIATIONPROTECTIONANDALARACONSIDERATIONS 18

9. FUEL HANDLING ACCIDENT AND CASK DROP ACCIDENT 20

10. RADIOACTIVE WASTE TRCATMENT -

22

11. SIGNIFICANT HAZARDS CONSIDERATION COMMENTS 22 l

12. FINAL NO SIGNIFICANT HAZARDS CONSIDERATION 25 l

13. ENVIRONMENTAL CONSIDERATIONS 31

14. CONCLUSIONS 31

15. REFERENCES 32 APPENDIX A: Technical Evaluation Report by Franklin Research Center t

1

1. INTRODUCTION 1.1 Licensee Submittal and Staff Review This report presents the NRC staff safety evaluation for the reracking of the spent fuel pools of the Diablo Canyon Nuclear Power Plant, Units 1 and 2. By letter dated October 30, 1985 (DCL-85-333, Ref. 1) the Pacific Gas and Electric Company (PG&E or the licensee) submitted, as License Amendment Request LAR-85-13, an application to increase the storage capacity of the spent fuel pools (SFP) for Unit 1 and Unit 2 of the Diablo Canyon Nuclear Power Plant (Diablo Canyon or the Plant), including the appropriate and necessary changes to the combined Technical Specifications for Units 1 and 2. The licensee requested the increase in storage capacity in order to be able to store onsite the spent fuel resulting from approximately 20 years of normal operation of both units.

The application is based on the licensee's " Report on Reracking of Spent Fuel Pools For Diablo Canyon Units 1 and 2" (the Reracking Report) which was submitted to the staff by letter dated September 19, 1985 (DCL-85-306, Ref. 2). During its review of the Reracking Report, including discussions at meetings with the licensee and audits of records, the staff requested additional information by letters dated January 8 (Ref. 3), January 15 (Ref. 4), February 18 (Ref. 5) and February 28, 1985 (Ref. 6). The licensee provided additional information supplementing its application by letters of December 20, 1985 (DCL-85-369, Ref. 7), December 24, 1985 (DCL-85-371, Ref. 8), January 28, 1986 (DCL-86-019, Ref. 9), January 28, 1986 (DCL-86-020, Ref. 10), March 11,1986 (DCL-86-067, Ref.11), April 24,1986 (DCL-86-108, Ref. 37) and April 24, 1986 (DCL-86 '109, Ref. 38). During its review of the application the staff met with the licensee and l representatives of the Joseph Oat Corporation, Camden, New Jersey, the designer I arid manufacturer of the fuel racks, to discuss in further detail the fuel racks and the fuel pool structure. Meetings were held by the staff with PG&E on December 5, 1985 (Ref. 12), January 8, 1986 (Ref. 13) and February 20, 1986 (Ref. 14).' In addition, the staff visited the Joseph Oat Corporation on January 30, 1986 to tour the facilities and observe the manufacturing process of the fuel racks (note: as a result of this visit a request for additional information was sent to the licensee as Reference 13), and on March 24-25, 1986 (Ref.15) to audit the structural analysis procedures and calculation packages for the racks and the pool structure. On April 14, 1986 the staff toured the i Unit I fuel handling building at the Diablo Canyon Plant to observe the structural layout and features of the fuel pool (Ref. 16). This report was prepared by the staff of the Office of Nuclear Reactor Regulation (NRR). Technical assistance for the structural evaluation of the spent fuel racks and spent fuel pools was provided by the Franklin Research I

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g .;- .g - q 1 ,. - (p. Center, Philadelphia, Pennsylvania. The principal contributors to this report are: W. Brooks Reactor Systems Branch, PWR-A R. Fell Plant Systems Branch, PWR-A. H. Gilpin _ Plant Systems Branch, PWR-A C. Herrick Franklin Research Center (Consultant) D. Jeng Engineering Branch, PWR-A F. Rinaldi Engineering Branch, PWR-A H. Schierling Project Directorate #3, PWR-A R. Serbu Plant Systems Branch, PWR-A A. Singh Plant Systems Branch, PWR-A J. Wing Plant Systems Branch, PWR-A The NRC Project Manager for the Diablo Canyon Nuclear Power Plant is Mr. H. Schierling (Address: -U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation, Washington, D.C. 20555 - Telephone: 301-492-8856). 1.2 License Amendment Request The license amendment request (Ref.1) is for changes to the spent fuel pool of each unit and for the accompanying changes to the combined Technical Specifications for both units. Accordingly, this safety evaluation applies to two amendments, one for Unit 1 Facility Operating License DPR-80, the other for ! Unit 2 Facility Operating License DPR-82. The request for the amendments, including the staff's proposed "No' Significant Hazards Consideration", was noticed in the Federal Register on January 13, 1986 (Ref. 17). Further details are addressed in Section 11 of this report. 1.3 Sumary Description of Reracking The current spent fuel system is described and analyzed in the licensee's Final Safety Analysis Report (FSAR) Update (Ref. 48), Section 9 and was evaluated by the staff in its Safety Evaluation Report (SER) and supplements (Ref. 30). The Diablo Canyon Units 1 and 2 have separate, identical spent fuel ' handling -i and storage facilities. Both are located in a common fuel' handling building, separated by the hot shop and parts of the auxiliary building. In the current design the only component' shared by both units-is a portable backup pump in the' 2

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spent fuel pool cooling system as discussed in Section 6. The evaluation in this report applies equally to both units. The proposed amendments will allow the licensee to expand the storage < capacity of each spent fuel pool (SFP) from the current capacity for 270 fuel assemblies to 1324 assemblies. The expansion will be accomplished by removing the current racks in the SFPs and replacing them with high density racks.in which the I individual stainless steel cells are more closely spaced. Each cell provides storage for one fuel assembly. The spent fuel storage racks will be arranged in two discrete regions within each pool. In Region I storage racks with poisoned cells will be located, that is, cells surrounded with "Boraflex", a neutron absorbing material. Region 1 provides for storage for 290 fuel assemblies, which is adequate for approximately one and one-half core (a full core is 193 assemblies). It will normally be used for storage of fuel assemblies with an enrichment of equal to or less than 4.5 weight percent of Uranium-235 (U-235) at their most reactive point in life, i.e., as new fuel. . In Region 2 storage racks with unpoisoned cells will be located, that is cells without Boraflex, providing 1034 storage locations for spent fuel assemblies meeting specified burnup conditions. The existing fuel storage racks have a nominal center-to-center cell spacing of 21 inches. The new storage racks for Region 1 and Region 2 will have a nominal center-to-center cell spacing of 11 inches. The major components of the fuel racks are the cells, either poisoned or unpoisoned, gap channels welded between the cells, a rack base plate assembly, including support legs, and a girdle bar around the upper part of the rack assembly. In the case of the poisoned racks, each cell is surrounded by the Boraflex neutron absorber, held in place by.a stainless steel cover sheet which also permits venting of the Boraflex to the pool environment. The fuel racks are assembled as an essentially fully welded construction. There will be a total of 16 fuel racks of various sizes in each pool. The racks are free standing, not connected to the pool floor, walls or to each other. The proposed reracking will not involve major changes to the I spent fuel pool structures.

2. CRITICALITY CONSIDERATIONS The current capacity of each spent fuel storage pool is 270 fuel assemblies; the reracking will increase the capacity to 1324 assemblies in each pool. This is accomplished by replacing the present racks with high density racks, i.e.,

reracking. A two-region design has been applied (Technical Specifications, Figure 3.9-1, Ref. 1). Region 1 is designed to accept fresh fuel and will store the fresh fuel assemblies on a 10.93 inch center-to-center cell spacing. Each storage location will be surroundcd on all four sides by a neutron absorber material, Boraflex, with sufficient boron to result in an effective multiplication factor, k-effective, of a value of 0.95 or less. This region will have the capacity to store 290 fuel assemblies, approximately one and one-half core. Region 2 has racks of the same design except that no neutron absorber material is used. Storage in Region 2 is restricted to spent fuel 3

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assemblies which have achieved a minimum burnup, dependent on the initial . enrichment, as discussed below. Space for 1034 spent fuel assemblies is { available in Region 2. j The criticality analysis was perfomed by Southern Sciences, a company which has perviously performed such a analyses for other fuel pool expansion applications, including Rancho Seco (Docket No. 80-312), Quad Cities Units 1 and 2 (Docket Nos. 50-254 and 50-265), and Virgil Sumer (Docket No. 50-395). ! Three different calculational methods were used in the criticality analyses of the spent fuel racks as follows: , (1) The AMPX-NITAWL-KEN 0 code package, which is a multigroup Monte-Carlo calculation using both the 123 and 27 group cross section sets. This is the primary analysis method. (2) The CASMG-2E code, which is a two dimensional, multigroup transport theory method. It was used both as a primary calculational tool and as a means of evaluating small reactivity effects associated with j The statistical nature of the KEN 0 ' manufacturing calculations (tolerances. item 1 above) makes their use for such effects impractical. The CASMO-2E code was also used for performing the burnup calculations for the fuel. (3) Two multi-group diffus. ion codes, PDQ07 and SNEID, were used for the calculation of small reactivity effects that could not be obtained by CASMO-2E because of geometry limitations in'that code. Each code listed above h.as been qualified for use by comparison of its calculated k-effective value to that- of critical experiments. On the basis of such comparisons calculational bias and uncertainty values have been obtained for application to the calculated results for the spent fuel racks. The KENO code was developed by Oak Ridge National Laboratory (Ref. 45). The code has been widely used by Southern Sciences for the calculation of the reactivity of spent fuel pool racks and has been found acceptable by the staff. Its use is applicable and acceptable for the Diablo Canyon fuel racks. Use of non-statistical codes such as CASMO-2E, PDQ07 and SNEID for calculation of small reactivity effects is standard industry practice and is acceptable. The CASM0-2E and similar codes are routinely used for obtaining lattice physics l parameters as a function of burnup. The licensee's Reracking Report (Ref. 2) l gives a comparison of CASMO-2E with several similar codes for the reactivity of fuel assemblies as a function of burnup. The codes agree to within 0.8% for k-effective, with the CASMO code usually being conservative, that is, predicting a higher reactivity. The staff finds that the CASMO-2E code is. acceptable for calculating burnup effects. The following assumptions were made in the calculations to assure conservatism: (1) The moderator is water at a temperature which yields the highest reactivity. 4 l

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(2) The storage racks are assumed to, be. infinite in extent in all f directions. ,

(3) No credit is taken for abh;cepcion in minor. structural members, for example in spacer grids. , 4 s , :< g] w c p The staff concludes that the methods used for the analyses are acceptab7e. 5 The effect of. uncertainties in the Boraflex width, thickness and Baron-10 concentration in conjunction with fuel assembly dimensions, fuel enrichment and density, and off-center position of the fuel in the storag:. cell were i' determined. These represent the usual uncertainties considered and are appropriate and acceptable. 'For the Region 2 racks an additional allowance for uncertainty in burnup was included, a The analyses result in effective multiplication factors of 0.912 0.008 and 0.932 0.006 for the racks in Region 1 and Region 2, respec,tively. These

values meet the staff criterion of 0.95 or less for k-effectave and, therefore, are acceptable. i?

Thek-effectivevalueforRegion2wasobtainedwiththeCASMO-2Ecodeforkfuel g having an initial enrichment of 4.5 weight percent U-235 and a burnup of 34,500 MWD /MTU. -In order to determine the required burnup as a function of i initial enrichment a series of iterative calculations were' performed with CASMO. An initial enrichment was chosen and the burnup required to obtain.the target k-effective value was obtained.. The resulting curve was extrapolated to obtain the enrichment for which no burnup was required. The k-effective:for this enrichment was then calculated by KEND. Comparison of the two-calculations showed that the CASMO calculation is conservative. This procedure is commonly used for such an analysis and is. acceptable. , The effect of abnomal and accident conditions on k-effective included considerations of heatup of the pool weter, dropping of an assembly onto the racks, misplacement of an assembly, and lateral' motion of the racks. For Region I the effect of a moderator temperature increase is to decrease the reactivity. For Region 2 an increase in reactivity results and k-effective reaches a value of 0.95 at the boiling temperature after.which the' reactivity decreases. In the case of postulated boiling credit may be taken for the presence of soluble boron in the pool which would reduce the reactivity by as much as 0.2. The potential for pool boiling is discussed ip Section 6. Dropping a fuel assembly onto the top of the racks has a negligible effect on reactivity since the dropped assembly would be separated from the fuel racks by-approximately 12 inches of water. Lateral movement of the racks is prevented by design features from reducing the spacing by an amount sufficient to increase the k-effective value above 0.95. Credit for the presence of boron in the pool water was taken in the analysis of ~ the abnormal location of a fuel assembly in order to ensure the k-effective value does not exceed 0.95. Both the placement of an assembly outside the racks N 5

3.,- . g e and thet hisibding of a fresh fuel assembly into the Region 2 racks were analyzed. In both cases the k-effective value was below the staff acceptance criterion of 0.95. The staff has determined that the analysis of abnormal events is acceptable. Conclusions The staff has reviewed the licensee's proposed spent fuel pool reracking with respect to criticality and finds that: (1) Acceptable analysis methods were used which have been qualified for use by comparison with experiment. (2) Conservative input assumptions were made. (3) Mechanical and calculational uncertainties were included. (4) The results meet the staff acceptance criterion of 0.95 for

                                    ,   k-effective of the racks.

In stunary, the staff concludes that the criticality analysis for the proposed rerecking of the fuel pools of the Diablo Canyon Nuclear Power Plant, Units 1 m ) 2 is acceptable. The staff further concludes that up to and including 290 Westinghouse design fuel. assemblies of enrichment 4.5 weight percent U-235 can be stored in Region 1 of the pool and that up to and including 1034 of exposed assemblies meeting the burnup criteria in the proposed Technical Specifications (Figure 3.9-2) can be safely stored in Region 2.

                            '3 . MATERIAL COMPATIBILITY AND CHEMICAL STABILITY The Diablo Canyon Nuclear Power Plant, Units I and 2 each have a facility for the wet storage of spent fuel assemblies. The staff has reviewed the compatibility and chemical stability of the materials wetted by the pool water.

During its review the staff requested additional information regarding the s material compatibility of the rack structure with the spent fuel pool water environment (Ref. 3). The licensee provided the information in the submittal of January 28, 1986 (Ref. 9). The proposed spent fuel racks are constructed of Type 304 stainless steel, except for the nuclear poison material Boraflex. The spent fuel pool liner also is constructed of Type 304 stainless steel. Some of the spent fuel storage racks utilize Boraflex sheets as a neutron absorber (Ref.18) as discussed in Section 2. Boraflex consists of 42 weight percent of boron carbide powder in a rubber-like silicone polymeric matrix. The spent fuel { l storage racks are constructed of individual fuel storage cells interconnected by channels to form an integral welded structure. The major components of the fuel racks are listed in Section 1.3. The spent fuel pools are filled with oxygen-saturated demineralized water containing boric acid. The fuel pool liner and the fuel rack assemblies, except the Boraflex used in Region 1, are stainless steel which is compatible with the pool water and 6 .

radiation environment. In this environment of oxygen-saturated borated water, the corrosive deteriorate exceedadepthof6x10~pnofType304stainlesssteelisnotexpectedto inch per year (Ref. 35), which is negligible relative to the initial minimum thickness of 0.08 inch of the walls of the poisoned cells. Dissimilar metal contact corrosion (i.e., galvanic attack) between the stainless steel of the pool liner or rack assemblies, and the Inconel and the Zircaloy in the fuel assemblies stored in the racks, will not be significant because all these materials are protected by highly passivating oxide films which are at similar galvanic potentials. Boraflex has been used previously as a neutron poison material in a number of spent fuel pool expansions, including Prairie Island Units 1 and 2 (Docket Nos. 50-282 and 50-306) and Oconee Units 1 and 2 (Docket Nos. 50-269 and 50-270). It has been found acceptable by the staff as a neutron poison material. For each proposed storage cell assembly, the space between the outside of the fuel cell wall and the cover sheet which contains the Boraflex is vented to the pool. This will allow gas generated by chemical degradation of the silicone polymer binder of the Boraflex during heating and irradiation to escape to the pool and will prevent bulging or swelling of the cover sheet. Boraflex is composed of non-metallic materials and therefore will not . develop a galvanic potential in contact with the metal components. Boraflex has under-gone extensive testing to determine the effects of gamma irradiation in various environments and to verify its structural integrity and suitability as a. neutron absorbing material (Ref. 19). The evaluation tests have shown that Boraflex is unaffected by the pool water environment and will not be degraded by corrosion. Tests g re performed at the University of Michigan, exposing Boraflex to 1.03 x 10 . rads of gamma radiation with substantial concurrent neutron flux in borated' water. These tests indicate that Boraflex maintains its neutron attenuation capabilities after being subjected to an environment of borated water and gama irradiation. Irradiation will cause some loss of flexibility, but will not lead to breakup of the Boraflex. Long-term borated water soak tests at high temperatures were also conducted (Ref. 20). The tests show that Boraflex withstands a barated water imersion at 240*F for 251 days without visible distortion or softening. The Boraflex showed no evidence of swelling or loss of ability to maintain a uniform distribution of boren carbide. The spent fuel pool water temperature under normal operating conditions will be approximately 105'F which is well below the 240*F test temperature. In general, the rate of a chemical reaction decrease expontially with decreasing temperature. Therefore, the staff does not anticipate any significant deterioration of the Boraflex at the pool normal operating conditions over the design life of the spent fuel racks. The tests have shown that neither irradiation, environment, nor Boraflex j composition have a discernible effect on the neutron transmission of the Boraflex material. The tests also have shown that Boraflex does not possess leachable halogens that might be released into the pool environment in the presence of radiation. Similar conclusions are reached regarding the leaching of elemental boron from the Boraflex. Baron carbide of the arade normally 7

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q u . present in the Boraflex will typically contain 0.1 weight percent of soluble

           .       boron. The test results have confirmed'the encapsulation capability of the silicone polymer matri A to prevent the leaching of soluble species from the j                 boron carbide.

To provide added assurance that no unexpected corrosion or degruiation of the neutron abscrber materia 7 will compromise t.he long-term integrity of tFa. i Boraflex,' the licensee w?lkconduct a loog* term poison coupon surveillance l progra'n as described in Section 8 of the'Reracking Report. ' Surveillance 1,ampbes in the form of removable Bortflex sheets will be exposed to a realistic - environment in the pool. They will 'co periodically removed, and examined for

                 ' deterioration.         t' The staff recently identified concerns related to the terminology used for ..

various welding ' processes, the applicability of the ASME Code Section JX to the welding, and the long-term compatibility of certain structuril w. ids with a spent fuel pool envi m ment (Ref. 39). The staff had' discussions with:PG&E on May 7,1936, on these' concerns regarding the' proposed reracking' for the Diablo Canyon spent fuel rach end received addit;fonal information from PG&E (Ref. 40). Based on the information provided by the licensea these concerns have been acceptably resolved for the proposed reracking~ Conclusions i Gased en the'above discussion, the M aff concludes that'ths corrosion of the: spent fuel pool components due to the spent ~ fuel stcragt pool environment should be of littic significance during the life of inp facility. Components in the spent fuel storage pcol are constructed of 411oy's which have a low differential galvanic p.otential between them and have aJ11gh 1*sistance to general corrosion, localized corrosion, and galvanic ccirosion. Tests under irradiation and at elevated temperatures in borated. water indicate that the Boraflex material vill not undergo significant degradation during the projected service life of approximately 40 years for the racks. The staff further concludes that the environmental compatibility and stability , of the matet f als used in the spent fuel storage pools is adequate based on the ' A test data cited above and actual senice experience at operating reactor facilities. The staff has reviewed th? description of proposed surveillance program for monitcring the Boraflex in the spent fuel storage pools and concludes that the , program can reveal deterioration that might lead to loss of neutron absorbing , capability during the life of the spent fuel racks. The staff does not l l, anticipate that such deterioration will occur, which would be gradual. In the l unlirqiy event"of Jhrafier . deterioration in the pool enyf ronment, the monitoring program wili. detect such deterioration and the licensee /will have sufficient : time to take corrective action, for example, rep 7acement of the Soraflex sheets i in Region 1. i Thef staff, therefore, finds that implementation of'tne proposed monitoring ! program and the selection of appropriate materf41s of construction by the licensee meet the requiremsnu of 10 CFR 50, Appendix A Generat Design , 8 . l s

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a Criterion 61 regarding the. capability to permit appropriate periodic inspection and testing of components and General Design Criterion 62 regarding preventing criticality by ma w ining a structural integrity of components and of' boron poison, and are, erefore, acceptable.

4. STRUCTURAL DESIGN The staff's evaluation of the structural design aspects of the proposed spent fuel stcrage expansion is based on a review performed by the NRC staff and the Franklin Research Center (FRC), as NRC consultant. -The FRC Technical Evaluation Report TER-C5506-625 is appended to this Safety Evaluation as Appendix A. The staff accepts the findings and conclusions of the FRC Technical Evaluation Report and incorporates this report as part of its evaluation.

During the course of the review by the staff and its consultant the staff requested additional information (Ref. 3, 5, and 6) which was provided by the licensee (Refs. 7, 9, 11, 37 and 38). Structural design aspects were the- ~ subject of two meetings between the NRC and the licensee (Ref. 12 and 14). In addition, as discussed in Section 1.1, the staff and its consultant visited the rack manufacturing facilities (see at Reference 13) and audited structural analysis procedures and calculation packages (Ref. 15) at Joseph Oat Corporation, and toured the Unit 1 fuel handling building at the Diablo-Canyon Plant to observe the structural layout and features of the fuel pool (Ref. 16). 4.1 Description of the Spent Fuel pool and Racks . There are two spent fuel pools at the Diablo Canyon Plant, one for each unit. They pool are are 48 constructed feet wide, b reinforced of.'y 58concrete. feet long, Theand overall 46dimensient feet The deep of each walls of the pools are 6 feet thick, except near the fuel transfer tube. The foundation slabs have a minimum thickness of 5 feet and are founded on approximately 5 feet of lean concrete that rests on rock strata. The walls and floor of the spent fuel pool are lined with a stainless steel liner 1/8 inch and 1/4 inch thick, respectively. This liner serves only as a water tight boundary, not as a structural member. The proposed reracking for Diablo Canyon Units I and 2 will utilize 16 high-density fuel racks comprising 13 storage rack configuration designs as defined in Table 1 for each spent fuel pool.* The 16 racks are to be arranged

Tables and Figures mentioned in this Section 4 refer to the FRC Report, Appendix A.

9

l in the pool of each unit is shown in Figures 1 and 2. An elevation view of a typical rack module is shown in Figure 3. Fixed and adjustable mounting support legs are provided for each rack module to achieve leveled, freestanding positions on the pool floor (Figures 4-a and 4-b). Horizontal and vertical cross sections showing typical construction of a rack j module are providt .n Figures 5 and 6, which also show the neutrun absorbing material Boraflex required for the three (3) rack modules in Regidn 1 (Table 1). The construction details of the 13 rack modules in Region 2 (Table 1) are the same except for the deletion of the neutron absorbing I material. l Clearance between the rack modules and the pool walls is a minimum of 3.0 inches and generally greater for most modules. Clearance between adjacent modules is 2 1/4 inches, which includes 7/8 inch thick girdle bars on each rack j module for protection in the event of postulated rack-to-rack or rack-to-wall I impacts. The materials for both Region 1 and Region 2 rack modules are: stainless steel sheet and plate ASTM A-240-304L forging material ASTM A-182 weld filler material ASME SFA-5.9, Type 308L and 308 LSI Typical module storage cells of both Region 1 and Region 2 racks have an l 8.85 inch square cross section to ensure that fuel assemblies with the maximum expected axial bow can be inserted and removed from the storage cells without damage to the fuel assemblies or the rack module. 4.2 Aeolicable Codes, S'tandards and Specifications The staff has evaluated the acceptar,ce criteria for reracking used by the licensee with respect to those in " Staff Position for Review and Acceptance of Spent Fuel Storage and Handling Applications" (Ref. 34). These criteria were previously reviewed and accepted by the staff during other spent fuel pool expansion applications and, therefore, are also acceptable for this reracking design. The design adequacy of the existing Diablo Canyon Units 1 and 2 l concrete pool structures was evaluated for the new loads, which result from the ! proposed fuel racks, in accordance with the design criteria for Category I f concrete structures provided in the Diablo Canyon Final Safety Analysis Report (FSAR), Section 3.8.4. These criteria were previously reviewed during the licensing review stage and were found acceptable, j 4.3 Loads and Load Combinations Loads and load combinations considered fer the analysis and design of the racks l and the pool structure were reviewed and found to be in agreement with the l applicable portions of the staff position and the Diablo Canyon FSAR Section 4.2, which was previously reviewed and approved by the staff. Additional i details are provided in the FRC Technical Report (Appendix A). 1 10 4 6

'e 'c 4.4 Design and Analysis of Racks

a. Nonlinear Dynamic Analysis Model The licensee's nonlinear model of a spent fuel rack mcdule for dynamic displacement analysis considers the rack module as a single cell, or stick model on a rigid base with four supports. Additional impact springs that simulate impact with adjacent rack modules and/or the pool walls and the fuel assembly in a rack storage cell are used in the modelling of the racks.

Sprihgs acting through gap elements were used to represent movement of the fuel mass in clearance space before impacting the cell walls. The licensee's model includes two mass points conprising 8 degrees of freedom. Mass 1 is located in the rack module and has 6 degrees of freedom, (i.e., 3-dimensional space with three linear translations and three rotations). Mass 2 is located at the top of the fuel assembly and moves with 2 transnational degrees of freedom. The lower mass point of the fuel assembly is lumped with the rack module mass. The racks' storage cells provide clearance that accommodates possible bowed fuel assemblies. This situation is considered in the licensee's analysis for the impact of fuel assemblies on the storage cells. The analysis accounts for the hydrodynamic coupling effects of the fluid, ignores fluid and structural damping, but considers the effect of the Hosgri earthquake on all of the rack modules. These assumptions provide conservatism ir; the analyses and design of the racks. To validate the dynamic results from the simpler model, the licensee compared the computed dynamic response of the 8 degree of freedom model to that for a 32 degree of freedom model under the same acceleration time history. Both the 8 and 32 degree of freedom modules were based on a Diablo Canyon rack module with a coefficient of friction of 0.2 between the rack module supports and the pool floor liner using zero structural damping. The close agreement between x and y displacement, support loads, and stress ratios (see Section 3.1.4 of Appendix A) indicates that the 8 degree of freedom model based upon the assumption of a rigid body rack module provides satisfactory analysis of the dynamic displacements under seismic excitation.

b. Frictional Forces Between Rack Support and Pool Liner The licensee used a minimum value of 0.2 and a maximum value of 0.8 for the range of static friction coefficients between the rack support pads ard the pool liner. This range of friction coefficients is determined to be edequate to bound a realistic friction coefficient based on comparisons with available data (see Reference 5 in Appendix A).

c. Hydrodynamic Coupling The licensee has used a state-of-the-art analysis to account for fluid coupling. The assumptions consider a rigid boundary between two adjacent racks. This approach does not make a distinction for interaction of non-identical racks. This distinction is judged by the staff to be unnecessary in 11 I

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r . . . t light of the minimal difference in calculated results for identical and nonidentical racks. Furthermore, this effect would be amply accounted fo- by the fact that conservative assumptions are adopted in other aspects of modelling, analysis and design. d- Rack Loads Seismic loads for the rack design are based on the design acceleration response spectra for the floor calculated for the plant at the licensing stage. This was based on a design earthquake, a double design earthquake (DDE) and the postulated Hosgri earthquake. The Hosgri earthquake (0.75 g) controls the design of the racks and the evaluation of the spent fuel pool. The seismic' loads were applied to the model in three orthogonal directions.

e. Analysis Method The non-linear dynamic displacement analysis is accomplished with the use of the DYNAHIS computer code as discussed in Appendix A. This code has been used in the evaluation of previous rack designs. It utilizes a time-step integration method based upon the central-difference method. The solutions'by the use of DYNAHIS have been found acceptable, as stated in the Appendix A.

f. Results of Rack Analysis The licensee has considered the two limiting racks (10x11 and 6x11 cell configurations) that address the largest size rack and the one with the largest aspect ratio. The results indicate that the largest tipping or rocking response results for the rack with the largest aspect ratio.

The results are present'6d in terms of stress ratio.; (R) of actual-over-allowable stresses. There are six stress ratios categories,'as follows: , R3 = ratio of direct tensile or compressive stress on a net section to its allowable value R2 = ratio of gross shear on a net section to its allowable value R3 = ratio of maximum bending stress due to bending about the x-axis to its allowable value for the section R4 = ratio of maximum bending stress due to bending about the y-axis to its allowable value R5 = combined flexure and compressive factor R6 = combined flexure and tension (or compression) factor The results were reviewed to determine that they meet the acceptance criteria of less than one for the normal condition and less than two for severe accident conditions (see Appendix A, Tables 3 and 4). 12 e

7 . .

g. Audit of Rack Stress Analysis The staff and its consultant conducted an audit of calculations and results at Joseph Oat Corpor; tion, the rack manufacturer. The results were found to meet the acceptance criteria. However, some of the results were found slightly lower than reported earlier in the Reracking Report (Ref. 2) submitted to the staff. This is considered acceptable, because it indicates additional margin of safety incorporated in the design. This finding resulted from the combined change in the area of the four support legs for the rack assemblies and from the reduction of the material allowable stress. The fuel rack support legs were enlarged to provide a diameter of 9 inches. The material stress allowables were reduced to account for the reduction factor related to postulated temperature effects.

The results were validated by checking the change between the final and original results. The change in size of support legs would reduce the applicable stress level ratios, while the reduction in the value of allowable stress would increase the applicable stress level ratios. These conditions were verified by checking the results at specific locations in the fuel racks, and were found acceptable. 4.5 Desion and Analysis of Pool Structure

a. Pool Structural Analysis The existing pool structure was evaluated for postulated interactions of the rack modules with the structure as a result of a seismic event. This interaction results from the vertical dynamic response between the rack legs and the pool slab, the postulated lateral impacts on the pool walls, other i localized bearing loads'en the liner, leak-chase drains and pool slab, and l postulated fuel accidental impacts. In addition, the spent fuel pool structure with the increased loading of the additional spent fuel was analyzed for the previously approved seismic loadings as stated in the FSAR which showed that the increased fuel density did not make a significant change in the response of the pool structure (see Appendix A, Tables 5 through 11).

The vertical dynamic response evaluation considered the two representative l modules with regard to weight and aspect ratio (i.e., 10x11 and 6x11 rack ; modules). The large 10x11 module was evaluated for the potential impact on the j pool walls. The impact force would be applied to the wall through the girdle l bar as a line load. The other localized loads include bearing and sliding i l forces affected by the controlling Hosgri earthquake. The resulting stresses l are found acceptable because they meet the requirements identified in i respective design codes: ) concrete components ACI 318-63 liner ASME Section III Division 2 1983 For details on the spent fuel pool liner analysis results, refer to Table 12 of Appendix A. Significt.nt safety margins are incorporated in the design of the spent fuel pool liners by the licensee. 13 C-___ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _

b. Cask Pit and Cask Restraint The licensee has committed to install a removable cask lateral support prior to filling the pools with water and storing the spent fuel (Ref. 37 and 38). The lateral support was designed to prevent a shipping cask from tipping into the spent fuel racks during a postulated seismic event. Also, this restraint will remain in place when spent fuel is stored in the rack modules adjacent to the cask pit. This support will provide protection from a cask impact on the rack modules and at the same time will provide a physical barrier to prevent the spent fuel racks adjacent to the cask pit from falling into the cask pit. The staff finds this acceptable.

4.6 Fuel Handling Accident Analysis Loads due to a fuel assembly drop accident were considered in separate analyses. The postulited loads from these events and the assumptions used in the analyses were fou.id to be acceptable. Additional description and details are provided in the FRC Technical Report (Appendix A). The radiological aspect are addressed in Section 9. 4.7 Conclusions i The staff concludes, based on its evaluation of information provided by the licensee, discussions with the licensee at meetings, and information audited by the staff and its consultant, that the licensee's structural analyses of the spent fuel rack modules and the spent fuel pool are in compliance with the acceptance criteria set forth in the FSAR and are acceptable. The analyses ) indicate the rack modules and pool structure are satisfactory for high density i fuel storage. Therefore, it is concit ded that the proposed spent fuel rack installation will satisfy the requirements of General Design Criteria 2, 4, 61 and 62 of 10 CFR 50, Appendix A, as applicable to the structural components.

5. INSTALLATION OF RACKS AND LOAD HANDLING

     .         As of this time no spent fuel has been stored in either of the two fuel pools. i
   '           Unit i new fuel assemblies for one core load were stored in the Unit 1 spent     l fuel pool prior to the initial loading of the reactor, under wat0r from March l               1976 to May 1981 and dry from May 1981 to November 1983; Unit 2 new fuel         i assemblies for one core load were stored dry in the Unit 2 pool prior to        l loading the reactor, from May~1977 to May 1985. At the current time both pools are dry and no fuel is stored in either of the pools.

The proposed reracking is to be performed in a dry pool condition, prior to the first refueling of each unit, to preclude any potential contamination from spent fuel storage as discussed in Section 8 of this report. The final disposal of the existing racks has not been decided by the licensee. It will be performed in accordance with applicable state and federal requirements. 14

i The handling of both the old and the new racks fall into the category of heavy loads as defined in NUREG-0612, " Control of Heavy Loads at Nuclear Power Plants" (Ref. 21). Installation will be perfonned consistent with the licensee's previous responses to the NUREG-0612 guidelines. In SSER-27 of July 1984 and SSER-31 of April 1985 (Ref. 30) the staff concluded that the Diablo Canyon program for the control of heavy loads is in compliance with the guidelines of NUREG-0612. The staff concludes that an installation in a dry l and uncontaminated condition will not result in an accident with the potential l release of radioactivity.

6. SPENT FUEL POOL COOLING SYSTEM The spent fuel pool cooling and cleanup system for each of the two units is designed to remove decay heat generated in the stored spent fuel assemblies and maintain the cleanliness of the spent fuel pool water. The cleanup function of the system is addressed in Section 7. During its review of the Reracking Report (Ref. 2) the staff discussed with the licensee the spent fuel pool cooling system, in particular the system operation, redundancy of components, water makeup sources and heat loads to the pool for various spent fuel loading conditions in the pool (Ref. 12 and Ref. 13). The staff requested additional l information in its letter of January 8,1986 (Ref. 3) which was provided by the l licensee in two letters of January 28, 1986 (Ref. 9 and Ref. 10).

The current spent fuel pool cooling portion of the system is a single loop with one pump and one heat exchanger. Provisions exist for connecting a portable pump, shared by both units, as a backup to the permanently installed pump. The ! i staff discussed this design feature in detail with the licensee regarding the ! requirements of 10 CFR Part 50, Appendix A, General Design Criterion 44 In i its letter of January 28,1986 (Ref.10) the licensee committed to pennanently I install in parallel a s'econd, redundant, 100% capacity pump, in the spent fuel I pool cooling system for each of the two units. The arrangement for the ) portable pump was found acceptable for the current capacity of 270 fuel l assemblies (Ref. 30). The installation of the additional pump will be i completed prior to exceeding the current spent fuel capacity for 270 fuel j assemblies. The staff finds this acceptable. '

    -      The spent fuel pool cooling system heat exchanger is cooled by the component                    I cooling water system which in turn is cooled by the auxiliary saltwater system                   ,

which rejects the waste heat to the Pacific Ocean. The fuel pool water is ' pumped from the pool through the tube side of the heat exchanger and returned > to the pool. The pump suction line is protected by a strainer and is located 4 l feet below the normal spent fuel pool water level. The connections to the spent fuel pool are provided with antisiphon devices to preclude possible ! draining of the pool water. The piping of the spent fuel pool cooling system , is arranged so that failure of any pipe will not drain the spent fuel pool I below the level required for acceptable radiation shielding. The spent fuel ; pool cooling system, including the additional pump to be installed, is designed to Seismic Category I criteria and is powered by a Class 1E electrical power system. 15 r _ ___.__ _ __- - ~

l l t The licensee has calculated the heat load to the pool resulting from a spent fuel discharge in accordance with Standard Review Plan Section 9.1.3, " Spent Fuel Pool Cooling and Cleanup System "(Ref. 22) and Branch Technical Position ASB 9-2, " Residual Decay Energy for Light Water Reactors for Long Term Cooling" (Ref. 23)7 The maximum normal heat load was calculated to be 2.28 x 10 Btu / hour. The maximum normal heat load is for the last normal off-load of 76 fuel assemblies (a normal off-load of 76 assemblies is larger than a one-third core discharge of 64 assemblies). This heat load will result in a maximum bulk pool temperature of 140*F at 17 hours after the transfer of the last assembly. This meets the the Standard Review Plan acceptance temperature. The maximum abnormal heat criterion load wasof 140*F fartothe calculated be bulk 4.38 poo} x 10 Btu / hour. The maximum abnormal heat load is for the offloading of 193 fuel assemblies (i.e., one full-core) into the pool after the last normal offload. This abnormal heat load results in a maximum bulk pool temperature of 174*F which meets the Standard Review Plan acceptance criterion of no bulk pool boiling for this condition. The licensee has also considered the complete loss of the SFP cooling system for an extended period of time. For this condition natural surface' cooling , will maintain the water temperature at or below the boiling point. Boiling l would commence after 9 hours for the maximum normal heat load condition and after 2.5 hours for the maximum abnormal heat lead condition. The , corresponding boil-off rates are 22,145 lb per hour and 43,676 lb per hour, I respectively. Four sources of make-up water are available in the event of such complete loss of cooling, two of which meet the criteria for Seismic Class I. These sources provide adequate make-up for the boil-off rates. l The staff has evaluated the spent fuel pool cooling sy' stem design with respect to the criteria of the Standard Review Plan, including the performance of independent calculations, and has found they comply with these guidelines. Conclusions i Based on the above the staff has concluded that the proposed spent fuel pool storage reracking to accommodate the storage of 1324 fuel assemblies in each pool is acceptable with respect to the expected maximum heat loads, pool water temperatures, and the spe't fuel pool cooling and support system capabilities. The staff finds the licensee's commitment to install a redundant pump in the cooling system for each unit acceptable.

7. SPENT FUEL FOOL CLEANUP SYSTEM I The function of the spent fuel pool cleanup system is to maintain water clarity and purity. A portion of the spent fuel pool water can be diverted through the demineralized, the resin filter, and the mechanical filter (strainer) in the clean-up system. A check valve in the piping to the demineralized prevents backflushing demineralized resins to the spent fuel pool. Transfer canal water may also be circulated through the same demineralized and filter by opening the gate between the canal and the spent fuel pool. This purification loop will I remove fission products and other contaminants which could be introduced into l

16 i

                                                                                          .__-__-_____A

 ,                                                                                                          1 i

l I the p' col water if a fuel assembly with defective cladding is transferred to the spent fuel pool. Radioactivity and impuri ,, levels in the water of a spent fuel pool increase primarily during refuelin; ;perations as a result of fission product leakage from defective fuel elements being discharged into the pool and to a lesser degree during other spent fuel handling operations. The reracking of the spent fuel pools at the Diablo Canyon Plant will not increase the refueling frequency ) and fraction of the core replaced after each fuel cycle. Therefore, the frequency of operating the spent fuel pool cleanup system likewise is not expected to increase. Similarly, the chemical and radionuclides composition of the spent fuel pool water will not change as a result of the proposed reracking. Following the discharge of spent fuel from the reactor into the pool, the fission product inventory in the spent fuel and in the pooi vater i One year after the refueling (i.e., after will decrease by) radioactive decay. reactor shutdown , which is less than the 18 months f cycle, the radioactive inventory has decreased by 99% (Ref 41). Furthermore, experience also shows that there is no significant leakage of fission products from spent fuel stored in pools after the fuel has cooled for several months. Thus, the increased quantity of spent fuel to be stored in the Diablo Canyon fuel pools will not increase significantly the total fission product activity l in the spent fuel pool water during the operation of the pools. l l The staff has evaluate ~d the infonnation provided by the licensee. Based on this l evaluation and its experience with oth'er high density spent fuel storage facilities, including evaluation of operating data, the staff has determined that the proposed reracking of the spent fuel pools at Diablo Canyorc , vill not i adversely affect the performance capability or capacity of the spent fuel pool I cleanup system. The ra, bioactivity and impurities in the pool water are not expected to increase as a result of the reracking. Replacement of. filters or demineralizers would offset any unanticipated increase of the radioactivity and , impurity level of the water in the event of a reduction of the decontamination I effectiveness. ) l The staff has determined that for the proposed fuel storage expansion the existing spent fuel' pool clean-up system (1) provides the capability and capacity to remove radioactive materials, corrosion products, and impurities from the pool water, and thus meets the requirements of 10 CFR Part 50, Appendix A General Design Criterion 61 as it relates to appropriate filtering ! l systems for fuel storage; (2) is capable of reducing occupational exposurers to l l radiation by removing radioactive products from the pool water, and thus meets i the requirements of 10 CFR Part 20, Section 20.1(c) as it relates to i maintaining radiation exposures as low as is reasonably achievable; (3) j cor. fines radioactive materials in the pool water to the demineralized and filters, and thus meets Regulatory Position C.2.f(2) of Regulatory Guide 8.8 ! l (Ref. 24), as it relates to reducing the spread of contaminants from the source; and (4) removes suspended impurities from the pool water by filters, i and thus meets Regulatory Position C.2.f(3) of Regulatory Guide 8.8 (Ref 24), ' as it relates to removing crud from fluids through physical action. 17 , l

    . ~.

Conclusions On the basis of the above evaluation, the staff concludes that the spent fuel pool cleanup system meets the requirements of 10 CFR Part 50, Appendix A, General Design Criterion 61, 10 CFR Part 20, Section 20.1(c) and the appropriate sections of Regulatory Guide 8.8 and, therefore, is acceptable for the proposed reracking of the spent fuel pool.

8. RADIATION PROTECTION AND ALARA CONSIDERATIONS The staff has evaluated the radiological aspects of the proposed reracking of the and spent fuel Guide Regulatory pool using(the criteria 8.8 Ref. 24). Duringofitsthe Standard review Review Plan, Section 12 of the proposed reracking the staff requested additional information regarding radiation protection and ALAPA considerations (Ref. 3). The licensee responded in a letter of January 28, 1986 (Ref. 9).

The licensee has proposed to perform the reracking to the Diablo Canyon spent fuel pools with the pools in a dry and radiologically clean condition, that is prior to the first refueling. This will preclude the need for contamination and airborne radioactivity controls, and will result in essentially no occupational dose incurred from the reracking process or from handling radioactive materials. This aspect is also addressed in the Environmental Assessment on the proposed reracking (Ref. 25). During the design review process for the spent fuel pool reracking effort, the licensee utilized ALARA design considerations consistent with Regulatory Guide 8.8. These included the following: (1) ALARA design" review meetings during the design and layout phase. (2) Identification and incorporation of ALARA design changes for post-reracking operations, such as improvement of cell alignment capability to reduce spent fuel handling time.

    .           (3) Evaluation of concrete and water shielding to determine dose rate impacts on plant areas and spent fuel pool operations from increased spent fuel storage.

(4) Utilization of utility and consultant personnel experienced in spent fuel pool reracking. modifications. (5) Incorporation of experience and lessons learned from other utilities in the reracking process and post-reracking operations. (6) Overall evaluation of the radiological impact on occupational exposure and spent fuel pool and other plant operations. (M Evaluation of changes in potential radiological releases to the environment. j j 1 1 18 j i 1 i

4

o , (8) Evaluation of potential changes in radioactive waste generation and handling at the facility. 3 (9) Evaluation of the radiological impact of increased fuel storage on the ventilation system and spent fuel pool water cleanup system capacity and function. The impact on the occupational dose from spent fuel pool operations following , reracking is expected to be minimal. The licensee has conservatively ' calculated increases in doses and dose rates which could result from the 3 increased spent fuel storage. The increase in annual occupational dose from j spent fuel pool operations for both units is estimated not to exceed 4 2.4 person-rem per year, and 96 man-rem over the plant life. This is less than 1% of the anticipated occupational dose for the Diablo Canyon facility, and , much less than 1% of the overall average dose for PWR's in the United States. I The water level above the spent fuel will remain unchanged at 10 feet during refueling operations and at 23 feet during storage. This continues to meet the l staff criteria and requirements for water shielding for spent fuel pools. As l l discussed in Section 7 the proposed spent fuel pool expansion will not increase the frequency of refueling operations and, therefore, the dose rates associated l with this activity will not change as a result of the reracking. Also, as i discussed in Section 7, the fission product leakage from defective spent fuel l into the pool water is highest at the time of the refueling operations and decreases significantly after the fuel has cooled for several months in the pool. The activity in the spent fuel and in the pool water significantly decreases by radioactive decay during the first year after refueling. There-fore, the dose rates from stored spent fuel and the pool water activity will not change significantly as a result of the proposed reracking. Finally, since there will be no significant increase in spent fuel pool water activity as a I result of the reracking there will be no significant increase in the dose rates I and doses associated with spent fuel pool water cleanup system filter change outs. Radioactive waste volumes are not expected to change. I l l The licensee has considered that, as a result of the reracking, some radiation

         .zonas next to the spent fuel pool walls could experience temporary dose rates above the FSAR designated zone dose rates; however, this could occur only if freshly discharged fuel elements were placed in certain storage locations close to the pool walls and decay is not considered. These affected zones are already designated and controlled as radiation areas and are typically low l

occup5ncy areas. The applicable radiological control. requirements would remain

the same within these zoned areas. Rezoning these areas would not accurately l reflect the normal zone radiation levels that will exist after only a short j decay period and for most of the plant life. i Since the airborne radioactivity and water activity will not significantly j increase, no design or capacity changes in the spent fuel pool ventilation system or spent fuel pool water cleanup systems are needed for radiological reasons. Similarly, the radiation monitoring system for the spent fuel pools as described in the FSAR are adequate and remain unchanged.

I 19 1

The licensee has comitted that only fuel from the Diablo Canyon Plant will be stored in the spent fuel pools. Fuel from other facilities will not be received or stored at the facility. The radiation protection program at the-Diablo Canyon Plant incorporates ALARA measures which include procedures and training for radiation protection, application of ALARA engineering controls such as temporary shielding, and ALARA work control procedures. These will be applied to spent fuel pool i operations at the Diablo Canyon Plant. The licensee also provided a description of contained and airborne radioactivity sources related to the spent fuel pool water which may become airborne as a result of failed fuel and evaporation. The staff has reviewed these source terms and finds them acceptable. Conclusions Based on its review of the licensee's submittals, the staff concludes that the projected activities and exposure (person-rem) estimates resulting from the fuel pool expansion are reasonable. The licensee has factored ALARA considerations into the design of the reracking and into post-reracking spent fuel pool operations. Previous staff evaluations of the licensee's radiation

protection programs and performance es documented in the Safety Evaluation Report arid in regional inspection reports iridicated that the licensee has the capability to implement adequate radiation protection measures and ALARA dose-reducing activities. The staff concludes that the licensee will be able to maintain individual occupational exposures within the applicable limits of 10 CFR Part 20, and maintain doses ALARA, c'onsistent with the guidelines of Regulatory Guide 8.8. Therefore, the radiation protection aspects of the proposed spent fuel pool reracking are acceptable.

9. FUEL HANDLING ACCIDENT AND CASK DROP ACCIDENT The staff has reviewed and evaluated the proposed spent fuel pool reracking

    -        with regard to a postulated fuel handling accident and cask drop accident. The review was performed to identify any change in radiological consequences of these accidents as a result of the reracking.

Calculations of offsite radiological consequences due to a fuel handling accident (i.e. the dropping of a spent fuel pool assembly onto stored spent fuel assemblies) were performed using the standard assumptions given in Regulatory Guide 1.25 (Ref. 26) and Standard Review Plan Section 15.7.4 (Ref. 27). Regardless of how many spent fuel assemblies are stored in the pool, all fuel rods in one entire, freshly discharged fuel assembly were assumed to be breached, releasing their entire gap activity in accordance with i conservative assumptions in Regulatory Guide 1.25. The assumption that all , fuel rods in one assembly rupture is conservative because the. kinetic energy available for causing damage to a fuel assembly dropped through water is fixed by the drop distance. The kinetic energy associated with the maximum drop 20

                                                                          .                  I

         ,                                            _    a.        .

height for a fuel handling accident is not considered sufficient to rupture the equivalent number of fuel rods of one assembly in both the dropped assembly and the impacted assembly. The accident evaluation conservatively assumes all rods in one entire assembly rupture and release their gap activity. Because the energy that causes the rupture of the fuel rods is fixed, impacting other fuel assemblies will not increase the total number of fuel rods ruptured. In addition, the spent fuel racks protect the stored fuel assemblies from impact and absorb some of the kinett: energy from the dropped assembly. Therefore, the assumption that one: full assembly releases its entire gap contents during a fuel handling accident is not affected by the proposed spent fuel pool expansion. In the analysis of a fuel handling accident the staff postulates the rupture of-all fuel rods in one full, freshly discharged assembly (i.e., no decrease in radioactivity due to decay after removal from.the reactor) and the subsequent release of the entire fuel clad fission product gap activity from the assembly. This assumption remains unchanged for the proposed spent fuel pool expansion. As stated by(the licensee in a response to a staff request for additionalRef. 9) the ne information initial enrichment (up to 4.5% U-235) and longer burnup times than the current licensing evaluation basis of 25,000 MWD /MTU for the average burnup, as stated in the FSAR, Section 15.5.22. The lice.,see, accordingly, performed the fuel handling accident analysis for a U-235 enrichment of ?.5% and 4.5% and for a burnup of 30,000 MWD /MTU and 50,000 MWD /MTU. The licensee has provided further information (Ref. 45) regarding the application of Regulatory Guide 1.25 (Ref. 26) assumptions in calculating the consequences of a postulated fuel handling accident involving spent fuel with an extended burnup of up to 50,000 MWD /MTV for Diablo Canyon Units 1 and 2. The additional informat' ion is largely based on the Westinghouse Topical Report WCAP-10125 (Ref. 46) which supports operation of Westinghouse fuel beyond 50,000 MWD /MTU with accumulated gap activities of less than the values in-Regulatory Guide 1.25. The staff has reviewed the Westinghouse Topical Report (Ref. 47) and concludes that it is acceptable for the Diablo Canyon analysis. The staff has reviewed the additional information and concludes, based on the maximum rod average linear heat generation rate of 8.43 kw/ft, that the w licensee's basis for using the gap activity consistent with Regulatory Guide 1.25 is acceptable for the analysis of the fuel handling accident involving extended burnup fuel. The offsite doses remain well within the guideline values of 10 CFR Part 100 and meet the acceptance critoria of Standard Review Plan Section 15.7.4 (Ref. 27). The offsite doses calculated in accordance with staff methodology will not significantly change as a result of increasing the j number or burnup of assemblies stored in both pools. The cask drop accident was also reviewed. The licensee has proposed administrative controls by Technical Specifications which would preclude the movement of a spent fuel shipping cask in an exclusion zone over and in the vicinity of stored spent fuel that could result in a cask drop or tipping accident damaging stored spent fuel. The change to Technical Specification 3/4.9.13 prohibit cask handling operations near the spent fuel pool while fuel 1 is stored in the spent fuel cask exclusion zone. 1 21 i l ______ ___________ _ _______ _ _j

 ,                                ,,                                                      . _a
  ~

Conclusions The proposed reracking will not change the consequences of a fuel handling l accident from those previously reported in the Diablo Canyon Safety Evaluation Report. The administrative controls proposed to be placed on cask movement will prevent a cask drop or tipping accident that could impact spent fuel assemblies . For these reasons, the staff finds the proposed reracking and associated changes to the Technical Specifications acceptable. i I

10. RADI0 ACTIVE WASTE TREATMENT The Diablo Canyon Plant includes a separate radioactive waste treatment system for each of the two units, designed to collect and process gaseous, liquid and solid waste that might contain radioactive material. The radioactive waste treatment system was evaluated by the staff in the Final Environmental Statement (FES), dated May 1973 (Ref. 29) and in the Safety Evaluation Report (SER), dated October 1974 (Ref. 30), in support of the issuance of operating licenses for the two units. There will be no change in the waste treatment j systems as a result of the proposed reracking. Further details of the waste i treatment are addressed in Section 2 of the Environmental Assessment (Ref. 25) on the proposed reracking.

11. SIGNIFICANT HAZARDS CONSIDERATION COMMENTS The licensee's request for these amendments was individuals'y noticed on January 13, 1986 (Ref. 17) followed by a bi-weekly notice on May 21, 1986 (Ref. 44). Separate coments, request for a hearing, and a petition for leave to intervene were filed by (1) the Mothert for Peace (Ref. 31), (2) Consumers Organized for Defense of Environmental Safety (CODES), and (3) the Santa Lucia Chapter of the Sierra Club (Sierra Club) (Ref. 33). The relevant coments of these groups are identified as M, L and S respectively and addressed below.

M-1: Mothers for Peace state "By creating more storage capacity, the radioactive inventory will be greatly increased and would therefore, present a significant hazard to our comunity in the event of an accident caused by an earthquake." The highest levels of radioactivity in a spent fuel pool occur imediately after the offloading of spent fuel elements. As discussed in Section 7, after one year of storage the initial radioactivity inventory has decreased by 99% due to radioactive decay (Ref. 41). There would be a gradual buildup over the potential storage period of long lived radioisotopes in the inventory stored in the pool. However, this is principally activity contained in the fuel matrix within the cladding. Potential accidents have been evaluated (see Section 12) and there are no sources of failure of the pool or its cooling system that I could provide a mechanism for the dispersal of the fuel pool inventory. The slight increase in inventory of long lived fission products has no effect on the potential for any such accident, and in the absence of such potential has no effect on consequences to the public. The only radioactive gas of significance due to the storage of additional spent fuel would be the noble gas Krypton-85 (Kr-85), which has a half life of 10.8 years (Ref. 25). However, experience 22 l l _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . . _ _ _ _ . . _ _ _ _ _ _ l

,/

has demonstrated that after spent fuel has been moved from the reactor to the spent fuel pool and has cooled for 4 to 6 months, the gaseous fission products, including Kr-85, have been released from the stored spent fuel with cladding detects. The release of. gaseous fission products, including Kr-85, from non-defective fuel is insignificant in comparison with the overall releases from routine plant operations. The staff and its consultant have evaluated the seismic design of the proposed changes to the spent fuel pool and storage of additional fuel assemblies as discussed in Section 4 of this report and sumarized in Section 12 for the l seismic event. The racks and the pools were designed to seismic Category 1-requirements and meet the structural acceptance criteria and seismic design i bases for Diablo Canyon. The staff concludes that the spent fuel pool and spent fuel racks have been designed to withstand the postulated seismic loadings and, therefore, there will be no significant hazard resulting from a postulated seismic event. M-2: Mothers for Peace refer to the past seismic design evaluation for Diablo Canyon and to the currently ongoing Long Term Seismic Program (LTSP) for the plant. In sumary, it is stated that, "Without a complete seismic record, the margin of safety could be seriously jeopardized, and thus reracking would again pose a significant hazard." i In April 1984 the Comission included in the Diablo Canyon Unit I low-power license a condition requiring the licensee to develop and implement a program to reevaluate the seismic design bases for Diablo Canyon (Ref. 36). The background for this condition was-addressed in SSER-27 of July 1984. The staff concluded that there is no reason to modify previous conclusions on the seismic design bases while the. program was being carried out. Additional details were provided in the license condition when restated in the Unit 1 full.-power license in November 1984 by including four specific elements. The applicability of the long term seismic program (LTSP) to Diablo Canyon Unit 2 was addressed in SSER-31 of April 1985. The LTSP was established to reevaluate the seismic design bases for Diablo Canyon, taken into consideration all relevant information that has become i available since ths review by the ACRS in 1978 (Ref. 42). The program was not 'I required because the staff questioned the adequacy and applicability of the l current seismic design bases. This position was a basis in the staff's determination that low and full power licenses be issued for both units, a view which was ebnfirmed by the Comission itself (Ref. 43). The current seismic ! design basis for the Diablo Canyon Plant is therefore appropriate for the I proposed reracking. The staff concludes that it is not necessary to complete the LTSP prior to the reracking. S-1: Sierra Club does not agree with the licensee's evaluation that the i proposed reracking does not involve a significant increase in the consequences i of a seismic event. ! 23

9 This concern is'the same as raised by Mothers for Peace, in part, and is addressed at M-1, second paragraph. The staff concludes there will be no significant change in the consequences resulting from the postulated' seismic event. 5-2: Sierra Club does not agree with the licensee's evaluation that the proposed reracking does not involve a significant increase "in the consequences of a loss of spent fuel cooling." The staff has evaluated the loss of spent fuel pool cooling as discussed in Sections 6 and 12 of this report. The staff has concluded that the-consequences will not be significantly increased form those previously evaluated. S-3: Sierra Club disagrees with the licensee's finding the "the proposed l reracking does not involve a significant reduction in a margin of safety." The staff has evaluated the change in margin of safety as a result of the proposed reracking as discussed in detail in Section 12 of this report.- The staff has concluded that the reracking does not result in a significant reduction in a margin of safety with respect to criticality, cooling or' structural considerations. C-1: C.0.D.E.S. believes that the License Amendment Request "does involve a significant increase in the consequences previously evaluated." The licensee has evaluated certain accidents and events (Re? 1). The staff has reviewed and evaluated the information provided by the licensee and has concluded, as discussed,,in Section 12 of this report that the selection of accidents and events was appropriate and that the consequences of previously analyzed accidents and events will not be significantly increased as the result l of the proposed expansion of the spent fuel pool' capacity and stor9ge of spent fuel as't:amblies. C-2: C.O.D.E.S. believes that the License Amendment Request "does create the.

         , possibility of a new or different kind of accident from any acrident previously evaluated."

The licensee has evaluated the possibility of a new or different kind of any accident previously evaluated (Ref. 1). The staff has reviewed and evaluated the information provided by the licensee and has concluded, as discussed in Section 12 of this report, that the proposed reracking dces not create the 4 possibility of a new or different kind of accident from any previously I evaluated for the Diablo Canyon spent fuel storage facility. ! C-3: C.0.D.E.S. believes that the License Amendment Request "does involve a significant reduction in the margin of safety." This concern is the same as raised by Sierra Club and is addressed at S-3 above. I 24 i _ - _ -_- a

C-4: C.0.D.E.S. states that 'the licensee has neglected entirely the part of human error or foibles may play in the probability or consequences of an accident involving the re-racked spent fuel ponds." The high density spent fuel racks are being manufactured by the Joseph Oat Corporation in accordance with the company's quality assurance program and control procedures. This process was audited by the licensee. The spent fuel pool modification and installation of the new racks will be performed in accordance with the licensee's quality assurance program. Operation of the spent fuel pool including loading and unloading of fuel elements will be performed in accordance with the Technical Specifications and plant operating procedures. The licensee has evaluated a spent fuel element drop accident and spent fuel cast tipping and drop accident, which are not expected to occur, but is postulated as a result of equipment design, mechanical or human error. As discussed in Section 9 the consequences will not be changed as a result of increasing the number of assemblies stored in the pool. The licensee also evaluated the misplacement of a fresh fuel assembly into the unpoisoned Region 2 and the placement of an assembly outside the racks as discussed in Section 2. In both cases the k-effective value was below 0.95 and the staff determined the analysis to be acceptable.

12. FINAL NO SIGNIFICANT HAZARDS CONSIDERATION The licensee's request for amendments to the operating licenses for Diablo Canyon Units I and 2, including a proposed determination by the staff of no significant hazards consideration was individually noticed in the Federal Register on January 13, 1986 (Ref. 17) followed by a bi-weekly notice on May 21, 1986 (Ref. 44). This is the staff's final determination of no significant hazards consideration.

The Commission's regulations in 10 CFR 50.92 include three standards used by the NRC staff to arrive at a determination that a request for amendment (s) involves a no significant hazards consideration. These regulations state that the Comission may make such a final determination if operation of a facility in accordance with the proposed amendment (s) would not (1) involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety. The proposed spent fuel pool expansion amendments are similar to more than 100 earlier requests from other utilities for spent fuel pool expansions. The majority of these requests have already been granted by the NRC, others are under staff review. The knowledge and experience gained by the NRC staff in reviewing and evaluating these similar requests were utilized in this evaluation. The licensee's request does not use any new or unproven technology in either the analytical techniques necessary to support the expansion or in the construction process. l The staff has determined that the licensee's request for amendments to expand the spent fuel pool storage capacity for Diablo Canyon Units 1 and 2 by 25 L

reracking to allow closer spacing of spent fuel assemblies does not significantly increase the probability or consequences of accidents previously evaluated; does not create new accidents not previously evaluated; and does not result in any significant reduction in the margins of safety with respect to criticality, cooling or structural considerations. The following staff evaluation in relation to the three standards demonstrates that the proposed amendments for the SFP expansions do not involve a significant hazards consideration. First Standard

              " Involve a significant increase in the probability or consequences of an accident previously evaluated."

The beenfollowing identifiedpostulated previouslyaccidents and events by the licensee in the involving FSAR Ref. (spent fuelwere

48) and storage had evaluated by the staff (Ref. 30) during the operating license review.

(1) Spent Fuel Assembly Drop Accident (2) Spent Fuel Shipping Cask Drop Accident (3) Loss of Spent Fuel Pool Cooling ) (4) Seismic Event (5) Tornado Generated Missile (6) Criticality A'ccident (7) Reracking Installation The licensee has considered these accidents and events as part of the reracking amendment request (Ref. 1) and the Reracking Report (Ref. 2). The staff has evaluated the accidents and events as discussed below. In general, as discussed in Section 7 and elsewhere in this report, the increase in the number of spent fuel assemblies to be stored in the spent fuel pool (i.e., from the l current 270 assemblies to the proposed 1324 assemblies) will not result in a ! corresponding increase in radioactive fission product inventory in the stored ! spent fuel assemblies or in the pool water throughout the life of the plant. l The total increase in the number of stored spent fuel elements by approximately l a factor of five will not be reached until about the year 2007 through successive normal offloads of 76 assemblies at each discharge and one full-core discharge. The release of fission product gap activity from defective fuel elements into the fuel pool decreases significantly after the elements cool down for several months. Furthermore, the radioactivity decreases by about 99% due to decay within about one year after reactor shutdown. Therefore, most of the spent fuel stored in the pool will be aged and will not contribute signifi-cantly to the total radioactive inventory of the spent fuel or the pool water. 26 d

                                                                                                                               ]
                                      ~
                                                           .m. ..

I The major contribution to the total inventory of radioactivity at any time is the most recently discharged spent. fuel.

1. Spent Fuel Assembly Droo Accident The spent fuel assembly drop in a spent fuel pool is evaluated in Section 9 of this report. The frequency of refueling operations will not change as a result of the proposed reracking. Therefore, the probability of occurrence for this accident will not increase. As with the existing rack configuration, it was conservatively assumed that the equivalent of all rods in one spent fuel assembly would be breached releasing the entire gap activity. There exists the potential for an increase in offsite doses as a result of an increase in fuel burnup. However, as discussed in Section 9, such increase would be insignifi-cant. The staff concludes that there is no significant increase in the con-sequences of this accident.

2. Spent Fuel Shipping Cask Drop Accident The spent fuel shipping cask drop is addressed in Section 9 of this report.

The implementation of the proposed Technical Specifications will preclude the movement of the spent fuel cask over any spent fuel and would prevent cask handling operations near the spent fuel pool while fuel is stored in the spent fuel cask exclusion zone. Therefore, the probability of occurrence for this accident will not increase significantly as a result of the spent . fuel pool expansion. The postulated dropping or tippirig of the cask outside the exclusion zone will not impact any spent fuel assembly. Therefore, the consequences of this accident will not increase from those previously evaluated and reported in the FSAR. ,

3. Loss of Soent Fuel.* Pool Cooling The spent fuel pool cooling system is addressed in Section 6 of this report with respect to its redundancy, water makeup sources, postulated maximum heat loads, and the potential for boiling. The licensee has committed to replace the portable second pump, currently shared by the spent fuel pool cooling systems for Units 1 and 2, by installing a permanent second pump in each system. This will increase the reliability of the system. The expansion of the spent fuel pool capacity will not increase the probability of a loss of spent fuel pool cooling previously evaluated by the staff in Section 9.3.2 of 1 the SER (Ref. 30). The most severe consequences of a loss of spent fuel pool I cooling would occur with the full-core offload when the pool is nearly filled !

as discussed in Section 6. The increase in consequences will be an increase in l the pool bulk water temperature which will remain below or at the boiling point. Adequate source of water are available to make-up for evaporation losses. The staff concludes that the increase in spent fuel stored in the pools will not significantly increase the consequences of loss of the spent , fuel pool cooling. The licensee also considered the offsite radiological ! consequences in the event that, in addition to the postulated loss of spent i fuel cooling system, all water make-up sources are postulated to be lost. In this case pool boiling and evaporation would occur, however, the radiological consequences are insignificant. 27 1

                                                                        -                                                         l
                                                                            --   _ _ _ . _ _ _ _ _ _ _ _ . __________________a

4. Seismic Event The structural design of the spent fuel rack assemblies and the spent fuel pool structures with respect to a postulated seismic event have been evaluated as discussed in Section 4 and Appendix A of this report. The spent fuel pool l expansion does not increase the probability of such an event. The racks and the pools were designed to seismic Category I requirements. The results of the ,

structural analyses show that the racks and the fuel pool structure meet the structural acceptance criteria for the Diablo Canyon Plant. The seismic loads resulting from a seismic event will not result in a failure of the racks or pool structuro, thus their integrity will be maintained. The' staff concludes that there will be no significant change in the consequences resulting from a postulated seismic event from those previously determined.

5. Tornado Generated Missile The spent fuel pool expansion does not increase the probability of occurrence of a tornado, the probability of generating a missile or the probability of such missile striking the fuel pool as previously evaluated by the staff in Section 3.3 of its SER (Ref. 30). Similarly, the consequences of a postulated tornado generated missile impacting the fuel pool structures have been previously analyzed by the licensee and evaluated by the staff. The spent fuel )

storage pools have adequate protection against tornado forces and tornado j generated missiles. The proposed reracking does not affect the evaluation. i The consequences of tornado generated missiles will not be significantly increased from those previously determined.

6. Criticality Accident Considerations regardin'g a criticality accident have been evaluated in Section

2. The effect of various abnormal and accident conditions on criticality were considered. With the inclusion of administrative controls in accordance with the proposed amended Technical Specifications to maintain the boron concentra-tion in the spent fuel pools at a minimum of 2000 ppm, and to limit the storage in Region 2 spent fuel racks to spent fuel assemblies based on initial enrich-ment and cumulative exposure, none of the postulated conditions will result in a criticality accident. There is no significant increase in the probability of a criticality accident due to the proposed reracking.

The change to the two-region spent fuel pool required the performance of an additional evaluation to ensure that the criticality criterion, k-effective less or equal to 0.95, is maintained. This included the evaluation of the limiting criticality condition, caused by dropping or~ misplacing of an unirradiated fuel assembly of 4.5 weight percent U-235 enrichment into a Region 2 storage cell (i.e., unpoisoned) or outside and adjacent to a Region 2 rack module. The evaluation for this case showed that, for the required boron concentration of 2000 ppm in the fuel pool water in accordance with the proposed Technical Specifications, the criticality criterion is maintained and there is no significant increase in the consequences of this accident. 28 f

7. Reracking Installation The staff evaluation of the proposed spent fuel pool expansion by removing the existing racks and installing the new racks is presented in Section 5 of this report. The reracking is performed prior to the first refueling in a dry and uncontaminated pool. Therefore, the staff concludes that the probability of an accident involving the release of radioactivity and its consequences are insignificant.

In summary, therefore, based on the above discussion, the probability or consequences of previously analyzed accidents and events for Diablo Canyon Units 1 and 2 will not significantly increase as the result of the proposed expansion of the spent fuel pool capacity and storage of spent fuel assemblies. Second Standard

                                                                             " Created the possibility of a new or different kind of any accident previously evaluated" The proposed reracking of the spent fuel pools has been evaluated in accordance with the design bases specified in the Diablo Canyon FSAR, the guidance contained in NRC position paper "NRC Position for Review and Acceptance of Spent Fuel Storage and Handling Applications" (Ref. 34), appropriate NRC Regulatory Guides and Standard Review ~ Plans, and appropriate industry Codes and Standards as listed in the Reracking Report (Ref. 2) and listed in this report.

In addition, several previous NRC safety evaluations for similar spent fuel pools expansions at other nuclear facilities have been considered. No unproven techniques and methodol,ogies were utilized in the analysis and design of the proposed high density racks and the reevaluation of the spent fuel pool structure. No unproven technology will be utilized in the construction and installation process of the new racks. The basic reracking technology in this instance has been developed and demonstrated in numerous applications for a fuel pool capacity increase which have already received NRC staff approval. As a result of the safety evaluation in this report and based on its e evaluations of similar spent fuel pool expansions, the staff concludes that the proposed reracking does not create the possibility of a new or different kind , l of accident from any accident previously evaluated for the Diablo Canyon spent l l fuel storage facilities, j Third Standard l

                                                                              " Involve a significant reduction in margin of safety".

The NRC staff safety evaluation review process has established that the issue of " margin of safety," when applied to a spent fuel pool modifications will i need to address the following areas. , I (1) nuclear criticality considerations, 4 29 i l 1

(2) thermal-hydraulic considerations, (3) material, structural and mechanical considerations. The established acceptance criteria used to assess the adequacy of facilities assure maintenance of the necessary margins of safety. This safety evaluation by the staff addresses the three areas identified above. The margin'of safety that has been' established for nuclear criticality considerations is that the effective neutron' multiplication factor j (k-effective) in the spent fuel pool is to be less than or equal to 0.95, 1 including all reasonable uncertainties and under all postulated conditions. As noted in Section 2 of this report, the criterion is met for all normal and abnormal conditions for the storage of spent fuel in the proposed configuration. The proposed amendments, therefore, do not significantly reduce the margin of safety for criticality. . The criteria used to evaluate the margin of safety with respect to thermal-hydraulic considerations of the storage of spent fuel are the methodologies, assumptions, and requirements identified in the staff's Standard Review Plan ' (SRP) Section 9.1.3 (Ref. 22), including Branch Technical Position ASB 9.2 (Ref. 23), to assure that the fuel pool temperature does not exceed the criteria of 140*F for no'rmal reload conditions (i.e., unloading of 76 assemblies) and 212'F for abnormal reload conditions (i.e., full-core discharge). As noted in Section 6 of this report these criteria are met for the normal reload conditions and the abnormal full-core discharge conditions for the bulk pool temperatures. The proposed reracking, therefore, does not j significantly reduce the margin of safety for spent fuel cooling with respect ' to thermal-hydraulic aspects. The criteria structural used and to evaluate mechanical the margin of considerations aresafety (with that 1) the respect to material, compatibility and chemical stability of the materials wetted by the pool water be demonstrated and no significant corrosion occur, and (2) the structural and mechanical design of the spent fuel pool and the storage racks maintain the fuel assemblies in a safe configuration for all environmental and abnormal loadinas using the codes, standards and specifications identified in Section 4 of the report. As noted in Section 3 of this report, the corrosion that will occur in the spent fuel pool environment will be of little significance for the life of the plant and the environmental compatibility and stability of the materials used is adequate based on test data and actual service experience in operating reactors. As noted in Section 4 and Appendix A of this report, the structural and mechanical design of the spent fuel pools and storage racks can withstand the environmental and abnormal loadings and the pool structure can sustain the ! increased floor loadings with adequate margin. The proposed reracking of both ' pools, therefore, does not significantly reduce the margin of safety with regard to materials, structural, and mechanical integrity. 30 j i

                                                                                                       -7t
                        ~

As a result of the review and evaluation by the staff as reported in this document, the staff concludes that the proposed reracking of the Diablo Canyon Units 1 and 2 spent fuel pools to expand the capacity of the pools does not result in a significant reduction in a margin of safety with respect to criticality, cooling or structural considerations. Sumary In sumary, based on the foregoing and the fact that the reracking technology in this instance has been well developed and demonstrated, the Comission has concluded that the standards of 10 CFR 50.92 are satisfied. Therefore the Comission has made a final determination that the proposed amendment for spent fuel pool expansion does not involve a significant hazards consideration. l

13. ENVIRONMENTAL CONSIDERATIONS A separate Environmental Assessment has been prepared pursuant to 10 CFR Part 51. The Notice of Isruance of Environmental Assessment and Finding of No Significant Impact was published in the FEDERAL REGISTER on May 29, 1986 (Ref.

25). 14 CONCLUSIONS The staff has reviewed and evaluated the licensee's request for amendments for the Diablo Canyon Nuclear Power Plant Unit I and Unit 2, operating licenses regarding the expansion of the spent fuel pools. Based on the considerations j discussed in this safety evaluation the staff concludes that: (1) these amendments will not (a) significantly increase the probability or consequences of ac'cidents previously evaluated, (b) create the possibility , of a new or different accident from any previously evaluated of (c) { significantly reduce a margin of safety and, therefore, the amendments do not involve significant hazards considerations; (2) there is reasonable assurance that the health and safety of the public will not be endagered by operation in the proposed manner, and (3) such activities will be conducted in compliance with the Comission's regulations and the issuance of these amendments will not be inimical to the comon defense and security or to the health and safety of the public. Dated: May 30, 1986 , l l l I l 31

15. REFERENCES

1. PG&E Letter.No. DCL-85-333, October 30, 1985 from D. A. Brand (PG&E) to H.

R. Denton (NRC),

Subject:

License Amendment Request 85-13, Reracking of Spent Fuel Pools.

2. PG&E Letter No. DCL-85-306, September 19 -1985 from J. D. Shiffer (PG&E) to G. W. Knighton (NRC),

Subject:

.' Report on "Reracking of Spent Fuel Pools for Diablo Canyon Units 1 and 2."

3. U.S. Nuclear Regulatory Comission, letter dated January 8,1986, from H.

E.- Schierling (NRC) to J.. D. Shiffer (PG&E),

Subject:

Spent Fuel Pool Reracking, Request for Additional Information.

4. U.S. Nuclear Regulatory Comission, letter dated January 15, 1986 from.

l H. E. Schierling (NRC) to J. D. Shiffer (PG&E),

Subject:

Spent Fuel Pool Reracking, Request for Additional Information.

5. U.S. Nuclear Regulatory Comission, letter dated February 18, 1986 from H. E. Schierling (NRC) to J. D.' Shiffer (PG&E),

Subject:

' Spent Fuel Pool Reracking, Request for Additional Information.

6. U.S. Nuclear Regulatory Comission, letter dated February 28, 1986 from H. E. Schierling (NRC) to J. D. Shiffer (PG&E),

Subject:

Sumary of. Meeting with PG&E held on February 20, 1986 on Reracking of Spent Fuel Pools for Diablo Canyon Units 1 and 2 including a Request for Additional Information. ,

7. PGE&E Letter No. D.CL-85-369, December 20, 1985 from J. D. Shiffer (PG&E) to S. A. Varga (NRC),

Subject:

Reracking-of Spent Fuel Pools - Peak Decay Heat Loads and Water Temperatures.

8. PG&E Letter No. DCL-85-371, December 24, 1985 from J. D. Shiffer (PG&E) to l- S. A. Varga (NRC),

Subject:

Reracking Reference Material.

9. PG&E Letter No. DCL-86-019, January-28, 1986 from J. D. Shiffer (PG&E) to S. A. Varga (NRC),

Subject:

Spent Fuel Pool Reracking - Additional Information.

10. PG&E Letter No. DCL-86-020, January 28, 1986 from .1. D. Shiffer (PG&E).to S. A. Varga (NRC),

Subject:

Supplement to the Spent Fuel Pool Reracking-Report.- Spent Fuel Pool Cooling System.

11. PG&E Letter No. DCL-86-067, March 11, 1986 from J. D. Shiffer (PG&E) to S.

A. Varga (NRC)

Subject:

Response to Questions on Spent Fuel Racks. 12. U.S. (NRC)Nuclear to J. D.Regulatory (Comission, Shiffer PG&E),

Subject:

January Sumary6,1986, fromwith of Meeting H. PG&E E. Schierling held on December 5, 1985 on Reracking of Spent Fuel Pools for Diablo Canyon Units 1 and 2. 32

_m

              ..                    .                                         7-t                                                                                                    s
                                                                                       'il

13. U.5 Nuclear Regulatory Comission, January 29, 1986 from H. E. Schierling -

(NRC) to J..D. 3hiffer (PG&E),

Subject:

Sumary of Meeting with PG&E held on January 8,1986 on Detailed Control Room Design Review >and. Three Other Subjects (Note: Spent Fuel' Pool .Reracking was one of'the other thne subjects). ' 4

14. U.S. Nuclear Regulatory Comission. February 28, 1986,.from- ,

H. E. Schier11ng -(NRC) to J. D. Sbiffer (PG&E),

Subject:

Sumary of- > Meeting with PG&E held on February 20,.1986 on Reracking of Spent Fuel Pools for Diablo Canyon UM ts'I and 2. 15. U.S. (NRC)Nuclear Regulatory to J. D. Shiffer (Comis[fon, PG&E),

Subject:

May.9.1986, Sumary of NRC Audit onfrom MarchH. 24 E. Schierl and 25, 1986 at Joseph Oat Corporation, Camden, New Jersey, p

16. 23, 1986, from H. E. Schierling U.S. Nuclear Regulatory (Comission, (NRC)toJ.D.Shiffer PG&E),

Subject:

May Sumary of Site Visit on Spent Fuel Pools on April 14, 1986. 17, Federal Register Notice, Vol. 51, No. 8,1451, January 13, 1986, Pacific Gas and Electric Company; Consideration of Issuance of Amendments to- 3 Facility Operating Licenses DPR-80 and DPR-82 for Diablo Canyon Nuclear i Power Plant, Units 1 and 2, Respectively, and Proposed No Significant Hazards Detennination and Opportunity for Hearing. ,4

18. J. S. Anderson, "Boraflex Neutser, Shielding Material Product Perfonrance s

Data," Brand Industries, Inc..iReport 748-30-1, August 1979. ; J. S. Andersor, " Irradiation Study of Boraflex Neutron Shielding il

19. 4 Materials," Brand' industries, Inc., Report 748-10-1, July 1979.

20. J. S. Anderson, "A Finkl Report on the Effects of High Temperature Borated Water Exposure on BISCO Boraflex Neutron Absorbing Materials," Brand I Industries, Inc., Report 748-21-1, August 1978. l 3

  -       21. U.S. Nuclear Regulatory Commission, Report NUREG-0612. " Control of Heavy Loads at Nuclear Power Marts, Resolution of Generic Technical Activity A-6," July 1980.

22. U.S. Nuclear Regulatory Comission, Report NUREG-0800, " Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants,"

Rev. 1, July 1981, Section 9.1.3, " Spent Fuel Pool Cooling and Cleanup Systems."

23. U.S. Nuclear Regulatory Comission Branch Technical Position ASB 9-2,

                  " Residual Decay Energy for Light-Water Reactors for Long-Term Cooling",

included in Reference 22, Section 9.2.S. "Ultimat.e Heat Sink". j

24. U.S. Nuclear Regulatory Comission, Regulatory Guide 8.8, "Information !

Relevant to Ensuring that Ocr.upational Radiation Exposures at Nuclear !

33 1 ;

                                                                                               . - - _ ~ . - _ _ - - - - _ _ . _ . _ , _ _ _ . , _ _ . _ _ _ _ _ _ _ _             _

s, . ,, ' 9 Power Stations Will Be As Low As Is Reasonably Achievable," Rev. 3. June 1978. 1

25. 21, 1986 from H. E. Schierling U.S.

(NRC) Nuclear to J. D.Regulatory (.Comission, Shiffer PG&E)

Subject:

May Environmental. Assessment and:

                      - Finding of No Significant Impact - Spent Fuel Pool Expansion, Diablo Canyon Nuclear Power Plant, Units 1 and 2.                                   I Also: Federal Register Notice, 51 FR 19430, May 29, 108,6.

26._ U.S. Atomic Energy Comission, Regulatory Guide (Safety Guide) 1.25,

                       " Assumptions Used for Evaluating the' Potential Radiological Consequences of a Fuel Handling Accident in the' Fuel Handling and Storage Facility for Eciling and Pressurized Water Reactors," March 1972.-                        ,
               $7h U.S. Nuclear Regulatory Comission, Report NUREG-0800, " Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants,"

l July 1981, Section 15.7.4, " Radiological Consequences of Fuel Handling i Accidents."

28. U.S. Nuclear Regulatory Comission, Report NUREG-0575, " Final Generic Environmental Impact Statement on Handling and Storage of Spent Light Water Power Reactor Fuel," Volumes 1, 2 and 3, August 1979.

29. U.S. Atomic Energy Comission, " Final Environmental Statement Related to the Nuclear Generating Station Diablo Canyon, Units 1 and 2," May 1973.

30. V.S. Atoric Energy Comission, " Safety Evaluation of the Diablo Canyon Nuclear Power Station, Units 1 and 2" NUREG-0675, October 1974 and Supplements No I through No. 33, as applicable.

31. Mothers for Peace to U.S. Nuclear Regulatory Comission, Secretary of the Comission, letter dated February 7,1986.

32. Consumers Organized for Defense of Environmental Safety (C.O.D.E.S.) to U.S. Nuclear Regulatory Comission, Secretary of the Comission, letter dated February 12, 1986.

33. Sierra Club - Santa Lucia Chapter to U.S. Nuclear. Regulatory Commission, Secretary of the Comission, letter dated ' February 10,-1986.

34 U.S. Nuclear Regulatory Comission, "0T Position for Review cnd Acceptance ! of Spent Fuel Storage and Handling Applications," April 14, 1978, and. i

                        " Modifications to the OT Position," January 18, 1979 letters from B. K.

Grimes (NRC) to All Power Reactor Licensees. 2

35. E.' G. Br'sh u and W. L. Pearl, " Corrosion and Corrosion Product Release in i I

                       ' Neutral Feedwater, " Corrosion, Vol. 28, p.129 (April 1972).

l s 1 34 y i

                                                                         ~
 '.o    .

e

36. U.S. Nuclear Regulatory Comission, " Amendment No. 9 to Facility Operating License DPR-76",-Letter from D. G. Eisenhut (NRC) to J. O Schuyler (PG&E),

dated April 18, 1984.

37. PG&E Letter No. DCL-86-108, April 24,1986 from'J. D. Shiffer (PG&E) to S. A. Varga (NRC),

Subject:

Additional Information on Spent Fuel Racks. ,

38. PG&E Letter No. DCL-86-109, April 24,1986 from J. D. Shiffer (PG&E) to S. A..Varga (NRC),

Subject:

. Supplemental Information-Regarding Spent Fuel Racks.

39. U.S. iuclear Regulatory Commission, April 25, 1986, from T. V. Wambach (NRC) to K. W. Berry (Consumers Power Company),

Subject:

Expansion of Spent Fuel Storage Cariacity at Pa111sades Plant.

40. PG&E Letter No. DCL-86-126, May 9, 1986 from J. D. Shiffer to S. A. Varga (NRC),

Subject:

Supplement Infortnation on Spent Fuel Racks.

41. U.S. Nuclear Regulatory Comission Information Report SECY-83-337.. cated August 15, 1983, from Executive Director of Operations to Commissioners,

Subject:

Study of Significant Hazards.

42. U.S. Nuclear Regulatory Comission July 14, 1978, from S. Lawroski, Chairman - ACRS, to J. Hendrie, Chairman - NRC,

Subject:

Report on Diablo

Canyon Nuclear Power Station Units 1 and 2.

43. U.S. Nuclear Regulatory Comission (Diablo Canyon Nuclear Power Plant, Units 1 and 2):

a.) Memorandum an'd Order CLI-84-5, dated April 13, 1984, 19.NRC 953, 961-2(1984). j b.) Decision CLI-84-12, dated August 10, 1984, 20 NRC 249, 251 n.3 (1984). c.) Memorandum and Order CLI-84-13, dated August 10, 1984, 20 NRC 267, 276(1984).

           '4   Federal Register Notice, Vol. 51, No. 98, 18699, May 21, 1986, Pacific Gas and Electric Company; Consideration of Issuance of Amendments to Facility Operating Licenses DPR-80 and DPR-82 for Diablo Canyon Nuclear Power Plant, Units I and 2, respectively, and Proposed No Significant Hazards Determination and Opportunity for Hearing.

45. L. M. Petrie and N. F. Cross, " KENO-IV, An Improved Monte Carlo Criticality Program," ORNL-4938, Oak Ridge National Laboratory, November 1985.,

46. Westinghouse Electric Corporation, Report WCAP-10125 (proprietary and non-proprietary version), " Extended Burnup Evaluation of Westinghouse Fuel," July 1982.

35

                                                                                               ~
           ,.             ;; -.   ~         .,           ,               ,
     .        j,                                     ;
             ,      47. U.S. Nuc? ear Regulatery Commission, Letter from C. O. Thomas (NRC) to E. P. Rane.(Westinghouse), dated October 11, 1985,

Subject:

Acceptance

                 .>         for Referencing of Licensing Topical Report WCAP---10125 (P), " Extended Burnup Evaluation of Westinghouse Fuel."

48. PG K Letter No. DCL-85-308, September 20, 1985 from J. D. Shiffer'(PG&E) to H. R. Denton (NRC),

Subject:

.FSAR Update Rev. 1.

49. PG&E Le'tter No; DCl-86-149, May 28,1986 from J. D. Shiffer (PG&E) to S. A. Varga (MC),

Subject:

Spent Fuel Pool Reracking Additional Information. , V 4 l l 1 I t 36 [

APPENDIX A TECHNICAL EVALUATION REPORT FRANKLIN RESEARCH CENTER DATED APRIL 30, 1986 1 1 i e e

                                                         . _ _ _ _ _ _ _ _ _ . _ _  ___m_____________.__

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    .

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4 i FRANKLIN RESEARCH~ CENTER DIVISION OF ARVIN/CALSPAN IVALUATICN CT SPENT TUEL RACKS STRUCTURAL ANAI,YSIS PACITIC GAS & ELECTRIC COMPAh7 l DIABI,0 CANYON UNITS 1 AND 2 TER-C5506-625 TECHNICAL REPORT i

i l l l l l l i 20fM & AACI STR!rts PHitac(LPwa PA 19103 TWX 710 6701889 TEL. (2151 u81000 1 i _ _ _ _ _ _ _ _ _ _ _ _ _ __________.___________D

b l TECHNICAL EVALUATION REPORT NRC DOCKET NO. 50-275, 50-323 FRC PROJECT C5506 I NRC TAC NO. -- FRC ASSKINMENT 26 NRC CONTRACT NO. NRC 03-81 130 FRC TASK 625 I I i EVALUATION OF SPDIT TUIL RACKS STRUCTURAL ANALYSIS PACITIC GAS & ELECTRIC COMPANY I DIABLO CANYON UNITS 1 AND 2 j i TER-C5506-625 . J Prepared for Nuclear Regulatory Commission FRC Group Leader: R. C. Herrick Washington, D.C. 29555 NRC Lead Engineer: T. Rinaldi April 30, 1986 { I l !

              ! ..                                                                                                                                                 This report was prepared as an account of work sponsored by an agency of the United States               l Government. Neither the United States Government nor any agency thereof, or any of tholt employees, makes any warranty, expressed or implied, or assumes any legal lisellity or responsibility for any third party's use, or the results of such use, of any information, appa-         i retus, product or process disetened in this report, or represents that its use by such third             !

party would not infringe privately owned rights. j Prepared by: Approved by: fM4xsl~44' Principal Author Ar#k~

                                                                                                                                                                                                                                          ' O'epartment ire or
                                                                                                                                                                                                                                                                            \

Date: 4-3#-N Date: 4 - D--{N l

           .                                                                                                                                                                                                                     FRANKLIN RESEARCH CENTER                  I I
           .                                                                                                                                                                                                                     DMS3ON OF ARVIN/CALSPAN                   '

sovn a aAct rrrerrs Puswisumena.ps ses

1 I ZER-C5506-62 5 ! i CONTD175 Section Title Pace 1 INTRODUCTION .. . . . . . . . . , . . . 1

  -.                    1.1 Purpose of the Review   .     .  .  .   .-  .-  .   .-   .             .                I 1.2 Generic Background.     .    .   .  .  .    .  .    .
                                                                                     ....-                          1 2     ACCEPTANCE CRITERIA.    .   .    .   .  .   .   .   .- .     .           ..                 3.

2.1 Applicable Criteria . . . . . . . . . .; 3 2.2 Principal Acceptance Criteria . 3 l . . . . . . . .. 3 TECHNICAL REVIEW . . . . . . . ,, . . . .- 6 3.1 Modeling of Spent Tuel Rack Modules for Seismic Analysis . 6 3.2 Evaluation of the Nonlinear Dynamic Displacement Analysis . 28 3.3 Rack Module Impact and Stress Analysis. . . . . . 31 3.4 Spent Puel pool Analysis for High Density ruel Racks . . 40 3.5 Puel Handling Accident Analysis

                                                                                     .             .             68 4     CONCLUSIONS.    .   .   .   .    .   .- .  .   .   .   .     .             .             72 5     RD fRDJCES . ,
                                        .   .   .   .    .   .. .  .   .   .   .     .            .              74 l

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TER-C5506-625 FCRrdORD This Technical Evaluation Report was prepared by Franklin Research Center under a contract with the U.S. Nuclear Regulatory Coceni. sion (Office of Nuclear Reactor Regulation, Division of Operating Reactors) for technical assistance in support of NRO operating reactor licensing actions. The technical evaluation was conducted in accordance with criteria established by this NRC. 1 i l a i i V ' e

1 i

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TER-C5506-625 i

1. INTRODUCTION 1.1 PURPOSE OF THE REVIEW 1

This technical evaluation report (TER) covers an independent review of the Pacific Gas and Electric Company's licensing report (1) on high-density ) spent fuel racks for Diablo Canyon Units 1 and 2 with respect to the evaluation of the spent fuel racks' structural analyses, the fuel racks ' j design, and the pool's structural analysis. The objective of this review was to determine the structural adequacy of the Licensee's high-density spent fuel racks and spent fuel pool. 1.2 GENERIC BACKGROUND 1 Many licensees have entered into a program of introducing modified fuel

   !       racks to their spent fuel pools that will accept higher density loadings of spent fuel in order to provide additional storage capacity. However, before the higher density racks may be used, the licensees are required to submit rigorous analysis or experimental data verifying that the structural design of the fuel rack is adequate and that the spent fuel pool structure can accoccodate the increased loads.

The analysis is complicated by the fact that the fuel racks are fully immersed in the spent fuel pool. During a seismic event, the water in the pool, as well as the rack structure, will be set in motion resulting in fluid-structure interaction. The hydrodynamic coupling between the fuel assemblies and the rack cells, as well as between adjacent racks, plays a significant l role in affecting the dynamic behavior of the racks. In addition, the racks are freestanding. Since the racks are not a.nchored to the pool floor or the I 1 pool walls, the motion of the racks during a seismic event is governed by the l 1 static / dynamic friction between the rack's mounting feet a.nd the pool floor, I and by the hydrodynamic coupling to adjacent racks and the pool walls. l l Accordingly, this' report covers the review and evaluation of analyses l submitted for Diablo Canyon Units 1 and 2 by the Licensee, wherein the l structural analysis of the spent fuel racks under seismic loadings is of primary concern due to the nonlinearity of gap elements and static / dynamic 1

                                                                                                                              'I TER-C5506-625 1

I friction, as well as fluid-structure interaction. In addition to the .q j evaluation of'the dynamic structural analysis for seismic loadings, the design ef the spent fuel racks and the analysis of the spent fuel pool structure I l under the increased fuel load are reviewed. I l i I I l l i i \ \ \ i 4 i

TER45506-625

2. ACCEPTANCE CRITERIA 2.1 APPLICABLE CRITERIA The criteria and guidelines used to determine the adequacy of the high-density spent fuel racks and pool structures are provided in the following documents:

o OT Position for Review and Acceptance of Spent Puel Storage and Handling Applications, U.S. Nuclear Regulatory Cecruission, January 18, 1979 [2] o Standard Review Plan, NUREG-0800, U.S. Nuclear Regulatory Cocunission ; l Section 3.7, Seismic Design Section 3.8.4, other Ca agory I Structures i l Appendix D to Section 3.8.4, Technical Position on Spent Fuel l Per Racks Sectwn 9.1, Fuel Storage and Handling o ASME Boiler and Pressure Vessel Code, American Society of Mechanical Engineers, Section III, Division 1 i

o. Building Code Requirements for Reinforced Concrete (ACI 318-63),

American Concrete Institute o Regulatory Gui, des, U.S. Nuclear Regulatory Cocunission l 1.29 - Seismic Design Classification 1.60 - Design Response Spectra for Seismic Design of Nuclear Power Plants l 1.61 - Damping Values for Seismic Design of Nuclear Power Plants 1.92 - Corr.bining Modal Responses and Spatial Components in Seismic Response Analysis o Other Industry Codes and Standards i American Institutes of Steel Construction. I 2.2 PRINCIPAL ACCEPTANCE CRITERIA i The principal acceptance criteria for the evaluation of the spent fuel racks' structural analysis for Diablo Canyon Units 1 and 2 are set forth by the NRC's OT Position for Review and Acceptance of Spent Puel Storage and J p ___ . . _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _

67 ,3.' , TER-C5505-625 Handling Applications (OT Position Paper) [2]. Section IV of the document describes the mechanical, material, and structural considerations for the fuel racks and their anclysis. The main safety function of the spent fuel pool and the fuel racks, as stated in that document, is "to maintain the spent fusi assemblies in a safe configuration throv;h all environmental and abnormal loadings, such as earth-quake, and impact due to spent fuel cask drop, drop of a spent fuel assembly, or drop of any other heavy object during routine spent fuel handling."- Specific applicable codes and standards are defined as follows:-

                                                                                              " Construction materials should conform to Section III, Subsection NF of the ASME* Code. All materials should be selected to be compatible with the fuel pool environment to minimize corrosion and galvanic effects.                                               -

Design, fabrication, and installat, ion of spent fuel racks of stainless steel materials may be performed based upon the AISC** specification or Subsection NF requ!raments of Section III of the ASME B&PV Code for Class 3 component supports. Once a code is chosen its provisions must be followed in entirety. When the AISC specification procedures are adopted, the yield stress values for stainless steel base metal may be obtained from the Section III of the ASME B&PV Code, and the design stresses defined in the AISC specifications as percentages of the yield stress may be used. Permissible stresses for stainless steel welds used in accordance with the AISC Code may be obtained from Table NF-3292.1-1 : of ASME Section III Code. Other materials, design procedures, and fabrication tcchniques will be reviewed on a case-by-case basis." Criteria for seismic and impact loads are provided by Section IV-3 of the

OT Position Paper, which requires the following: o Sv;ismic excitation along three orthogonal directions should be imposed simultaneously. o The peak response from each direction should be combined by the square root of the sum of the squares. If response spectra are available for vertical and horizontal directicns only, the same horizontal response spectra may be applied along the other horizontal direction.

American Society of Mechanical Engineers Boiler and Pressure Vessel Codes, Latest I' aion.

                                                                                       " American Institute of Steel Construction, Latest Edition.

p_ *e TER-C5506-625 o Increased damping of fuel racks due to submergence in the spent fuel pool is not acceptable without applicable test data and/or detailed analytical results. o Local impact of a fuel assembly within a spent fuel rack cell should be considered. Temperature gradients and mechanical load combinations are to be l considered in accordance with Section IV-4 cf the OT Position Paper. The structural acceptance criterka are prcvided by Section IV-6 of the OT ' Position Paper. For sliding, tilting, and rack impact during seismic events, Section IV-6 of the OT Position Paper provides the following:

                "For impact loading the ductility ratios utilized to absorb kinetic energy in the tensile, flexural,' compressive, and shearing modes should be quantified. When considering the effects of seismic loads, factors .of safety against gross sliding and overturning of racks and rack modules under all probable service conditions shall be in accordance with the

! Section 3.8.5.II-5 of the Standard Review Plan. This position on factors l of safety against sliding and tilting need not be met provided any one of the following conditions is met: l (a) it can be shown by detailed nonlinear dynamic analyses, that the

amplitudes of sliding motion are minimal, and impact between l adjacent rack modules or between a rack module and the pool walls is prevented provided that the factors of safety against tilting are within the values permitted by Section 3.8.5.II.5 of the Standard Review Plan' (b) it can be shown that any sliding and tilting motion will be contained within suitable geometrie ::amtraints such as thermal clearances, and that any impact due to the clearances is incorporated."

i 1' l' )<

TER-C5506-625

3. TECHNICAL REVIEW 3.1 MODELING OF SPDC TVEL RACK MODULES TCR SEISMIC ANALYSIS 3.1.1 Overview of Rack Module Dynamic Analysis The fully submerged high-density fuel rack modules, as reviewed herein, are freestanding on the spent fuel pool floor, i.e., they are neither anchored to the pool floor liner, attached to the pool walls, nor fastened to each other, but are constrained by friction between the module support pads and the l pool liner. As a consequence, the modules may exhibit highly cocplex motions under seismic excitation, thus requiring comprehensive, nonlinear, structural i

dynamics analyses to simulate, or predict, their maximum displacements and associated stresses. Sources of nonlinearity in the analyses include the following: o Hydrodynamic coupling through the water between each fuel assembly and the storage cell walls of a module, as well as that between adjacent rack modules ar.d between rack modules and the pool walls. , i o Friction (static and sliding) between the base of the rack module and the pool floor liner, which acts together with hydrodynamic coupling forces between adjacent codules and the pool walls to both excite and l restrain the module during seismic events. o Impact between the fuel assemblies and the storage cells of a rack module, wherein each fuel assembly moves through its clearance space to impact, repeatedly, the cell walls. o Impact of a rack sodule with adjacer:.t codules and/or the spent fuel

   ,              pool walls.

o Lift-off of one or tuore support pads of a rack module from the pool floor and the subsequent impact of that pad (or pads) with the pool floor. To determine the marinum rack module displacements, impact forces, and stresses, the Licensee modeled selected rack modules, performed the nonlinear dyruunic displacement analysis, and, using the results of the displacement analysis, computed the maximum stresses in the rack module and compared the stresses to allowable values. The nonlinear displacement analysis was a full three-dimensional time-history analysis Msed on earthquake acceleration time histories derived from the accepted earthquake spectrums defined in the updated

1. ~6-1 l

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                                                                                                                                                                    -TER-C5506-625 Final Safety Analysis Report (3). The use of three-dimensional' analysis resolved any question regarding adequate su:xnation of separate two-dimensional analyses through the square-root-of-the-sum-of-squares technique.

3.1.2 Rack Module Orientation in the Soent Puel Pool Diablo Canyon Units 1 and 2 plan to utilize 16 high-density fuel racks comprising 13 rack module storage configuration designs as defined in Table 1 for each spent fuel pool. The 16 rack modules are to be arranged in the pools of Units 1 and 2 as shown in Figures 1 and 2 (which were adapted from Reference 1). s Clearance between the rack mo&les and the pool walls is a minimum of 3.0 inches and generally greater for most a dules. Clearance between adjacent modules is 2 1/4 inches, which includes 7eS-inch-thick girdle bars on each rack module for rack-to-rack and rack-to-wall impacts. Note that the rack modules shown as Region 1 in the spent fuel pools of Units 1 and 2 are provided with neutron absorbing material (Boraflex) to accept spent fuel that has not achieved sufficient burnup for Region 2. The fuel rack enodules of Region 2 do not contain the neutron absorbing material. 3.1.3 Rack Module Construction i As provided by the Licensee [1], an elevation view of a typical rack module is shown in Figure 3. Fixed and adjustable mounting supports are shown in Figures 4-a and 4-b. Horizontal'and vertical cross sections showing typical construction of the rack modules are provided in Figures 5 and 6. Note that Figures 5 and 6 include the neutron absorbing material required for rack modules in Region 1. Typical construction details of Region 2 rack modules were stated by the Licensee to be the same except for the deletion of the neutron absorbing material. Typical module storage cells of both Region 1 and Region 2 racks have an 8.85-inch-square cross section. The Licensee stated that this cross sectional' dimension is sufficient to ensure that fuel assemblies with the maximum expected axial bow can be inserted and removed from the storage. cells without damage to the fuel assemblies or the rack module [1].

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4. o . !

TER-C5506-625 .i Table 1. Rack Module Storage Configuration (Table 2.2 of Ref. 1) 1 l Approximate ! Module Quantity Cells per Array weight Region Type Per Unit Module Sise * (1b/ module)- 1 A 2 100 10 x 10 21,500 1 3 1 to 9 x 10 19,500 2 C 1 100 10 x 10 25,500 2 D 3 90 9 x 10 23,000 2 E. 1 66 11 x 6 17,000 2 F 1 72 9x8 18,500

2 G 1 30 10 x 8 20,500.

1 2 a 1 24 , 10 ~~~ 15,000 2 J 1 96 9 x 10 + 6 24,500 2 K 1 54 6x9 14,000 2 L 1 31 9x9 21,000

2 E' 1 110 11 x 10 28,000 2 N 1 al 9x9 21,000 j i

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4 ________________________i

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7, ., a TER-C5506-625 , 1 i Materials for both Region 1.and Region 2 rack modules were stated to be: Stainless steel sheet and plate ASTM A-240-304L Forging material ASTM A-182 , Weld filler material ASME STA-5.9, Type 308L and 308 LSI 3.1.4 Nonlinear Dynamic Displacement Model The Licensee's nonlinear model of a spent fuel rack module for dynamic displacement analysis is shown in Figures 7, 8,-9, and 10. Figure.7 represents the rack module as a single-cell, or stick, model en a rigid base and four sppports. Figure 8 shows che addition of impact springs acting through gap elements to simulate compression-only contact (impact) with adjacent rack modules and/or the pool walls. The springs registered the impact force of such contacts in the analysis. Sixteen contact springs were used, two at each top and bottom corner, to resolve tho' horizontal forces at the corner. Figure 9 represtnts the modeling of a fuel assembly in a rack storage cell, Springs acting through gap elements were used to represent movement of the fuel mass in clearance space before impacting the cell walls. Although the Licensee showed the presence of fluid dampers, fluid damping was not used. The damper elements, however, may be used to represent hydrodynamics coupling between the fuel assembly mass and the cell walls as evaluated in a later section of this report. Figure 10 (Figure 6.3.1 of Reference 1) was used by the Licensee to clarify the model concept by showing a two-dimensional view of the three-dimensional model. Note the relationship of the fuel assembly mass (mass 2*) to the model, and that the fuel assembly masses (1* and 2*) may be offset from the rack module centroid, as shown in Figure 7, to investigate the effects of spent fuel being placed on one side of a rack module. Figure 10 also provides greater detail in the modeling of the rack module base and support. At the base plate, springs X ,are the rack-to-rack or rack-to-wall impact springs. Springs K6 model the vertical stiffness of the lower rack < and supports. They are connected to gap elements to acconrnodate lif toff.

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y ZER-C5506-625 l 1 Springs Kg are special elements that, like springs, sense loads up to some predefined limit, after which motion is permitted to take place with resist- q ance at that limiting value, and thus simulate sliding friction at a constant coefficient of friction. Springs K are the torsional counterpart of linear r friction springs K and g are used by the Licensee to include any torsional I frictional moment that may be developed at a support due to rotational dis-placement of the rack module. It was noted that the Licensee's model includes two mass points comprising 8 degrees-of-freedom (DOT). Mass 1 is located in the rack module and has 6 degrees- of-freedom, i.e., three-dimensional space with three linear transla-tions and three rotations. Mass 2* is located at the top of the fuel assembly and moves with 2 transnational degrees-of-freedom. The lower mass point of the fuel assembly is lumped with the rack module mass. The Licensee's 2-mass, 6-degree-of-freedom model was formulated in an effort to reduce the degrees-of-freedom to a minimum to facilitate simul-taneous analysis in three-dimensional space, while addressing impacts between the fuel assemblies and the storage cell walls, and between adjscent rack modules. With the rack module modeled as one mass point with 6 degrees +of-freedom, the Licensee assured that the rack module was a rigid body, and justified this assumption by citing the high natural frequen- cies l of the rack module in contrast with the much lower seismic excitation frequencies. The mass of the fuel assemblies (the major portion of mass of a loaded rack module) was divided into one mass at the top of the rack model and one lumped into the base. Kinetic energy principles were used first to estimate

the portion,of the total fuel assemblies mass that was assigned to the top fuel mass. Refinements to the proportioning procedure were made by comparisons with a parallel solution using a 32 degree-of-freedom model that j included a riumoer of mass points assigned to the fuel assembly. The j

        .        comparison to the 32 degree-of-freedom model served both as a refinement of the procedure pnd a check on the magnitude of mass assigned to the top fuel          I assembly mass point.

In order to validate the dynamic results from the limited model, the i Licensee coepared the computed dynamic response of the 8 degree-of-fr9edom model to that for a 32 degree-of-freedom model under the sa.me acceleration I

TER-C5506-625 time history. The following table was supplied by the Licensee in response to a request for more justification of the 8 degree-of-freedom model [4): 32 DCF 8 DCF Time Step 1 x 105 seca 5 x 10-6 sec* 5 x 10-6 ,,e Max. z-displacement 2.595 in 2.710 in 2.793 in Max. y-displacement 2.147 in 2.213 in 2.278 in Max, support load 1.464 x 105 lb 1.484 x 105 lb 1.22 x 105 lb on the foot Max. R6 in support 0.272 0.279 0.280 1eg { Max. R6 in the 0.115 0.106 0.123 rack body

                                                 *Two solutions, using different integration time steps Both the 8- and 32 degree-of-freedom models were based on a Diablo Canyon rack module with a coefficient of friction of 0.2 between the rack module supports and the pool floor liner using zero structural damping. Note that R 6 cenotes the ratio of th.e resulting stress to the allowable value as discussed later in this report. The close agreement between x and y displacement, support loads, and stress ratios indicates that the 8 degree-of-freedom model based upon the assu=ption of a rigid-body rack module provides satisfactory analysis of the dynamic displacements under seismic excitation.

3.1.5 Frictional Force Between Rack Supports Pads and the Pool Liner The Licensee used a maximum value of 0.8 and a minimum value of 0.2 for the range of static friction coefficient between the rack support pads and the pool liner [1]. Rabinowicz, in a report to the General Electric Company [5), focused attention on the mean and the lowest coefficient of friction to be used in these circumstances. While Rabinowicz supported the range of co-efficient used by the Licensee, he also indicated that the dynamic, or sliding, coefficient of friction is inversely proportional to velocity. The Licensee did not indicate whether the analysis used an initial static coefficient of friction and a dynamic coefficient of friction once sliding motion began. However, the range (0.8 and 0.2) of friction coefficient used by the Licensee l

        , _ = _ = , , -
                                                                                                             .=
        .        <.a                                                                                            1 l

i TER-C5506-G25 is considered to be sufficient to bound the influence of friction coefficient. Thus, the Lien.ntee's use of friction coefficient between the l support pads and the pool floor lineti is acceptable. t l 3.1.6 Hydrodynamic __C:' apling ' Hydrocyaamic coupling is the effect upon adjtcent rack arMules, or upon a fuel assembly within a fuel cell of a rack module, resulting from fluid (water) motion induced by the movement of the racks, or fuel within racks. As

                   ,    rack nodules ace?1'etate" terward each other, or ns a luel assembly accelerates toward the cell Valls, the fluid between the bodies 19 accel,arated to flow out
       /                                                                                                        1 7

of that spaca, The fluid mass causes fluid pressures to develop upon both ; i bodie. bounding the fluid. As large bodies approach each other across a small i 1 - ! I fluid space, these fluid forces become large. l The effective fluid masses, or coupling coefficient 2..were considered by the LicenseA for motion between adjacent rack redules, between rack modules I and the pcol wallri and betveen fuel assemblies and the storage cell walls. ! Fritz (6) provided data on the coupling coefficients for various body shapes and arrangements, as well as provided limited experimental verification of the technique. Fritz's data were used by the Licensee for .tnalysis of the fluid coupling in the equations of notion for fuel racks and fuel assemblies. It; l should be noted that fluid coupling, as discuss 9d here and used by the j i Licensee, f.: coupling based upon fluid inartial forces and does not constitute jg damping. , , l , Fritz's work, as applied to the rack modules, is limited in that his ! experiinental work, which ccepared well with his theory, was accceplished for l infinitesimal vibratory displacements and for relatively larg3 fluid spaces between the vibrating body and the fluid boundary wall. While this is opposite to the conditions that prevail for spent fuel rr.ek modules, the. technique is based upon well establishi..i principles in fluid mechanics and y serves to provide a lower bounding estimate of the fluid coupling for rack l ecdule analysis. I The Licensee's consideration of hydrodynamic coupling assumes that each adjacent rack redule is vibrating in a manner equal and opposite to the rack ,1 t

                 ,      module under ant. lysis. This assumption permits the consideration of a rigid 4                                                                    >

h. ,

                                          ./      ,

u 9

                                                       -                               A      W

TER-C5506-625 boundary halfway between the two adjacent racks, and permits the analysis to proceed using one-half of the real clearances between modules. While techniques such as this are useful in making problems tractable, they isolate the rack module under analysis so that the differences in rack response with adjacent dissimilar racks, or racks with different fuel loadings, are not simulated by the analysis. The Licensee's use of state-of-the art analysis to estimate the hydrodynamic mass and coupling is acceptable for analysis of the impact forces: the method includes the virtual mass of the water in the analysis of dynaraic response, and conservatively underestimates the coupling forces to yield higher impact forces with adjacent structures. 3.1.7 Fuel Assembiv Imoact Within a Storage Cell Clearance is provided between a fuel assembly and the storage cell walls to permit bowed fuel assemblies to be inserted and removed safely. Thus, under seismic excitation, the fuel assembly will vibrate within the clearance

     ,    space and, when excited to higher amplitudes, will move through the clearance space before impacting the storage cell walls.

The model for fue1 assembly dynamics is shown in Figure 9 (Figure 6.2.3 of Reference 1). The stiffness of the impact springs was derived by the Licensee from the stiffness of only a small portion of the storage cell wall, instead of including the lesser stiffness of t;he more limber fuel assembly. Thus, the impact springs used by the Licensee in the analysis were exceedingly stiff. Fluid damping was not used, but hydrodynamic coupling as discussed in the preceding section was employed. Thus, the fluid coupling forces acted upon both the rack module and the fuel assemblies prior to impact with cell walls, j Structural or impact damping was not used in the analysis of fuel assembly impact within a storage cell. l The use of stiff impact springs and no impact damping resulted in very high computed impact forces that were moderated only by the hydrodynamic coup-ling. Actually, the limber fuel assemblies usually have considerable hystere-sis in cross-sectional deformation from which the Licensee could have derived , i I

1 TER-C5506-625 considerable damping and a much lower stiffness for the impact springs. The usa of high stiffness impact springs and no damping were cited by the Licensee as added conservatism in the analysis. While conservatism is warranted, undue conservatism in a dynamic simulation analysis should generally be avoided in order to simulate the dynamic interactions of the structural components as q realistically as possible. . Evaluation of the Licensee's use of stiffer imget springs indicated, in this case, that the major effect was to predict higher is: pact forces which acted for a shorter period of time, but which did not J alter, significantly, the energy transferred during the impact. < Another aspect of fuel assembly impact, within a storage cell, was that the consideration of a single cell, or stick, model dictated that only one fuel asseebly, or one equivalent fuel assembly, could b; considered in the analysis. This was tantamount to assuming that all fuel asse blies in a given rack moved in harmony. However, when one considers that the analysis of the racks is primarily concerned with large displacements and that large rack displacements are usually associated with liftoff and tipping of a rack in a { definite mode (say, rocking about an axis parallel to its longer dimension), it is more easily accepted that the fuel assemblies can move in near harmony. However, impacts with other rack modules, especially if impact occurs at a corner, can distort the more uniform rocking motion. While rack module impacts are predicted for a Hosgri earthquake at the Diablo Canyon plant, the ! I assumption of uniform motion of fuel assemblies adds a small degree of con-l servatism to the maximum impact forces predicted. 3.1.8 Damping , I l With respect to damping in the nonlinear dynamic displacement analysis, I l the Licensee supplied the following [1): l "In reality, damping of the rack motion arises from material hysteresis j (material damping), relative intercomponent action in structures i (structural damping), and fluid drag effects (fluid damping). In the l l- analysis, a maximum of 4% structural damping is imposed on elements of the rack structure during HE seismic simulations. This is in accordance with the TSAR and NRC guidelines.* Materials and fluid damping are

              *US NRC Regulatory Guide 1.61, " Damping Values for Seismic Design of Nuclear Power Plants," 1973.

25 l

TER-C5506-62 5 conservatively neglected. The dynamic model has the provision to incorporate fluid damping effects:-however, no fluid damping has been used for this analysis." 3.1.9 Rack Modale Impact Modeling Wh'en initial design analysis, and early verification runs of the non-linear dynamic displacement analysis, by the Licensee indicated that inter-rack impacts were likely, the Licensee designed the racks to accommodate such impacts and incorporated impact springs and gap elements at each corner into the dynamic model. Elements 10 through 24 of Table 2 list the inter-rack impact elements. The following statement by the Licensee is taken from Reference 1:

            "During these verification runs, it became apparent that due to the high level of slab acceleration asbociated with the Hosgri event, inter-rack impact could be anticipated to occur, especially for low values of the friction coefficient between the support and the pool liner. To account for this potential, yet still retain the simplicity of simulating only a single rack, gap elements were located at the corners of the rack at the top and at the baseplate. [ Figure 8) shows the location of these gap elements. Loads in these elements, computed during the dynamic analys.is, are used to assess rack integrity if inter-rack impact occurs. The 8 DCF model is used to avoid possible numerical problems due to the large numer of nonlinear elements that would be required to model inter-rack impact with the 32 DCF model."

3.1.10 Seismie Leading the acceleration time-histories used to input seismic excitation to the nonlinear dynamic displacement analyses are shown as Figures 6.1.1 through , 6.1.6 of Reference 1. While the updated FSAR for Diablo Canyon Units 1 and 2 indicates that the design of structures must satisfy three earthquake conditions (DE = design earthquake, DDE = double design earthquake, and HE = Hosgri earthquake), studies performed on a typical rack module indicated that the displacements and stresses were more severe for the Hosgri event [1]. Therefore, the Licensee used DE and HZ for the design for the rack modules. The acceleration time-histories used in the Licensee's anclysis included horizontal x, horizontal y, and vertical components for the DE and HE events. The three simultaneous earthquake acceleration time-history components were input to the three-dimensional, nonlinear displacement analysis of a rack module to produce the results evaluated later in this report.

1

 ',.                                                                                                   j i                                                                                                 \

i 4 1 TER-C5506-625 'I Table 2. Gap Elements and Friction Elements (*able 6.3.1 of Ref. 1) I. Nonlinear Serines (Car Clements) (24 total) Number Node Location De s c r ip t io n 1 Support 51 I compression only element ! 2 Suppert 52 I compression only element l 3 Support 53 I compression only element l 4 Support 54 I compression only element 1 5 2 ,.2

X rack / fuel assemoly impact element j 6 2,2* x rack / fuel assembly Lapact element 7 2,2* Y rack / fuel assemcly impace element 8 2,2' Y rack / fuel assembly Lapact element 9 *op cross-section Inter-rack impact elements of rack (corners) 10
Inter-rack imptet elements 11 Inter-rack impact elements .

12 Inter-rack Lapact elements ; 13 Inter-rack impact elements 14 Inter-rack Lmpact elements 15 g,, Inter-rack impact elements 16 Inter-rack Lapact elements 17 sottom cross-section Inter-rack impact elements 18 of rack (corners) Inter-rack impact elements 19 Inter-rack Lapact'elsmants 20 Inter-rack impac t e16ments 21 Inter-rack Lapact elements 22 Inter-rack Lapact elements 23 , g, Inter-rack impact elements 24 Inter-rack inpact elements II. Priction tienents (16 total) Nemoer Node Location De s c rip t io n j l 1 Support 51 X direction support friction 2 Support 51 Y direction friction 3 Support 52 1 direction friction l 4 Support Support 52 53 Y x direction direction friction friction

  ]           5 6       Support    53        Y direction    friction

7 Support 54 X direction friction j 8 Support 54 Y direction friction ,

I 9 51 x Slab moment 10 51 Y 51ab moment 11 52 x Slab moment 12 52 Y Slab moment , l

  ,         13           53                x Slab moment Y Slab moment                                                l 4         14           53 15           54                x Slab moment                                                 l 16           54                Y Slab soment                                                i I

i I l l

TER-C5506-625 3.1.11 Selection of Rack Modules for Analysis The Licensee selected two rack modules for dynamic displacement analysis: the largest module (10 x 13 cell configuration? and tne module with the largest aspect ratio of its horizontal dimensions (6 x 11 cell configuration) [1]. The 6 x 11 rack module was also the rack with the shortest horizontal dimension. Experience in the review of spent fuel rack dynamic analysis [7, 8) has indicated that the largest displacement response in a tipping, or rocking, mode generally is realized for rack modules that have the highest ratio of height to width. Since all of the spent fuel rack modules for Diablo Canyon Units 1 and 2 are of the same height, then the racks having the shortest cross-sectional dimensional should be analyzed. The reason for this is that the natural frequency of a rack in a rocking, or tipping, mode, with liftoff of the support pads from the pool floor 1,.ner, is lowest for racks that have the highest ratio of height to width. Two rack modules of the Diablo Canyon set have the same high ratio of height to width, the 6 x 11 and 6 x 9 module designs. The Licensee chess the 6 x 11 rack for analysis. The S x 9 design, by contrast, would be expected to have similar dynamic response from the same f horizontal earthquake acceleration component directed across its 6 cell direction. However, the 6 x 9 design could suffer somewhat greater response f in the 9 cell direction. Thus, the simultaneous response of the 6 x 9 design from the two horizontal and one vertical earthquake components could produce somewhat greater total displacements, with possibly more nonlinear interaction between the two linear and one rotational motions in the horizontal plane. In sumary, any differences between the responses'of the 6 x 11 and 6 x 9 rack module designs would not be expected to be large enough to cause any of the design criteria to be exceeded, given the conservative results that were presented by the Licensee. 3.2 EVALUATION CF THE NCNLINEAR DYNAMIC DISPLACDENT ANALYSIS 3.2.1 Evaluation of the DYNAHIS Analysis Method The time-history solution of the 3 degree-of-freedom tr. mathematical model of the rack modules was described by the Licensee as follows (1):

                                                                                                                                  .         {

TER-C5506-625 "Having asse , bled the structural model, the dynamic equations of motion corresponding to each degree-of-freedom can be written by using Newton's second law of motion; or by using Lagrange's equation. The system of equations can be represented in matrix notation as: [H] (q) = [Q] + (G} wher9 the vector [Q) is a function of nodal displacements and velocities, and 1G} depends on the coupling inertia and the ground acceleration. Pramultiplying the above equations by (M)~1 renders the resulting equation uncoupled in mass. We have: (q} = [M)-1 [Q) + [M]-1 (G). This equation set is mass uncoupled, displacement coupled, and is ideally suited for numerical solution using the central difference scheme. The computer program 'DYNAHIS' is utilized for this purpose. Stresses in various portions of the structure are computed from known element forces at each instant of time." In a later cocraunication (9), the Licensee added that the algorithm of the code DYNAHIS is documented in the book, "The Component Element Method in 1 Dynamics" by Levy and Wilkinson, McGraw-Hill (1976), and provided the following description:

            "There are no numerical simulations or modeling assumptions unique to DYNAHIS. Moreover, in contrast to a general-purpose finite element code, DYNAHIS does not internally generate the equations of motion. These l I          equations are written out explicitly and can be verified independently.

Such verifications have been carried out internally at Oat, by several A/E firms, and, at the h~rtG's request, by Franklin Research Center in 1984." A detailed review of DYNAHIS was made during the evaluation of two previous spent fuel rack sedule designs [7, 8). In that review, DYNAHIS was found to be categorized as an equation solver, as opposed to well known finite-element engineering analysis computer programs. As the Licensee stated above, DYNAHIS does not generate the equations of motion internally. Rather, in the course of the analysis, the analyst formulates the set of dynamic equations from the mathematical model that was constructed (as discussed in Section 3.1), and inputs those I equations, along with the initial conditions, boundary conditions, and l l excitation time-histories, to DYNAHIS for solution. l

i _________j

TER-C5506-625 DYNAHIS was determined, in the previous reviews (7, 8), to use a time-step integration method based upon the central-difference method. In association with those reviews, a number of cociputer solutions were performed to verify that the central difference solution method was both stable and convergent. It was determined that the integration method was both stable and convergent, provided that.the integration time step was maintained within certain bounds unique to the specific problem. Outside of those bounds, the displacement was seen to rise very rapidly in magnitude, as is character.;stic of central-difference integration. Identification of integration time-step values within the acceptable region for a stable, convergent solution was accomplished satisfactorily by comparing the displacement solutions for two or more time-step values that differed by at least a factor of 2. If the computed 1 displacements were the same, or nearly the same, then the solution was found to be acceptable. 1 In response to a request, the Licensee provided the following comparison l cf displacement computed using two time-step values (4):

             " Convergency and stabi,lity of the runs are assured by running the same problem at different time steps and observing the close agreement between the results thereof. The output data of an 8 DOF run for a fully loaded 10 x 11 rack with coefficient of friction (COF) = 0.8 for two time steps are shown below.

Maximum Maximum Max. R6 Maximum i Time Step x-displacement y-displacement in Rack Floor Load l 5 x 10-6 sec 0.3402 in 0.4780 in 1.49 6.735 x 105 lb 2.5 x 10-6 sec 0.3401 in 0.4783 in 1.52 6.727 x 105 id The above table confi ms the excellent convergence characteristics of the 8 DCF model." In support of DYNAHIS, the Licensee submitted the background of the analysis method as well as four problems in which the solution by DYNAHIS was verified by other recognized analysis methods (9). This supporting material is included with this review as Appendix A. Solutions by DYNAHIS for the spent fuel rack modules for Diablo Canyon Units 1 and 2 are considered to be acceptable. l l h

TER-C5506-625 3.2.2 Rack Displacements During the design and analysis of the rack modules, it was' determined that the earthquake excitation was sufficient to cause impact of the rack modules. Thus, instead of submitting displacement data, which is generally to show that the rack modules would not impact, the Licensee submitted an impact analysis for the racks that were analyzed and compared the impact forces and stresses to allowable values. The resulting impact forces are reviewed in Section 3.3.2 of this report. 3.3 RACK HODULE IMPACT AND STRESS ANALYSIS 3.3.1 Structural Design Criteria With respect to load and stress criteria, the Licensee developed the allowable loads and stresses in accordance with Appendix D, Section 3.8.4 of the USNRC Standard Review Plan [10). However, the NRC staff have indicated that operating reactors should be based upon the staff's OT position (2) rather than the Appendix D, Section 3.8.4 of the Standard Review Plan. In response, the Licensee submitted the following comparison of the j loading combinations and limiting values of the OT position and Appendix D [4):

                  The analysis presented in the Reracking Report meets OT Position guidelines. As stated in the Roracking Report (1), the stress levels in the rack feet greatly exceed those in the rack modules proper.

Therefore, in essence, support feet stresses govern the rack design. l Since there is no interaction between Tc,, Ta loads, and inertia loads l- (D, L, E, E') in the support feet, the Standard Review plan 3.8.4 (NUREG--0800) and OT Position criteria both yield the following effective limits: Leading D+L+E Normal limits as stated in Appendix XVII-200 D + L + E' Normal condition stress limits increased by Appendix F-1370. Thus, although there are differences in the stress criteria between the OT position paper and Standard Review Plan 3.8.4, Appendix D, they amount to identical limits for Diablo Canyon." I l 1 1 l .

TER-C5506-625 In reporting the computed stresses and their comparison to allowable values, the Licenses reported the ratio of the computed stress to the allowable value as R3 through R6 . The Licensee's definitions of Ry t. rough R6 follow (1]:

                 "R1=      Ratio of direct tensile or compressive stress on a net section to its allowable value (note support feet only support compression)

R2= Ratio of gross shear on a net section to its allowable value R3= Ratio of maximum bending stress due to bending about the x-axis to its allowable value for the section R4= Ratio of maximum bending stress due to bending about the y-axis to its allowable value RS= Combined flexure and compressive factor f R6= Combined florure and tension (or compression) factor." Evaluation of the Licensee's position indicated that it is acceptable, as are the allowable stresses derived from this position as shown in Reference 1. 3.3.2 Imract toads and Stresses 3.3.2.1 Impact Between Fuel Assembly and Cell Wall The Licensee provided the following description of acceptable loads (1):

                  "The local stress in a cell wall is estimated from peak impact loads obtained from the dynamic simulations. Plastic analysis is used to obtain the limiting impact load that can be tolerated. Including a safety margin of 2.0, we find that the total limit load for the number of cells (NC) is:

Ng Limit Load (1b) 66 530100 100 883501 From the results of the dynamic analyses, we find the actual impact loads do not exceed 251,000 lb for NC = 110 and do not exceed 136,000 lb for NC

                  = 66."

As discussed in Section 3.1.7, the Licensee overestimated the spring constants of the impact springs associated with a storage cell by considering I

1 TER-C5506-625 4 only the compliance of the cell wall without including the compliance of the fuel assembly. Thus, where impact does occur, the coreputed forces are higher than would be expected. Actually, the energy of the impacting fuel t.ssembly is the same; a lower spring constant would spread the impact over a longer time period at a lower net force. The Licensee's impact forces and their comparison to allowables are shown ; 1 in Tables 3 and 4. All impact forces are below the allowable values and are

i acceptable. l 3.3.2.2 Impacts Between Adjacent Rack Modules The Licensee provided for rack-to-rack impact as follows (1): I "All of the dynamic analyses assume, conservatively, that adjacent racks move coepletely out of phase. Thus, the highest potential for inter-rack impact is achieved. Based on the dynamic loads obtained in the gap elements simulating adjacent racks, we can study rack integrity in the { j vicinity of the impact point. The use of high-yield stress framing materia. around the top of the rack allows us to pennit impact loads of up to 175,000 lb. The maximum reported value in the tables in this report is 71,400 lb. Thus, impacts between racks can be accommodated without violating rack integrity. We also study the case where the-corner of one rack impacts an adjacent rack away from a corner. We show, under such a condition, that the stress levels remain below the yield value." .- The Licenset <,cepared s pact forces to allowable values as shown in Table. 3 and 4. In response to discussion with the NRC staff concerning the effect of partially filled fuel racks, the Licensee supplied the following (9):

           "The loading cases evaluated for the Diablo Canyon limiting rack modules as reported in the raracking report (PG and E letter DCL-85-306 dated September 19, 1985) are as follows:

o Racks fully loaded j l ) o Racks empty j o Racks half full (for the 6 x 11 rack module only) . 1 o Racks partially full (11 cells). j i

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 .. 7 TER-C5506-625
   .            Tables 6.8.1 and 6.8.2 in the reracking report'also show that significant design margins exist for the above load cases.

The selection of load cases was based on past experience of Oat with cellular rack'grometries (Fermi Unit 2, Quad Cities Units 1 ard 2, Rancho Seco, Grand. Gulf Unit 1, Oyster Creek, Pilgrim, and V. C. Summer). Specifically, the selection process included the following: o Review of licensing submittals from Quad Cities (Docket Nos. 50-254 and 50-265) o Table-1 [of Ref. 9) (a copy of Table 6.5 of a licensing submittal for l Quad Cities) shows the results of a number of cases evaluated for l Quad Cities. In all cases, a fully loaded rack with H.= 0.8-1 results in maximum stress. response.. This conclusion is valid for the 1 l- Diablo Canyon rack design. Accordingly, this load was evaluated for the Diablo Canyon racks. o The kinematic response (maximum displacement) could be greater for asymmetrically loaded racks (other than the cases evaluated for Diablo Canyon). However, e h loading cases are not critical for Diablo Canyon since the raeks are intentionally engineered to accommodate rack-to-rack impact with a high margin of safety against plastic deformation. As a result of discussion with the NRC staff on February 20, 1986, PG and E evaluated an additional loading case involving an asymmetrically loaded 10 x 11 rack module. One quadrant of the sedule is assumed to be loaded with fuel assemblies. The results are {1 I as follows: l MAXIMUM l

                          . DISPLACEMENT                               LOAD FACTORS (Top Corner)                  (Maximum values occur in support lees) 1 m       RUN    CCF     EQ         Dx     Dy         R1    R;     R3        R4       R5     R6 I           HO4d   .8      DE       .0748"  .0900"    .079 .070    .173'    .164      .303   .345 M01    .8      HD       .13'20" .2627"    .105  .125' .293      .240      .455   .519 M03d   .2      DE       .0292"  .0353"    .042 .014    .033     .021      -068
                                                                                     .      .073 M02    .2      NE      1.0920"  .0664"    .194 .064 .141        .099      .310   .333 A comparison of the above results with the displacements and load factors documented in Table 6.8.2 of the reracking report (1) confirms pGandE's judgmentthat a fully loaded rack with 4 = 0.8 provides maximum response."

0.- . . TER-C5506-625 9 3.3.3 Rack Weld Stresses The Licensee indicated that the maximum weld stress in the rack occurred in the supports as follows (1):

               "The critical weld locations under seismic loading are at .the connection -

of the rack to the baseplate and in the support leg welds. For the rack welds, the allowable weld stress is the ASME Code value of 18.520 psi. For the support legs, the allowable weld stress is 80% of the yield strength (EE conditions)

35,200 psi. The welds at the rack base are fillet welds. Accounting for skip welding in this location, the maximum principal stress l's 7151 psi. The support leg weld stress is found to be 27,494 psi under HE conditions, which is less than the allowetle value.-

Weld stresses due to heating of an isolated hot cell are also computed. The assumption used is that a single cell is heated, over its entire length, to a temperature above the value associated with all surrounding cells. No thermal gradiant in the vertical direction is assumed so that the results are conservative. Using the temperatures associated with this unit, we show that the spot welds along the entire cell length do not exceed the ' allowable value for a thermal loading condition. The maximum computed shear stress in the spot welds, at the most critical location, is less than 6,400 psi under the HE condition." Evaluation indicated that this is acceptable. 3.3.4 Rack Module Tipping Stability l The Licensee provided the following as evidence of rack module stability l in the tipping mode (8):

               "With regard to the safety margin for overturning for the limiting rack module, the OT position paper, by reference to the Standard Review plan, stipulates that a factor of safety of 1.1 over the extreme condition ground motion (Hosgri earthquake for Diablo Canyon racks) be provided against rack overturning. Towards this end, the rack module with the worst aspect ratio (6 x 11 module) was run with 1.1 times the Hosgri seismic excitation. A coefficient of friction of M = 0.8 was used to produce the maximum overturning condition. The x-aris is parallel to the short side and the y-axis is parallel to the long side. The x, y origin is at the rack centerline. Two cases of eccentric loading are considersd. Both cases have 50% of the cells filled with fuel assemblies.

Case 1 loads all fuel in the positive x half of the rack; Case 2 loads all fuel in the negative y half of the rack. The following table i summarizes the results of the two analyses. ' l i l

a TER-C5506-625 Rotation in x-z plane, degrees Rotaticn in y-z plane, degrees Value for Factor Value for Factor Load Calculated Incipient of Calculated Incipient of Cace Kaximum Tippine Safety Ma ximum Tipping Safety 1 0.58 19.36 33.4 1.15 35.94 21.2 2 0.77 21.14 27.4 1.26 33.38 26.5 These results show that a large margin o'f safety against kinematic instability exists in all rack modules." 3.4 SPDIT FUEL POOL ANALYSIS FOR HIGH DENSITY IVEL RACKS 3.4.1 Spent Fuel Pool Overview The Licensee provided the following description of the spent fuel pools at the Diablo Canyon Units 1 and 2 as well as a statement of the modifications made to the pool to accommodate the high density spent fuel rack modules (4): { l "The spent fuel pools are located on each side of the east end of.the auxiliary building. Figure 11 shows the plant layout and foundation j elevations. Generally, one half of the auxiliary building is symmetrical to the other, with each half of the structure containing equipment for one unit. The Unit 1 spent fuel pool is shown in Figure 12. The walls of the spent fuel . pool are 6 feet thick except for local areas around the fuel transfer tube, as shown on (Figure 12]. The foundation slab under the spent fuel pool has a minimum thickness of 5 ft and is founded on approximately 5 ft of lean concrete placed on rock. The spent fuel pool sides and'bottoo are lined with stainless steel, 1/4 in, thick on bottom and 1/8 in, nominal thickness on the sides. Figures 5 and 6 (of Ref. 4] show the typical layout and details of the linsr. 1 The spent fuel pool structural analysis, addressed in the FSAR, is affected by the high density fuel racks due to the following: o Weight of the proposed high-density racks o Thertal gradient across the walls o Interaction between the racks and the pool structure. There are no physical changes to the spent fuel pool concrete structure addressed in the FSAR analysis. However, minor modifications to the liner plate will be made to acconcodate the high-density racke. The modifications include relocation of spent fuel handling tool brackets to the east and of the north and south walls of Units.1 and 2, respectively. (Figure 11] shows typical details of the modified brackets." i TER-C5506-625

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l 1 I Figure 12. Plan of Spent Fuel Pool Walls (Figure 3 of Ref. 4) 4 TER-C5506-625 3.4.2 Soent Puel Pool Analyses The following analys,is, a part of the updated FSAR [3), was cited by the ! Licensee as pertinent to the spent fuel pool, although no change was required to these analyses (4):

                      "As described in the FSAR, Section 3.7.2.1.7.1, the seismic inertia loads were obtained using a time-history analysis of a spring and lumped mass model of the auxiliary building. Two horizontal models and a vertical model, shown in Figure 8, were used in the analysis.

A detailed analytical static model of the auxiliary building (Tigure 9) was used to distribute seismic inertia forces and mecents to various j walls, diaphragms, and columns, as described in the FSAR, Section I 3.8.2.4." l Other analyses were updated to consider the change in fuel mass and floor loads associated with the high density spent fuel racks. Included in analyses ; that were updated was the dynamic response of the spent fuel pool under se2smic excitation. While the chsnge in global mass was determined to be less l than 1%, the Licensee included analysis document the fact that changes in response were insignificant. Other analyses that were modified were described

by the Licensee as follows (4): I "The pool walls were analyzed for local out-of-plane effects due to hydrostatic, hydrodynamic, and seismic loads using U.S. Bureau of Reclamation tables for plates (Ref. 1). Hydrodynamic effects were determined based on the methods in TID 7024, " Nuclear Reactors and i Earthquakes" [Ref.11), and " Dynamics of Fixed-Base Liquid-Storage Tanks" l (Ref. 12). Thermal effects in the peol elements due to the increased j l ther=al gradient generated from the use of the high-density racks were analyzed considering cracked concrete section. - The liner has been evaluated for loads and load combinations described in the (following section]. The mathematical model used in the thermal evaluation consists of a set of springs representing the axial stiffness ! of the liner and flexural stiffness of the embedded anchors." l Other considerations included the following: l o pool floor vertical dynamic response o pool wall impact o pool floor bearing loads

             .       o    pool leak chase system o    cask pit.

For the analyses unchanged from the TSAR, the Licensee provided the following summary (4): I

TER-C5506-625

                " Global loads - Controlling regions in the walls of the auxiliary building due to in-plane effects are sumarized in FSAR Updated Tables 3.8-14, 15, and 16 for the DE, DDE, and Hosgri events respectively.        The pool structure wall stress ratios from these tables are tabulated in Table   5."

3.4.3 Lead Ccebinations and Acceptance Criteria The Licensee indicated that the following load combinations and acceptance criteria were used for the analysis of the spent fuel pool (4):

                " Concrete Structure Lead Combinations The loads and load combinations used for the eva'luation of the spent fuel pool structure are in accordance with the FSAR Update, Section 3.8.2.3, and are listed below.

A. Normal Conditions f Dead load, live load, loads from the DE, thermal loads, and pipe reactions are considered in all possible combinations. Inasmuch as working stress design is used for normal operating loads, the factored load approach is not used. For each structural member, the combination of these loads that produces the maximum stress { is used for design. Stated in equation form: ( C = D + L + DE + To+Ro (1) where: C = Required capacity of member based on working stress of ACI 318-63 as supplemented by Section 3.8.2.5.3 of the FSAR Update, which is described in response to Item 15d (of Ref. 4) w D = Dead load of structure and equipment loads including hydrostatic loads L = Live load DE = Loads resulting from the design earthquake To = Thermal loads during normal operating conditions (inside face of wall 176*F, exposed face 91*F, and bottom face of lean concrete 101*F) Ro = Pipe reactions during normal operating conditions. l B. Abnormal Conditions Dsad load, live load, earthquake loads, and loads associated with accidental pipe rupture are considered in the following combinations; for each structural member, the combination that producesthemaximumstresgi,susedfordesign:

ai . U R-C5506-625 Table 5. Auxiliary Building Spent Fuel Pool Concrete Walls l Stress Ratios (Table 1 of Ref. 4) thear' esi Memeat i 10 3 . k f t beacitv(b) beacitv(b) fg}' Mall teettien(a) M D,3,q 1 pemaad A. DESIGN EARTHQUAXE MLL AD IQQ . 11 220 141 3.4.1 2.3 85 85 220 164 664 2.6 NALL BC 100 45 155 20 113 3.4 MLL CD 100 to 150 54 146 2.7 MLL A8 100 80 240 54 -145 2.7

8. DOUBLE DESICM EARTW)UAXE MLL AD . 1Q0 110 210 221 111 L1 85 170 370 316 1.086 2.2 MLL BC 100 85 260 40 244 3.0 E LL CD 100 160- 420 105 280 2.6 MLL A8 100 160 410 106 268 2.5 .

C. HOSORI EARTHCUAKE MLL AD f 122 122 31Q Q 110 L1 85 230 320 493 979 1.4 M LL BC 100 130 190 60 197 1.5 M LL CD 100 330 480 220 339 1.4 MALL A8 100 270 460 199 324 1.6 (a) For wall identification, see Figure Counterparts in ttnit 2 are similar. (b) Axial demand effect is included in the capacities. I (C) Stress ratio . capacity / demand for shear or soment, whichever is smaller. l l 1 l

                                                                                                              'l j                                                        45_     .

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i TER-C5506-625 U=D+L+TA+RA + 1.5 PA (2) j U=D+L+TA+RA + 1.25 PA+ ' 1.0 (Yj + Ym+Y)r + 1.25 DE (3) U=D+L+TA+RA + 1.0 PA+ 1.0 (Yj + Ym+Y)r + DDE (4) U=D+L+TA+RA + 1.0 PA+ 1.0 (Yj + Ym+Y)r + HE (5) where: ; TA = Themal loads.on structure generated by a postulated pipo I break, including To (for TA, inside face of walls 214*F, I exposed face 93*F, and bottom face of lean concrete 113*F) l RA = Pipe reactions on structure from unbroken pipe generated I by postulated pipe break conditions, including R o i FA = Pressure load within or across a compartment and/or l building generated by a postulated pipe break, and including ! l an appropriate dynamic factor (DLF) to account for the dynamic nature of the load l ' Yj = Jet load on structure generated by a postulated pipe break, including an appropriate DLF l Ym = Missile impact load on a structure generated by, or during, a postulated pipe break, such as a whipping pipe, including an appropriate DLF ) Yr = Reaction on structure from broken pipe generated by a postulated pipe break, including an appropriate DLF U = Ultimate strength required to recist design loads based on l the methods described in ACI 318-63. For load cce.binations involving Y3 Ym, and Y r , local stresses due to those concentrated loads may exceed the allowable provided there is not loss of function. See Reference 13, Enclosure 3, Document (B), for more detailed information concerning the acceptance criteria for these load combinations. DDE = Loads resulting f rca the double design earthquake HE = toads resulting from a Hosgri earthquake. For the spent fuel pool structural evaluation, PA , Yj, Y me and Y r are not applicable. l

l

      -                                                                           l l.

IER-C5506-625 Liner Load Con 6inations To assure leaktightness of the liner, the following loads and load combinations are used in the evaluation of the liner: 1 D+L+T o (6) I D + L + T, + HE (7) l Concrete Structure Acceptance Criteria l l The acceptance criteria as describad in the FSAR Update, Section l 3.8.2.5, are su:snarized below. . For DE and DDE load ccebinations, the nominal design strength for j concrete and specified yield strength for reinforcing and structural j steel are considered. For load combinations including HE, however, j the actual material properties are used. j

1. Nomal Loads

] For nomal loads, the auxiliary buildirJ is designed for the j allowable working stresses of ACI 318-63, as supplemented by Item 3 below. f 2. Abnormal Loads For abnormal loads, the auxiliary building is designed for l overall elastic behavior. For concrete elements, the strength I design of ACI 318-63 applies, as supplemented by Item 3 below. l l Ym* ""d Yr as l For load ccebinations applicable, involving local stresses due to t Yj, hose concentrated loads may exceed the allowables provided there is no loss of function. See l Reference 13, Enclosure 3, Document (B), for more detailed : information concerning the acceptance criteria for these load l combinations.

3. In-Plane Loads on Concrete Elements The design of slab diaphrages and shear walls for in-plane forces l is not explicitly covered by ACI 318-63. Section 104 of ACI 318-63 allows criteria based on test data to be used for the design of elements not covered by its provisions. Consequently, the document titled "Recoceended Evaluation Criteria for Diablo l Canyon Nuclear Power Plant Auxiliary Building Walls and Diaphragms" (Ref. 14) is developed to provide criteria for evaluation of auxiliary building shear walls and floor diaphragms for in-plane seismic forces, including the simultaneous effects of out-of-plane forces.

Accordingly, the structural elements are evaluated as follows: 6

                                    .                                             -e                                          ;

e . N TER-C5506-625 g, (1) The columns are evaluated by the provisions of ACI 318-63 for all loading conditions. ; (2) The slabs and walls are evaluated for out-of-plane loads a'ecording to ACI 318-63, and for in-plane loads e.ccording to Reference 13. j Liner'Aegoptance Criteria The allowable concrete bearing pressure under rack dead loads is in accordance with ACI 318-ti3. The design for local impact loads is not j covered by ACI 318-63. ACI 349-80 is used to determine the allowable i bearing pressure on concrete due to rack impact. The allowable foundation pressure is in accordance with TSAR UpdaEn Section 3.8.4.1.4.

l The allowable liner strains and anchor displacements for normal j operating and accident thermal conditions are in accordance with ASME Section III, Division II, 1983." !l'

     ,     3.4.4    Pool Floor Vertical Dynamic Response                                                                      .

The Licensee provided the following. summary of' the vertical impacts on 3 the pool floor structure (4):

                 "The interaction between the pool structure and the rack modules is primarily betweenfthe support feet and the pool slab. The table below shows the maximum cu=ulative support reaction (sum of 4 feet) for the two representative modules studied (Hosgri earthquake):

Module Maximal value of Impact Type Dead Lead (b) cumulative reaction (a) f, actor = a/b g 10 x 11 150,676 lb 677,000 4.27 6 x 11 97,652 lb 405,500 4.153 i An impact factor f of 4.5 can be safely taken to bound the' data for all ! modules. Since there are 13 types of modules in a pool containing 16 modules, these maximal reactions will occur at different instants of i time, and statistical sucznation of these reactions is appropriate. Dynamic amplification of loads due to soil-structure interaction is . determined to be insignificant because the structure is supported on bedrock." 3.4.5 Pool Wall Impacts Analysis by the Licensee'datemined that (4): I q

1 q 4 L l TER-C5506-625 i

                           "Those empty racks which are located on the outer periphery of the pool          !

J to within 4 in of the pool wall may impact the wall under the Hosgri j earthquake. The associated impact loads are smail: the 10 x 11 module j y i (empty) develops a maximum load of 38,5".v Ib with a total duration of 1

1. .001 sec. The wall impact loads for all y m. mate modules may be j conservatively estimated to be 80,000 lb un(- HE conditions, applied as I a line load over the 1e.ngth of the girdle bar facing the wall." j I

In response to discussions about the frequency of wall impacts by the spent fuel rack modules, the Licensee submitted the following (9): , I "The ro ning frequencies of the racks as determined by Cat

are less than 10 Hz. The fun &. mental frequencies of the pool walls are greater than 30 ,

Ez. The two frequencies are far apart, hence the potential interaction i effect between the rack modules and the pool wall is insignificant."

With the loadJ and stresses lumped into the values shown in Tables 6, 7, and 8, evaluation in this review indicated that the response of the wall to l impacts is satisfactory. 3.4.6 Pool Floor Bearing Loads Maximum bearing loads on the spent fuel floor were investigated by the Licensee as follows (4):

                           "The maximum vertical impact load under the support leg of rack modules is 189 kips for the Hosgri load combination. This load is transmitted to foundation slab and bedrock via bearing. Two cases are considered:

(1) 6-1/2 in. diameter adjustable rack foot over leak chase 8 (2) 6-1/2 in diameter adjustable rack foot away from leak chase l The bearing pressure is coeputed by discounting the area of leak chase drains below the rack leg. The maximum bearing pressure computed is 7.5 ksi, which in below the concrete (8.4 ksi) and steel (24 ksi) allowables. However, for case of installation, provisions are being made to provide bearing plates with minimum dimensions of 12.5 in x 12.5 in x 1 in under each foot of the racks. This would further reduce the bearing pressure to 4.5 ksi and increase the factor of safety against the allowable value. The effect of the impact load on the foundation bearing pressure is i detemined to be insignificant because the maximum coeputed contact l pressure underneath the fuel pool is 7.2 ksi, which is significantly I lower than the allowable foundation bearing pressure of 80 ksf used in design of adjacent structures (TSAR Update Sectir. 3.8.4.1.4). I

Joseph Oat Corporation, Camden, NJ, spent fuel rack module vendor.

6 TER-C5506-625 Table 6. Auxiliary Building Spent Puel Pool Concrete Structure Stress Ratios (DE)(a) (Table 2 of Ref. 4)

n. mand vate , canneity vitu i Moment Shear Moment- $ hear Stress llLil(b) L:,1fif,,t M.f t_ k-ft/ft LLf_t_ litig.,

A8 195 31 220 56 1.1 l SC 220 35 245 56 ~ 1.1 AD 225 42 245 56 1.1 CI 145 20 220 56 1.5 ED 180 27 250 65 1.4

                                                                                                           &           215                       27                              260             46        1.2 Easemat     210                       small                           275             1arge     1.3                                                                     ,

I i (a) Stress ratio = capacity / demand for shear or mecent, whichever is ; smaller.

                                                                                                                         *                                                                                                                                                       :{

(b) For wall identification, see Figure 12. l l _ - - _ _ _ _ - _ _ . . - - . - - _ - - - - - - -- _ _ . - - _ _ - - . - _ _ ~ - - _ - - - - _ _ _ _ _ _ _ _ _ - - - . - . - _ - . - - _ _ _ _ _ -- _- ._ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ - _ _ . _ _ - _ _ _ _ - _ .

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yable 7. . A.millary Building Spent Puel Pool Concrete Structure

7 Stress Ratios (DDE)(A) I (Table 3 of Ref. 4) l

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l P,pa nf' g !g n canaeitv Valuet l l Moser f, ' Shear Moment Shaar Stress k U.( > > V111!1 i LLt1 k.11Lt1 Lt.t1 LLL11. l A8' .765 35- 420 85 1.6 K 295 , 42 460 85 1.6

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AD 305 52 4 'O 85 1.5

                                                         ' C!                115                          25           ' 420             85       2.2 ED               235                          30     ,..

470 99 ' 2.0 CF 285

32 870 71 1.7 s ,,

Basemat 265 small 520 targe 2.0

( t (a) Stress ratio = capacity / demand for shear or moment,' whichever is smaller. I ,s

                               ,   (b) For wall identification, see Figure 12.
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n. anne valuet canaH tv valuet

                                                                                                                                .1 kl.1(b)                    b          b            b     b A8            300          45         500            92  1.7 BC            355          55         550            92  1.5
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AD 365 64 550- 92 1.4 CE 200 31 500 92 2.1 ED 245 34 570 107 2.3

                     &             295          37         590          .76   2.0 l                    Sasemat        345          small      630          large 1.8 l

l. c l I l 1 (a) Stress ratio = capacity / demand for shear or morrent, whichever is smaller. .-

(b) For wc11 identification, see Figure 12. I i l l l l

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o TER-C5506-625 The maximum horizontal sliding force on a support leg of rack modules is estimated to be 151 kips. This sliding force is resisted by frir:tional resistance between the steel liner and. concrete floor slab, and by bearing of the stiffener angles embedded in concrete. Two loading cases are considered: ) (1) Sliding force resulting from four legs af interior rack modules clustered together 1 (2) Sliding force resulting from one leg of a module, j The liner plate weld seams are checked for the maximum axial force developed as a res' ult of diference between the rack sliding forces and i the resisting forces. The maximum force computed in the liner weld seam is 1.5 k/in, compared with allowable force of 6 k/in. 3.4.7 Sumary of Concrete Loads and Stresses , I i The Licensee provided e. sumary of computed loads and their comparison to allowable values in Tables 6, 7, and 8 [4). It was noted that Tables 6, 7, and 8 include loads that changed as a result of introducing higher density fuel storage as opposed to Table 5 that sumarized the unchanged TSAR position, i Evaluation in this review indicated that the load values reported are satisfactory. ' l 3.4.8 Pool Dynamic Response Under the Increased Spent Fuel Mass Although the change in mass due to the increased density of fuel storage only increased the r. ass of the spent fuel pool dynamic system by less than 1%, the Licensee choose to demonstrate by the following that the change in response was exceedingly small [9): i "As described in FSAR Section 3.7.2.1.7.1 and in PGandE's previous response to Item 15, the seismic inertia loads for the auxiliary building ! were obtained using a time-history analysis of a spring and lumped mass j model. [ Figure 13) shows the mathematical sodels used in f.he previous ) analysis, j 1 The added weight of the racks and fuel assemblies represents approxi-mately 1% of the global mass of the auxiliary building. In order to evaluate the effect of this added weight, the original matham tical model was revised to include the additional mass consistent with its geometric location in the building. [ Figure 14) illustrates the revised seismic model. The effects of the Hosgri earthquake was evaluated using the Newmark time-histories which govern the building response. The results of the evaluation are as follows:

 .o TER-C5506-625.

i LEGEND: 2 - peODE NUMBER

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N$MCDEL ~ Em MODEL VERTICAL WCDE L l l l I I Figure 14. Auxiliary Building Revised Mathematical Models I To Include High Density Racks-(Figure 12 of Ref. 9) I

l TER-C5506-625 l l o The dynamic characteristics of the auxiliary building, in general, remain unchanged as showr: in [ Tables 9 and 10). o (Table 11) compares the significant shear forces and overturning moments of the revised analysis with the previous analysis. The ; revised analysis results show an increase of approximately 1% over the original forces which is consistent with the increase in the building mass. As indicated in PGandE's previous response to Item 15b, these minor increases are accoranodated by the design l; margins. o (Figures 15 and 16] show the coeparison of typical response ! spectra at the location of the spent fuel pools (elevation 100 l ft). The original RRS was conservatively developed by adding the { transnational and torsional spectra, neglecting the phase i relationship of the two spectra at the location of the pools. l These spectra used the Blume and Newmark enveloped time-histories. I The revised response spectra were developed using two approaches: (1) using the same methodology as described above - this resulted in an identical RRS: and (2) using the time-histories at mass ; point 7 (Figure 15) of the revised model - this result is l enveloped by the originai %S. I Based on the above, it is concluded that the effect of the added weight ! of racks and fuel assemblies is insignificant on the building response." ' 3.4.9 Analvsis of the Spent Fuel pool Liner { Thermal analysis of the liner was performed by the Licensee in accordance i with the analysis model shown in Figure 17 (4). Liner strains and their com-parison to allowable values as performed by the Licensee are shown in Table 12. The effects of sliding forces on the liner were investigated by the l Licensee as follows (4):

              "The maximum horizontal sliding force on a support leg of rack modules is          !

estimated to be 151 kips. This sliding force is resisted by frictional ' resistance between the steel liner and concrete floor slab, and by bearing of the stiffener angles embedded in concrete. Two loading cases are considered: ! 1

1. Sliding force resulting free four legs of interior rack modules clustered together

2. Sliding force resulting from one leg of a module.

The liner plate weld seams are checked for the maximum axial force developed as a result of diference between the rack sliding forces and l the resisting forecs. The maximum force computed in the liner weld seam is 1.5 k/in, compared with allowable force of 6 k/in."

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Ne th.$euth Weri?eetal freut f Orieimal Aa21vtit Revised Analytig Overturning Overturning Element Shar Noment Shear Moment (Kips : 10 3) (Kip-ft 106 ) (Kips s 103 ) (Kip-ft 106 ) 5 3.177 0.1525 3.178 0.1525 1 13.710 0.2949 13.750 0.2957 2 71.930 2.0476 72.220 2.0547 3 115.200 3.7320 115.700 3.7481 4 90.160, 5.11'01 91.090 5.1391 I tant-West He*4 rental f reut Deieimal Araivtit Revited Analysis Overturning Overturning Element Shear Mcment Shear Moment l (Kips : 1,03 ) (Kip-ft 106 ) (Kips : 103) (Kip-ft : 10 6) i 5 2.854 0.1370 2.855 0.1370 1 13.550 0.2913 13.580 0.2919 l 2 78.150 2.1104 76.250 2.1130 i 3 122.500 3.9026 122.700 3.9059 4 80.210 5.1456 80.f10 5.1617 l l l l I 1 l l ! l :

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                                $=Ancsees O!SPL ACOdpff se , - sum <ss a,.snaress or u oi eu.ct 4 a,,.starewss or u cn eusco e' Figure 17. Spent Puel Pool Liner Model for Thermal Analysis (Figure 10 of Ref. 4) l L

 ,:     s IEP-C5506-625 Table 12.        Auxiliary Building Spent Puel Pool Liner Strains.and Displacements (Table 5. of Ref. 4)

LINER PLATE STRAINS Liner Governing teentien tend tee $ M A11ewable(l) tafety Tacte+(1) Mall D+Ta+HE 0.00112 0.005 4.5 Floor 0+Ta+HE 0.00156 0.005 3.2 ! LINER ANO40R DISPLACEMENT 5 Liner Governing Locatien tend camb gg A11ewable(3) tafety Faetee(2) Mall D+Ta+ME 0.01220 0.06200 5.2 l Floor D.Ta+ME 0.01375 0.06200 4.5 (1) From ASME SLPV Code Section III. Division 2,1983 Edition. Table CC-3720-1. (2) Safety factor is computed relative to the code allowable. (3) From ASME't&PV Code Section III. Division 2.1983 Edition. Table CC-3730-1 using Su . 0.125 in. based on test results given in SC-TOP-1 s Appendis B (Ref. A .

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l TER-C5506-625 When the possibility of an air gap forming.between the liner and the-concrete pool floor was questioned, the Licensee' investigated its effect on the liner with rack sliding forces. Th. Licensee's. response follows (9):

               "The assumption of an airspace underneath the'11ner was' evaluated and found to have no impact on PGandE's previous response to Item 15f.(of
              'Ref. 4]. The floor' liner plate is 1/4 inch thick and is welded to-embedded angles oriented in east-west and north-south directions as shown in (Figure 18). The fuel pool was constructed using standard construction -

practices for mass concrete construction which included:- I o control of pour rates to minimise,the effect of plastic shrinkage o controlled curing to minimize the effect of drying shrinkage o ample reinfo coment to minimize shrinkage o a surface finish having a uniform texture that does not contain voids and is free of projecting aggregrate, high spots, and depressions. Based on the above construction features, shrinkage of the concrete' slab is expected to be minimal. However, it'is conservatively assumed that the average shrinkage is 700 x 10-6 in/in for the entire depth (5 feet thick) of ths' slab. This results is a postulated air gap of 0.0(2. inches. The simplified model shown'in (Figure 19]- was used to determine the deformation of the liner under a hydrostatic head of 38 feet of water. The load that would cause' contact of the liner with the concrete was-determined to be 4.6 psi which is small compared with'the hydrostatic load of 16.5 psi. The results, therefore, indicated that the liner would be in contact with the concrete under a hydrostatic head alone. To check the liner contact under thermal effects, the same mathematical model as described in PGandE's previous response to item 15b was used. -l The analysis indicated that in addition to hydrostatic load, a normal { load of less than 0.5 kip per rack leg.is needed to achieve full liner ' contact with the concrete under the rack leg. .This additional normal load is insignificant compared with the leg reaction load of 189 kips. Therefore, under thermal conditionsi the: liner will remain in contact I with the concrete under the rack leg. The liner strains were also determined and found to be within the allowable value. Under hydrostatic loads, the liner strain resulting. , from the liner contacting the concrete due to a postulated air-space is -! 0.00006. The combined strain resulting from the governing' load 1 combination, D + Ta + HE, is 0.00156 compared with an allowable value j of 0.005 (previously reported in PGandE's response to Item 15 (of Ref. I 4]). This provides a factor of safety of 3.2. i 1 64-4 '

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Figure 18. Pool Floor Liner Details (Figure 15 of Ref. 9) _ l. _ _ _ _ _ _ . _ _ _ _ ._.._______m_m.___m_-.

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TER-C5506-625 Based on the above, it is concluded that the liner will remain in contact with the concrete to develop adequate frictional resistance." i i I Evalustion of the Licensee's liner analysis indicated that the analysis is acceptable. l 3.4.10 Spent Fuel Pool Leak Chase System In response to discussions concerning the leak chase system of the spent fuel pool, the Licensee provided the following response (9):

                 "The spent fuel storage pool for each unit is a reinforced concrete structure with a seam-welded stainless steel plate liner. Monitoring             l trenches approximately 1 inch square are located behind the liner on each side of all seam welds connecting adjacent liner plates in both the walls and floor of the pool. These trenches lead to one of six collection pipes which are valved and terminated at a conr.cn sump. Leakage past th'e liner would be detected by opening these leak detection shutoff valves and observing any water accumulation or flow. These valves are normally.

closed. Prior to placing irradiated fuel in the spent fuel pool, Surveillance f Test Procedure (STP) I-1C, " Routine Weekly Checks" will be. revised. STP I-1C will require the operators to open each leak detection shutoff valve on a weekly basis and determine if any water has accumulated, i Plant Technical Specification 3/4.9.11 requires that at least 23 feet of water be maintained over the top of irradiated fuel assemblies seated in the storage racks. One assurance of this is provided by a low water i level alarm which annunciated in the conctrol room when the pool level ! goes below 137'-8". If and when a low level alarm occurs, operators ' would refer to Volume 16, "Annuciator Response" cf the plant manual. Volume 16 would direct the operators to open each leak detection shutoff valve and determine if any water had leaked through the fuel pool liner. If leakage past the liner is detected, additional testing would be performed and the leak would be repaired as appropriate. There are no l safety consequences to this since leakage past the liner would be l terminated by closing the leak detection shutoff valves, and adequate ! makeup capability exists which exceeds any credible leakage through the i liner (discussed in PGandE letter DCL-86-020 dated January 28, 1986)." 1 l l No detrimental effect of 1.he higher density spent fuel loading was noted for the leak chase system. I 1

. i a O TER-C5506-625 3.4.11 Cask Pit and Cask Restraint The spent fuel pools of Units 1 and 2 each include a cask pit area in a corner of the pool; the cask pit is recessed below the floor of the spent fuel pool. Because spent fuel rack sodules E, F, and J are adjacent to the cask pit, the Licensee is providing a cask restraint, in accordance with the FSAR (3), that will prevent the cask froa impacting with the adjacent spent fuel racks during cask operations [16). During seismic events, a cask in the cask pit will be restrained from lateral movement toward the fuel racks by the walls of the recessed cask pit. Via Reference 17, the Licensee provided a sk' etch of the cask restraint (Figure 20) plus an installation, plan (Figure 21) showing the relationship of the cask pit, the cask restraint, and the adjacent fuel rack modules. Reference 17 confirms that the Licenses is mounting the cask restraint so that the centerline of the horizontal macter of the restraint is at the elevation of the impact girdle bars on the rack modules. Thus, should the fuel rack modules impact the cask restraint during a sesimic event, the impacts will take place on the girdle bars of the rack modules that are provided for that purpose, as discussed earlier in this report. 3.5 FUEL HANDLING ACCIDDIT ANALYSIS In the Licensee's approach to fuel handling accident analysis, the following assu=ptions were used (1):

              "1. The virtual mass of the body is conservatively assumed to be equal to its displaced fluid mass. Evidence in the literature (Ref. 12),

indicates that the virtual mass can be many times higher.

2. The minimum frontal area is used for evaluating the drag coefficient.

3. The drag coefficients utilized in the analysis are the lower bound values reported in Lhe literature (Ref.13). In particular, at the beginning of the fall when the' velocity of the body is small, the corresponding Reynolds number is low, resulting in a large drag tne Hir.ient.

4. The falling bodies are assumed to be rigid for the purposes of impact stress calculation on the rack. The solution of the immersed body motion problem is found analytically. The impact velocity thus coeputed is used to determine the maximum stress generated due to stress wave propagation."

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   ,   o TER-C5506-625 f'

Using these assumptions, the Licensee provided the following with respect to handling accident considerations (1):

                    " Dropped Fuel Accident I A fuel assembly (weight = 1616 pounds with control rod assembly) is dropped from 36 inches above the module and impacts the base. The final velocity of the dropped fuel assembly (just prior to impact) is calculatsd and, thus, the total energy at impact is known. To study baseplate integrity, we assume that this energy is all directed toward punching of the baseplate in shear and thus transformed into work done by the supporting she'ar stresses. It is determined that. shearing deformation of the baseplate is less than the thickness of the baseplate so that.we conclude that loca1' piercing of the baseplate will not occur.

Direct impact with the pool liner does not occur. The suberiticality of the adjacent fuel assemblies is not violated. - Dropped Fuel Accident II One fuel assembly drops from 36 inches above the rack and hits the top of j the rack. permanent deformation of the rack is found to be limited to j the top region such that the rack cross-sectional geometry at the level j of the top of the active fuel (and below) is not altered. The region of j local permanent deformation does not extend below 6 inches from the rack l top. An energy balance approach is used here to obtain the results. Jamed Fuel-Handling Equioment and Horizontal Force A 4400-pound uplift force and an 1100-pound horizontal force are applied at the top of the rack at the ' weakest' storage location; the force is l assumed to be applied on one wall of the storage cell boundary as an' l upward shear force. The plastic deformation is found to be limited to the region well above the top of the active fuel." . For Drop Fuel Accident I, a possible mode of rack module deformation was not addressed by the Licensee. In this mode, the kinetic energy of the falling fuel assembly is dissipated in the strain energy associated with the ! shearing of welds between the base plate and the storage cell walls and, as a minimum, would shear the welds of all four walls of the imediate cell. In i the extreme, the worst that could happen would be the deflection of the base plate to che floor. Since the fuel assembly does not extend through the base plate, and since there are no protrusions on the bottom of the base plate, there would be no damage to the pool liner should the base plate touch the liner. 4 Review of the fuel handling accident analyses indicates that they are l acceptable.

g TER-C5506-625

4. CONCLUSIONS Based upon the review end evaluation, the following conclusions were reached:

o The Licensee used three-dimensional, nonlinear dynamic displacement analyses with three simultaneous, independent, orthogonal, earthquake acceleration time histories to provide greater resolution of the rack module displacements and impacts than is possible with two-dimen-sional analyses combined by the sguare root of the sum of the squares method. o Seismic loadings included the simultaneous application of one vertical and two horizontal acceleration time histories derived from the design, double design, and Hosgri earthquake excitations specified by the updated Final Safety Analysis Report (FSAR). o Impact analysis of fuel assemblies within storage cells as well as rack-to-rack and rack-to-wall impacts were integrated into the Licensee's dynamic displacement analysis when design studies indicated that impacts would take place under the specified earthquake excitations. o Fuel assembly impact forces within a rack storage cell were over- J estimated by use of impact springs based upon the' storage cell wall 1 compliance, neglecting the greater compliance of the fuel assembly. However, the resulting stresses were within allowable values.

o Two rack modu 1es were selected by the Licensee as representative'of bounding the range of 13 design configurations used in each spent fuel pool of Units 1 and 2. These involved the 10 x 11 cell rack module as the largest and heaviest loaded module, and the 6 x 11 cell design as the module with the largest ratio of horizontal dimensions l and, therefore, most likely to undergo the most displacement in the tipping mode. While dynamic displacement analysis was not perfomed for the 6 x 9 cell design, its displacement response would be erpected to be somewhat greater than the 6 x 11 cell design, but not enough greater to cause any of the design criteria to be exceeded, given the conservative results presented by the Licensee. o An 8 degree-of-freedom mathematical model was used by the Licensee for dynamic displacement analysis af ter it was validated by compri-son of results with that of a 32 degree-of-freedom model. The 8 degree-of-freedom model was used to accomodate the added nonlinear l complexity associated with the increased number of gap elements required for rack-to-rack impact. o Rack module stresses are within the allowable values. o Rack module stability in the tipping mode under a Hosgri earthquake indicated a wide margin of safety in tipping. \ ' 4

y

                                                                                              ]  J

0. TER-C5506-625 o

o A cask restet. int structure was provided'to protect the fuel' rack- . modules from impact by the spent fuel cask during cask operations. Lateral movement'of a cask in the cask pit toward the fuel rack modules during a seismic event is restrained by the walls of the recessed cask pit. In summary, it was concluded that the Licensee's structural analyses of-the spent fuel rack modules and the spent fuel pool are adequate and are. acceptable: they indicate the rack modules and pool structure.to be , satisfactory for high density fuel storage. ] l l e i

l 1 i I l

. =

TER-C5506-625 , 9

5. REFERENCES

1. Pacific Cas and Electric Company, Licensing Report on High-Density Spent Fuel Racks, "Raracking of Spent Fuel Pools, Diablo Canyon Units 1 and 2," i September 1985 l

2. U.S. Nuclear Regulatory Comission j CTI Position for Review and Acceptance et Spent Fuel Storage and Handling !

Applications January 18, 1979

3. Pacific Gas and Electric Company, Updated Final Safety Analysis Report ;

(FSAR) for Diablo Canyon Units 1 and 2 j

4. Pacific Gas and Electric Company, Letter No. DCL-86-019 to U.S. Nuclear Regulatory Cocaission, January 28, 1986

5. E. Rabinowicz

          " Friction Coefficient Value for a High-Density Fuel Storage System" Report to General Electric Nuclear Energy Programs Division I

November 23, 1977

6. R. J. Fritz "The Effect of Liquids on the Dynamic Motions of Imersed Solids,"

Journal of Engineering for Industry, pp. 167-173, February 1972 ,

7. Franklin Research Center, Evaluation of Spent Fuel Racks Structural l Analysis, Oyster Creek Nuclear Generating Station, Technical Evaluation Report TER-C5506-125, July 5,1984

8. Franklin Research Center, Evaluation of Spent Fuel Racks Structural

         . Analysis, Virgil C. Sumer Nuclear Station, Technical Evaluation Report                                      i TER-C5506-526, August 15, 1984

9. Pacific Gas and EAectric. Company, Letter No. DCL-86-067 to U.S. Nuclear Regulatory Comission, March 11, 1986

10. U.S. Nuclear Regulatory Commission, Standard Review Plan, Section 3.8.4, Appendix D, NUREG-0800, Revised July 1981

11. U.S. Department of Coccerce (NTIS), " Nuclear Reactors and Earthquakes,"

TID 2025

12. A. S. Veletsos and J. Y. Yang, " Dynamics of Fixed-Base Liquid-Storage Tanks," Proceedings of U.S.-Japan Seminar on Earthquake Engineering Research with Emphasis on Lifeline Systems, November 1976

13. Letter (Docket Nos. 50-275 and 50-323) from A. Giambusso of the U.S.

Atomic Energy Commission to F. T. Searls of the Pacific Gas and Electric Company, dated August 23, 1973, including enclosurer 4 74-

                                                                                   .____.______-_______m.___.___J

              ,                                  a                                                                                                                                                                        l TER-C5506-625

14. " Recommended Evaluation Criteria for Diablo Canyon Nuclear Power Plant l Auxiliary Building Walls and Diaphragms," Jack R. Benjamin & Associates, Inc., February 11, 1933 4 l

15. " Containment Building Liner Plate Design Report, BC-TOP-1," Bechtel Power Corporation, San Francisco, California

16. ' Pacific Gas and Electric Ccapany,. Letter DCL-86-109 to U.S. Nuclear Regulatory Commission, April 24, 1986

17. Pacific Gas and Electric Company, Letter DCL-86-108 to U.S. Nuclear Regulatory Coasnission, April 24, 1986 l

I- l l l I I . l l ll l a

                                                                                                                                                                                                                        'I I

i I , I i l , l

1

                                     '\
 ,                         n     .

Y' APPENDIX A

                                                                                                   -hel I,ICENSEE RESPONSE, PER REFERENCE 9, FOR BACKG700ND AND                                        l VERIFICATION OF DISPLACEMENT ANALYSIS METHOD INNAHIS                                           l

I l ;'

I i N I ll ! s s s i N 3 l l I i N I ; FRANKLIN RESEARCH CENTER cxvison or ArvmeAtspau : 20 n a aAca srassrs.rwitAostesia.pa isies I 3 l

                                        <                                                 \
                                                                   .                   .o w_

R PGandE Le%%er No.: DCL-86-067 i s cl .

-                                                                 ENCLOSURE
                                                                                                                 \

i i PGandE RESPONSES TO NRC REQ' JEST FOR ADDITIONAL INFORMATION i l Letter Item 26 Hith respect to the computer code, DYNAHIS, discuss specific efforts performed j to verify that the code provides realistic seismic responses of the spent fuel i racks subject to a set of complex loading conditions and geometric ' constraints. For example, indicate if any experimental data were used as the basis for demonstrating the validity of the code. Provide a discussion of the assumptions and limitations which are unique to the code. Meetine Item 26a ' Regarding I c126 on the validation of the computer coh DYNAHIS, the specific benchmark problems should be identified and appropriately referenced. The lv results, including response comparisons, of problems run with the DYNAHIS code .) and the ANSYS code should be provided. Discuss the extent of use of the i DYNAHIS code in raracking applications by other utilities. Reseense 26 and 26a , The algorithm of the code DYNAHIS is documented in the book 'The Component l Element Method in Dynamics" by Levy and Hilkinson, McGraw Hill (1976). The I application of the code to a vice variety of linear and nonlinear problems is i illustrated in that teit. The Joseph Oat Corporation (Oat) version of this a code contains the added attribute of diagonalizing the mass matrix. However, l the code has been tested thoroughly at Oat's computer center. The i

            '               benchmarking effort'consistsd of running several nonlinear problems for which        '

classical solutions exist.' A dynamic problem involving a nondiagonal mass l matrix was also constructed and analytically solved. DYNAHIS was run to verify its accuracy versus the analytical solution. Its accuracy has been i further confirmed by comparing results of sample problems run on DYNAHIS and l on general purpose codes such as ANSYS. In fact, CYNAHIS has received a level I. of scrutiny which rivals that given to any general-purpose finite element code used for nonlinear analysis of dynamic systems, including reviews of reracking applications for Fermi II, Quad Cities I and II, Rancho Seco, Grand Gulf Unit 1, Oyster Creek, Pilgrf s, and V.C. Summer. Since no pertinent data on the dynamics of rack structures exist, benchmarking against experiments is not possible. Instead, Dat has taken the proven , recourse of using a mathematical model that is highly conservative in all respects, e.g., amount of damping and the mass lumping scheme. A detailed description of'the calculation method of DYNAHIS is presented in Section 6 of the reracking report. There are no numerical simulations or modeling assumptions unique to DYNAHIS. Moreover, in contrast to a general-purpose finite element code, DYNAHIS does not internally generate the equations of rotion. These equations are written out explicitly and can be

              ,            verified independently. Such verifications have been carried out internally tt Oat, by several A/E firms, and, at the NRC's request, by Franklin Research I                           Center in 1984.

07195/0042K

 .        y          -       .    .                           +..
                                                                                      .                  'I.

As discussed in meetings with the Staff..one DYNAHIS verification was conducted as part of Cat's work for Detroit Edison Company.on the Enrico Fermi Unit II raracking application. - As part of this DYNAHIS validation. several problems that would test the analytical features of DYNAHIS'related to hydrodynamic. response of rack modules were run on DYNAHIS by Oat and on'ANSYS. by Detroit Edison's architect /en '

             -combinations of masses, springs,:gineer.

dampers, and These problems gap elements. consisted of Comparisons of-various- l the DYNAHIS and ANSYS response predictions ~ confirmed that the results from the i two computer codes were in excellent agreement. An additional, more complex

            - problem, a 32-DOF model of a rack module, was also analyzed on ANSYS by the architect / engineer. These results were also compared with 04t's DYNAHIS response predictions and the comparisons were in good agreement. These.

evaluations were not formally documented or submitted on the Fermi II docket. The following is a description of some specific verification runs that have

been made. These benchmark problems used to verify DYNAHIS solutions fall l into three classes: (1) comparison with classical closed form itnear solutions.which exercise the features of DYNAHIS (e.g., decoupling of mass matrix and input force history from external files); (2) nonlinear problems which utilize the gap elements and friction elements and compare results with ,

textbook numerical analyses; and (3) problems which compare DYNAHIS resu?ts to-corresponding results obtained from general-purpose finite element codes. j Problem l f i 2 Degree of Freedom (DOF) 1.inear System with Ma'ss Coupling 2mxi + mI 2 + KIl

I sin at l

ax1 + 2a52 + KX2

0 The table below shows results obtained for x'(t) and x2(t) from the analytically derived solution and from DYNAHMS for the particular values of a = 750 lb-seed /in., K = 106 lb/in., c = 6.28 rad /sec, and I = 386 lb.

10WTf0N COMPARISON x1 x 10+3 x2 x 10+3 i Time Analytical (DYMAHIS) Analytical- (DYNAHIS) (sec) Solution Numerical Solution Numerical '

                  .1                .204               .208              .063              .064 2.1                .188               .189              .022              .021 3.6                .252               .256              .054              .054 4.1                .330               .332              .040              .041 5.7                .457               .460              .033              .033 8.7                .433               .433-             .084              .088 10.8                 .423               .423              .109.             .109 11.2                 .398               .398              .061              .063 13.3                 .407               .408              .090              .091 14.4                 .230               .229              .081              .081 As can be seen, there is good agreement between the analytical solution and the DYNAHIS results.

07195/0042K - s e

f, # r Preblem 2 Nass Dropped on a Beam ("The Component Element Method in Dynamics," 5. Levy, McGraw Hill, 1976, pp.,151-155). See Figure 1 for-details of the medel. One half of the beam 9s cedeled 'Jy a 7 degree of freedom system (including rotary inertia in the1beun, and one DOF for the dropped mass). . The DYNAHIS results are plotted and co'npared with the text results. The contact force time-history maximum values are predicted with good agreement. The additional

                     .I.higher frequency oscillations in the gap element forces are due to the fact that DYNAHIS includes the effect of higher frequency rotary inertia degrees of

! : freedom. The plots in Figure 2 show the textbook results, and Figures 3 and 4 show the DYNAHIS result.s. The agreement is found to be excellent. Problem 3 Dropped Hass with Friction Elements (From Levy text, pp. 55-56). As shown in Figures 5 and 6 the DYNAHIS output plot is in good agreement with the curve (F = 103 lb) from the text. This prob 1cm demonstrates that the DYNAHIS code correctly accounts for friction effects. Problem 4 Comparison of ANSYS with DYAAHIS. The comparison of results was carried out using ANSYS example problem E6.2 which is a pendulum system as shown in Figure 7. The same numerical values

 -                      for masses, springs, gaps, etc. were used in the DYNAHIS model. Since the time-history algorithms are different in the two codes, the excellent l               agreement obtained verified the suitability of the integration algorithm in DYNAHIS.      Figures 8:and 10 are reproductions from the ANSYS examples manual showing the mass displacements and contact forces. Figure 9 shows the DYNAHIS l               mass displacements. The DYMAHIS results for contact forces at selected time 1                 intervals are superimposed on the ANSYS output in Figure 10. The two analyses produce results that are in close agreement.

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         ,,  u 7590-01 UNITED STATES NUCLEAR REGULATORY COMMISSION PACIFIC GAS AND ELECTRIC COMPANY DOCKET NOS. 50-275 AND 50-323 NOTICE OF ISSUANCE OF AMENDMENTS TO FACILITY OPERATING LICENSES AND FINAL DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATION The U.S. Nuclear Regulatory Comission (the Comission) has issued Amendment No. 8 to Facility Operating License Nos. DPR-80, and Amendment No. 6 to Facility Operating License No. DPR-82, issued to Pacific Gas and Electric Company (the licensee), which revise the combined Technical Specifications for the' Diab 1'o Canyon Nuclear Power Plant, Unit Nos. I and 2 (the facilities) located in San Luis Obispo County, California.

The amendments permit the expansion of the spent fuel storage capacity for the Diablo Canyon Nuclear Power Plant, Units 1 and 2, increasing the storage capacities from 270 to 1324 spaces for each of the two spent fuel pools. This expansion would be accomplished by reracking the existing spent fuel storage pools with new spent fuel racks composed of individual cells made of stainless steel. The new fuel storage racks will be arranged in two discrete regions within each pool. Region 1 will provide 290 poisoned cell locations which will normally be used for new fuel with an enrichment equal to or less than 4.5% U-235. Region 2 will provide 1034 unpoisoned cell locations which will provide normal storage for spent fuel assemblies with an initial enrichment equal to or less than 4.5% U 235 and meeting required burnup considerations in accordance I with the amended Technical Specifications. The existing fuel storage racks

2-

I have a nominal center-to-center spacing of 21 inch. The new racks will have l a nominal 11 inch center-to-center spacing for both regions. The major components of the fuel racks are the fuel assembly cells, gap channels, rack base plate assembly including support legs, a girdle bar around the upper part of the rack assembly, and, in Region 1 only, neutron absorbing (poison) material Boraflex, including cover sheets. The fuel racks are designed to maintain the required I subcriticality of K equal to or less than 0.95 for Regions 1 and 2. 3 eff The letter requesting the license amendments, dated October 30, 1985 (LAR-85-13), includes the requested Technical Specification changes and the licensee's determination on significant hazards consideration. The supporting report on "Reracking of Spent Fuel Pools for Diablo Canyon Units 1 and 2" had been submitted to the staff by letter, dated September 19, 1985. The application for these amendments complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations. The Comission has made appropriate findings as required by the Act and the Commission's rules and regulations in 10 CFR Chapter I, which are set forth in these license amendments. Notice of Consideration of Issuance of Amendments and Proposed No Signifi-

 ,       cant Hazards Consideration Determination and Opportunity for Hearing in connection with this action was initially published in the FEDERAL REGISTER on January 13, 1906, (51 FR 1451) and in the bi-weekly publication on May 21, l          1986, (51 FR 18699). A request for hearing was filed by (1) Mothers for Peace on February 7,1986,(2) Sierra Club-Santa Lucia Chapter on February 10, 1986, and (3) Consumers Organized for Defense of Environmental Safety (C.O.D.E.S.)         ,

l on February 12, 1986. i 4

   . , , s Under its regulations, the Comission may issue and make an amendment imediately effective, notwithstanding the pendency before it of a request for a hearing from persons, in advance of the holding and completion of any required hearing, where it has det armined that no significant hazards consideration is involved.

The Comission has applied the standards of 10 CFR 50.92 and has made a final determination that these amendments involve no significant hazards consideration. The basis for this determination is contained in the Safety Evaluation related to this action. Accordingly~, as described above, these amendments have been issued and made imediately effective and any hearing will be held after issuance. } A separate Environmental Assessment has been prepared pursuant to 10 CFR Part 51. The Notice of Issuance of Environmental Assessment and Finding of No Significant Impact was published in the FEDERAL REGISTER (51 FR 19430) on May 29, 1986. For further details with respect to the action see (1) the application for the amendments dated October 30, 1985 and additional infonnation provided by

the licensee in letters dated September 19, December 20, cnd December 24, 1985 and January 28 (2 letters), March 11 and April 24 (2 letters),1986, (2) Amendment Nos. 8 and 6 to Facility Operating License Nos. DPR-80 and DPR-82, (3) the Comission's related Safety Evaluation and (4) Environmental Assessment and Notice of Issuance of Environmental Assessment and Finding of No Significant Impact. All of these items are available for public inspection at the Commission's Public Document Room,1717 H Street, N.W. , Washington, D.C. ,

and at California Polytechnic State University Library, Documents and Maps

            ?h ~                                                                                                           l l

_4-1 Department, San Luis Obispo, California 93407. A copy of items (2), (3) and (4) may be obtained upon request addressed to tne U.S. Nuclear Regulatory Commission, Washington, D.C. 20555, Attention: Director, Division of PWR Licensing-A. Dated at Bethesda, Maryland, this 30th day of May 1986. FOR THE NUCLEAR REGULATORY COMMISSION cla, e% PWR Project Directo a e No. 3 Division of PWR Lice ing-A, NRR ! I i i ! l I l l t l

 '_ - - _ - _ _ _ . . _ . - _ . _ _ _ _ _ _ _ - - -}}

ML20237K380

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