ML20052G127
| ML20052G127 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 05/10/1982 |
| From: | Sacramo R CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.), NUS CORP. |
| To: | |
| Shared Package | |
| ML20052G119 | List: |
| References | |
| ISSUANCES-OLA, NUDOCS 8205140333 | |
| Download: ML20052G127 (8) | |
Text
~
w UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICFNSING BOARD In the Matter of
)
) Docket No. 50-155-OLA CONSUMERS POWER COMPANY
) (Spent Fuel Poc2.
)
Modification)
(Big Rock Point Nuclear Power Plant)
)
TESTIMONY OF RAYMOND F.
SACRAMO CONCERNING POSSIBLE DISTORTION OF THE SPENT FUEL POOL RACKS (O'NEILL CONTENTION IIE-3)
My name is Raymond F.
Sacramo.
I reside at 17204 Chiswell Road, Poolesville, Maryland.
I have been employed with NUS Corporation, an engineering consulting firm in Gaithersburg, Maryland, since October 1, 1977.
My educational background and work experience are explained fully in Attach-ment A to this testimony.
I am technically qualified to address, in part, the issue which is discussed by the Licens-ing Board at page 5 of its February 5, 1982, Memorandum and-Order concerning O'Neill Contention IIE-3, namely:
Whether or not the spent fuel pool racks for the Big Rock Point spent fuel pool might be distorted to the point of adversely affecting criti-cality because of either a drop of a fuel assembly or heating of the pool.
8205140333 820510 PDR ADOCK 05000155 T
. My testimony will set forth my conclusions as to the nature of the distortion of the racks that could occur as a result'of a fuel assembly drop or heating of the pool.
This information will be provided to Dr. Y.
S. Kim who will determine the effect, if any, on the criticality analysis for the proposed expansion of the Big Rock Point spent fuel pool.
Figure 1 describes the spent fuel racks to be used at the Big Rock Point plant.
A rack consists of square stainless steel cans made of 1/4-inch thick type 304 stainless steel.
Each can is approx 4mately 7 feet long.
A 1/2-inch thick plate is welded at the bottom of each can to provide support for the fuel assembly.
A five-inch diameter opening in this fuel support plate allows cooling water to flow upward through the fuel assembly to provide for the removal of the decay heat from the fuel element.
These stainless steel cans are welded to a gridded base comprised of a lattice of stain-less steel beams 1-1/4 inches thick and 4 inches wide.
The storage cans are structurally tied at the top of the rack with an over-under stainless steel grid system.
Adjustable legs are provided to accommodate unevenness in the pool floor a-d to allow natural circulation flow between the base of the rack and the floor.
The region of a fuel assembly containing the uranium pellets is stored within and along the " active length" (70
,c x
. inches) of the storage can as shown on Figure 1.
The active length area of the various storage cans and their spacing affects the criticality analysis.
For Big Rock Point, the design parameter of interest is the 9-inch spacing on center-to-center distance between the fuel storage cans over their active length.
I vil.1 first determine the extent, if any, of distor-tion of the rack over the active length of the stored fuel assembly due to a drop of a fuel assembly.
Two postulated fuel assembly drop cases were analyzed.
I have re-examined this evaluation, which appears on pages 5-8, 5-9, 5-10, and 4-7 of Revision 1 to Consumers Power Company's application, entitled " Consumers Power Company, Big Rock Point Spent Fuel Rack Addition, Consolidated Environmental Impact Evaluation and Description and Safety Analysis," dated April 1982.
I have confirmed the adequacy of this evaluation.
The first case postulates a drop of a fuel assembly (383 lbs.) from the auxiliary hook of the overhead crane, a maximum drop height of 42 feet and 4 inches (508 inches).
The fuel assembly would fall into the fuel pool, and it is postu-lated to pass through an empty storage can and impact the fuel assembly support plate at the bottom of the can.
The kinetic energy of the falling fuel assembly at impact is 195,000 in-lbs.
This energy would cause significant distortion in the
. fuel assembly support plate.
However, stresses above the support plate in the active fuel length region, when combined with other design load conditions, would be below design allowables.
Thus, the center-to-center distance between the storage cans is maintained.
The second case postulates the drop of a fuel assem-bly from the upper limit of the crane auxiliary hook (424 inches) to the top of the fuel rack lead-in guides between two adjacent storage cans.
162,400 in-lbs. of energy is released when the fuel assembly strikes the lead-in guides.
The lead-in guides would be crushed, and the storage cans above the top grids would also be crushed by approximately 7/8 ' inch.
Based on my analysis and re-evaluation, the forces transmitted below the crushed area through the storage can side plates are well below those which would cause lateral distortions in the storage cans.
After striking the lead-in guides, the fuel assembly would fall flat on top of the fuel rack.
The addi-tional energy to be dissipated, 15,300 in-lbs., is very small compared to that already discussed and would not cause any additional damage.
As in the first impact case, local distortions to the rack would occur, but under both cases only small anounts of energy are transmitted to the active length of stored fuel regions, resulting in no change in the rack geometry in these regions.
-4
. I will now discuss the effect of heat on the racks.
The free-standing stainless steel racks will expand as the water temperature increases in the pool.
The maximum:expan-sion can be determined by assuming a worst-case temperature differential of 167*F, that is, the difference between an ambient temperature of 70*F and the mos4 extreme temperature under pool boiling conditions of 237*F.
The 167 F temperature differential, when multiplied by the coefficient of thermal expansion for stainless steel (9.60 x 10-6*F in/in) and the center-to-center nominal distance of 9 inches, conservatively results in an increase to the center-to-center distance between stored fuel assemblies of approximately 0.015 inches over the nominal pitch of 9 inches.
The foregoing results of my evaluation have been provided to Dr.
Y.
S. Kim.
His testimony will discuss the significance of this information on the criticality analysis prepared for the expansion of the Big Rock Point spent fuel pool.
RAYMONiiF. SACRAMO EDUCATION University of Pittsburgh, M.S., Mechanical Engineering,1977 Penn Stain University, B.S., Mechanical Engineering,1974 Orexel University. Civil Engineering,1967 EXPERIENCE NUS CORPORATION,1977-Present Westinghouse Electric Corporation, 1974-1977 United States Army, 1968-1970 Beaumont Birch Company, 1966-1967 NUS - Responsible for directing the analytical efforts required to support Engineering Division design projects. These projects cover a variety of modifications and new system incorporations at more than 25 nuclear power plants resulting from TMI, retrofit, or NRC 1.E. Bulletin nuclear-industry-related programs. Associated analytical efforts include various levels of finite element modeling using both static and dynamic analysis techniques, as required, to investigate normal, upset, and emergency loading conditions imposed on mechanical or electncal systems and associated support structures.
Responsible for developing finite element models and performing structural integrity stress anal-yses for ASME B&PV Code Class 11 and ill piping systems; safety-related components; and high-density spent fuel racks and spent-fuel-pool concrete structures. Analyses have employed static response spectrum and nonlinear displacement time-history analysis techniques. Arialyses are performed in accordance with the rules and requirements of the appropriate ASME, AISC, or ACI Codes. Also responsible for the fabrication followup efforts for high-density spent-fuel racks designed by NUS.
Nestinghouse Electric Company - On the Clinch River Breeder Reactor Project, was respon-sible for assessing the mechanical reliability and integrity analyses on mechanical components interfacing with the shutdown system. Responsible for the development of design support docu-ments for the reactor upper internal structure, core support structure, bypass flow and lower inlet modules, head heating and cooling systems, and riser assemblies. The documents considered statistical evaluations of thermal, seismic, and accident-condition transients and predicted the failure probabilities of structural-integrity related failure modes resulting from these transients.
The documents were presented at final design reviews and were responsible for design changes that decreased the probabilities associated with component failure modes, which affected the mechanical shutdown system reliability.
l United States Army - Assigned to perform mechanical design and drafting.
Beaumont Birch Company - Performed drafting and mechanical design.
MEMBERSHIP American Society of Mechanical Engineers c.
E' e
NUS COAPO AATION
RAYMOND F. SACRAMO Page Two PUSUCATIONS
" Statistical Fatigue Testing and Analysis of 304 Stainless Steel " Thesis, University of Pittsburgh, 1977.
" Reactor Scram Experience for Shutdown System Reliability Analysis" (coauthor) American Nuclear Society Technical Paper, June 1976.
"Probabilistic Assessment of Primary Piping Integrity" (coauthor) American Nuclear Society Technical Paper, April 1976.
NUS COAPOAATICN
7%* 4(AD=/A/ GU/Dd s1 13 CL
/
fast CAAf l 4{,
I*,N+T
'N 1
.=.
8 l
y 1
g'-
D1hf' g
5 g
a Si sy-b N s,s, pI#,V Y,
A
""la"l'
%ng,
,/
u
,,nar part on asu sanansxem p, ggg,
f/Q AoCK 0/dy pf/M NEL SM
- -... - -. - - -