ML20216J826
| ML20216J826 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon  |
| Issue date: | 06/29/1987 |
| From: | PACIFIC GAS & ELECTRIC CO. |
| To: | |
| References | |
| CON-#387-3936 OLA, NUDOCS 8707070042 | |
| Download: ML20216J826 (45) | |
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UNITED STATES OF AMERIC/L NUCLEAR REGULATORY COMMISSd JJL -2 P4 :15 2
BEFORE THE ATOMIC SAFETY AND LICENSING BOARD 3
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Docket Nos. 50-275 OLA In the Matter of
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50-323 5
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PACIFIC GAS AND ELECTRIC COMPANY )
(Reracking of Spent Fuel Pools) 6
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7 (Diablo Canyon Nuclear Power
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Plant Units 1 and 2)
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June 29, 1987 8
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REVISED PROPOSED FINDINGS OF FACT AND jg CONCLUSIONS OF LAN SUBMITTED BY PACIFIC GAS AND ELECTRIC COMPANY RELATING TO AN AMENDMENT REQUEST TO RERACK jj THE SPENT FUEL POOLS AT THE DIABLO CANYON NUCLEAR POWER PLANT 12 13 I.
INTRODUCTION 14 By Memorandum and Order dated April 9, 1987, the Atomic Safety and 15 Licensing Board (ASLB) ordered that the parties file with the Board and serve 16 on each other proposed findings of fact and conclusions of law on or before 17 June 5, 1987. The Board also stated that the parties will be given an 18 opportunity to revise the proposed findings and conclusions after the 19 hearing. Hearings were held from June 15 to June 18, 1987, at the San Luis 20, Bay Inn, Avila Beach, California. At the conclusion of the hearing, the Board 21 required that the Intervenors and Pacific Gas and Electric Company (PGandE) 22 file revised proposed findings and conclusions of law within ten days of the 23 end of the hearing. The Board also required that the NRC Staff file revised 24 proposed findings and conclusions within 20 days of the end of the hearing.
25 In accordance with the ASLB order, PGandE and the Staff filed draft 26 Proposed Findings of Fact and Conclusions of Law dated June 4, 1987. The e707070042 B70629 DR ADOCK 050 5
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Intervenors, Sierra Club, filed draft Proposed Findings of Fact and 2
Conclusions of Law dated June 3, 1987. PGandE hereby files Revised Proposed 3
Findings of Fact and Conclusions of Law.
4 While there were a number of contentions admitted by the Board for 5
hearing, several were deleted by prior stipulation or withdrawal of the 6
petitioning parties. The contentions that were litigated at the hearing were 7
raised by the Sierra Club and relate primarily to the structural integrity of 8
the high density racks when subjected to the ground motions associated with 9
the postulated Hosgri earthquake. One contention, Sierra Club Contention 10 I(B)9, was withdrawn by the Sierra Club at the beginning of the hearing. The 11 specific contentions which were litigated are set forth in Appendix A attached 12 hereto.
13 Both the NRC Staff and PGandE offered testimony and sponsored 14 witnesses for each of the contentions.
To address the Sierra Club Conten'tions 15 I(A)3 and 4; I(B)2, 8, and 9; II(A)l-9 and II(B), the Staff presented the 16 testimony of Hansraj G. Ashar, a Structural Engineer in the Office of Nuclear 17 Reactor Regulation (NRR), NRC; Howard Martin Fishman, Section Head, Structural 18 Engineering Section, Engineering Department, Franklin Research Center (FRC);
19 Dr. Walter L. Brooks, a Nuclear Engineer in the Reactor Systems Branch, NRR, 20 NRC; and Amarjit Singh, a Reactor Operations Engineer, Inspection, Licensing 21 and Research Integration Branch, Program Management, Policy Development and 22 Analysis Staff, NRR, NRC. To address the limited area of multi-rack impacts, 23 the Staff presented the testimony of Mr. Giuliano DeGrassi, a Research 24 Engineer in the Structural Analysis Division, Department of Nuclear Energy, 25 Brookhaven National Laboratory. Messrs. Ashar, Fishman, Brooks, Singh, and 26 DeGrassi were presented by the Staff as a single panel of witnesses, i
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To address the Sierra Club Contention I(B)7 concerning alternativei 2
to the proposed amendment, the Staff presented the testimony of 3
Donald P. Cleary, Senior Task Manager, Reactor and Plant Safety Issues Branch, 4
Division of Reactor and Plant Systems Office of Nuclear Regulatory Research, 5
NRC.
6 PGandE presented a single panel which addressed all of the Sierra 7
Club's contentions. PGandE's witnesses consisted of J. D. Shiffer, Vice 8
President of Nuclear Power Generation at PGandE; K. P. Singh, President of g
Holtec International; S. Bhattacharya, Senior Civil Engineer at PGandE; 10 S. E. Turner, Consultant and Project Manager at Black and Veatch; 11 E. E. DeMario, Advisory Engineer at Westinghouse Electric Corporation; and H.
12 H. White, Chief Civil / Structural Engineer at Bechtel Hestern Power Corporation.
13 Mr. Shiffer has been involved with nuclear plant operations for more 14 than 25 years and sponsored the testimony on Contention I(B)7 related to the 15 evaluation of alternatives. The remaining panel members sponsored those 16 portions of the testimony regarding rack adequacy and structural and safety 17 issues.
K. Singh and S. Turner have been involved with the design and 18 evaluation of spent fuel racks at several nuclear facilities.
S. Bhattacharya lg and H. White have had extensive experience with the structural design and 20 analyses of nuclear power plants.
E. DeMario has been involved with fuel 21 assembly designs for various Westinghouse fuel concepts.
22 Both the Staff and PGandE witnesses are highly qualified in their 23 respective fields of endeavor and their qualifications were not challenged by 24 the Intervenors during the hearing.
25 The sole sponsoring witness for the Sierra Club was 26 Dr. Richard B. Ferguson, who offered testimony on each of the contentions.,
h 1
However, Dr. Ferguson admitted that he is not an expert in the following 2
technical areas related to the design and analysis of spent fuel racks:
3 nuclear engineering; nuclear systems; nuclear criticality; seismic design; and 4
federal laws, codes, and regulations. He further stated that he has no 5
academic training (i.e., he has never taken an undergraduate or graduate 6
course) in the following areas: nuclear engineering; nuclear systems; finite 7
element analysis; and spent fuel storage technologies.
8 Dr. Ferguson's testimony and his professional qualifications clearly 9
indicate that, other than his recent involvement with the proposed reracking 10' at Diablo Canyon, he has little or no experience with any of the technical 11 subjects at issue in this proceeding. Ferguson, Tr. 424-426 and 431-435.
12 l Accordingly, his testimony should be accorded little or no weight.
i 13 Each party introduced a number of exhibits to be marked for 14 identification at the hearing. By stipulation of all parties, all' except four 15!
of the exhibits were consolidated and admitted into evidence. The four 16!
exhibits which were not admitted as evidence were introduced by the Sierra 17l Club and marked for identification only:
Exhibit No.1, NRC Board I
l 18l Notification 87-05 dated March 27, 1987; Exhibit No. 2, NRC letter to PGandE 19 dated June 2,1987, regarding Boraflex deficiencies at Quad Cities; Exhibit 20l No. 3, a letter to NRC dated February 11, 1987, from Hisconsin Electric Power 21 Company regarding Boraflex data at Point Beach; Exhibit No. 4. Affidavit of R.
22 l C. Herrick of Franklin Research Center, dated June 30, 1986. A list of the 23 exhibits introduced by each party is contained in Appendix B attached hereto.
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1 II. PROPOSED FINDINGS OF FACT 2
A.
Desian and Analysis of Hiah Density Racks 3
1 Diablo Canyon Units 1 and 2 have separate spent fuel handling and 4
storage facilities. Each unit has a spent fuel pool with storage capacity for 5
270 spent fuel assemblies.
Each pool currently contains spent fuel from the 6
first refueling outage, which occurred in late 1986 for Unit I and mid-1987 7
for Unit 2.
Based upon operating schedules and the desirpility of 8
maintaining full core discharge capability, it is necessary that the spent 9
fuel storage capacity for both units be increased. Shiffer, et al, ff. Tr.
10 179 at 8-9.
11 2
Reracking with high density racks was chosen because it is a safe 12 method of increasing onsite storage. Further, high density reracking is the 13 most prudent, reasonable, economical, and timely method of the various 14 techniques available to provide increased storage. Shiffer, et al, ff. Tr.
15 179 at 9-10.
16 3
The high density spent fuel racks to be installed in the Diablo 17 Canyon fuel pools consist of a total of 16 racks of varying sizes for each 18 pool, with a total of 1324 fuel assembly storage cells plus 10 miscellaneous 19 storage locations. The number of cells range from 24 to 110 per rack and the 20 individual storage cells have an 8.85-inch (nominal) square cross section.
21 Each cell is sized to contain and protect a. single Westinghouse-type PHR 17x17 22 fuel assembly and the cells are arranged with a 10.93-inch center-to-center 23 spacing in the rack modules. Stainless steel gap channels are welded between 24 the cells to provide a " honeycomb" type structure which provides considerable 25 rigidity and resistance to impact as well as to seismic loads. Shiffer, et 26 al, ff. Tr.179 at 12-13. -
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1 4.
The rack modules are freestanding, with no connection to the pool 2
floor, walls, or adjacent rack modules. The rack support feet rest on bearing 3
plates on the pool floor. Each module is equipped with a girdle bar on each 4
side near the top. The girdle bars serve as a designated impact location and 5
are designed to accommodate impact loads.which may occur during a seismic i
6 event, and they maintain a specified minimum gap between the cell walls of 7
adjacent rack modules for all loading conditions. Shiffer, et al, ff. Tr.179 8
at 11.
9 5
The rack modules are specifically designed for storage of spent fuel 10 with varying amounts of burnup. There are three modules (290 cells) 11 '
designated as Region 1; these utilize a neutron-absorbing material, Boraflex, 12 on all four sides of the individual storage cells in the rack modtale. These 13l cells are designed for storage of new fuel assemblies with enrichments up to 14' 4.5 weight percent U-235 and spent fuel that has not achieved a specified 15 minimum burnup. There are 13 modules (1034 cells) cesignated as Region 2; 16 spent fuel stored in this region will be required (by Technical 17 Specifications) to have a specified minimum burnup and, thus, no Boraflex is i
18 used in this region. Shiffer, et al, ff. Tr. 179 at 13-14.
19 5
The spent fuel pools at Diablo Canyon are located at each end of the 20, east side of the auxiliary building. Each pool is approximately 35 feet wide, 21 37 feet long, and 40 feet deep. The normal water level in the pool provides a 22 minimum of 23 feet of water above the top of the stored fuel. The concrete 23 pool valls are 6 feet thick except around the fuel transfer canal where the 24 wall is 5 feet thick. The concrete foundation of the pool has a minin;um 25 thickness of 5 feet and is founded on approximately 5 adcitional feet of lean 26 concrete placed directly on rock. The pool walls and floor are lin?d with
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1 stainless steel plate with a thickness of 0.25-inch on the floor and 2
approximately 0.125-inch on the walls. Shiffer, et al, ff. Tr.179 at 14.
3 2
The high density spent fuel racks, when fully loaded with spent fuel, 4
will increase the overall mass of the auxiliary building by less than one 5
percent. The liner plate and pool structures were evaluated for these new 6
loading conditions and found to be adequate to support and transfer the high 7
density rack reaction loads. Shiffer, et al, ff. Tr.179 at 14-15.
8 B.
The NRC has established acceptance criteria and design guidance for 9
safe storage of spent fuel.
The seismic design criteria and guidance are 10 primarily contained in Section 9.1.2 and Section 3.8.4, Appendix D of the 11 '
Standard Review Plan (SRP), and in the NRC Position Paper, "0T Position for 12 i Review and Acceptance of Spent Fuel Storage and Handling Applications,"
13, (Position Paper) dated April 14, 1978 (PGandE Exhibit 12). Shiffer, et al, 14 ff. Tr. 179 at 15.
15; 2
SRP Section 9.1.2, Paragraph III.3.a. requires that spent fuel 16I storage racks be classified and designed to Seismic Category I requirements.
17 The criteria for seismic design and fuel assembly impact loads are provided in 18 Section IV (3) of the Position Paper.Section IV (5) of the Position Paper 19 states that SRP Section 3.8.4 provides acceptable procedures for modeling and 20!
analyzing the seismic responses of the spent fuel racks.
Further,Section IV 21 l (2) of the Position Paper identifies either of two industry codes,Section III 22 of the ASME Code or the AISC Specification, as being acceptable for deriving 23 the allowable stress criteria for the racks. Other codes are acceptable based l
on a case-by-case review. Structural acceptance criteria are provided in 24 25 l Section IV (6) of the Position Paper. The criteria permit rack sliding and 26 j rack-to-rack impacts and provide specific guidance on how such impacts are to a
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1 be incorporated in the rack design. Shiffer, et al, tf. Tr.179 at 15-16; 2
Fishman, et al, ff. Tr. 519 at 12-13; Ashar, Tr. 591, 595, and 596.
3-10.
In this connection, Dr. Ferguson admitted that it is possible that 4
his testimony, which stated that "all Category I equipment is prohibited from 5
sliding...." is incorrect.
In fact, he admitted that he is not familiar with 6
different kinds of Seismic Category I equipment and the fact that some Seismic 7
Category I equipment is permitted to slide. Therefore, while the spent fuel 8
racks are Category I equipmer.t. they are not prohibited from sliding.
9 Ferguson, Tr. 468-470; Ashar, Tr. 591.
10
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Contrary to Dr. Ferguson's interpretation, the NRC Staff interprets 11 the Position Paper to allow "the possibility of collision of the racks with 12 each other and with the spent fuel pool walls." The Staff's position, as 13 stated in the Position Paper, is that " impact loading should be quantified and 14 that sliding and tilting motions will be contained within suitable geometric 15I constraints." The Staff further testified that there is no dispute that the i
16' Position Paper permits sliding, tilting, and impacts of racks, including 17 rack-to-rack, rack-to-wall, and rack-to-floor impacts.
Fishman, et al, ff.
18 Tr. 519 at 12-13; Ashar. Tr. 591-592 and 595-596.
19 12 The Diablo Canyon high density racks comply with the applicable 20j seismic design criteria in that:
The racks were designed as Seismic Category I components in accordance 21 22 with SRP Section 9.1.2, Paragraph III.3.a.
23 The allowable stress criteria for the racks were derived from the 24 Section III, Subsection NF requirements of the ASME Code for Class 3 25 component supports. Construction materials conform to Subsection NF of 26 i the ASME Code and were selected to be compatible with the fuel pool l
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environment.
The seismic excitation was simultaneously applied in three orthogonal 2
3 directions.
Increased damping of fuel racks due to submergence in the 4
spent fuel pool was not considered for conservatism.
Local impact of the 5
fuel assemblies within the spent fuel rack cells was considered in a 6
manner which maximized forces acting on a rack module.
The procedures used for modeling and analyzing the seismic responses of 7
8 the Diablo Canyon spent fuel racks were consistent with the requirements 9
of the Position Paper. The models were developed based on current 10l engineering practices.
11 ll The possibility of gross sliding, tilting, and rack impacts under the 12 !
postulated Hosgri event were evaluated in accordance with the acceptance i
13, criteria specified in Section IV (6) of the Position Paper. Shiffer, et I
14 al, ff. Tr. 179 at 16-17.
15; 13 No exceptions to acceptance criteria were taken for the design of the 16i Diablo Canyon high density spent fuel racks. The racks were designed and I
17-constructed using the approved acceptance criteria to maintain the spent fuel i
- 8l assemblies in a safe configuration for normal and abnormal loads, including 19 potential impacts between racks and between the racks and the fuel pool walls, 20,
which may occur during a Hosgri event. Shiffer, et al, ff. Tr. 179 at 17.
21 14 The analytical model developed by PGandE for high density rack 22 analysis was a nonlinear dynamic model, and appropriately considered the 23 potential effects of the following possibilities: movement of the fuel 24 assemblies, frictional resistance at the base of the rack, rack sliding and 25 ll rocking behavior, rack uplift and subsequent isapact on the bearing plate, and 26 rack impacts with adjacent racks and pool walls. In addition to the potential
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1 rack movernents addressed in the analysis, fluid effects, known as hydrodynamic 2
coupling, were also considered. Shiffer, et al, ff. Tr.179 at 17.
3 15 Potential impact of the fuel assemblies within storage cells was 4
simulated by impact springs (designated as K ).
These impact spring y
5 constants were selected based on a series of parametric studies conducted to 6
assess the impact of a fuel assembly on the cell wall of the rack. The 7
cumulative impact loads of the fuel assemblies on the rack module were 8
calculated by assuming that all fuel assemblies in the rack move in unison; 9
1.e., they will impact the cell walls of the rack at the same instant and in 10 the same direction. Shiffer, et al, ff. Tr.179 at 18.
11 16 Rack sliding behavior was addressed in the model. A set of springs, K, was included in the model near the base to simulate the sliding friction 12 f
13 of the rack, whereas other springs, K, combined with gap elements, 6
14 simulate lift-off resulting from rocking. Another set of springs, K, was R
15 included to capture the rotation of the leg in the vertical plane. A further 16 set of springs, K, was included in the model to determine rack-to-rack and g
17 rack-to-wall impact forces.
Shiffer, et al, ff. Tr. 179 at 18.
18 12 Friction coefficients of 0.8 and 0.2 which bound the experimental 19 data were used in the analysis to maximize the inertial force and horizontal 20, displacement of the racks. This wide range of friction values is typically i
21 '
used in the industry for rack design. Shiffer, et al, ff. Tr.179 at 18.
22 18 Fluid inertial effects, produced by rack motion, were also addressed 23 in the model.
In particular, the accelerating fluid mass results in two types 24 of inertial effects. As a rack starts to slide, the water inside and 25 surrounding the rack is set in motion. This produces an additional inertial 26 force on the rack, which was addressed in the analysis by adding an i
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I appropriate amount of water mass, known as "virtua' mass," to the mass of the 2
rack and fuel assemblies. The second effect of the accelerating fluid mass is j
3 hydrodynamic coupling. As the space between moving racks or between the racks 4
and adjacent walls is reduced, the fluid between the b'odies is expelled from 5
that space. This causes fluid pressures to develop on the surfaces bounding 6
the fluid mass, which retards the seismic motion of the racks. The effects of 7
the fluid motion on rack displacements are determined by the kinetic energy of 8
the fluid. By underestimating the kinetic energy of the fluid, one 9
necessarily overestimates the rack displacements.
If the kinetic energy of 10f the fluid is ignored completely (e.g., assuming the absence of fluid), one I
11 l will grossly overestimate the rack displacements.
The calculation method used I
12 for rack analysis includes fluid motion but underestimates the fluid kinetic 13l energy and, accordingly, overestimates rack displacements; i.e., the 1/, I calculation method is conservative. PGandE's use of virtual mass and 15 hydrodynamic coupling in the analysis is based on the fundamental principles 16 of fluid dynamics. Shiffer, et al, ff. Tr. 179 at 18-19.
17 10 Hydrodynamic coupling coefficients used in the analysis made use of 18l theoretical results that were within their region of applicability, as 19 indicated by NRC Staff witnesses:
20!
"I reviewed the applicant's formulations presented in the reracking report and provided in greater detail in the 21 l Cited papers.
22 I personnaly verified some of the derivations of the equations.
I checked the specific numbers and I looked at 23 some of the computer output that the licensee provided that indicated that the fluid coupling input were consistent 24 with their assumptions.
25 j And I also made -- performed some numerical calculations based on general fluid coupling theory to convince myself 26 i that this is a meaningful application of the fundamental 1 l
theories of fluid." Fishman, Tr. 597.
j 2
"What I had done is review the methodology which is based on the same method as used by Fritz in his paper.
I know 3
that the -- I know from personal experience that the particular methodology is widely used and accepted in the industry.
4 5
And I felt that it was appropriate for this application.
I then reviewed thc equations that were presented by the 6
' licensee and they followed the methodology of Fritz.
7 I did not check numbers, check computer input numbers, specifically.
I spot-checked a few cases, but not every 8
one.
But, basically, I felt the methodology was acceptable and would give appropriate results." DeGrassi, Tr. 598.
"And based on complete picture of the parameters used by 10 the licensee, and the way the racks are designed, we made a
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determination that what is being used is appropriate and reasonable." Ashar, Tr. 599.
jj j 12l 20 The analytical process used in the design of the racks consisted of:
I 13 Development of a non11near dynamic model of a rack module consisting of 14 inertial mass elements, hydrodynamic coupling, and gap and friction' elements.
15; 16 Generation of the equations of motion and inertial coupling and solution j7 of the equations using a computer program, DYNAHIS, to determine rack 18 orces, moments, and displacements, Computation of the detailed stress field in the rack (at the critical jg 1 cations) and in the support legs using the forces, moments, and 20 displacerants calculated in the previous step. Shiffer, et al, ff. Tr.
21 179 at 19-20.
22 21 Using the methodology described above, PGandE calculated the 23 potential loads on the racks. These calculations were performed in conformity 24 25 ]
with the loading combinations and acceptance criteria specified in the NRC 26 j Staff's Position Paper and Section 3.8.4, Appendix D, of the Standard Review h I i
1 Plan. The loading combinations included the combined effects of dead load, 2
live load, thermal interaction within the pool, and inertia loads due to 3
seismic events. A series of rack loading cases (fully loaded, partially full) 4 was considered in order to establish the design loads. The resulting stresses 5
in the racks were determined to be lower than the allowable stress values 6
permitted by acceptance criteria. These allowable values provide a sufficient.
7 factor of safety when compared with the ultimate capacity of the racks.
8 Shiffer, et al, ff. Tr.179 at 20.
9 22 The design basis analysis was performed with a single-rack model.
10 Conservatisms were built into the evaluations performed for the high density 11 racks in terms of modeling assumptions, postulated loadings, and safety 12 margins on stress allowables. Several of the conservatisms inherent in the 13 design basis analysis are:
Adjacent racks were assumed to move in a manner equal and opposite (out 14 15 of phase) to the rack module being analynd, thereby maximizing the 16 potential for rack-to-rack impact.
A value of 4 percent damping was used.between the fuel assemblies and 17 18 racks, between adjacent racks and between racks and walls. A value of 10 19 percent for impact damping (in addition to structural damping) has been 20' used at other plants licensed by the NRC. The analyses neglected fluid i
21 damping.
The impacts between cell walls and the fuel assemblies were assumed to 22 l
23 occur in phase.
In reality, the fuel assemblies exhibit complex and l
24 random behavior. However, they were all assumed to move in unison so 25 l that the maximum response could be obtained.
1 26 l The form drag opposing the motion of the racks within the pool water was li l I
1 conservatively neglected as discussed by PGandE witness K. P. Singh:
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2 "He model the effect of water in a highly conservative manner. He leave out the effects of water that actually 3
tires us when we swim, the drag effects. All the energy dissipation properties of water that come about that we 4
collectively call fluid damping, form drag, and so forth, we neglect it completely.
Even the effects of water that 5'
one calculates using classical incompressible fluid mechanics relations, which are known for centuries, even 6
tinose we model in a very conservative way. He leave an awful lot of margin. The energy dissipation, that is 7
available to us in the actual rack configuration in the pool, we don't take credit for."
Singh, Tr. 197.
8 The fluid coupling coefficients were calculated based on the conservative 9
e 10 assumption that the adjacent rows of racks are an infinite distance away 11 (the distance measured perpendicular to the direction of rack movement).
12 This reduces the " cross-coupling effect" of the adjacent rows of racks 13; and yields conservative displacements and impact forces.
Fishman, et al, 14 ff. Tr. 519 at 21.
The calculation of fluid inertial effects included an underestimate of 15 16 the fluid kinetic energy and resulted in a conservative overestimate of 17 rack displacement.
Hydrodynamic coupling coefficients used in the analysis neglected certain 18 19 nonlinearities of the motion. Studies in the literature show that 20, incorporation of these nonlinear effects would significantly lower rack 21 l response. Shiffer, et al, ff. Tr.179 at 20-22.
l 22 !
23 Dr. Ferguson admitted that his testimony regarding experimental 23 validation of fluid coupling effects in rack movement was incorrect in that he 24 erroneously assumed that Dr. Singh's work (Sierra Club Exhibit 8) recommended 25 I that such experimentation be performed. He corrected his understanding of the s
situation when he stated that:
26 i 0 l
j "I guess - it is true, upon rereading, I see that, in that paper, he specifically expressed the desire to have 2
experimental verification of fluid damping."... "I no longer would maintain that Kris (Singh] said in public that 3
fluid coupling theory,..., to have that experimentally tested." Ferguson, Tr. 476 and 478.
]
2_4 Industry practice with regard to high density rack design and-analysis was reviewed by PGandE for 10 other U.S. nuclear plants that have freestanding spent fuel racks. The rack suppliers for these plants included Joseph Oat Corporation; Exxon Nuclear Company; GCA Corporation, par Systems; Nuclear Energy Services, Inc.; Westinghouse Electric Corporation; and General Electric Company.
The analytical techniques used for designing those racks were similar to the techniques used for the Diablo Canyon rack analysis.
"l Shiffer, et al, ff. Tr.179 at 22-23.
12 I3I B.
Contfntions Contention I(A)3 I
It is the contention of the Sierra Club, Santa Lucia 16j Chapter (Sierra Club), that the report submitted to the NRC entitled Rerackina of Soent Fuel Pools. Diablo Canyon 17 Units 1 and 2 and other communications between Pacific Gas and Electric Company (PGandE) and the NRC, which are 18 available to the public on the same subject (the Reports),
fail to contain certain relevant data necessary for i
19' independent verification of the claims made in the Reports regarding consistency of the proposed reracking with the l
protection of the public nealth and safety, and the 2 0
environment. In particular, the Reports fail to contain 21 data regarding:
23 3)
The expected velocity and displacement of the spent fuel pools (pools) as a function of time in three 24 dimensions during the postulated Hosgri earthquake (PHE);
25 26 25 Data regarding the velocity and displacement of the fuel pools as a d
4
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function of time in three dimensions for the postulated Hosgri earthquake is 2
not necessary for rack analysis or review by the NRC Staff in evaluating the 3
technical adequacy of the rack design since the acceleration time-histories l
)
4 are used for that purpose. Consequently, the velocity and displacement 5
time-history data for the fuel pools were not included in the Reracking Report 6
because a record of such data was not required during the design process.
7 Shiffer, et al, ff. Tr.179 at 24; Fishman, et al, ff. Tr. 519 at 6-7.
8 Zfi.
The design process for the racks utilized the postulated Hosgri 9
earthquake acceleration time-histories for the base of the spent fuel pool.
10 i Velocity and displacement information can be derived from the acceleration 11 '
tin:e-histories used in the design, which are contained in the Reracking l
12 i Report, Figures 6.1.1, 6.1.2, and 6.1.3.
Shiffer, et al, ff. Tr.179 at 24; I
13j Fishman, et al, ff. Tr. 519 at 6-7.
i 14 I
15 Contention I(A)4 l
It is the contention of the Sierra Club, Santa Lucia 16 [:'
Chapter (Sierra Club), that the report submitted to the NRC 17!
entitled Rerackina of Soent Fuel Pools. Diablo Canyon i
Units 1 and 2 and other communications between Pacific Gas 18l and Electric Company (PGandE) and the NRC, which are available to the public on the same subject (the Reports),
19 fall to contain certain relevant data necessary for independent verification of the claims made in the Reports 20' regarding consistency of the proposed reracking with the j
protection of the public health c.nd safety, and the 21 i environment. In particular, the Reports fail to contain data regarding:
22 23
)
4)
The expected maximum velocity and displacement of the 24 [
racks obtained from the computer modeling of rack behavior during the PHE; j
i 25 !)
26 jj ZZ.
The maximum velocity of the racks is not documented in the Reports l
d l
I since it is not a value needed for design of the racks. However, the maximum 2
displacement for a loaded rack module is included in the Reracking Report in 3
Table 6.8.2.
Shiffer, et al, ff. Tr.179 at 24-25.
4 18 In the design basis analysis, the maximum displacement relative to 5
the pool structure of a loaded rack module is approximately 2.8 inches. The l
6 maximum displacement relative to the pool structure of a nearly empty module y
is aparoximately 4.2 inches. Fishman, et al, ff. Tr. 519 at 25; Shiffer, et 8
al, ff. Tr.179 at 25.
9 10 Contention I(B)2 11 It is the contention of the Sierra Club that the F:eports fail to include consideratico of certain relevant 12 '
conditions, phenomena and alternatives necessary for l
independent verification of claims made in the Reports regarding consistency of the proposed reracking with public 13 f
health and safety, and the environment, and with federal 14 law. In particular, the Reports fail to co.nsider:
15 2)
The resonant behavior of the spent fuel assemblies in the racks in response to the PHE and the consequences 16l of such behavior; 17 18 29 The rack analysis performed by PGandE considered potential resonant 19 behavior of fuel assemblies. The design basis analysis performed to evaluate 20, the fuel racks utilized a mathematical representation of the various 21 components and their response behavior. Since resonant behavior is a 22 I fundamental condition described by the equations of motion, and since the 23 equations of motion were appropriately represented, the analysis considered 24 l the possibility of resonant behavior. Shiffer, et al, ff. Tr.179 at 26.
25ll 30 The design basis analysis demonstrated that, due to the specific 26 p conditions present, the fuel assemblies do not experience resonant behavior.
d l
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1 These conditions include the nonlinearities of the system (including the 2
presence of water, the movement of the fuel assemblies within the fuel racks, 3
and the presence of friction at the fuel rack base). The analysis 4
appropriately represented these physical conditions and demonstrated that the 5
integrity of the racks is maintained. As a practical matter, resonance will 6
c.ot occur since the amplitude cannot increase beyond the 0.302 inch clearance 7
between the fuel assembly and cell wall. Shiffer, et al, ff. Tr.179 at 27; 8
Fishman, et al, ff. Tr. 519 at 10-11.
9 10 Contention I(B)2 11 It is the contention of the Sierra Club that the Reports fail to include consideration of certain relevant l
12 conditions, phenomena and alternatives necessary for independent verification of claims made in the Reports 13l regarding consistency of the proposed reracking with public health and safety, and the environment, and with federal 14 law.
15 In particular, the Reports fail to consider:
16' 1
7) alternative on-site storage facilities including:
17l 18 (1) construction of new or additional storage facilities and/or; 19 (ii) acquisition of modular or mobile spent nuclear 20, fuel storage equipmerit, including spent nuclear fuel storage casks; 21 31 A discussion of alternatives is documented in PGandE's Reracking 22 l 23 Report, Chapter 9.
In particular Diablo Canyon was designed to store spent 24 fuel for a nominal period of one year and then ship the fuel offsite for 25 l
reprocessing or disposal. Due to the unavailability of fuel reprocessing 26 facilities and of permanent disposal sites, the spent fuel must now be stored.
1 for an extended period of time at Diablo Canyon. Therefore, the alternatives 2
that must be considered, in addition to onsite storage, consist of various 3
methods of storing the spent fuel offsite or shutting down the reactor. The 4
consideration of alternatives, including offsite shipment of spent fuel and.
5 shutdown of the reactor, was documented in the Reracking Report, Chapter 9 6
(PGandE Exhibit 2).
While the onsite storage alternative was chosen, there 7
are no regulations which specify the nature of onsite storage methods that 8
must be considered or documented. The discussion included in the Reracking 9
Report was sufficient to comply with NRC requirements. However, the Reracking 10 Report did not specifically address PGandE's evaluation of other onsite storage methods because the expansion of the spent fuel pool capacity through 11 i
12 reracking was clearly superior. Shiffer, et al, ff. Tr. 179 at 28-29.
13 E.
PGandE evaluated the two methods of onsite storage facilities 14 mentioned in the contention. The evaluation was brief since these two 15 specific methods, additional storage facilities and acquisition of modular 16!
storage equipment, do not offer any increase in safety over high density 17 racks, and they involve technical, regulatory, and other disadvantages when 18 compared with high density racks. Shiffer, et al, ff. Tr.179 at 28; Cleary, 19 ff. Tr. 604 at 2-3.
20, M.
An additional storage pool was considered less attractive because it 21 would not provide any added safety for spent fuel storage than with properly 22 designed high density racks in the existing pools. Moreover, the costs of 23 constructing a new seismically qualified structure and auxiliary support 24 systems would obviously be very high compared with roracking. Finally, this I
25 H would involve increased handling of the spent fuel. Shiffer, et al, ff. Tr.
26 j 179 at 29; Cleary, ff. Tr. 604 at 3-5. l
i l
1 E
Acquiring modular storage equipment was considered less attractive 2
because such equipment would not provide any added safety over and above 3
properly designed high density racks. Further, modular equipment such as dry 4
cask storage was not a licensed concept at the time the reracking decision was 5
made by PGandE, and casks were still being tested. In any event, dry cask 6
storage is not a viable option for Diablo Canyon based upon the design of the 7
dry casks currently available. The dry casks are designed to store only fuel 8
that has been discharged from the reactor at least five years prior to cask 9
storage. Thus, this storage method could not be used for at least five years 10 following the first refueling outage. Cleary, Tr. 617. See 10 CFR 72.
11 The existing low density racks at Diablo Canyon were originally 1
12 designed, in accordance with early NRC guidelines, to accommodate spent fuel 13 discharged from one refueling (roughly 70 assemblies), plus a reserve capacity 14 ef a full core offload (193 assemblies) in the event a full core discharge is 15 necessary. Shiffer, et al, ff. Tr. 179 at 29.
I 16 The storage space associated with one refueling discharge is 17 currently occupied at Diablo Canyon Units 1 and 2 after the first refueling 18l outages. Based upon operating schedules and the desirability of maintaining 19 full core discharge capability, it is necessary that the spent fuel storage 20; capacity for both units be increased.
Further, the cost of the casks, 21 assuming their availability, which would be required for the needed capacity 22 at Diablo Canyon would be high compared with the reracking alternative. At 23 the time that PGandE made the reracking decision, there were no plants in the 24 United States using modular storage facilities for spent fuel storage.
25 l Subsequently, two plants were licensed to use modular storage facilities such 26 as dry casks, but these plants did so only when all of the storage space in i
I l
l 1
existing pools had been filled after they had previously reracked with high 2
density racks. Shiffer, et al, ff. Tr.179 at 29-30; Cleary, ff. Tr. 604 at 3
5-9.
4 E.
The Sierra Club did not present "any concrete evidence" to show that 5
PGandE failed to consider other alternatives to reracking. Rather, this 6
contention is based only on opinion. Dr. Ferguson admitted as much when he
?
stated that the particular contention "is just un opinion" he has reached.
8 Ferguson, Tr. 443.
9 M.
The Sierra Club's testimony on Contention I(B)7 was amended by its 10 only witness, Dr. R. Ferguson, who admitted that PGandE did, in fact, consider i
11 other alternatives to reracking, though not in his opinion " seriously." He 12 stated that he wished to amend his testimony to say that PGandE " failed to 13 consider them [other alternatives) seriously." Moreoever, Dr. Ferguson 14 admitted that duririg the discovery process, the Sierra Club received documents 15!
from PGandE which considered other alternatives. Specifically, he admitted h
16 ll that "[t]here were some documents provided related to cask storage."
I 17l Ferguson, Tr. 444.
18l 32 PGandE produced evidence (PGandE Exhibit 13) which showed that it did 19 review "four or five alternatives" before selecting reracking. Dr. Ferguson 20f admitted that Exhibit 13 contains "a brief summary of descriptions of some 21 factors involved with the alternatives" considered.
Ferguson, Tr. 446-447.
I 22 23 Contention I(B)8 24 It is the contention of the Sierra Club that the Reports fail to include consideration of certain relevant 25 g conditions, phenomena and alternatives necessary for independent verification of claims made in the Reports 26 ll regarding consistency of the proposed reracking with public 1,
j health and safety, and the environment, and with federal law.
In particular, the Reports fail to consider:
2 8) the use of anchors, braces, or other structural 4
members to prevent rack motion and subsequent damage during the PHE; 5
6 M.
The use of anchors, braces, or other structural members to prevent 7
rack motion is not discussed in the Reports since freestanding racks meet 8
safety requirements, without the need for use of these items. Shiffer, et al.
9 ff. Tr. 179 at 31; Fishman, et al, ff. Tr. 519 at 11-12.
10 H.
Structural anchors, braces, or other structural members are not jj required to prevent rack motion and potential subsequent rack damage. The 12 freestanding racks satisfy NRC criteria and guidance applicable to spent fuel 13 storage racks; Fishman, et al, ff. Tr. 519 at 11-12. The design accommodates 14 the calculated rack motion during the postulated Hos,gri earthquake and stiows 15 that the racks have sufficient safety margins.
In addition, freestanding 16j racks have several advantages over anchored or braced racks. Particularly, 17
-freestanding racks reduce the stress on the liner caused by thermal loads from 18 the heat generated by the spent fuel. Further, sliding provides a very 39 effective means to dissipate energy. A freestanding rack is, therefore, 20, considered a better design to absorb seismic energy and, thus, has a distinct 21 advantage over anchored or braced racks. Further, no welding is required to 22 install the freestanding racks.
Finally, inspection and/or replacement of 23 racks, if necessary, is simplified by the use of freestanding racks. Shiffer, 24 et al, ff. Tr.179 at 31.
///
25 ll Ill 26 i l
1 I
i i
1 Contention II(A)l. 2. 3 l
2 It is the contention of the Sierra Club that the proposed reracking is inconsistent with the protection of the public 3
health and safety, and the environment, for reasons which include the following:
4 A) during the PHE, collisions between the racks and the 5
pool walls are expected to occur, resulting in:
6 1) impact forces on the racks significantly larger than those estimated in the reports; 7
2) impact forces on the racks significantly larger than those expected to damage the racks; 8'
3) significant permanent deformation and other damage to the racks and pool walls; 9
10 4.Q.
During the postulated Hosgri event, there will be no collisions 11 between the racks and the pool walls that will result in impact forces on the 12 racks significantly larger than those estimated in the Reports. The Reports describe the analyses performed for a wide range of collision scenarios and 13 14 the resulting impact forces are expected to bound those that could reasonbly 15 occur. Further, there will be no collisions between the racks and pool walls 16l that will result in impact forces that could cause significant permanent 17 deformation and other damage to the racks and pool walls. The calculated i
18l forces and stresses are less than those which would cause significant 19 permanent deformation or other damage to the racks and pool walls. Hinor 20, local damage to the liner, concrete pool wall, or racks may occur, but such 21 damage does not adversely affect the integrity of the fuel pool, racks, or 22 fuel. Shiffer, et al, ff. Tr.179 at 34; Fishman, et al, ff. Tr. 519 at 13-14.
l 23
_41 The design basis analysis and the techniques used to determine 24 potential impact forces were based on a conservative mathematical 25 h representation of the racks to ensure an upper-bound impact force for 26 rack-to-rack and rack-to-wall impact. Fishman, et al, ff. Tr. 519 at 13-14 j
l
i e
l 1
and 18. Some of the conservative assumptions include:
k 2
Adjacent racks were assumed to move in a manner equal and opposite (out 3
of phase) to the rack module t 'ng analyzed.
A value of 4 percent damping was used between the fuel assemblies and 4
5 racks, between adjacent racks, and between racks and walls. A value of 6
10 percent for impact damping (in addition to structural damping) has 7
been used at other plants licensed by the NRC. The analyses neglected 8
fluid damping.
The impacts between cell walls and the fuel assemblies were assumed to 9
10j occur in phase.
In reality, the fuel assemblies exhibit complex and i
11[l random behavior. However, they were all assumed to move in unison so i
12 i that the maximum response could be obtained.
I The form drag opposing the motion of the racks within the pool water was 13.
!i 14 l' conservatively neglected.
The fluid coupling coefficients were calculated based on the conservative 15 4
16 d assumption that the adjacent rows of racks are an infinite distance away 4
17 j (the distance measured perpendicular to the direction of rack motion).
I 18l This reduces the " cross-coupling effect" of the adjacent rows of racks 19 and yields conservative displacements and impact forces.
20 '
The calculation of fluid inertial effects included an underestimate of 21 h the fluid kinetic energy and resulted in a conservative overestimate of I
22 ;
rack displacement.
I 23 l Hydrodynamic coupling coefficients used in the analysis neglected certain 24 nonlinearities of the motion.
Studies in the literature show that n
25 [
incorporation of these nonlinear effects would significantly lower rack i
26,;
response.
'l
_ 24 -
l
1 4._2 Several parametric studies were performed by PGandE that included 2
both simplified and complex two-dimensional, single-and multi-rack analytical 3
models, as well as enhancements to the original design basis, 4
three-dimensional, single-rack model. The results of these studies confirm in 5
all cases that rack impact loads and stresses due to the postulated Hosgri 6
earthquake are below allowable values.
Fishman, et al, ff. Tr. 519 at 22.
7 Therefore, the design basis evaluation was conservative and the high density 8
spent fuel racks satisfy acceptance criteria and will maintain their integrity 9
for the postulated Hosgri event. Shiffer, et al, ff. Tr. 179 at 34-36.
10 D.
While impact forces are important to the design process, of more 11 significance are the stress ratios in that they better reflect the effect of 12 j impacts on the racks. The controlling stress ratios for the racks have an 13' allowble value of 2.0.
The highest stress ratio for the impacts determined 14 from the design basis anaysis was 1.436. For the impacts determined from'the 15 parametric studies, the highest stress ratio was 0.743. Thus, the design 16 basis evaluations were shown to be conservative and bounding, and the racks 17 were shown to accommodate the impact with acceptable margins. Shlffer, et al, 18 ff. Tr. 179 at 36; DeGrassi, Tr. 526-527.
19
_4_4 In evaluating the walls and the rack components, impact loads were 20, conservatively assumed to be static. No credit was taken for the short 21 duration of the loading. Stresses derived from these calculated forces were 22 significantly smaller than the stresses the racks and walls are capable of 23 withstanding without any adverse effect. Shiffer, et al, ff. Tr.179 at 36-37.
24 M.
Because of the conservative assumptions and methods used to analyze 25 j rack-to-rack and rack-to-wall impact forces, the resulting impact forces on 26 [
the racks bound those that might occur during the postulated Hosgri event. I
1 Shiffer, et al, ff. Tr.179 at 37; Fishman, et al, ff. Tr. 519 at 15-16.
2 afi.
If a rack should impact an adjacent rack or the wall, the impact 3
force would occur at the girdle bar or at the baseplate. The fuel rack 4
strength at the girdle bar level is significantly greater than that required 5
to resist the design loads. As the rack impacts the wall, the rack girdle 6
bars perpendicular to the wall would be loaded in compression by direct 7
bearing.
These bars can sustain a direct impact load greater than 175,000 8
pounds each before the onset of yielding, and incipient failure occurs at a 9
load of at least twice the yield force. The impact resistance along the 10 girdle bar which impacts flat against the wall is greater than 20,000 pounds 11 per storage cell. Hith regard to the baseplate, its resistance is 12 substantially greater than that for the girdle bars. Shiffer, et al, ff. Tr.
13l 179 at 37.
14 AZ.
Notwithstanding Dr. Ferguson's unsupported conclusion that rack 15 failure could occur with impact loads of any amount larger than the allowable 16 il loads, the NRC Staff agrees with PGandE in that such failure is highly 17 unlikely due to the reserve margin between the onset of yielding and incipient 18l failure. This yield-to-failure relationship is typical of ductile structural 19 l materials. This safety margin was recognized and confirmed by Mr. Fishman l
1 20!
when he stated that:
)
21 "Between this allowable load and the load required to cause large permanent deformation, there is a large reservoir of j
22 energy absorbing capacity in the rack modules." Fishman, et al, ff. Tr. 519 at 15-16.
23 24 He further concluded that:
25 h "A summary of the impact analysis results... clearly indicates that in all cases no significant permanent e
26 ij deformation or corresponding damage at the impact locations
}u
!l l
f 1
is expected (i.e., stress levels remain below the yield value)." Fishman, et al, ff. Tr. 519 at 16.
2 3
M.
Dr. Ferguson conceded that the Sierra Club Contention II(A) which 4
alleged possible impact forces higher than those reported by PGandE is unfounded. In particular, Dr. Ferguson admitted that the calculations that he 5
6 performed to substantiate the contention were inaccurate and unreliable.
7 Consequently, there were no material facts presented in his testimony to show rack failure.
Ferguson, Tr. 479.
8 9
B.
Impact forces result in loads being applied to the rack girdle bars 10 and baseplates. These forces were compared with the calculated capacities of jj !
the girdle bars and baseplates. For all cases, including straight-on impacts 12 and those where the corner of one rack impacts an adjacent rack away from the corner, the calculated capacities are greater than the maximum expected impact 13 l,-
14 forces.
In addition, impact forces generate stresses in the rack body an'd support feet. For all important structural members, stress levels were 15l 16l determined to be within acceptance criteria. Finally, the effects of impacts 17 between racks and the pool walls were also evaluated. Local and overall 18 stresses were evaluated for both the pool liner and concrete pool walls.
jg While minor local liner or concrete deformations may occur, overall stresses 20, were found to be within allowables. Thus, based on conformance with NRC 21 l acceptance criteria, the structural integrity of the spent fuel pool and racks is assured. Shiffer, et al, ff. Tr. 179 at 37-38; Fishman, et al, ff. Tr. 519 22 23 at 16-17.
24 j EQ.
The racks are rugged, stainless steel, honeycomb-type structures.
2c j Although minor deviations from manufacturing tolerances may exist for such 26 il components, the effects of such minor deviations are not significant and are 1
n 1 i
1 accommodated by conservatisms in the analysis methodology.
In addition, the j
2 racks are fabricated from a very ductile steel which makes minor deviations 3
insignificant. Shiffer, et al, ff. Tr. 179 at 38.
4 51 While there may be minor local deformation to the racks or pool walls 5
during the postulated Hosgri event, there would be no permanent deformation or 6
other damage that would lead to criticality, radiological releases, damage to 7
the fuel, increases in heat generation, or otherwise adversely affect the 8
public health and safety. Shiffer, et al, ff. Tr.179 at 38-39.
9 10 Contention II(A)4 11 It is the contention of the Sierra Club that the proposed reracking is inconsistent with the protection of.the public 12 health and safety, and the environment, for reasons which include the following:
13 A) during the PHE, collisions between the racks and the 14 pool walls are expected to occur resulting in:
15 16' 4) reduction of the spacings between fuel assemblies; 17, i
18l 52 Each Diablo Canyon fuel assembly consists of a 17 x 17 array of 19 cylindrical rods of which 264 rods contain fuel pellets. The assembly is 20!
approximately 8.4 inches square and 13.3 feet in length. Each fuel rod is a 21 Zircaloy tube containing uranium dioxide fuel pellets. Grids are positioned l
22 at vertical intervals along the length of the fuel assembly to maintain the 23 rod spacing. Shiffer, et al, ff. Tr. 179 at 39.
24 53 The active fuel region is the region within the fuel assembly which 25 h contains fuel pellets. This region extends 144 inches, from approximately 26 ;
3 inches above the bottom of the fuel assembly nozzle, which rests on the rack
'i
'I 5
I baseplate, to approximately 10 inches below the rack girdle bars. Shiffer, et 2
al, ff. Tr.179 at 39.
3 S.
The maximum forces generated by the postulated Hosgri earthquake will 4
not result in a reduction of the design spacing of 10.93 inches between the 5
active fuel region within any rack module.
Fishman, et al, ff. Tr. 519 at 6
17-18, 32, and 33. This spacing, with its tolerances, was used in the 7
criticality analysis. Shiffer, et al, ff. Tr. 179 at 40.
8 Contention II(A)5 9
It is the contention of the Sierra Club that the proposed reracking is inconsistent with the protection of the public 10' health and safety, and the environment, for reasons which include the following:
11 L l'
A) during the PHE, collisions between the racks and the 12 pool walls are expected to occur resulting in:
13 14 5) increase in the nuclear criticality coefficient k(eff) above 0.95; 15 16i M.
Criticality analyses were performed for the Diablo Canyon high 17 density spent fuel storage racks to assure that a k,ff equal to or less than 18 0.95 is maintained when the racks are fully loaded with fuel of the highest 19 anticipated reactivity in each of two regions and when the pool is flooded 20.
with unborated water at a temperature corresponding to the highest 21 reactivity. Fishman, et al, ff. Tr. 519 at 32. The maximum calculated 22 reactivity includes a margin for uncertainty in reactivity calculations and in 23 mechanical tolerances, statistically combined, such that the k,77 will be 24 equal to or less than 0.95 with a 95 percent probability at a 95 percent 25ll confidence level. Shiffer, et al, ff. Tr.179 at 40.
26 l H.
The Diablo Canyon spent fuel pools will be continually maintained at 1 I
I a boron concentration of at least 2000 ppm as required by the plant Technical 2
Specifications. This soluble boron not only provides an additional and very 3
large subcriticality margin under normal storage conditions, but precludes the 4
possibility of exceeding a k,ff of 0.95 under credible abnormal conditions, 5
including the postulated Hosgri event. Shiffer, et al, ff. Tr.179 at 40-41; 6
Fishman, et al, ff. Tr. 519 at 31.
7 EZ.
The spacing requirement to maintain k,7f less than 0.95 without 8
borated water is essentially the fuel assembly spacing in the rack design 9
(10.93 inches), based upon the criticality analysis described in Section 4.0 10 of PGandE's Reracking Report. With borated water normally present in the 11 l spent fuel pool, the k,7f would not reach 0.95 until the water gap between 12l storage cells in Region 1 (nominally 1.786 inches) has been reduced to less 13 '
than 0.1 inch uniformly everywhere, an extremely implausible condition. While 14 analyses have demonstrated that significant rack deformation would not occur, 15 even if it were assumed that there was zero gap between storage cells, the 16 resulting configuration would still not be critical.
In Region 2, reducing 17 the gap between storage cells to zero from the nominal 1.9 inches would not 18 result in k,77 exceeding 0.95.
Shiffer, et al, ff. Tr.179 at 41.
19 18 Hith unborated water in the spent fuel pool, the highest k gff, l
20 including an allowance for uncertainties and manufacturing tolerances, was 21 calculated to be 0.920 in Region 1 and 0.938 in Region 2.
Both calculations 22 are based upon conservative specifications of fuel enrichment and burnups and 23 provide subcriticality margins greater than that required by NRC regulations.
24 Hith the normal concentration of soluble boron present (2000 ppm), the safety 23 margin below criticality is much larger, with the maximum k,77 being less 26 j than 0.75 in both regions.
There are no postulated collisions or plausible l
41 I i
l 1
I j
reductions in spacing that could result in k,7f exceeding the limit of 2
0.95.
Shtffer, et al, ff. Tr. 179 at 41-42; Fishman, et al, ff. Tr. 519 at 3
34-35.
4 5
Contention II(A)6 6
It is the contention of the Sierra Club that the proposed reracking is inconsistent with the protection of the public 7
health and safety, and the environment, for reasons which include the following:
8 A) during the PHE, collisions between the racks and the 9
pool walls are expected to occur resulting in:
10 11 6) release of large quantities of heat and radiation; l
12*
l 13!
53 Any postulated condition that would cause the release of radiation 14 would require the fuel cladding to rupture; however, fuel cladding rupture 15 cannot occur unless the fuel assembly grids are crushed.
For Diablo Canyon, 16j the calculated impact forces are not large enough to cause crushing of the 17l grid and rupture of the cladding. Shiffer, et al, ff. Tr. 179 at 42.
18 ftQ.
During the postulated Hosgri event at Diablo Canyon Units 1 and 2, 19 due to the motion of the rack module relative to the motion of the fuel 20l assemblies, the fuel assemblies in the spent fuel pool storage racks could 21 contactthestainigsssteelwallsofthestoragecells. However, the maximum 22 impact force on a fuel assembly grid has been calculated to be only 23 approximately 1700 pounds and the maximum fuel rod bendirg stress has been 24 calculated to be only approximately 800 psi. Shiffer, et al, ff. Tr.179 at 25 [
42-43.
26 q fil.
The structural integrity of the fuel assembly was evaluated by 4
] i
I comparing the calculated forces against capacity determined from analytical 2
and experimental data. Specifically, the maximum impact force on the grid, 3
the fuel rod bending stresses due to flexure, and the fuel rod local contact 4
forces at the grid supports were evaluated. Shiffer, et al, ff. Tr.179 at 43.
5 12 Dynamic impact tests have been performed by Westinghouse on fuel 6
assembly grids for all Hestinghouse 17 x 17 fuel assembly designs to determine 7
their ultimate strength, which is the load at which incipient plastic 8
deformation of the grid cells occurs. The evaluation showed that the safety 9
factor for the grids, which is defined as the ratio of the ultimate grid 10 strength (i.e., greater than 3400 pounds) divided by the maximum impact force 11 applied to the grid, was greater than 2.
The evaluation of fuel rod bending 12l stresses showed that the ratio of the fuel rod allowable stress limit 13' (i.e., greater than 16000 psi) to the maximum calculated stress during the 14 Hosgri event is greater than 20 for all Hestinghouse 17 x 17 fuel assembi'y 15 designs.
16 The maximum local contact fccce that a fuel rod can sustain without 17 cladding failure was calculated by Westinghouse employing finite element 18 analysis methods. A finite element model of the fuel rod was formulated which 19 consists of discrete elements, each of which has stdess and deflection 20' characteristics defined by stress-strain theory. Th3 calculated local stress 21 levelscausedbythereactionforcewerewellbelow[heallowablestress 22 levels in the fuel rods, ensuring that the integrity [of the fuel cladding will 23 be maintained during the Hosgri event. Thus, the integrity of fuel assemblies 24 l stored in the high density spent fuel racks at Diablo Canyon will be 25 mainta b i ind there can be no resulting release of large quantities of heat and radioactive material. Shiffer, et al, ff. Tr.179 at 43-44; Fishman, et 26 j l
b I
al, ff. Tr. 519 at 31.
2 i
3 Contention II(A)7. 8. 9 4
It is the contention of the Sierra Club that the proposed reracking is inconsistent with the protection of the public 5
health and safety, and the environment, for reasons which include the following:
6 A) during the PHE, collisions between the racks and the 7
pool walls are expected to occur resulting in:
8 9
7)
. radioactive contamination of the nuclear power plant and its employees above the levels 10 permitted by federal regulations; 11 8) radioactive contamination of the environment in the vicinity of the nuclear power plant above the 12 levels permitted by federal regulations; and 13l 9) radioactive contamination of humans and other living things in the vicinity of the nuclear 14 power plant above the levels permitted by federal regulations.
15 16 fi3 The racks have been qualified to withstand the intpact loads which may 17 result from collisions between racks and pool walls during the postulated 18 Hosgri earthquake. Therefore, no damage to the fuel would occur, and there 19 can be no resulting releases of large quantities of heat and radioactive 20 material. Additionally, the racks will maintain the fuel assemblies in a 21 subcritical configuration even during any such collisions, and releases due to 22 criticality in the pools cannot occur. Consequently, no radioactive 1
23 contamination of h6 mans and other living things in the vicinity of the plant 24 above the levels permitted by federal regulations could result from collisions 25, j between the rteks and the pool walls during the postulated Hosgri earthquake.
26 j' Shiffer, et al, ff. Tr.179 at 45.
I
! l
1 Contention II(B) i 2
It is the contention of the Sierra Club that the proposed reracking is inconsistent with the protection of the public 3
health and safety, and the environment, for reasons which include the following:
4 5
8) during the PHE, collisions between groups of racks 6
with each other and/or with the pool walls are expected to occur with results similar to those i
7 described in II(A) above.
8 9
63 Because of the dissimilarity of the racks (in terms of geometry, I
10 tolerances, and gap spacings) it is highly unlikely that groups of racks would 11 move as a unit under a random seismic motion. Shiffer, et al, ff. Tr.179 at 12 i 46; Fishman, et al, ff. Tr. 519 at 19.
I 13j 55 PGandE did, however, conduct several multi-rack parametric studies, I
14 which confirmed that the design basis analysis had resulted in a conservative 15l rack design. Shiffer, et al, ff. Tr. 179 at 46.
16i 55 The parametric studies on multi-rack interactions utilized realistic i
17 modeling assumptions and evaluated variations of all key parameters that might 18l affect the qualification of the racks. Some of these parameters include 19 loading of the racks, hydrodynamic coupling coefficients as they apply to the 20i specific location of the rack, manufacturing tolerances, and friction 21 coefficients. These studies show that the loads on the racks are comparable 22 to those predicted by the design basis analysis, and, in all cases, these 23 loads are signficantly lower than the allowables. Thus, the parametric' 24 studies confirm that PGandE's modeling assumptions in the design basis 25 h analysis adequately represent potential group behavior of the racks. All 26 j potential collision conditions under the postulated Hosgri event are bounded i
Ii i
I by the loads for which the racks have been qualified. Shiffer, et al, ff. Tr.
2
'179 at 46-47.
3 4
III. CONCLUSIONS OF LAH 5
The Board has considered all the evidence submitted by the parties 6
and the entire reco'rd in this proceeding. That record consists of.the 7
Commission's Notice of Hearing, the pleadings and testimony filed by the 8
parties, the transcript of the hearing, and the exhibits received into 9
evidence. All issues, arguments, or proposed findings presented by the 10 parties, but not addressed in this decision, have been found to be without 11 merit or unnecessary for this decision.
Based upon the foregoing findings 12 which are supported by reliable, probative, and substantial evidence as 13' required by the Administrative Procedure Act and the Commission's Rules of 14 Practice, and upon consideration of the entire evidentiary record in this 15 proceeding, the Board, with respect to the issues in controversy before.us, 16 concludes that:
17 1
Pacific Gas and Electric Company has fully met its burden of proof on I
18 each of the contentions decided in this Decision.
19 2
There is no reasonable alternative to the proposed reracking of the 20 Diablo Canyon Units 1 and 2 spent fuel pools if shutdown of Diablo Canyon 21 Power Plant is to be avoided in the near future. Further, there is no 22 assurance that any of the alternatives can or will become available in a time 23 frame such that shutdown could be avoided.
24 3
There is reasonable assurance that the analysis and design of the 25 proposed high density spent fuel racks for Diablo Canyon Units 1 and 2 were 26 l performed in accordance with regulatory requirements and Commission 1
-I l 1
1 guidelines, and that the racks and pool structures meet applicable licensing 2
criteria.
3 4.
There is reasonable assurance that the Diablo Canyon Power Plant 4
spent fuel pools can be reracked without endangering the health and safety of 5
the public, as authorized by License Amendment Nos. 8 and 6 to License Nos.
6 DPR-80 and DPR-82, respectively, issued by the NRC Office of Nuclear Reactor 7
Regulation on May 30, 1986. Accordingly, the Board affirms the issuance of 8
said Amendment Nos. 8 and'6, and additionally concludes that no modifications 9
thereof or additional conditions are required.
(
10 5
The activities authorized by License Amendment Nos. 8 and 6 are not 11 inimical to the common defense and security or to the health and safety of the 12 public.
13
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14
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15
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16
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17
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18
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4 I
19'
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l 20'
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21
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22
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i 23
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24
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25
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26 l
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l I l
i 1
IV. ORDER 2
HHEREFORE, in accordance with the Atomic Energy Act of 1954, as 3
amended, and the Commission's regulations, and based on the findings and 4
conclusions set forth herein, IT IS ORDERED that the Director of Nuclear 5
Reactor Regulation make immediately effective License Amendment Nos. 8 and 6 6
to License Nos. OPR-80 and DPR-82, respectively, consistent with the Board's 7
decision in this case.
8 9
10 THE ATOMIC SAFETY AND LICENSING BOARD l
11 12 B. Paul Cotter, Jr., Chairman 13 ADMINISTRATIVE JUDGE
~
14 15 Glenn O. Bright ADMINISTRATIVE JUDGE 16' 17 4
Jerry Harbour 18l ADMINISTRATIVE JUDGE 19 I
20:
Dated at Bethesda, Maryland 21 this day of
, 1987 22 23 24 25 26 i
! l
APPENDIX A Sierra Clu' Contentions b
Contention I I(A).
It is the contention of the Sierra Club, Santa Lucia Chapter (Sierra Club), that the report submitted to the Nuclear Regulatory Commission (NRC) entitled Rerackina of Soent Fuel 4
Pools Diablo Canyon Units 1 and 2 and other communications between Pacific Gas and Electric Company (PGandE) and the NRC which are av'ailable to the public on the same subject (the Reports) fail to contain.certain relevant data necessary for independent verification of the claims made in the Reports regarding consistency of the proposed reracking with the protection of the public health and safety, and the environment.
In particular, the Reports fail to contain data regarding:
- 3) the expected velocity and displacement of the spent fuel pools (pools) as a function of time in three dimensions during the postulated Hosgri earthquake (PHE);
- 4) the expected maximum velocity and displacement of the racks obtained from the computer modeling of. rack behavior during the PHE; I(B).
It is the contention of the Sierra Club that the Reports fail to include consideration of certain relevant conditior.s, phenomena l
A-1
and alternatives necessary for independent verification of claims made in the Reports regarding consistency of the proposed reracking with public health and safety, and the environment, and 1
with federal law.
In particular, the Reports fail to consider:
- 2) the resonant behavior of the cpent fuel assemblies in the racks in response to the PHE and the consequences of such behavior;
- 7) alternative on-site storage facilities including:
(i) construction of new or additional storage facilities and/or; (ii) acquisition of modular or mobile spent nuclear fuel storage equipment, including spent nuclear fuel storage casks;
- 8) the use of anchors, braces, or other structural members to prevent rack motion and subsequent damage during the PHE; Contention II II.
It is the contention of the Sierra Club that the proposed reracking is inconsistent with the protection of the public health and safety, and the environment, for reasons which include the following:
i A-2
A) during the PHE, collisiens between the racks and the pool walls are expected to occur resulting in:
I
- 1) impact forces on the racks significantly larger than those estimated in the reports;
- 2) impact forces on the racks significantly larger than those expected to damage the racks;
- 3) significant permanent deformation and other damage to the racks and pool walls;
- 4) reduction of the spacings between fuel assemblies;
- 5) increase in the nuclear criticality coefficient k(eff) above 0.95;
- 6) release of large quantities of heat and radiation;
- 7) radioactive contamination of the nuclear power plant and its employees above the levels permitted by federal regulations;
- 8) radioactive contamination of the environment in the vicinity of the nuclear power plant above the levels permitted by federal regulations; and
- 9) radioactive contamination of humans and other living things in the vicinity of the nuclear power plant above the levels permitted by federal regulations.
B) during the PHE, collisions between groups of racks with each other and/or with the pool walls are expected to occur with results similar to those described in II(A) above.
A-3
APPENDIX B PGandE Exhibits Introduced Into Evidence 1.
PGandE Letter DCL-85-333, October 30, 1985; License Amendment Request 85-13, Reracking of Spent Fuel Pools.
2.
FGandE Letter DCL-85-306, September 19, 1985; Reracking Report.
3.
PGandE Letter DCL-86-019, January 28, 1986; Additional Information - Spent Fuel Pool Reracking.
4.
PGandE Letter DCL-86-067, March 11, 1986; Response to Questions on Spent Fuel Racks.
5.
PGandE Letter DCL-87-022, February 6, 1987; Rack Interaction Studies.
6.
PGandE Letter DCL-87-072, April 9,1987; Additional Information on Rack-to-Rack Interactions (Proprietary and Nonproprietary).
7.
PGandE Letter DCL-87-082, April 23, 1987; Three-Dimensional Studies.
8.
PGandE Letter DCL-87-115. Hay 18, 1987; Additional Information on Reracking Analysis.
9.
Seismic Analysis Report, Rev. 3, September 3, 1986.
- 10. NRC Standard Review Plan, Section 9.1.2, NUREG-0800.
- 11. NRC Standard Review Plan, Section 3.8.4, Appendix D, NUREG-0800.
- 12. NRC "0T Position for Review and Acceptance of Spent Fuel Storage and Handling Applications," April 14, 1978 (Supplemented January 18, 1979).
- 13. Distribution of Job Estimate, PGandE, NPG-041, February 14, 1984, i
B-1
f APPENDIX B (Continued)
NRC Staff Exhibits Introduced Into Evidence 1.
Safety Evaluation By the Office of Nuclear Reactor Regulation Relating Relating to the Reracking of the Spent Fuel Pools Amendment No. 8 to Unit 1 Facility Operating License No. DPR-80 Amendment No. 6 to Unit 2 Facility Operating License DPR-82, Pacific Gas and Electric Company, Docket Nos. 50-275-and 50-323.
2.
Evaluation of Spent Fuel Racks, Structural Analysis, Pacific Gas &
Electric Company, Diablo Canyon Units 1 and 2, Ter-C5506-625, Revised May 28, 1987.
3.
Evaluation of the Structural Adequacy of the Diablo Canyon High Density Spend Fuel Racks in Accommodating Multiple fuel Rack Impacts During the Postulated Hosgri Earthquake, May,1987.
4.
Environmental Assessment by the Office of Nuclear Reactor Regulation Relating to the Expansion of Spent Fuel Fools Facility Operating License Nos. DPR-80 and DPR-82 Pacific Gas and Electric Company, Diablo Canyon Units 1 and 2 Docket Nos. 50-275-and 50-323.
B-2
APPENDIX B (Continued)
Sierra Club Exhibits Introduced tilto Evidence I
I 1.
Board Notification 87-05, March 2*i,1987.
(Marked for Identification only)
J 2.
NRC Letter to PGandE on Boraflex June 2, 1987.
(Marked for Identification only) 3.
Misconsin Elesctric Letter to NRC on Boraflex, Jur.e 11, 1987.
(Marked for Identification only) 4, Affidavit of R. Clyde Herrick Regarding the Intervenor's Application for a Stay, June 30, 1986.
(Marked for Identification only) 5.
Meeting Summary - Structural Analysis of Spent Fuel Pool Expansion, Byror.
Station Units 1 and 2. USNRC, January 12, 1987.
6.
USNRC, " Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants," NUREG-0800, Revised July, 1961 ("SRP"),
Section 3.8.5.
7.
A. I. Soler and K. P. Singh, " Seismic Response of a free Standing Fuel Rack Construction to 3-D Floor Motion", Nuclear Engineering and Design 80 (1984), pp. 315-329 (" Seismic Response").
8.
A. I. Soler and K. P. Singh, " Dynamic Coupling in a Closely Spaced Two-Body System Vibrating in a Liquid Medium: The Case of Fuel Racks",
Proc. of the Third Conf. on Vibration in Nuclear Plants, British Nuclear Energy Soc., (1983), pp. 815-834.
9.
R. J. Fritz, "The Effect of Liquids on the Dynamic Motion of Immersed Solids", Jour. of Eng for Ind., February 1972, pp. 167-173 ("Fritz").
- 10. Summary Notes, Meeting of December 5, 1985. PGandE and the NRC et al.,.
i
- 11. Summary, Diablo Canyon Project Telephone Calls, March 8, 1985 -
June 30, 1986.
B-3
r
.}
P'
$I UNITED STATES OF AMERICA y.q P 4 '.16 NUCLEAR REGULATORY COMMISSION
)
DocketNos.50-hh!
"J In the Matter of
)
50-323
)
FACIFIC GAS AND ELECTRIC COMPANY )
(Reracking of Spent Fuel Pools)
)
(Diablo Canyon Nuc1 car Power
)
Plant Units 1 and 2)
)
_)
CERTIFICATE OF SERVICE I hereby certify that on June 29, 1987, copies of the following document in the above-captioned proceeding have been served on the following by deposit in the United States mail, first class, or as indi:ated by an asterisk through delivery by Federal Express: REVISED PROPOSED FINDINGS OF FACT AhD CONCLUSIONS OF LAH SUBMITTED BY PACIFIC GAS AND ELECTRIC COMPANY RELATING TO AN AMENDHENT REQUEST TO RERACX THE SPENT FUEL POOLS AT THE DIABLO CANYON NUCLEAR P0HER PLANT.
B. Paul Cotter, Jr., Chairman
- Docketing and Service Branch Administrative Judge Of fice of the Secretary Atomic Safety and Licensing U.S. Nuclear Rep;ulatory Comission Board Panel Hashington DC 20555 U.S. Nuclear Regulatory Comission (1 original plus 3 copies) 4350 East West Highway 4th Floor Bethesda HD 20814 Glenn 0. Bright
- Lawrence J. Chandler Esq.*
Administrative Judge Bejamin H. Vogler Esq.
Atomic Safety and Licensing Office of Executive Legal Director Board Panel U.S. Nuclear Regulatory Commission U.S. Nuclear Regulatory Comission Maryland National Bank Building 4350 East West Highway 4th Floor Room 9604 Bethesda HD 20814 7735 Old Georgetown Road Bethesda HD 20814 Dr. Jerry Harbour
- Lewis Sho11enberger Administrative Judge Regional Counsel Atomic Safety and Licensing U.S. Nuclear Regulatory Commission Board Panel Region V U.S. Nuclear Regulatory Comission 1450 Maria Lane, Suite 210 4350 East West Highway 4th Floor Halnut Creek CA 94596 Bethesda HD 20814 Atomic Safety and Licensing Edwin F. Lowry, Esq.*
Board Panel Grueneich fx Lowry U.S. Nuclear Regulatory Comission 380 Hayes Street, Suite 4 Hashington DC 20555 San Francisco CA 94102
4 Atomic Safety and Licensing Managing Editor Appeal Board Panel San Luis Obisco County U.S. Nuclear Regulatory Commission Telearam-Tri bung Washington DC 20555 1321 Johnson Avenue San Luis Obispo CA 93406 Mr. Lee M. Gustafson Richard E. Blankenburg Pacific Gas and Electric Company Co-publisher 1726 M Street NH Suite 1100 Hayne A. Soroyan, News Reporter Hashington DC 20036-4502 South County Publishing Company P. O. Box 460 Janice E. Kerr, Esq.
Arroyo Grande CA 93420 Public Utilities Commission 5246 State Building 350 McAllister Street San Francisco CA 94102 Dr. Richard B. Ferguson Sierra Club / Santa Lucia Chapter Rocky Canyon Star Route Creston CA 93432 Richard F. Locke Pacific Gas and Electric Company 77 Beale Street, 27th Floor San Francisco, CA 94106 Dated at San Francisco, California, this 29th day of June,1987.
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