ML20214T102
| ML20214T102 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon  |
| Issue date: | 06/04/1987 |
| From: | Atomic Safety and Licensing Board Panel |
| To: | |
| References | |
| CON-#287-3683 OLA, NUDOCS 8706100119 | |
| Download: ML20214T102 (78) | |
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1 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION
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3 BEFORETHEATOMICSAFETYANDLICENSINGBdkND4u,.
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)
Docket Nos. 50-275 otf).
5 In the Matter of
)
50-323
)
6 PACIFIC GAS AND ELECTRIC COMPANY )
(Reracking of Spent Fuel Pools)
)
7 (Diablo Canyon Nuclear Power
)
Plant Units 1 and 2)
)
June 4, 1987 8
)
9 10 TESTIMONY OF PACIFIC GAS AND ELECTRIC COMPANY'S 11 HITNESS PANEL ADDRESSING SIERRA CLUB CONTENTIONS I AND II 12 13 14 15 16 HITNESSES:
H0HARD V. GOLUB 17 RICHARD F. LOCKE J. D. Shiffer Pacific Gas and Electric Company 18 K. P. Singh P. O. Box 7442 S. Bhattacharya San Francisco, California 94120 19 E. E. DeMario (415) 781-4211 S. E. Turner 20 H. H. White BRUCE NORTON c/o R. F. Locke 21 P. O. Box 7442 San Francisco, California 94120 22 (415) 781-4211 23 Attorneys for Pacific Gas and Electric Company 24 25 26 0706100119 070604 AnoCKOD00gD DH
~
1 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION 2
BEFORE THE ATOMIC SAFETY AND LICENSING BOAR _D 3
4
)
Docket No. 50-275 5
In the Matter of
)
50-323
)
6 PACIFIC GAS AND ELECTRIC COMPANY )
(Reracking of Spent Fuel Pools)
)
7 (Diablo Canyon Nuclear Power
)
Plant Units 1 and 2)
)
June 4, 1987 8
)
9 TESTIMONY OF PACIFIC GAS AND ELECTRIC COMPANY'S 10 WITNESS PANEL
. ADDRESSING SIERRA CLUB CONTENTIONS I AND II 11 12 I.
INTRODUCTION AND BACKGROUND 13 Q1 Please state your name, affiliation, and qualifications.
14 Al My name is James D. Shiffer.
I am Vice President of Pacific Gas and 15 Electric Company's (PGandE's) Nuclear Power Generation Department.
I am 16 responsible for the operation, maintenance, and related construction 17 activities at Diablo Canyon.
I have more than 25 years of experience in 18 the nuclear power industry and I am a Registered Nuclear and Mechanical 19 Engineer in the State of California.
I am particularly familiar with 20 Diablo Canyon as a result of my involvement with the plant since.its 21 inception.
Prior to Diablo Canyon, I was involved with operational 22 activities for other nuclear plants including Vallecitos, Humboldt Bay, 23 and Ginna. A summary of my professional qualifications and experience is 24 provided in Attachment I and is incorporated herein by reference.
25 My name is Krishna P. Singh.
I am the President of Holtec 26 International and a Registered Professional Engineer in the State of m
f 1
Pennsylvania and Michigan.
I am responsible for the analyses and 2
qualification of the high density spent fuel storage racks at Diablo 3
Canyon.
Prior to November 1986, I was Vice President of Engineering at' 4
Joseph Oat Corporation, where I was responsible for all project 5
engineering and applied research and development activities. While at 6
Joseph Oat, my responsibilities included supervision of design, research, 7
and development in the area of pressure vessels, heat exchangers, and 8
related structures.
I also supervised the design and manufacture of high 9
density spent fuel storage racks for nuclear power plants, including 10 those to be used for Diablo Canyon. More specifically, I was responsible 11 for the design and licensing of high density racks used at the.following 12 nuclear plants: Quad Cities Units I and 2 Fermi Unit 2. V.C. Summer, 13 Oyster Creek, Rancho Seco, Pilgrim, and Grand Gulf Unit 1.
These racks 14 have been reviewed and licensed by the NRC.
I.have authored or 15 co-authored numerous publications on the design and analysis of various 16 structures and components of power plants, including spent fuel storage 17 racks.
I have also been granted a patent on the design of a spent fuel 18 storage rack. A summary of my professional qualifications and experience 19 is provided in Attachment 2 and is incorporated herein by reference.
20 Hy name is Shankar Bhattacharya.
I am a Senior Civil Engineer at 21 Pacific Gas and Electric Company. My responsibilities include the 22 seismic analysis and design related to the Diablo Canyon Power Plant, 23 including review of the spent fuel pool and the new high density spent 24 fuel racks. Additionally, my responsibilities include providing 25 assistance to the Engineering Group Supervisor of Civil Engineering in 26 supervision, direction, and execution of tasks related to the Diablo,
f 1
Canyon Power Plant.
Prior to my involvement with Diablo Canyon, I was a 2
design supervisor at Bechtel Corporation, where my responsibilities 3
included supervision and direction of engineering analysis and design 4
related to construction of power plants, including nuclear facilities.
5 Hhile at Bechtel, I supervised the design and licensing of high density 6
spent fuel racks for a nuclear power plant.
I am a Registered Civil and 7
Hechanical Engineer in the State of California. A summary of my 8
professional qualifications and experience is provided in Attachment 3 9
and is incorporated herein by reference.
10 Hy name is Edmund E. DeMario.
I am employed by Westinghous5 11 Electric Corporation as an Advisory Engineer in the Commercial Nuclear 12 Fuel Division.
I joined Hestinghouse in 1969, where I was responsible 13 for designing advanced fuel assemblies and performing analyses and tests 14 to evaluate fuel performance under various reac. tor conditions.
15 Particularly, I have been responsible for the mechanical design of 16 advanced fuel assemblies, including the 17 x 17 fuel assembly (the type 17 currently used at Diablo Canyon), the Vantage-5 fuel assembly, and the 18 Optimized fuel assembly.
I am also responsible for many new concepts in 19 the design of fuel assemblies and for the training of engineers in fuel 20 assembly design.
I have been granted six patents related to fuel 21 assembly design and have made approximately 20 additional patent 22 disclosures in the same area.
I am a Registered Professional Engineer in 23 the State of Pennsylvania. A summary of my professional qualifications 24 is provided in Attachment 4 and is incorporated herein by reference.
25 My name is Stanley E. Turner.
I am an employee of Black & Veatch, 26 Consulting Engineers, working as a Consultant and Project Manager in its,
1 Southern Science Office located in Dunedin, Florida.
I am a Registered 2
Nuclear Engineer in the State of Florida with over 30 years experience in 3
the nuclear fleid, encompassing the analyses and safety evaluations of a 4
wide variety of reactor types and configurations.
I was a member of the 5
American Nuclear Society (ANS) Standards Committee ANS-5, on Energy and 6
Fission Product Release.
I was also Chairman of the ANS-5.4 Standards 7
Committee on Fission Product Release from Oxide Fuel. At the present 8
time, I am a member of ANS Standards Committee 8.17 (Nuclear Criticality 9
Safety of Reactor Fuel Elements) and Chairman of the ANS 5.3 Committee 10 (Fission Product Release from Failed Fuel Elements).
I have been 11 actively engaged in the safety evaluations of new and spent fuel storage 12 facilities for twelve nuclear power stations. A summary of my 13 professional qualifications and experience is provided in Attachment 5 14 and is incorporated herein by reference.
15 Hy name is H1111am H. Hhite.
I am Chief Civil / Structural Engineer 16 for Bechtel Hestern Power Corporation and a Registered Civil Engineer in 17 the State of Oregon.
I was an Assistant Project Engineer in the Diablo 18 Canyon Project organization. My responsibilities included supervision 19 and direction of seismic-related engineering analyses for the Diablo 20 Canyon Project.
Prior to my involvement with Olablo Canyon, I was an 21 engineering specialist working in the area of seismic analysis for 22 several Dechtel projects.
Earlier, I was a structural engineer with the 23 Tennessee Valley Authority, where I was responsible for seismic analyses 24 of all Category I structures for a twin-unit nuclear power plant. A 25 summary of my professional qualifications and experience is provided in 26 and is incorporated herein by reference..
s
1 Q2 What contentions will you address?
2 A2 We will address Sierra Club Contentions I and II, which are:
3 4
Contention I 5
6 I(A).
It is the contention of the Sierra Club. Saata Lucia Chanter 7
(Sierra Club). that the report submitted to the Nuclear 8
Regulatory Commission (NRC) entitled Rerackina of Soent Fuel 9
Pools Diablo Canyon Units 1 and 2 and other communication!
10 hetween Pacific Gas and Electric Company (PGandE) and the NRC 11 which are available to the oublic on the same subjttt (the 12 ReRotts) fail to_Contain certain relevant data necessary for 13 indspandant verification of the claims made in the Reporti 14 regardina consistency of the croposed rerackiilg__with the 15 prolaction of the opblic health and safety. mid the environment.
16 17 In_Darticular. the Regotti fail to contain data reoardina:
18 19 3) the exgstted velocitv and dispheement of the spent fuel 20 gqols_(goolgl_A1_A_[ unction of time in three dimensions during 21 the_gottyhted_Rosgri earthquake (PHE);
22
- 4) the_HDeCigd maximum velocity and disolacement of the racts 23 oktAlagd from the_cnegater_m0dellina of rack _ behavior during the 24 Elig; 25
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26
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1 I(B).
It is the coritention of the Sierra Club that the Regorts fall to 2
include co,ntderation of certain relevant conditions. chenomena 3
and alternatives necessary for indeoendent verification of claims 4
made in theLRecorts regardina consistency of the proggitd 5
rerackina with oublic health and safety. and the environment. and 6
with federal law.
7 8
In carticular. the Reports fail to consider:
9 10 2) the resonant behavior of the spent fuel allemblies in the 11 Iteks in resDante to the PHE and the consiguengts_of_ingh 12 hahaylor; 13 7) alternative on-site storage facilities includina:
14 (1) construction of new or additional storagt_ facilities 15 and/or; 16 (11) ag_guisition of modular or mobile spent nuclear _luti 17 storage equioment. including sgent nuclear fuel storggt 18 ggib; 19 8) the use of anchors. braces. or--other structural member 1_to 20 orevent rack motion and suhitauent_ damage _durino the PHE; 21 9) the use of "bstaflex"_ntutron ahtorbing material for all 22 joent fuel _rAch.
23 24 Cofttention II 25 26 It is the coattatloftof._the _ Sierra Clyh_that_the_procated 1
reracking is inconsistent with the orQ1ertion of the oublir 2
health and safety. and the environment. for regions which includa 3
the following:
4 5
A) durina the PHE. collisions betyeen the racks and the_nggl 6
Walls are_ expected.to. gicur resultina in:
7
- 1) halet forces on_the_ racks sianificantly larger _than 8
those estimated in the reports; 9
- 2) impact forces on thLracks significaativ larger than 10 thote_exggs.ted_to_ damage the racht; 11 3)
.s i g n i f I e lat_Demancat_d eformAllon_and_o_ther_dAmagf_to 12 lbf_ tach 1_and_Dool walls; 13
- 4) redu c t i o n of_the_s D Ac i n a s b glyg en_[g el_Aistmbligt ;
14
- 5) intrtAs e_la_the_nD Cle ar_ Critic a l i t v_ c offf_tc ent_M efD 15 ahost_0 M ;
16
- 6) relcale of large_nuantities of heit_and_radlation; 17
- 7) radioattlynggatadnation of thLancleALpower_Dlaul 18 a n d_i t s_employteLAbov e t h e t elell_p cImitt e d_by_ff d eral 19 EcoulAtloni; 20 0) radlo a ttiyLcontamia a11oLof_the_e nv i ro nment_in_t he 2i y1einity_of_the_ayclear_poteLniant_ahayt_the_Letelt 22 parmttted_by_fsderal regulations; and 23 9)
CAdloactl1LCoatadnation__of humaas_aad_olher living 24 th}3 g s _1. n_the _ v i c i n i t y_ of_the_au cit ar_D owir_ plaat_A b oYe 25 the_Letels _p ermttle d_by_fid eraLtc oulA11oni.
26
- 0) du rinLihe_&lL_to111sions_b etxt en_ stoup s_o f_rA cks>1th.
1 nch other and/or with the cool walls are exgected to ottyr 2
with results similar to those dutdhtLin II(A) above.
3 4
Q3 What is the purpose of your testimony?
5 A3 The purpose of our testimony is to provide responses to the above 6
contentions.
This testimony will demonstrate that the models developed 7
and analyses performed to evaluate the adequacy of the proposed high 8
density spent fuel racks conservatively represent the response of the 9
fuel racks to the postulated Hosgri event, demonstrate the safety of the 10 storage racks, and confirm that the design of the racks meets applicable 11 regulatory requirements.
In addition, the testimony shows that 12 appropriate parameters have been considered or are bounded by those used 13 in the analyses, and that the results confirm that the fuel racks, pool 14 structures, and fuel assemblies satisfy appItc@le licensing criteria for 15 postulated loading conditions.
Since the fuel array remains subcritical 16 and no fuel cladding damage occurs for all postulated loading conditions, 17 there will be no resulting release of large quantitles of heat or 18 radioactivity, and therefore, the health and safety of the pubile will be 19 assured. The testimony also includes a discussion of the various methods 20 of increasing onsite storage capacity considered by PGandE.
21 Q4 Explain why Olablo Canyon Units 1 and 2 need increased spent fuel storage 22 capacity.
23 A4 Diablo Canyon Units 1 and 2 have separate spent fuel handling and storage 24 facilities. As presently configured and licensed, each unit has a spent 25 fuel pool (SFP) with storage capacity for 270 spent fuel assemblies.
The 26 existing low density racks at Diablo Canyon were designed, in accordance.
1 with AEC guidelines, to accommodate spent fuel discharged from one 2
refueling (roughly 70 assemblies), plus a reserve capacity of a full core 3
offload (193 assemblies) in the event a fu11 core discharge is 4
necessary.
Each pool currently contains spent fuel from the first 5
refueling outage, which occurred in late 1986 for Unit I and mid-1987 for 6
Unit 2.
Based upon operating schedules and the desirability of 7
maintaining full core discharge capability, it is necessary that the 8
spent fuel storage capacity for both units be increased. (Finding 1) 9 QS Why were the original racks designed and constructed to store only'270 10 spent fuel assembIles?
11 A5 Olablo Canyon, like most American nuclear power plants, was designed to 12 store a one-third core of spent fuel and maintain a full core discharge 13 capability, because it was expected that after an initial storage period 14 the spent fuel would be transferred to another. location for either 15 reprocessing or permanent disposal. However, due to a change in federal 16 policy in the 1970s, there is currently no operating commercial spent 17 fuel reprocessing facility in the United States.
In addition, because no 18 federal permanent disposal facility has yet been established, nuclear 19 plants throughout the country must increase their onsite storage capacity 20 to allow storage of spent fuel untti a disposal or reprocessing facility 21 is available.
22 06 Why was reracking using high density racks in the existing SFPs chosen as 23 the method to increase storage capacity?
24 A6 Reracking with high density racks was chosen because it was determined to 25 be a safe method of increasing onsite storage. Further, high density 26
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[
I reracking was determined to be the most prudent, reasonable, economical, 2
and timely method of the various techniques available to provide 3
increased storage. (Finding 2) 4 Q7 What safety, environmental, and design aspects did PGandE review and 5
ant. lyze in designing the high density spent fuel racks?
l 6
A7 PJandE analyzed and considered such issues as design, fabrication, 7
nuclear criticality, thermal /hydraulle analysis, structural analysis 8
including parametric studies, environmental impacts, and the cost-benefit 9
appraisal of the high density spent fuel racks.
In particular, PGandE 10 documented the analysis of the racks and pool structures under the specified combinations of inertial, seismic, and mechanical loads and 12 thermal gradients, and demonstrated compilance with acceptance criteria I3
(
in accordance with NRC guidelines.
Furthermore, PGandE described the I4 rack in-service surveillance program and the quality assurance program for the manufacture of the racks.
I0 In support of its application dated October 30, 1985 (PGandE Letter I7 No OCL-85-333), to the NRC seeking license amendments for Units 1 and 2 18 to increase spent fuel storage capacity (Exhibit 1), PGandE submitted a I9 report entitled "Reracking of Spent Fuel Pools--Olablo Canyon Units 1 20 and 2" (Reracking Report or Report), dated September 19, 1985 (PGandE 21 Letter No. DCL-85-306) (Exhibit 2).
PGandE subsequently submitted 22 further information pertinent to the issues at hand supplementing that 23 Report in PGandE Letters DCL-86-019 January 28, 1986 (Exhibit 3);
1 24 DCL-86-067, March 11, 1986 (Exhibit 4); OCL-87-022, February 6, 1987 25 (Eahlblt 5); DCL-87-072, April 9, 1987 (Exhibit 6); OCL-87-082, April 23 j
26 j
1937 (g,,1tbit 7); and DCL-87-Il5, May 18, 1987 (Exhibit 8).
PGandt also 1 I
I produced, as part of discovery in this proceeding, the Seismic Analysis 2
Report, Rev. 3, September 3, 1986 (Exhibit 9) (see PGandE Response to 3
Interrogatories from Sierra Club, October 3, 1986).
The analyses 4
described in Exhibits 2 and 9 are hereinafter referred to as the " design 5
basis analysis." Exhibits 1 through 9 are referred to as the " Reports."
6 h8 Hhat were the results of the safety and environmental analyses?
7 A8 PGandE's analyses demonstrated that the new racks in the spent fuel pools 8
meet all applicable NRC and ASHE criteria with acceptable margins of 9
safety.
The analyses also demonstrated that reracking was safe and the 10 most reasonable method for increasing capacity, and that neither'the Il reracking operation nor the increased onsite storage of spent fuel posed 12 an undue health or safety hazard to plant personnel, the public, or the 13 environment.
i 14 Q9 Old the NRC Staff concur with those results?
15 A9 Yes.
In accordance with its regulations governing procedures for review 16 of license amendment appilcations, the NRC Staff published a notice of 17 the requested amendments in the Federal Register and also published a 18 proposed finding of "no significant hazards consideration" (51 Federal 19 Register 1451, January 13, 1986).
Thereafter, the Staff issued its 20 Environmental Assessment and Finding of No Significant Impact on 21 May 29, 1986 (51 Federal Register 19430), and Final Determination of No 22 Significant Hazards Consideration on May 30, 1986 (51 rederal Register 23 20725-26).
In addition, the Staff issued its Safety Evaluation Report 24 (SER), dated May 30, 1986, discussing in detall the license amendments, 25 including the technical basis for the Staff's determination that the 26 proposed activities were acceptable. -
1 Those documents demonstrate that the NRC Staff concurred with 2
PGandE's conclusions that (1) there is reasonable assurance that the 3
health and safety of the public will not be endangered. (2) the proposed 4
reracking will be conducted in compilance with the Commission's 5
regulations, and (3) the issuance of the amendments will not be inimical 6
to the common defense and security or to the health and safety of the 7
public.
Since the Staff concluded that there were no significant 8
radiological or nonradiological impacts associated with the reracking 9
proposal, and that it would have no significant impact on the human 10 environment, an Environmental Impact Statement was not required to be 11 prepared.
12 13
!!. GENEBAL:_D ES IGLAM0EAL111S_Of_HIGILD ERSULRACKS 14 15 010 Describe the arrangement of the high density spent fuel racks proposed 16 for installation in the spent fuel pools.
17 A10 The high density spent fuel racks to be installed in the Olablo Canyon 18 fuel pools consist of a total of 16 racks of varying sizes for each 19 pool.
The racks have a total of 1324 fuel assembly storage cells, plus 20 10 miscellaneous storage locations.
The arrangement of the racks in the 21 pools is shown in Figures 1 and 2 for Units 1 and 2, respectively.
22 (Finding 3) 23 011 Decribe the high density spent fuel racks.
24 All The rack modules consist of a number of storage cells for storing fuel 25 assemblies, with the number of cells ranging from 24 to 110 por rack.
26 The individual storage cells have an 8.05-inch (nominal) square cross -
I section.
Each cell is sized to contain and protect a single 2
Westinghouse-type PHR 17 x 17 fuel assembly.
The storage cells are 3
arranged with a 10.93-inch center-to-center spacing in the rack modules.
4 Stainless steel gap channels are welded between the cells to provide a 5
" honeycomb" type structure.
The gap channels add considerable rigidity 6
and resistance to impact as well as to seismic loads resulting from the 7
postulated Hosgri earthquake.
8 The modules are freestanding, with no connection to the pool floor, 9
walls, or adjacent rack modules. The rack support feet rest on bearing 10 plates on the pool floor.
Each rack module is equipped with a girdle bar 11 on each side near the top. A typical rack module is shown in Figure 3.
12 The girdle bars protect the rack module from potential impact loads 13 resulting from rack-to-rack and rack-to-wall 14 interactions during a seismic event.
The girdle bars also maintain a 15 speelfled minimum gap between the cell walls of adjacent rack modules for 16 all loading conditions. (Findings 3, 4) 17 012 Olscuss the differences between the rrck modules in Region 1 and Region 2.
18 A12 The rack modulas are specifically designed for storage of spent fuel with 19 varying amounts of burnup.
The three modules (290 cells) designated as 20 Region I utilize neutron-absorbing material (Boraflex) on all sides of 21 individual storage cells.
These cells are designed for storage of new 22 fuel assembitos with enrichments up to 4.5 weight percent U-235 and spent 23 fuel that has not achieved a specified burnup. With a specified minimum 24 burnup (in Technical Specifications) for spent fuel stored in the 13 25 modules (1034 cells) designated as Region 2, the 10.93-inch 26 conter-to-conter cell spacing has boon maintained without the requirement -
{
l 1
for Boraflex as a neutron absorber. (Finding 5) 2 Q13 Describe the fuel pool structures at Olablo Canyon.
3 A13 The spent fuel pools are located on each end of the east side of the 4
auxiliary building.
Figure 4 shows the plant layout and foundation 5
elevations.
Each pool is approximately 35 feet wide, 37 feet long, and I
6 40 feet deep.
The normal water level in the pool provides a minimum of 7
23 feet of water above the top of the stored fuel.
The concrete pool 8
l walls are 6 feet thick except around the fuel transfer canal, where the 9
wall is 5 feet thick. The concrete foundation of the pool has a minimum 10 thickness of 5 feet and is founded on approximately 5 additional
- feet of 11 j
lean concrete placed directly on rock.
The spent fuel pool walls and 12 4
floor are lined with a stainless steel plate, which is 0.25 inches thick 4
j 13 on the floor and approximately 0.125 inches thick on the pool walls. No 14 structural changes to the spent fuel pool concrete structure or liner i
i 15 will be required to accommodate the high density spent fuel racks.
16 (Finding 6) 1 17 Q14 Do the fuel pool structures satisfy the structural criteria in the FSAR 18 j
Update when the added mass of fuel to be stored in the high density racks j
19 is considered?
(
20 A14 Yes, the fuel pool structures satisfy the criteria described in the FSAR i
l j
21 Update.
The fully loaded, high density spent fuel racks increase the L
4 r
j 22 overall mass of the auxillary building by less than I percent, Rack l
l 23 reaction loads are generally applied to the floor liner and concrete l
24 foundation via the rack support legs as vertical bearing and hortrontal I
25 shear loads.
The vertical loads are transferred to the bedrock by l
4 l
26 bearing; the horizontal loads are generally transferred to bedrock l
l 1
i
l i
i through frictional resistance between the rack support legs, bearing 2
plates, the liner, and the concrete foundation.
The liner plate and l
3 foundation were evaluated for the new loading conditions and found to be 4
adequate to support and transfer the high density rack reaction loads 5
(see Exhibit 3), (Finding 7) 6 Q15 What documents contain the criteria and guidance followed by the nuclear 7
industry for seismic design of high density spent fuel racks?
i 8
A15 The NRC has established acceptance criteria and design guidance for safe I
9 storage of spent fuel.
The seismic design criteria and guidance are i
10 primarily contained in Section 9.1.2, " Spent Fuel Storage," (Exhibit 10) and Section 3.8.4, "Other Seismic Category I Structures," Appendix 0, I
i II j
12
" Technical Position on Spent Fuel Racks," (Enhibit 11) of the Standard i
l 13 Review Plan (SRP), NUREG-0800, and in the NRC position paper "0T Position 1
i 14 for Review and Acceptance of Spent Fuel Storage and Handling 15 Appilcations," dated April 14, 1978, issued by the NRC to all power l
16 reactor licensees (Position Paper) (Exhibit 12). (Finding 8) l i
l 17 Q16 Olscuss the applicable seismic design criteria, j
18 A16 SRP Section 9.1.2 Paragraph !!!.3.a. requires that spent fuel storage l
17 racks be classified and designed to Seismic Category I requirements (see j
20 Exhibit 10).
The criteria for seismic design and fuel assembly impact i
21 loads are provided in Section IV (3) of the Position Paper (see
[
22 Exhibit 12).
Section IV (5) of the Position Paper states that SRP l
)
l 23 Section 3.8.4 provides acceptable procedures for modeling and analyzing 24 the seismic responses of the spent fuel racks (see Exhibit 11).
I 25 (Finding 9)
L 4
1 26 017 Olscuss the applicable allowable stress and structural acceptance l
j [
i l
1 I
criteria.
2 A17 Section IV (2) of the Position Paper identifies either of two industry 3
codes.Section III of the ASME Code or the AISC Manual. as being 4
acceptable for deriving the allowable stress criteria for the racks.
5 Other codes are acceptable based on a case-by-case review.
Structural 6
acceptance criteria are provided in Section IV (6) of the Position 7
Paper.
This section permits rack impacts and provides specific guidance 8
on how such impacts are to be incorporated in the rack design. (Finding 9) l 9
Q18 How did PGandE comply with the criteria for the seismic design of the 10 racks?
II A18 The Diablo Canyon high density racks comply with criteria appilcable to l
12 the seismic design as follows:
13 The racks were designed as Seismic Category I components in i
14 accordance with SRP Section 9.1.2, Paragraph !!!.3.a.
l 15 The allowable stress criteria for the racks were derived from the i
16 Subsection NF requirements of the ASME Code for Class 3 component 17 supports. Construction materials conform to Section !!!, Subsection l
18 NF of the ASME Code and were selected to be compatible with the fuel 19 pool environment.
20 i
The seismic excitation was simultaneously applied in three 21 orthogonal directions.
Increased damping of fuel racks due to 22 submergence in the spent fuel pool was not considered for 23 conservatism.
Local Impact of the fuel assemblies within the spent 24 fuel rack cells was considered in a manner which maximized forces t
25 acting on a rack module.
l 26 The procedures used for modeling and analyzing the seismic responses,
1 of the Diablo Canyon spent fuel racks were consistent with the 2
requirements of the Position Paper.
The models were developed based 3
on current engineering practices.
4 The possibility of gross sliding, tilting, and rack impacts under 5
the postulated Hosgri event were evaluated in accordance with the 6
acceptance criteria specified in Section IV (6) of the Position 7
Paper.
8 No exceptions to the acceptance criteria were taken for the design 9
of the Diablo Canyon high density spent fuel racks.
The racks were 10 designed and constructed using the approved acceptance criteria to 11 maintain the spent fuel assemblies in a safe configuration for normal and 12 abnormal loads, including potential irrpacts between racks and between the 13 racks and the fuel pool walls, which may occur during the postulated 14 Hosgri event (Exhibit 3). (Findings 10, 11) 15 Q19 Describe how the analytical model used by PGandE accounts for potential 16 rack motion during a Hosgri earthquake.
17 A19 The nonlinear dynamic model of the rack modules appropriately considered 18 the potential effects of the following possibilities:
movement of the 19 fuel assemblies, frictional resistance at the base of the rack, rack 20 sliding and rocking behavior, rack upitft and subsequent impact on the 21 bearing plate, and rack impacts with adjacent racks and pool walls.
In 22 addition to the potential rack movements addressed in the analysis, fluid 23 effects, known as hydrodynamic coupling, were also considered.
Figure 5 24 shows the analytical model used for the design basis calculations 25 (Exhibit 2, 9). (Finding 12) 26 Q20 How is the behavior of the fuel assemblies addressed in the seismic model? - -
1 A20 Potential impact of the fuel assemblies within storage cells was 2
simulated by impact springs (designated as K in Figure 5).
The impact g
3 spring constants were selected based on a series of parametric studies to 4
assess the impact of a fuel assembly on the cell wall of the rack. The 5
cumulative impact loads of the fuel assemblies on the rack module were 6
conservatively calculated by assuming that all fuel assemblies in the 7
rack move in unison; i.e., they will impact the cell walls of the rack at 8
the same instant and in the same direction. (Finding 13) 9 021 How is rack sliding behavior addressed in the seismic model?
10 A21 Springs K were included in the model near the base to simulate fhe g
11 sliding friction of the rack, whereas the springs K, c mbined with 6
12 the gap elements, simulate lift-off resulting from rocking.
The springs 13 K were included to capture the rotation of the leg in the vertical R
14 plane. The springs K were included in the model to determine the g
15 rack-to-rack and rack.o-wall impact forces.
16 Friction coefficients of 0.8 and 0.2 bound the experimental data and 17 were used in the analysis to maximize the inertial force and horizontal 18 displacement of the racks.
This wide range of friction values is 19 typically used in the industry for rack design. (Findings 14, 15) 20 022 How are fluid inertial effects addressed in the seismic model?
21 A22 The motion of the racks produces fluid inertial effects.
In particular, 22 the accelerating fluid mass results in two types of inertial effects. As 23 a rack starts to slide, the water inside and surrounding the rack is set 24 in motion.
This produces an additional inertial force on the rack, which 25 was addressed in the analysis by addir.g an appropriate amount of water 26 mass, known as " virtual mass," to the mass of the rack and fuel f
I assemblies.
The second effect of the accelerating fluid mass is 2
hydrodynamic coupling. As the space between moving racks or between the 3
racks and adjacent walls is reduced, the fluid between the bodies is 4
expelled from that space.
This causes fluid pressures to develop on the 5
surfac'es bounding the fluid mass, which retards the seismic motion of the 6
racks.
7 These inertial terms were calculated to assure conservative values 8
of rack displacements. The effects of the fluid motion on the rack 9
displacement are determined by the kinetic energy of the fluid.
By 10 underestimating the kinetic energy of the fluid, one necessarily~
11 ov.arestimates the rack displacements.
If the kinetic energy of the fluid 12 is ignored completely (e.g., assuming the absence of fluid), one will 13 grossly overestimate the rack displacements.
The calculation method used 14 for rack analysis includes fluid motion but underestimates the fluid 15 kinetic energy and, accordingly, overestimates rack displacements; i.e.,
16 the calculation method is conservative.
PGandE's use of virtual mass and 17 hydrodynamic coupling in these analyses is based on the fundamental 18 principles of fluid dynamics. (Finding 16) 19 Q23 Describe the analytical process that was used in the design of the racks.
20 A23 The analytical process used in the design of the racks consisted of:
21 Development of a nonlinear dynamic model of a rack module consisting 22 of inertial mass elements, hydrodynamic coupling, and gap and 23 friction elements.
24 Generation of the equations of motion and inertial coupling and 25 solution of the equations using the computer program DYNAHIS to 1
26 determine rack forces, moments, and displacements.. _.
-w
1 Computation of the detailed stress field in the rack (at the 2
critical locations) and in the support legs using the forces, 3
moments, and displacements calculated in the previous step.
4 (Finding 17) 5 024 Describe the cases considered and the results of the analyses performed.
6 A24 Using the methodology described above, PGandE calculated the loads on the 7
racks.
These calculations were performed in conformity with the loading 8
combinations and acceptance criteria specified in the NRC Staff's 9
Position Paper and Section 3.8.4, Appendix D, of the Standard Review 10 Plan. The loading combinations included the combined effects of dead 11 load, live load, thermal interaction within the pool, and seismic inertia 12 loads due to seismic events. A series of rack loading cases (fully 13 loaded -half full, etc.) was considered in order to establish bounding 14 design loads (Exhibits 2, 3, 4, 9).
15 The resulting stresses in the racks were determined to be lower than 16 the allowable stress values permitted by acceptance criteria.
The 17 allowable values provide a sufficient factor of safety when compared with 18 the ultimate capacity of the racks.
19 In addition to the design basis analysis, additional parametric t
20
-studies were performed using single-and multi-rack configurations in 21 response to NRC questions.
These parametric studies provided further I
22 verification of the adequacy of the design and analysis of the racks, and 23 of the safety margins that exist compared with the ultimate capacities l
24 (Exhibits 5, 6, 7, 8). (Finding 18) l 25 Q25 What conservatisms exist in the design basis analysis?
l 26 A25 The design basis analysis was performed with a single-rack model.
l '
1 Conservatisms were Suilt into the evaluations performed for the Diablo 2
Canyon fuel racks in terms of mo& ling assumptions, postulated loadings, 3
and safety margins on stress allowables.
Several of the conservatisms 4
inherent in the design basis analysis are discussed below.
5 Adjacent racks were assumed to move in a manner equal and opposite 6
(out of phase) to the rack module being analyzed, thereby maximizing 7
the potential for rack-to-rack impact.
8 A value of 4% damping was used between the fuel assemblies and 9
racks, between adjacent racks, and between racks and walls. The 10 analyses neglected fluid damping. A value of 10% for impact damping II (in addition to structural damping) has been used at other plants 12 licensed by the NRC.
13 The impacts between cell walls and the fuel assemblies were assumed 14 to occur in phase.
In reality, the fuel assemblies exhibit complex 15 and random behavior. However, they were all assumed to move in 16 unison se that the maximum response could be obtained.
17 The form drag opposing the motion of the racks within the pool water 18 was conservatively neglected.
19 The fluid coupling coefficients were calculated based on the 20 conservative assumption that the adjacent rows of racks are an 21 infinite distance away (the distance measured perpendicular to the 22 horizontal ground motion). This reduces the " cross-coupling effect" 23 of the adjacent rows of racks and yields conservative displacements i
24 and impact forces.
25 The calculation of fluid inertial effects included an underestimate l
26 of the fluid kinetic energy and resulted in a conservative l
l !
I overestimate of rack displacement.
2 Hydrodynamic coupling coefficents used in the analysis neglected 3
certain nonlinearities of the motion.
Studies in the literature 4
show that incorporation of these nonlinear effects would 5
significantly lower rack response (see Exhibits 2, 9). (Finding 19) 6 Q26 Have your analyses been reviewed by other qualified individuals?
7 A26 Yes.
Both the NRC Staff technical reviewers and their consultants have 8
reviewed the models, assumptions, analyses, and results. Their 9
conclusions agree with the appropriateness and adequacy of the anafyses.
10 Q27 How does your design basis analysis compare with others which have been 11 approved by the NRC?
12 A27 Industry practice of high density rack design and analysis was reviewed 13 for 10 other U.S. nuclear plants that have freestanding spent fuel 14 racks. The rack suppliers for these plants included Joseph Oat 15 Corporation; Exxon Nuclear Company; GCA Corporation, par Systems; Nuclear 16 Energy Services, Inc.; Hestinghouse Electric Corporation; and General 17 Electric Company.
18 The analytical techniques used for designing those racks were 19 similar to the techniques used for the Diablo Canyon rack analysis.
Some 20 examples of similarities are:
21 The racks were analyzed using the nonlinear time-history approach.
22 For seismic analysis, each rack was modeled using more than two 23 degrees of freedom. Gaps between racks and between racks and walls 24 were incorporated in the models.
25 The racks were evaluated for bounding friction coefficients of 0.2 26 and 0.8. [
/
1
~
- The racks were analyzed considering hydrodynamic coupling effects.
2 Impact damping of 4 to 10 percent was used to estimate fuel assembly 3
to rack impact forces. The Diablo Canyon rack design used a percent
)
4 impact damping for conservatism.
5 j-For those plants where the potential for rack impact exists, impact 6
analysis was performed using a nonlinear time-history approach.
7 He conclude from this review that the analytical methodology used 8
for design of the Diablo Canyon high density racks is similar to th(
9 NRC-accepted methodologies used by others in the industry. (Finding 20) 10 11 III. DISCUSSION OF CONTENTIONS 12
{
13 Contention I(A) 14 It is the contention of the Sierra Club. Santa Lucia Chanter. (Sierra 15 Club) that the reoort submitted to the Nuclear Reaulatory Commission 16 (NRC) entitled Rerackina of Soent Fuel Pools Diablo Canyon Units 1 and 2 17 and other communications between Pacific Gas and Electric Comoany 18 j
(PGandE) and the NRC which are available to the oublic on the same l
19 subiect (the Reports) fail to contain certain relevant data necessary for 20 indeoendent verification of the claims made in the Reoorts recardina 21 consistency of the pronosed rerackina with the orotection of the oublic 22 health and safety. and the environment.
23 5
24 Contention I(A)3) 25 In oarticular. the Reports fail to contain data reaardina the expected 26 velocity and disolacement of the soent fuel cools (nools) as a function
I of time in three dimensions durina the oostulated Hosari earthauake (PHE).
2 Q28 Do the Reports contain data regarding the velocity and displacement of 3
the fuel pools as a function of time in three dimensions for the 4
postulated Hosgri earthquake?
5 A28 No. This data is not necessary for review by the NRC Staff in evaluating 6
the technical adequacy of the rack design because the acceleration 7
time-histories are used for that purpose.
The velocity and displacement 8
time-histories of the pools were not reported because a record of such 9
data was not required during the design process. Consistent with seismic 10 design practices throughout the industry, the design process for~the 11 racks utilized the postulated Hosgri earthquake acceleration 12 time-histories for the base of the spent fuel pool. Velocity and 13 displacement information, nevertheless, can be derived from the 14 acceleration time-histories used in the design., The acceleration 15 time-histories which were used in the design of the racks are contained 16 in the Reracking Report, Figures 6.1.1, 6.1.2, and 6.1.3 (Exhibi t 2).
17 (Findings 21, 22) 18 19 Contention I(A)4) 20 In earticular. the Reoorts fail to contain data regardina the exoected 21 maximum velocity and disolacement of the racks obtained from the comouter 22 modelina of rack behavior durina the PHE.
23 Q29 Are the maximum velocity and displacement of the racks calculated in the 24 analyses and documented in the Reports?
25 A29 The maximum velocity of the racks is not documented in the Reports since 26 it is not a value needed for design of the racks.
The maximum '
1 displacement for a loaded rack module is included in the Reracking Report 2
in Table 6.8.2 (Exhibit 2).
The computer program algorithm used to 3
determine the rack loads and stresses internally computes rack velocity 4
and displacement at each time increment as part of the solution of the 5
equations of motion. While maximum relative displacement of the racks is 6
of interest (to identify any possible impacts), velocity is only of 7
secondary interest in the determination of acceleration response and 8
resulting loads on the racks.
Therefore, maximum rack displacements 9
relative to the pool structure were provided as output from the computer 10 program and documented in the Reports, but maximum velocity was 50t.
11 (Finding 23)
~
12 Q30 What is the maximum displacement of the fuel racks relative to the pool 13 structure?
14 A30 In the design basis analysis, the maximum displacement relative to the 15 pool structure of a loaded rack module is approximately 2.8 inches; the 16 maximum displacement relative to the pool structure of a nearly empty (11 17 fuel assemblies) rack module is approximately 4.2 inches as indicated in 18 the Seismic Analysis Report (Exhibit 2, 9). (Finding 24) 19 20 Contention I(B) 21 It is the contention of the Sierra Club that the Recorts fail to include 22 consideration of certain relevant conditions. ohenomena and alternatives 23 necessary for indeoendent verification of claims made in the Reoorts 24 regardino the consistency of the orocosed rerackino with oublic health 25 and safety. and the environment. and with federal law.
26
///. -.
f I
Contention I(B)2) 2 In carticular. the Reoorts fail to consider the resonant behavior of the 3
scent fuel assemblies in the racks in resoonse to the PHE and the 4
conseauences of such behavior.-
5 Q31 Describe the concept or condition referred to as resonant behavior.
6 A31 Resonant behavior, or resonance, is associated with forced vibration of a 7
linear system, where the amplitude of the response is a function of the 8
frequency of the applied force. At resonance, the amplitude of the 9
response of the linear system is limited only by the amount of damping 10 present.
Resonant behavior can occur when the frequency of a fofcing Il function (e.g., earthquake motion) is close to the natural frequency of a 12 linear system. Resonant behavior occurs when a repetitive forced 13 vibration of a given frequency acts on a system with a common natural 14 frequency and results in increased system response.
In a nonlinear 15 system such as the racks and fuel assemblies, at any given frequency, an 16 increased res onse may occur but any such increase is limited regardless 17 ofthedampin].
- 18 Q32 How does the ack analysis consider potential resonant behavior of the fl9 fuel assemblies?
'20 A32 The design basis analysis performed to evaluate the fuel racks utilized a 21 mathematical representation of the iarious components and their response 22 behavior. Since resonant behavior is a fundamental condition described 23 by the equations of motion, and since the equations of motion were 24 appropriately represented, the analysis considered the possibility of 25 resonant behavior. (Finding 25) 26 033 Here the dynamic properties of the fuel assemblies and racks _ _ _ _ _
~_
(
i 1
appropriately incorporated in the analysis of the fuel racks?
2 A33 Yes. As discussed above, the design basis analysis appropriately and 1
}
3 conservatively represented the dynamic behavior of the fuel assemblies' 4
and fuel ~ rack system.
It did so by developing and solving the equations 5
of motion of the system.
This included representation of the fuel 6
assemblies, fuel racks, fuel assembly to storage cell wall interface, and 7
rack-to-rack and rack-to-wall interfaces.
8 In addition, the mathematical modeling of the fuel assemblies and
]
9 fuel racks was reviewed and accepted by the NRC and its consultants.
I 10 Their review concluded that the nonlinear analytical design basil model 11 appropriately represented the dynamic and structural characteristics of 12 the system.
13 Q34 Can the spent fuel assemblies experience resonant behavior during the 14 Hosgri event?
15 A34 The design basis analysis demonstrated that, due to the specific j
16 conditions present, the fuel assemblies do not experience resonant u
l1 17 behavior. These conditions include the nonlinearities of the system 18 (including the presence of water, the movement of the fuel assemblies t
19 within the fuel racks, and the presence of friction at the fuel rack 20 base). The analysis appropriately represented these physical conditions j
21 and demonstrated that the integrity of the racks is maintained. As a 4
22 practical matter, resonance will not occur since the amplitude cannot 23 increase beyond the 0.302 inch clearance between the fuel assembly and 24 cell wall. (Finding 26) 1 l
25 035 What conditions would have to exist to achieve resonant behavior of the 1
26 fuel assemblies within the fuel racks during the seismic event?
P,
j
--- =
,f I[
j l
A35 For resonance to occur, the fuel assemblies would have to exhibit a 2
linear response to the seismic motion.
This would necessitate the 3
elimination of the effects of the pool water, the fuel assembly moving 4
within the fuel rack, and the friction interface at the base of the rack i
5 modules. Therefore, the conditions necessary for resonant behavior to 6
occur do not exist.-
i 7
l_
8 Contention I(B)7) 9 In oarticular. the Report fails to consider alternative onsite storage 10 facilities includina:
(1) construction of new or additional storage.
f Il facilities and/or: (ii) acauisition of modular or mobile soent nuclear 12 fuel storage eauioment. includina soent nuclear fuel storaae casks.
i-13 Q36 Did PGandE consider various methods of onsite storage?
I4 A36 Yes.
15 Q37 Did PGandE evaluate the two methods mentioned in the contention?
16 A37 Yes.
The evaluation was brief, however, since we did not judge these two 17 specific methods to offer any increase in safety over high density racks 18 and, in fact, there are significant disadvantages to each of these.
19 (Finding 27)
I 20 038 Why was the consideration of various onsite storage methods not 21 documented in the Reracking Report?
22 A38 A discussion of alternatives is documented in PGandE's Reracking Report, 4
23 Chapter 9 (Exhibit 2).
In particular, Diablo Canyon was originally i
24 designed to store spent fuel for a nominal period of one year and then i
25 ship the fuel offsite for reprocessing or disposal. Due to the 26 unavailability of fuel reprocessing facilities and of permanent disposal 4
i e
L..._,_..._-._.
1 sites, the spent fuel must now be stored for an extended period of time 2
at Diablo Canyon.
The alternatives that must be considered in addition 3
to onsite storage consist of various methods of storing the spent fuel 4
offsite or shutting down the reactor. Accordingly, the consideration of 5
alternatives, including offsite shipment of spent fuel and shutdown of 6
the reactor, was documented in the Reracking Report, Chapter 9.
- However, 7
the Reracking Report did not specifically address PGandE's consideration
- 8 of various onsite storage methods which were evaluated because, in our l
9 view, the expansion of the spent fuel pool capacity through reracking was 10 clearly superior. (Finding 28) 11 Q39 Why was the construction of new or additional storage facilities less 12 attractive?
13 A39 An additional storage pool was considered less attractive because such a 14 pool would not provide any added safety'for spent fuel storage above that 15 from storage in properly designed high density racks in the existing 16 pools. Moreover, the costs of constructing a new seismically qualified 17 structure and auxiliary support systems would obviously be very high 18 compared with reracking.
Finally, this would involve increased handling 19 of the spent fuel. (Finding 29) 20 Q40 Why was the acquisition of modular or mobile spent nuclear fuel storage 21 equipment, including spent nuclear fuel storage casks, less attractive?
22 A40 Acquiring modular storage equipment was less attractive because, in our 23 judgment, such equipment would not provide any added safety over and 24 above properly designed high density racks.
Further, modular equipment 25 such as dry cask storage was not a licensed concept at the time the 26 reracking decision was made by PGandE, and casks were still being in
I tested.
In any event, dry cask storage is not a viable option for Diablo 2
Canyon based upon the design of the dry casks currently available.
The 3
dry casks are designed to store only fuel that has been discharged from 4
the reactor at least five years prior to cask storage.
Thus, this 5
storage method could not be used for at least five years following the 6
first refueling outage.
7 The existing low density racks at Diablo Canyon were designed, in 8
accordance with early NRC guidelines, to accommodate spent fuel 9
discharged from one refueling (roughly 70 assemblies), plus a rese'rve 10 capacity of a full core offload (193 assemblies) in the event a full core discharge is necessary. The storage space associated with one-refueling 12 discharge is currently occupied at Diablo Canyon Units 1 and 2 after the 13 first refueling outages.
Based upon operating schedules and the 14 desirability of maintaining full core discharge capability, it is 15 necessary that the spent fuel storage capacity for both units be 16 increased. Further, the cost of the casks, assuming their availability, I
which would be required for the needed capacity at Diablo Canyon would be 18 high compared with the reracking alternative.
I9 At the time that PGandE made the reracking decision, there were no 20 plants in the United States using modular storage facilities for spent 21 fuel storage. Subsequently, two plants were licensed to use modular 22 storage facilities such as dry casks, but these plants did so only when 23 all of the storage space in existing pools had been filled after they had 24 previously reracked with high density racks. (Finding 30) 25 pjj 26 pjj I
Contention I(B)8) 2 In oarticular. the Reoorts fail to consider the use of anchors, braces.
3 or other structural members to orevent rack motion and subseauent damage 4
durino the PHE.
5 041 Is the use of anchors, braces, or other structural members to prevent 6
rack motion discussed in the Reports?
7 A41 The use of anchors, braces, or other structural members to prevent rack 8
motion is not discussed in the Reports since our analysis has shown that 9
freestanding racks meet safety requirements. (Finding 31) 10 042 Are structural anchors, braces, or other structural members requtred to II prevent rack motion and subsequent rack damage?
12 A42 No.
The freestanding racks satisfy NRC criteria and guidance applicable 13 to spent fuel storage racks.
The design accommodates the calculated rack 14 motion during the postulated Hosgri earthquake and shows that the racks 15 have sufficient safety margins.
In addition, freestanding racks have 16 several advantages over anchored or braced racks. (Finding 32) 17 Q43 What are the advantages of freestanding racks?
18 A43 Freestanding racks reduce the stress on the liner caused by thermal loads 19 from the heat generated by the spent fuel. Additionally, sliding 20 provides a very effective means to dissipate energy. A freestanding rack 21 is, therefore, considered to be a better design to absorb seismic energy 22 and, thus, has a distinct advantage over anchored or braced racks.
23 Further, no welding is required to install the freestanding racks.
24 Finally, inspection and/or replacement of racks, if necessary, is 25 simplified by the use of freestanding racks. (Finding 32) 26 lll.- _
I Contention I(B)9) 2 In oarticular. the Reports fail to consider the use of "boraflex" neutron 3
absorbina material for all soent fuel racks.
4
-Q44 Do the Diablo Canyon racks utilize Boraflex neutron-absorber in all spent 5
fuel racks?
6 A44 No. The Diablo Canyon spent fuel racks utilize the two-region concept in which one region (Region 1) is designed for new fuel or low burnup fuel, and a second larger region (Region 2) is d'esigned for fuel of a specified minimum burnup. The spent fuel racks were arranged in a safe and 10 cost-effective design utilizing the space available in the storage pool.
II (Finding 33) 12 Q45 What are the advantages of using a neutron-absorbing material such as I3 Boraflex in the spent fuel racks?
14 A45 The use of neutron-absorbing material between cells in the storage rack 15 reduces the spacing required between fuel assemblies while still assuring 16 an acceptable margin to criticality.
Both water-spacing and 17 neutron-absorbing material (or a combination) serve the purpose of 18 reducing the neutron multiplication factor to a safe level, and the use I9 of Boraflex, as in Region 1, allows more fuel to be stored in a given 20 space.
In Region 2, Boraflex is not needed since the spent fuel to be 21 stored is restricted to lower reactivity fuel of a specified minimum 22 burnup. (Finding 34) 23 Q46.Hould the use of Boraflex in Region 2 have afforded any advantage in 24 operation or safety?
25 A46_ No.
The racks were designed to provide the spent fuel storage capability 26 needed by PGandE.
Further reduction in spacing by using Boraflex in I
Region 2 was unnecessary.
Furthermore, both regions were designed to 2
provide an adequate safety margin to criticality and to satisfy NRC 3
acceptance criteria for criticality safety. (Finding 35) 4 5
Contention II(A) 6 It is the contention of the Sierra Club that the crocosed rerackina is 7
inconsistent with the orotection of the oublic health and safety. and the 0
environment. for reasons which include the following:
9 10 Contention II(A)1)
II Durina the PHE. collisions between the racks and the cool walls are 12 exoected to occur resultina in imoact forces on the racks sianificantiv I3 larger than those estimated in the Reoorts.
14 15 Contention II(A)2)
I6 Durina the PHE. collisions between the racks and the cool walls are exoected to occur resultina in imoact forces on the racks sianificantiv IO larger than those exoected to damaae the racks.
19 20 Contention II(A)3) 21 Durina the PHE. collisions between the racks and the cool walls are 22 exoected to occur resulting in sianificant oermanent deformation and 23 other damaae to the racks and cool walls.
24 047 During the postulated Hosgri event, will there be any collisions between 25 the racks and pool walls that will result in impact forces on the racks 26 significantly larger than those estimated in the Reports?
I A47 No.
The Reports describe the analyses performed for a wide range of 2
collision scenarios, and the resulting impact forces are expected to 3
bound those that could reasonably occur. (Finding 36) 4 Q48 During the postulated Hosgri event, will there be any collisions between 5
the racks and pool walls that will result in impact forces that could 6
cause significant deformation and other damage to the racks and pool 7
walls?
8 A48 No.
The calculated forces and stresses are less than those needed to 9
cause significant permanent deformation or other damage to the racks and 10 pool walls. Minor local damage to the liner, concrete pool wall; or II racks may occur, but such damage does not adversely affect the integrity 12 of the. fuel pool, racks, or fuel. (Finding 36) 13 049 What studies did you perform to further support your position that no 14 collisions between the racks and the pool walls are expected to occur 15 which would result in impact forces significantly larger than those 16 estimated in the Reports?
I7 A49 The design basis analysis and the techniques used to determine potential 18 impact forces were based on a conservative mathematical representation of I9 the racks to ensure an upper-bound impact force for rack-to-rack and 20 rack-to-wall impact.
Some of the conservative assumptions include:
21 Adjacent racks were assumed to move in a manner equal and opposite 22 (out of phase) to the rack module being analyzed.
23 A value of 4% damping was used between the fuel assemblies and 24 racks, between adjacent racks, and between racks and walls. The 25 analyses neglected fluid damping. A value of 10 percent for impact 26 damping (in addition to structural damping) has been used at other.
I plants licensed by the NRC.
2 The impacts between cell walls and the fuel assemblies were assumed e
3 to occur in phase.
In reality, the fuel assemblies exhibit complex 4
and random behavior.
However, they were all assumed to move in unison so that the maximum response could be obtained.
0 The form drag opposing the motion of the racks within the pool water was conservatively neglected.
8 The fluid coupling coefficients were calculated based on the I
conservative assumption that the adjacent rows of racks are an 10 infinite distance a say (the distance measured perpendicular ~to the II horizontal ground motion). This reduces the " cross-coupling effect" 12 of the adjacent rows of racks and yields conservative displacements I3 and impact forces.
~
14 The calculation of fluid inertial effects included an underestimate e
15 of the fluid kinetic energy and resulted in a conservative 16 overestimate of rack displacement.
4 I7 Hydrodynamic coupling coefficients used in the analysis neglected e
18 certain nonlinearities of the motion.
Studies in the literature I9 show that incorporation of these nonlinear effects would 20 significantly lower rack response.
21 In addition, the dynamic interaction between the pool wall and the 22 peripheral racks was considered in the rack-to-wall interaction.
The 23 evaluation showed that the rocking frequencies of the rack (approximately 24 10 cycles per second) are significantly different from the natural 25 frequencies of the pool walls (greater than 30 cycles per second).
Such 26 a large difference in the frequencies precludes the possibility for any I
significant amplification of the impact force. (Exhibit 4) 2 The results of additional parametric studies performed in response 3
to questions from the NRC further support this position.
Several different studies were performed (Exhibits 5, 6, 7) that included both simplified and complex two-dimensional, single-and multi-rack analytical models, as well as enhancements to the original design basis, three-dimensional, single-rack model.
The results of these studies confirm in all cases that rack impact loads and stresses due to the Hosgri earthquake are below allowable values. He have concluded with a 10 high degree of confidence that the design basis evaluation was conservative and the high density spent fuel racks satisfy acceptance 12 criteria and will maintain their integrity for the postulated Hosgri I3 event. (Finding 37) 14 050 What is the significance of the design basis impact forces and stress 15 ratios on the design of the racks?
I6 A50 While impact forces are important to the design process, of more I7 significance are stress ratios in that they better reflect the effect of 18 impacts on the racks.
The.ontrolling stress ratios for the racks have I9 an allowable value of 2.0.
The highest stress ratio for the impacts 20 determined from the design basis analysis was 1.436.
For the inpacts 21 determined from the parametric studies described in A49, the highest 22 stress ratio was 0.743 (Exhibit 7).
Thus, the design basis evaluations 23 were shown to be conservative and bounding. (Finding 38) 24 QS1 How were the rack components and the walls evaluated for impact forces?
25 A51 In evaluating the walls and the rack components, the impact forces were 26 conservatively assumed to be static.
No credit was taken for the short I
duration of the loading.
Stresses derived from these calculated forces 2
were significantly smaller than the stresses the racks and walls are 3
capable of withstanding without any adverse effect. (Finding 39) 4 052 Do you consider these calculated impact forces to be bounding values for 5
the postulated Hosgri event?
6 A52 Yes.
Because of the conservative assumptions and methods used to analyze rack-to-rack and rack-to-wall impact forces, we believe that the calculated impact forces on the racks bound those that might develop during the postulated Hosgri event. (Finding 40) 0 Q53 Discuss the capacity of the fuel racks to withstand impact force ~s.
I A53 If a rack should impact an adjacent rack or the wall, the impact force 12 would develop at the girdle bar or at the baseplate.
The fuel rack I3 strength at the girdle bar level is significantly greater than that I
required to resist the design loads. As the rack impacts the wall, the 15 rack girdle bars perpendicular to the wall would be loaded in compression I6 by direct bearing. These bars can sustain a direct impact load greater I7 than 175,000 pounds each before the onset of yielding, and incipient 18 failure is at least twice the yield force.
The impact resistance along I9 the girdle bar which impacts flat against the wall is greater than 20,000 20 pounds per storage cell.
The resistance of the baseplate is 21 substantially greater than that for the girdle bars. (Finding 41) 22 054 How were the effects of impact forces on the fuel racks and pool walls 23 evaluated?
24 A54 Impact forces result in loads applied to the rack girdle bars and 25 baseplates.
These forces were compared with the calculated capacities of 26 the girdle bars and baseplates.
For all cases, including straight-on 1
impacts and those where the corner of one rack impacts an adjacent rack 2
away from the corner, the calculated capacities are greater than the 3
maximum expected impact forces.
In addition, impact forces generate 4
stresses in the rack body and support feet.
For all important structural 5
members, stress levels were determined to be within acceptance criteria.
6 Finally, the effects of impacts between racks and the pool walls were 7
also evaluated.
Local and overall stresses were evaluated for both the 8
pool liner and concrete pool walls. While minor local liner or concrete 9
daformations may occur, overall stresses were found to be within 10 allowables. Thus, based on conformance with NRC acceptance crit ~eria, the II structural integrity of the spent fuel pool and racks is assured.
12 (Finding 42)
I3 055 Did you use as-built dimensions of the racks in evaluating impact forces?
14 A55 Yes.
The racks are rugged, stainless steel, honeycomb-type structures.
15 Although minor deviations from manufacturing tolerances may exist for I0 such components, parametric studies have shown that the effects of such I7 minor deviations are not significant and are accommodated by 18 conservatisms in the analysis methodology.
In addition, the racks are I9 fabricated from a very ductile steel which makes minor deviations i
20 insignificant. (Finding 43) 21 056 Does the design basis analysis indicate that the postulated Hosgri 22 earthquake could result in damage to the racks or fuel which would 23 adversely affect public health and safety?
24 A56 No. The design basis analysis shows that while there may be minor 25 deformation to the racks or pool walls, there would be no permanent 26 deformation or other damage that would lead to criticality, radiological. _ _ _ _ _ _ _ _ _ _ _.
I releases, damage to the fuel, increases in heat generation, or otherwise 2
adversely affect the public health and safety. (Finding 44) 3 4
Contention II(A)4) 5 Durina the PHE. collisions between the racks and the cool walls are 6
noected to occur resultino in reduction of the spacings between fuel 7
assemblies.
8 9
Contention II(A)S) 10 Durina the PHE. collisions between the racks and the cool walls ire II exoected to occur resultina in increase in the nuclear criticality 12 coefficient k(eff) above 0.95 I3 057 Describe the Diablo Canyon fuel assemblies.
14 A57 Each Diablo Canyon fuel assembly consists of a 17 x 17 array of 15 cylindrical rods of which 264 rods contain fuel pellets.
The assembly is I6 approximately 8.4 inches square and 13.3 feet in length.
Each fuel rod I7 is a Zircaloy tube containing uranium dioxide fuel pellets. Grids are 18 positioned at vertical intervals along the length of the fuel assembly to I9 maintain the rod spacing. (Finding 45) 20 058 What is the active fuel region within the fuel assembly and what is its 21 significance?
22 A58 The active fuel region is the region within the fuel assembly which 23 contains fuel pellets.
This region extends 144 inches, from 24 approximately 3 inches above the bottom of the fuel assembly nozzle, 25 which rests on the rack baseplate, to approximately 10 inches below the 26 rack girdle bar. (Finding 46).
I 059 Describe the fuel assembly spacings used in the criticality analyses and 2
the resulting spacings predicted by the seismic analysis.
3 A59 The center-to-center spacing of fuel assemblies in the racks is 4
maintained by the rigid rack structure at 10.93 inches, and this spacing, 5
with its tolerances, was used in the criticality analysis. (Finding 47) 6 Q60 Is there any reduction in spacing between fuel assemblies due to 7
collisions between the racks and the pool walls during the Hosgri event?
8 A60 No.
The design basis anatpis has confirmed that the maximum forces 9
generated by the postulated Hosgri earthquake will not result in a 10 reduction of the design spacing of 10.93 inches between the actife fuel II regions within any rack module. (Finding 47) 12 061 Discuss the criticality evaluations performed.
I3 A61 Criticality analyses were performed for the Diablo Canyon high density I4 spent fuel storage racks to assure that a k,ff equal to or less than 15 0.95 is maintained when the racks are fully loaded with fuel of the 16 highest anticipated reactivity in each of two regions and when the pool I7 is flooded with unborated water at a temperature corresponding to the 18 highest reactivity.
The maximum calculated reactivity includes a margin I9 for uncertainty in reactivity calculations and in mechanical tolerances, 20 statistically combined, such that the k,77 will be equal to or less 21 than 0.95 with a 95% probability at a 95% confidence level.
The details 22 and results of the criticality analyses are provided in Section 4.0 of 23 PGandE's Reracking Report (Exhibit 2). (Finding 48) 24 062 Does Diablo Canyon use borated water in the fuel pool?
25 A62 The Diablo Canyon spent fuel pools will be continually maintained at a 26 boron concentration of at least 2000 ppm, as required by the plant I
Technical Specifications.
This soluble boron not only provides an 2
additional and very large subtriticality margin under normal storage 3
conditions, but precludes the possibility of exceeding a k,77 of 0.95 4
under credible abnormal conditions, including the postulated Hosgri 5
event. (Finding 49) 6 063 Discuss the fuel assembly spacing requirements to maintain k,77 less 7
than 0.95 without borated water and with borated water.
8 A63 The spacing requirement to maintain k,77 less than 0.95 without borated 9
water is essentially the fuel assembly spacing in the rack design (10.93
~
10 inches), based upon the criticality analysis described in Section 4.0 of II PGandE's Reracking Report. With borated water normally present in the 12 spent fuel pool, the k,77 would not reach 0.95 until the water gap I3 between storage cells in Region 1 (nominally 1.786 inches) has been 14 reduced to less than 0.1 inch uniformly everywhere, an extremely 15 implausible condition. While analyses have demonstrated that significant 16 rack deformation would not occur, even if it were assumed that there was 17 zero gap betweer, storage cells, the resulting configuration would still 18 not be critical.
In Region 2, reducing the gap between cells to zero I9 from the nominal 1.9 inches would not result in k,77 exceeding 0.95.
20 (Finding 50) 21 Q64 Discuss the k,77 calculated for Diablo Canyon fuel racks with unborated 22 and borated water..
23 A64 Hith unborated water in the spent fuel pool, the highest k,77, 24 including an allowance for uncertainties and manufacturing tolerances, 25 was calculated to be 0.920 in Region I and 0.938 in Region 2, both based 26 upon conservative specifications of fuel enrichment and burnupt, and both _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _.
I providing subtriticality margins greater than that required by NRC 2
regulations.
With the normal concentration of soluble boron present 3
(2000 ppm), the safety margin below criticality is much larger, with the 4
maximum k,77 being less than 0.75 in both regions. (Finding 51) 5 065 Have you identified any collisions that could result in k(eff) exceeding 6
0.957 7
A65 No, there are no postulated collisions or plausible reductions in spacing 8
that could result in k,77 exceeding the limit of 0.95. (Finding 51) 9 10 Contention II(A16).
~
II Durina the PHE. collisions between the racks and the cool walls are 12 exoected to occur resultina in release of large cuantities of heat and _
13 radiation.
14 066 What conditions must exist before large quantities of heat and radiation 15 could be released?
16 A66 Any postulated condition that would cause the release of radiation would 17 require the fuel cladding to rupture; however, fuel cladding rupture 18 cannot occur unless the fuel assembly grids are crushed. (Finding 52) 19 067 Could the fuel assembly grids be crushed or the fuel cladding rupture 20 during a postulated Hosgri Earthquake?
21 A67 No.
The calculated impact forces are not large enough to cause crushing 22 of the grid and rupture of the cladding. (Finding 52) 23 068 During the postulated Hosgri event, what is the calculated impact force 24 on the fuel assemblies?
25 A68 During the postulated Hosgri event at Diablo Canyon Units 1 and 2, due to 26 the motion of the rack module relative to the motion of the fuel -
I assemblies, the fuel assemblies in the spent fuel pool storage racks 2
could contact the stainless steel walls of the storage rack cells.
The 3
maximum impact force on a fuel assembly grid has been calculated to be 4
approximately 1700 pounds, and the maximum fuel rod bending stress has 5
been calculated to be approximately 800 psi. (finding 53) 6 069 How was the structural integrity of the fuel assemblies evaluated?
7 A69 The structural integrity of the fuel assembly was evaluated by comparing 8
the calculated forces against capacity determined from analytical and 9
experimental data. Specifically, the maximum impact force on the grid, 10 the fuel rod bending stresses due to flexure, and the fuel rod 15 cal II contact forces at the grid supports were evaluated. (Finding 54) 12 070 What were the results of this evaluation?
I3 A70 Dynamic impact tests have been performed by Westinghouse on fuel assembly 14 grids for all Hestinghouse 17 x 17 fuel assembly designs to determine 15 their ultimate strength, which is the load at which incipient plastic 16 deformation of the grid cells occurs.
The evaluation showed that the I7 safety factor for the grids, which is defined as the ratio of the 18 ultimate grid strength (i.e., greater than 3400 pounds) divided by the I9 maximum impact force applied to the grid, was greater than 2.
20 The evaluation of fuel rod bending stresses showed that the ratio of 21 the fuel rod allowable stress limit (i.e., greater than 16000 psi) to the 22 maximum calculated stress during the Hosgri event is greater than 20 for 23 all Hestinghouse 17 x 17 fuel assembly designs.
24 The maximum local contact force that a fuel rod can sustain without 25 cladding failure was calculated by Hestinghouse employing finite element 26 analysis methods. A finite element model of the fuel rod was formulated -
I which consists of discrete elements, each of which has stress and 2
deflection characteristics defined by stress-strain theory.
The 3
calculated local stress levels caused by the reaction force were well 4
below the allowable stress levels in the fuel rods, ensuring that the 5
integrity of the fuel cladding will be maintained during the Hosgri event.
6 Thus, the integrity of fuel assemblies stored in the high density 7
spent fuel racks at Diablo Canyon will be maintained, and there can be no 8
resulting release of large quantitles of heat and radioactive material.
9 (Finding 55) 10 II Contention II(A)7) 12 Durina.the PHE. collisions between the racks and the cool walls are I3 exoected to occur resultina in radioactive contamination of the nuclear I4 power olant and its emolovees above the levels oermitted by federal 15 r12ulations.
16 II Contention II(A)8)_
IO Durina the PHE. collisions between the racks and the cool walls are I9 expected to occur resultina in radioactive contamination of the 20 environment in the vicinity of the nuclear oower olant above the levels 21 permitted by federal regulations.
22 23 Contention II(A)9) 24 Durina the PHE. collisions between the racks and the cool walls are 25 s tteted to occur resultina in radioactive contamination of humans and 26 other living things in the vicinity of the nuclear oower olant above the.
I levels oermitted by federal regulations.
2 Q71 Could collisions between the racks and the pool walls during the 3
postulated Hosgri earthquake result in radioactive contamination of 4
humans and other living things in the vicinity of the plant above the 5
levels permitted by federal regulations?
6 A71 No.
The racks have been qualified to withstand the impact loads which 7
may result from collisions between racks and pool walls during the 8
postulated Hosgri earthquake.
Therefore, no damage to the fuel would 9
occur, and there can be no resulting releases of large quantities of heat 10 and radioactive material. Additionally, the racks will maintain the fuel II assemblies in a subcritical configuration even during any such-12 collisions, and releases due to criticality in the pools cannot occur.
13 Consequently, no radioactive contamination of humans and other living 14 things in the vicinity of the plant above the levels permitted by federal 15 regulations could result from collisions between the racks and the pool 16 walls during the postulated Hosgri earthquake. (Finding 56) 17 l9 fantention II(8)
I9 Durina the PHE. collisions between arouos of racks with each other and/or 20 with the cool walls are exoected to occur with results similar to those 21 described in (II)(A) above.
22 072 How did PGandE's design basis analysis consider the group behavior of 23 racks?
24 A72 PGandE selected a comprehensive three-dimensional model of the rack being 25 analyzed with predefined motions of adjacent racks to simulate group 26 behavior.
To obtain upper-bound values of the rack seismic response, I
conservative (out-of-phase) motions of the adjacent racks were utilized 2
in conjunction with several previously mentioned conservative assumptions 3
in other aspects of the model.
This approach to bound rack loads and 4
stresses is consistent with industry practice.
5 073 Do you believe that movement of groups of racks as a unit is likely?
6 A73 Because of the dissimilarity of the racks (in terms of geometry, 7
tolerances, and gap spacings), it is highly unlikely that groups of racks 0
would move as a unit under a random seismic motion. (Finding 57) 9 074 Has PGandE performed any additional studies to confirm the adequacy of 10 the design basis analysis?
II A74 Yes.
The validity of the design basis analysis was confirmed by PGandE 12 by performing multi-rack studies (Exhibits 5, 6).
These parametric I3 studies demonstrated that the design basis analysis resulted in a 14 conservative rack design. (Finding 58) 15 075 Describe the parametric studies.
16 A75 PGandE has performed numerous parametric studies to investigate I7 multi-rack interactions. The studies utilized realistic modeling 18 assumptions and evaluated variations of all key parameters that may I9 affect the qualification of the racks. Some of these parameters include 20 t
loading of the racks, hydrodynamic coupling coefficients as they apply to 21 the specific location of the rack, manufacturing tolerances, and friction 22 coefficients.
The results of these studies show that the loads on the 23 racks are comparable to those predicted by the design basis analysis, 24 and, in all cases, these loads are significantly lower than the 25 allowables.
Thus, the parametric studies confirm that PGandE's modeling 26 assumptions in the design basis analysis adequately represent potential _.
I group behavior of the racks. (Finding 59) 2 076 Are there any collisions which could result in impact forces exceeding 3
those for which the racks were qualified?
4 A76 No. He believe that all potential collision conditions under the 5
postulated Hosgri event are bounded by the 1 cads for which the racks have 6
been qualified. (Finding 59)
Q77 Does this conclude your testimony?
8 A77 Ye5-9
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d 4
LIST OF EXHIBITS 1.
PGandE Letter DCL-85-333, October 30, 1985; License Amendment Request 85-13, Reracking of Spent Fuel Pools.
]
- 2..
PGandE 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.
E 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, May 18, 1987;. Additional Information on l
Reracking Analyses.
9.
Seismic Analysis Report, Rev. 3, September 3, 1986.
4
- 10. NRC Standard Review Plan, Section 9.1.2, NUREG-0800.
11.
NRC Standard Review Plan, Section 3.8.4, Appendix D, NUREG-0C00.
- 12. NRC OT Position for Review and Acceptance of Spent Fuel Storage and i
Handling Applications, April 14, 1978 (Supplemented January 18, 1979).
1 r
N 4
l l-
ATTACHMENT 1 PROFESSIONAL QUALIFICATIONS OF JAMES D. SHIFFER e
JAMES 0. SHIFFER VICE PRESIDENT, NUCLEAR POWER GENERATION EDUCATION a.
B.S. Chemical Engineering, Stanford University, 1960.
b.
H.S. Nuclear Engineering, Stanford University, 1961.
c.
Registered Professional Engineer, Hechanical and Nuclear, California.
EMPLOYMENT HISTORY - Joined PGandE in September 1961 a.
Summer 1959 and Summer 1960 - Employed by PGandE in summer engineer program. Assigned to Vallecitos Boiling Hater Reactor and Central Chemical Laboratory, b.
September 1961 to April 1962 - Engaged in Humboldt Bay Unit No. 3 startup preparation.
c.
April 1962 to July 1969 - Various assignments in power plant nuclear engineering, chemical engineering, and other technical operations at Humboldt Bay.
d.
July 1969 to July 1970 - Engaged at Humboldt Bay and at Company's General Office in Diablo Canyon startup preparation.
Includes a seven-month assignment in Rochester, New York, during startup and initial testing of R. E. Ginna PHR Plant.
e.
July 1970 to August 1971 - Engaged in Diablo Canyon startup preparation on Diablo Canyon Task Force.
f.
August 1971 to October 1978 - Assigned to Diablo Canyon as Power Plant
- Engineer, g.
November 1978 - Assigned to Diablo Canyon as Technical Assistant to the Plant Superintendent.
h.
February 1980 - Assigned to General Office as Manager of Nuclear Plant Operations.
1.
October 1984 - Assigned as Vice President of PGandE's Nuclear Power Generation Department.
NUCLEAR EXPERIENCE a.
Education - Masters degree thesis research involving operation of the Stanford Swimming Pool Reactor, irradiation of foils to determine reactor parameters, flux wire counting, and radiochemical work.
b.
Vallecitos - Assigned to Vallecitos for two summers.
Participated in the startup of the AVBWR plant.
c.
Humboldt Bay - Participated in prestartup activities including preparation of training material, initial loading, and low-level testing procedures and power testing procedures. Trained operating personnel for AEC license examinations.
Received an AEC Senior Operator's License.
Participated in preoperational testing of equipment and systems.
Directed initial loading and testing programs as shift nuclear engineer.
Directed the preparation of all reactor refueling procedures subsequent to initial fueling and directed the performance of this work on shift.
Responsible for the theoretical analyses of reactor core nuclear and thermal-hydraulic performance, plus evaluation of the performance of plant safeguard and other auxiliary equipment.
Provided technical advice and guidance for the chemical and radiation protection engineers and participated in the establishment and implementation of the chemical radiochemcial, and radiation protection programs at the plant.
d.
R. E. Ginna - Assigned to Ginna for seven months from July 1, 1969, to February 1970. Conducted training program for operators taking the AEC Operator's License examination.
Participated in the preparation and review of procedures and programs for initial loading, low-level physics testing, power operation testing, and radiochemical control.
Participated in initial loading, low-level physics testing, and power operation testing programs, e.
Diablo Canyon - Participated in the preparation and review of licensing material for Units 1 and 2 including the PSAR, FSAR, and Technical Specifications. Supervised staff of engineers (including persons experienced in nuclear engineering, instrumentation, radiation protection, and chemical engineering) engaged in the preparation of equipment operating and testing procedures, emergency plans, administrative procedures, and equipment specifications, and related material required prior to the startup of the plant.
FORMAL TRAINING COURSES a.
Stanford University Nuclear Engineering Curriculum as required by AEC Scholarship Program.
b.
Digital Computer Applications for Nuclear Reactor Calculations, UCLA Extension; Spring 1963.
c.
Diablo Canyon Design Lecture Series - Series of lectures given by designers of Diablo Canyon systems and equipment, Westinghouse APD; Winter 1971.
d.
In-Place Filter Testing Workshop, Harvard School of Public Health; fall 1971.
f' e.
Refresher Training in Radiological Engineering, General Electric Vallecitos Nuclear Center; Summer 1972.
f.
Short Course in Reactor Noise Analysis, University of Tennessee; Fall 1976.
g.
Simulator Training - Westinghouse Nuclear Training Center, Zion, Illinois. Option III (three-week course, 1974) and Option II (one week course, 1978).
h.
Metallurgy for Non-Metallurgists, one week course given by Center for Professional Advancement, August 1980.
4 98 o..
9 ATTACHMENT 2 PROFESSIONAL QUALIFICATIONS OF KRISHNA P. SINGH e
w
KRISHNA P. SINGH Address:
416 Gatewood Road, Cherry Hill, NJ 08003 Telephone:
(609) 428-8408 (Res.)
(609) 234-8668 (Office)
Education:
1968 to 1972 - Graduate School of Arts and Sciences. University of Pennsylvania. Philadelphia. Pennsylvania Ph.D. in Mechanical Engineering and Mechanics, (1972).
M.S. in Engineering Mechanics (1969).
Ph.D. Thesis " Contact Stresses in Elastic Bodies with Arbitrary P ofiles".
Scholastic Average of Graduate Studies:
Straight A (4.00 out of 4.00).
Support:
Fellowship of University of Pennsylvania.
1963 to 1967 - B.I.T. Sindri: University of Ranchi India B.S. In Mec'hanical Engineering Scholastic average of undergraduate studies:
First Class with Distinction (equivalent to straight A), ranked second in the Institute.
1961 to 1963 - St. Xavier's College. Ranchi. India.
B. Sc. I. in Science Grade average:
First Class with honors.
(Ranked second in the university among nearly 4,000 students).
Work Exoerience:
1967 - 1968:
Lecturer in the Applied Mechanics Department, Engineering College, Allahabad.
Taught courses in " Strength of Materials" and " Theory of Machines".
1969..(Systrl Engineer at Westinghouse Electric Corporation, Lester, Pennsylvania.
Responsibility:
Investigation of the effect of stator leakage in gas turbines.
19tv 1971 - November 1974:
Principal Engineer: Joseph Oat Corporation, Camden, New Jersey.
Responsibility: Stress analysis of pressure vessels, columns and heat exchangers. Authored several technical reports for static and dynamic stress analysis of process equipment used in nuclear power plants.
November 1974 to 1979:
Chief Engineer:
Joseph Oat Corporation. Supervision of design, research and development in the area of pressure vessels, heat exchangers and structures.
The group produced nearly 50 custom designs and specialized analyses per year.
November 1979 to 1986:
Vice President of Engineering: Joseph Oat Corporation.
Responsible for all project engineering, and applied research and development activities of the company.
November 1986 - Present:
President, Holtec International, Mount Laurel, N.J.
S.cholastic and Extracurriculars:
Award for "Overall Best Scholarship; at St. Xavier's College, Ranchi (1963)
Certificate of Merit from Bihar Secondary School Examination Board for finishing second in nearly 100,000 students (1961).
Elected Chairman, Debating Society, B.I.T. Sindri (1966-67)
Ranked seventh in India in the Indian Engineering Services competitive examination taken by several thousand engineers (1967)
Eunded Research:
Four contracts (totalling over $800,000) irom Electric Power Research Institute, (EPRI), Palo Alto, California to conduct research in various problems in power plant operation and maintenance. Hork performed on these contracts has not been published for contractual reasons.
Continuing Education Courses Taughj;:
Five day course given at I.I.T., Bombay, India on " Heat Exchanger Mechanical Design".
(1979)
Three courses in " Heat Exchanger Technology" given at Duke Power Company. (1902-83)
Three day course given at National Italian Reactor Authority, e
Genoa, Italy on " Nuclear Plant Equipment Design". (1984)
Professional Activities. Associations. etc.
a.
Registered Professional Engineer, Pennsylvania (1974), Michigan (80) b.
Fellow, A.S.M.E.
c.
Member, Pressure Vessel and Piping Subcommittee of A.S.M.E. Nuclear Engineering Division (1976) d.
Member, TEMA Technical committee (1976-1986) e:
Reviewer, A.S.M.E: transaction Journals f.
Member, Technical Committee on Nuclear Heat Exchangers of Heat Exchange Institute (1977-)
g.
Member, American Nuclear Society (1979-)
h.
Listed in: "Who's Who in Technology" (1979-)
1.
Member, American Association for Advancement of Science (1980-)
Eatents. etc.
a.
" Heat Exchanger for Hithstanding Cyclic Changes in Temperature" (with M. Holtz and A. Soler), Patent No. 4,207,944 (1980) b.
" Radioactive Fuel Cell Storage Rack" (with M. Holtz), U.S. Patent No. 4,382,060 (May, 1983)
Backi (authored or edited) a.
Mechanical Deslan of Heat Exchanaers and Pressure Vessel Comoonents, with A.I. Soler, Arcturus Publishers, Cherry H111, NJ, 1100 pages, hardbound (1984).
b.
"Feedwater Heater Workshop Proceedings", with Thomas Libs, EPRI 78-123 (1979) c.
"Feedwater Heater Procurement Guidelines", EPRI-CS-4155 (1985).
Publications 1.
"A Method for Solving Ill-Posed Integral Equations of the First Kind", (with B. Paul), Computer Methods in Applied Mechanics and Engineering 2 (1973) 339-348, 2.
Numerical Solutions of Non-Hertzian Elastic Contact Problems. (with B. Paul), Journal of Applied Mechanics, Vol. 41, No. 2, 484-490, June, 1974.
3.
"On the Inadequacy of Hertzian Solution of Two Dimensional Line Contact Problems", Journal of the Franklin Institute, Vol. 298, No. 2, 139-141 (1974).
4.
"How to Locate Impingment Plates in Tubular Heat Exchangers",
Hydrocarbon Processing, Vol. 10, 147-149 (1974).
4 1
4 5.
" Stress Concentration in Crowned Rollers", (with 8. Paul), Journal of Engineering for Industry, Trans. ASME, Vol.
97, Series B, No. 3, 990-994 (1975).
6.
" Application of Spiral Wound Gaskets for Leak Tight Joints",
Journal of Pressure Vessel Technology Trans. ASME, Vol. 97, Series J No. 1, 91-83, (1975).
7.
" Contact Stresses for Multiply-Connected Regions - The Case of Pitted Spheres", with B. Paul and W.S.
Woodward, Proceedings of the IUTAM Symposium on Contact Stresses. August 1974. Holland, Delft University Press, 264-281, (1976).
8.
" Design of Skirt-Mounted Supports", Hydrocart,on Processing, Vol. 4, 199-203, April 1976.
9.
" Predicting Flow Induced Vibration in U-Bend Regions of Heat Exchangers - An Engineering Solution", Journal of the Franklin Institute, Vol. 302, No. 2, 195-205, Aug. 1976.
- 10. "A Method to Design Shell-side Pressure Drop Constrained Tubular -
Heat Exchangers", with M. Holtz, Journal of Engineering for Power, Trans. of the ASME, Vol. 99, No. 3 July 1977, pp 441-448.
- 11. "An Efficient Design Method for Obround Pressure Vessels and Their End Closures", International Journal of Pressure Vessel and Piping, Vol. 5, 1977, pp 309-320.
- 12. " Analysis of Vertically Mounted Through-Tube Heat Exchangers",
Journal of Engineering for Power, Trans. ASME, Vol.100 No. 2, April 1978, pp 380-390.
- 13. " Study of Bolted Joint Integrity and Inter-Tube-Pass Leakage in U-tube Heat Exchangers: Part I - Analysis", Journal of Engineering for Power, Trans. ASME, Vol. 101, No. 1, pp 9-15 (1979)
- 14. " Study of Bolted Joint Integrity and Inter-Tube-Pass Leakage in U-Tube Heat Exchangers, Part II - Applications",~ Journal of Engineering for Power, Trans. ASME, Vol. 101, No. 1, pp 16-22 (1979).
- 15. "On Thermal Expansion Induced Stresses in U-Bends of Shell-and-Tube Heat Exchangers", (with Maurice Holtz); Trans. ASME, Journal of Engineering for Power, Vol. 101, No. 4, October, 1979, pp. 634-639,
- 16. " Heat Transfer Characteristics of a Generalized Olvided Flow Heat Exchanger", Proceedings of the Conference on Industrial Energy Conservation Technology, Houston, Texas, pp 88-97 (1979)
- 17. "An Approximate Analysis of Foundation Stresses in Horizontal Pressure Vessels", (with Vincent Luk), Paper No. 79-NE-1, Trans.
ASME, Journal of Engineering for Power, Vol.102, No. 3, pp 555-557, July, 1980. 1
- 18. " Generalization of the Split Flow Heat Exchanger Geometry for Enhanced Heat Transfer", (with Michael Holtz), AIChE. Symposium series 189, Vol. 75, pp 219-226 (1979)
- 19. " Analysis of Temperature Induced Stresses in the Body Bolts of Single Meeting, Paper No. 79 HA/NE-7.
- 20. " Optimization of Two-Stage Evaporators for Minimizing Rad-Waste Entrainment", (with Maurice Holtz) Journal of Hechanical Design, Trans. of the ASME, Vol.102, No. 4, pp 804-806 (1980)
- 21. "A Comparison of Thermal Performance of Two and Four Tube Pass Designs for Split Flow Shells", (with H.J. Holtz), Journal of Heat Transfer, Trans. of the ASME, Vol. 103, No. 1, pp 169-172, February, 1981
- 22. "A Method for Maximizing Support Leg Stress in a Pressure Vessel Mounted on Four Legs Subject to Homent and Lateral Loadings",
International Journal of Pressure Vessels and Piping, Vol. 9, No. 1, pp 11-25 (1981)
- 23. " Design, Stress Analysis and Operating Experience in Feedwater Heaters", (with Tom Libs), Proceedings of the Conference on Industrial Energy Conservation Technology, Houston, pp 113-118 (1980)
- 24. "On the Necessary Criteria for Stream Symmetric Tubular Heat Exchanger Geometries", Heat Transfer Engineering, Vol. 3, No. 1 (1981)
- 25. "Some Fundamental Relationships for Tubular Heat Exchanger Thermal Performance", Trans. ASME, Journal of Heat Transfer, Vol. 103, pp 573-578 (1981)
- 26. " Transient Swelling of Liquid Level During Pool Bolling in an Emergency Condenser", with J.P.
Gupta, Letters in Heat and Hass Transfer, Vol. 8, No. 1, pp 25-33, Jan/Feb., 1981.
- 27. "An Approximate Method for Evaluating the Temperature Field in Tubesheet Ligaments Under Steady State Conditions", with H. Holtz, Journal of Engineering for Power, Trans. ASME, Vol. 104, pp 895-900 (1982).
- 28. " Feasibility Study of a Multi-Purpose Computer Program to Optimize Power Cycles for Operative Plants", with Y. Henuchin and N. Hirota, Proceedings of the Conference on Industrial Energy Conservation Technology, Houston, (1981).
- 29. " Design Parameters Affecting Bolt Load in Ring Type Gasketed Joints", with A.I. Soler, Trans. ASME, Journal of Pressure Vessel Technology, Vol. 105, pp 11-13 (1983).
- 30. "A Design Concept for Minimizing Tubesheet Stress and Tubejoint Load in Fixed Tubesheet Heat Exchangers", with A.I. Soler, Trans.
ASME (c. 1982).
- 31. " Dynamic Coupling in a Closely Spaced Two-Body System Vibrating in Liquid Medium:
The Case of Fuel Racks", with A.I. Soler, Proceedings of the Third International Conference on " Vibration in Nuclear Plant", Keswick, England, May, 1982, pp. 815-834.
- 32. "Effect of Nonuniform Inlet Air Flow on Air Cooled Heat Exchanger Performance", with A.I. Soler and Lee Ng, Proceedings of the Joint ASME-JSME Heat Transfer Conference, 1983, pp. 537-542,
- 33. " Seismic Response of Free Standing Fuel Rack Constructions to 3-D Motions", with A.I. Soler, Nuclear Engineering and Design Vol. 80 (1984), pp. 315-329.
- 34. "A Method for Computing Maximum Hater Temperature in a Fuel Pool Containing Spent Nuclear Fuel", Nuclear Technology, American Nuclear Society (c. 1984).
- 35. "On Minimization of Radwaste Carry-Over in an N-stage Evaporator" with Maurice Holtz and Vincent Luk. Heat Transfer Engineering, pp.
68-73, Vol. 5, No. 1-1 (1984).
- 36. Feedwater Heater Procurement Guidelines - Some New Performance Criteria, Symposium on state-of-the-art Feedwater Heater Technology, EPRI (c. 1984),
- 37. " Method for Quantifying Heat Duty Derating due to Inter-Pass Leakage in Bolted Flat Cover Heat Exchangers", Heat Transfer Engineering, pp. 19-23, Vol. 4, No. 3-4 (1983).
- 38. "On Some Performance Parameters for Closed Feedwater Heaters, Proc.
of PVP Conference, ASME (c.1985).
- 39. "A Design Procedure for Evaluating the Tube Axial Load due to Thermal Effects in Multi-Pass Fixed Tubesheet Heat Exchangers",
(with A.I.
Soler), Journal of Pressure Vessel Technology, Trans.
ASME (c. 1986).
- 40. "An Elastic-Plastic Analysis of the Integral Tubesheet in U-tube Heat Exchangers - Towards an ASME Code Oriented Approach", Proc. of PVP Conference, ASME (c. 1985).
- 41. "Feedwater Heaters", invited paper, Proc. of the Joint NSF-ASI Conference, Poona, India (c. 1986).
- 42. " Surface Condensers", invited paper, Proc.
of the Joint NSF-ASI Conference, Poona, India (c. 1986). -
- 43. " Flow Induced Vibration", invited paper, Proc. of the Joint NSF-ASI Conference, Poona, India (c. 1986).
44 "rechanical Design of Heat Exchangers", invited paper, Proc of the Jair.t NSF-ASI Conference, Poona, India (c. 1986).
e 7-
V ATTACHMENT 3 PROFESSIONAL QUALIFICATIONS OF SHANKAR BHATTACHARYA a
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PROFESSIONAL 00ALIFICATIONS OF SHANKAR BHATTACHARYA My name is Shankar Bhattacharya.
I am a Senior Civil Engineer at Pacific Gas and Electric Company. Currently I am the deputy group supervisor of the Civil Engineering section of the Diablo Canyon Project. My responsibilities include providing assistance to the group supervisor in supervision, direction, and execution of all technical and administrative tasks related to the Diablo Canyon Power Plant.
I am a Registered Professional Civil (No.
C22884) and Mechanical (No, M016656) Engineer in the State of California and am a member of the American Society of Civil Engineers.
I am also a member of Standards Committees (Committee No. 2.2, 2.10 and 2.23) for the American
~
Nuclear Society.
My educational background includes:
- 8. Tech. (Hons.) Civil Engineering, Indian Institute of Technology; MS, Civil Engineering, Bucknell University, Pennsylvania; Doctoral Fellow in Civil Engineering at University of California, Berkeley; and MBA, University of California, Berkeley.
Prior to my involvement in the Olablo Canyon Project, I was a design supervisor at Bechtel Power Corporation's San Francisco Power Division office. My responsibilities for Bechtel included supervision and direction of engineering analysis and design related to construction of power plants, including nuclear facilities.
Earlier I was an associate civil engineer for the East Bay Municipal Utility District in Oakland, California, where I was responsible for static and dynamic analysis work associated with miscellaneous hydro projects.
ATTACHMENT 4 PROFESSIONAL QUALIFICATIONS Or EDMUND E. DEMARIO 5
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PROFESSIONAL QUALIFICATIONS AND EXPERIENCE OF E0fiUND E. DEMARIO Hy name is Edmund E. DeMario and my business address is P.O. Drawer R.
Columbia, S.C.
29250.
I am employed by the Hestinghouse Electric Corporation (Westinghouse) as an Advisory Engineer in the Commercial Nuclear Fuel Olvision (CNFD)
I graduated from the Stevens Institute of Technology with a Degree in Mechanical Engineering in 1960.
I also completed 20 credits of graduate study in Chemical Engineering at the Stevens Institue of Technology.
In September,1961. I joined General Dynamics at the Vandenburg Air Force Base as a Test Engineer,on the Atlas Missile program and in August, 1963 I joined Lockheed at the Vandenburg Air Force Base as a Test Engineer in the Agena Space Vechicle Program.
In February 1966, I joined Bell Aero Systems in Hheatfleid NY where I worked on the design and development of new rocket engines.
In January, 1969, I joined the Nuclear fuel Division of the Westinghouse Electrical Corporation as a Design Engineer where I was responsible for designing advanced fool assemblies and performing analyses and tests to evaluate the fuel performance under the various reactor conditions.
Af ter being promoted to the position of Fellow Engineer in 1976, I was subsequently promoted to the position of Advisory Engineer in December,1982, with the responsibility for the mechanical design of advanced fuel assemblies.
In addition I am responsible for the training of engineers in fuel assembly design.
I have been granted six patents related to the design of fuel assemblies and in addition I have made approximately twenty patent disclosures in the same area.
I have been responsible for the mechanical design of advanced fuel assemblies including the 17 x 17 fuel assembly (the type currently used at Diablo Canyon), the Vantage-5 and the Optimized Fuel Assembly.
I am also responsible for the design and development of nany new concepts in the mechanical design of fuel assemblies.
I am a Professional Engineer in the State of Pennsylvania.
ATTACHMffiI_5 i
PROFESSIONAL QUALIFICATIONS i
0F STANLEY E. TURNER T
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PROFESSIONAL QUALIFICATIONS OF DR. STANLEY __E.
TURNER EduC411on:
University of South Carolina. B.S. Chemistry, 1945 University of Texas, Ph.D., Nuclear Chemistry, 1951 PIofesilonal Expatiencei Southern Science Office of Black & Veatch, Engineers-Architects ProjectManager/ Consultant (1977-Present)
Dr. Turner is responsible for a wide range of scientific projects, including reactor nuclear analyses, radiological monitoring systems, assessment of alternate nuclear fuel cycles, combustible gas generation and control, and safety evaluations.
Over the past few years, Dr. Turner has been involved in the design, evaluation and licensing of high density spent fuel storage racks, including both criticalit assessment of radiological consequences.y analyses and Additionally, he has evaluated the core physics performance and isotope production rates of research, test, training, production, and power reactors. As Project Manager for several U.S. Arms Control and Olsarmament Agency programs, Dr. Turner has investigated posssible modifications to reactors for improved fuel utilization and has evaluated advanced PHR reactor concepts, involving extended fuel burnup, increased core regionalization and alternate methods of reactivity control.
NUS_ Corporation _ _ienior_.Contullaat_u923-1977)
Dr. Turner was Project Manager en numerous assignments. Among them were the assessment of post-LOCA hydrogen generation, and methods of control and development of specialized radiological monitoring systems; a survey of European nuclear fuel cycle plans and capabilities; generic review of public issues in the nations's nuclear power program; investigation of Halon-1301 for fire control and inhibition of hydrogen burning; a study of radiolytic decomposition of Halon; and a survey of U.S.
nuclear plant practice for foreign clients. His other work dealt with such activities as analytical physics support, evaluation of catalyst performance, and fission gas release and inventory calculations.
SoutherfLNucleat_Inginet11DQ,_ lac.. - Vice President,_Physict G26h19731 During his association with this company, Dr. Turner managed and participated in a number of projects which involved assessing tritium production and control methods; performing calculations of heavy isotopo production; reviewing Itcensing
I documents; preparing operating procedures; performing safety assessment of large, special purpose reactors; evaluating consequences of industrial sabotage in nuclear power plants; assessing reactors for maritime application; and evaluating fuel cycle economics.
GengIg1 Nuclear Enaineerina - Senior Reactor Physicist (1957-11ft41 Dr. Turner performed or directed most of the fuel cycle cost evaluations, heavy isotope analysis, and fuel management work l
performed by this company. He planned and coordinated various experiments and testing programs, and managed research and development activities related to advanced nuclear fuel elements.
In addition, he participated in plant licensing actions and safety reviews, and served as a member of the Safety Committee for an operating nuclear power plant.
Sotony-flobil_Rei.eAECh_LaboILtory - Physicist (1952-1957)
Dr. Turner performed research in radiological methods for oil >
exploration,' including radiation measurements and field tests.
~
lLSJay3_Ridiological Defense Laboratory - Physicist (1951-11521 Dr. Turner performed research in the consequences and methods of defending against nuclear bomb detonations, including fleid tests and radiological measurements, l
Honorary Sotittica:
Sigma Pi Sigma, Phi Lambda Epsilon, Blue Key, Sigma Xi P_rofestional 6f filiations:
Dr. Turner is a member of the ANS Standards Committee 8.17 on Nuclear Criticality Safety, and Chairman of ANS 5.3 and 5.4 Working Groups on Fission Product Release. He was formerly a member of the ANS 5 Committee on Decay Heat and contributed to the formulation of the standard on fission product decay heat.
Registered Professional Nuclear Engineer:
Florida, No. 22862.
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ATTACHMENT 6 PROFESS 10NAL QUAllFICAT10NS OF HILLIAH H. NHITE e
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PROFESSIONAL 00ALIFICATIONS OF HILLIAM H. HHITE My name is Hilliam H. White.
I was an Assistant Project Engineer in the Diablo Canyon integrated organization consisting of Pacific Gas and Electric Company and Bechtel Power Corporation employees. My responsibilities included supervision and direction of seismic-related engineering analyses for the Diablo Canyon Unit 1 Project Engineering Organization. Currently I am the Chief Civil / Structural engineer for the San Francisco office of Bechtel Western Power Corporation.
I am a Registered Professional Civil Engineer in Oregon and member of the American Society of Civil Engineers.
My educational background includes:
BS, Civil Engineering, University of Idaho; MS, Civil Engineering, Univerrity of Colorado; PhD, Civil Engineering, University of Colorado.
Prior to my involvement in Diablo Canyon, I was an engineering specialist with Bechtel's San Francisco Power Division working with the Chief Civil Engineer's staff in the area of seismic analysis for several Bechtel projects.
Earlier, I was a Structural Engineer with the Tennessee Valley Authority, where I was responsible for seismic analysis of all Category I structures for a twin-unit nuclear power plant, including seismic input for the design of the nuclear steam supply system.
N
'E, b
g.,
s I was an Assistant Professor at Oregon State University, where I taught undergraduate and graduate courses in structural mechanics and analysis and computer applications.
I performed a special study for Bechtel on soil-structure interaction for the proposed Mendocino nuclear power plant while teaching at Oregon State Un'iversity.
, s y,
While employed at the Bettis Atomic Power Laboratory, I was a Senior s.s Engineer working on shock analysis of nuclear reactors aboard submarines and was involved in programs to assess the shock resistance of reactor internals subjected to long-term irradiation damage.
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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION Docket Nos. 50-275 g7 g -8 P3 :28
)
In the Matter of
)
50-323' (RerackingofSpentF6$ Pools):~hNi PACIFIC GAS AND ELECTRIC COMPANY )
)
pe (Diablo Canyon Nuclear Power
)
Plant Units 1 and 2)
)
)
CERTIFICATE OF SERVICE I hereby certify that on June 4,1987, 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 indicated by an asterisk through delivery by Federal Express: TESTIMONY OF PACIFIC GAS AND ELECTRIC COMPANY'S HITNESS PANEL ADDRESSING SIERRA CLUB CONTENTIONS I AND II.
B. Paul Cotter, Jr., Chairman
- Lawrence Chandler, Esq.*
Administrative Judge Benjamin H. Vogler, Esq.
Atomic Safety and Licensing Office of Executive Legal Director Board Panel U.S. Nuclear Regulatory Commission U.S. Nuclear Regulatory Commission Maryland National Bank Building 4350 East West Highway 4th Floor Room 9604 Bethesda MD 20814 7735 Old Georgetown Road Bethesda MD 20814 Glenn 0. Bright
- Diane M. Grueneich*
Administrative Judge Grueneich & Lowry Atomic Safety and Licensing 380 Hayes Street, Suite 4 Board Panel San Francisco CA 94102 U.S. Nuclear Regulatory Commission 4350 East Hest Highway 4th Floor Docketing and Service Branch Bethesda MD 20814 Office of the Secretary U.S. Nuclear Regulatory Commission Dr. Jerry Harbour
- Administrative Judge Atomic Safety and Licensing Board Panel U.S. Nuclear Regulatory Commission 4350 East West Highway 4th Floor Bethesda MD 20814 16I M
Richard F. Locke Pacific Gas and Electric Company 77 Beale Street, 31st Floor San Francisco, CA 94106 Dated at San Francisco, California, this 4th day of June,1987.
.