ML19352B217
| ML19352B217 | |
| Person / Time | |
|---|---|
| Site: | Fermi  |
| Issue date: | 06/01/1981 |
| From: | Tedesco R Office of Nuclear Reactor Regulation |
| To: | Tauber H DETROIT EDISON CO. |
| References | |
| NUDOCS 8106030428 | |
| Download: ML19352B217 (11) | |
Text
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g nen.,Ig UNITED STATES 2
NUCLEAR REGULATORY COMMISSION o
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WASHINGTON, D. C. 20555
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51 1%1 Docket No. 50-341 Mr. Harry Tauber Vice President Engineering & Construction Detroit Edison Company 2000 Second Avenue Detroit, Michigan 48226
Dear Mr. Tauber:
Subject:
Requests for Additional Infonnation in Fermi 2 Operating License Application As a result of our continuing review of -the operating license application for the Enrico Fermi Atomic Power Plant Unit 2, we hive developed the enclosed requests for additional informatirq.
Request No. 130.8 of the enclosure responds to your February 26, 1981 letter regarding the Mark I Containment Long Term Program - Schedule for submittal of the Plant Unique Analyf s and provides the staff's position regarding this submittal. A discussed with your representatives, the Plant Unique Analysis
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must be submitted by May 1, 1982 to allow adequate time for review.
Please amend your application to comply with the requirements listed in the enclosure. Our review schedule is based on the assumption that the additional information requested in Request Nos. 130.7, 130.6A and 130.9 will be available for our review by June 8,1981.
If you wish clarification of the requests or if you cannot meet these dates, please telephone the Licensing Project Manager, L. Kintner, within 7 days after receipt of this letter.
Sincerely, dh Robert L. Tedesco, Assistant Director for Licensing i
Division jggi g
Enclosures:
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Request for Additional
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l Mr. Harry Tauber Vice President Engineering & Construction Detroit Edison Company 2000 Second Avenue Detroit, Michigan 48226 cc: Eugene B. Thomas, Jr., Esq.
David E. Howell, Esq.
LeBoeuf, Lamb, Leiby & MacRae 21916 John R 1333 New Hampshire Avenue, N. W.
Hazel Park, Michigan 48030 Washington, D. C.
20036 Mr. Bruce Little Peter A. Marquardt, Esq.
U. S. Nuclear Regulatory Cannission Co-Counsel Resident Inspector's Office Thh Detroit Edison Company 6450 W. Dixie Highway 2000 Second Avenue Newport, Michigan 48166 Detroit, Michigan 48225 Dr. Wayne Jens Mr. William J. Fahrner Detroit Edison Company Project Manager - Fermi 2 2000 Second Avenue The Detroit Edison Company Detroit, Michigan 48226 2000 Second Avenue Detroit, Michigan 48226 Mr. Larry E. Schuerman
. Detroit _ Edison. Company _ _ _
3331 West Big Beaver Road Troy, Michigan __4.8084
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ENCL.0SURE REQUEST FOR ADDil!0NAL INFORMATION IN THE SAFETY REVIEW ENRICO FERMI ATOMIC POWER PLANT UNIT 2 DOCKET NO. 50-341 Request by the following branches in NRC are included in this enclosure.
Requests and pages are numbered sequentially with respect to previously transmitted requests.
Branch Page No.
Structural Engineering Branch 130 through 130-16 O
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130.0 Structural Engineering Branch 130.7 'Se'ismic Reassessment As a result of the change in the seisaic input response spectrum for the site, it is required that a reassessment of seismic Category I structures and bur'ied pipes and conduits be made to determine their capacity to resist Y
additional earthquake loads as follows:
1.
Even though NRC does not require the simultaneous considervtion of LOCA and the SSE in load combinations, it is requested that the-applicant present the information on the respective stress components due to LOCA, if applicable, and SSE and the relation of each with the l
specified allowable values.
2.
The three components of earthquake should be considered as specified I
in SRP 3.7.2.
However, if two components were used, they should be i
combined by the absolute sum method.
f 3.
Damping values as specified in Regulatory Guide 1.61 may be used, provided that the resulting stresses are at or near the yield stress l
1e rel.
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4.
The minimum as-built strengths of structures can be used, if these are supported by test data:
for concrete, the cylinder compressive test i
l results, for reinforcing stee' and structural steel the mill test results.
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130-10
130.8 Mark I Containment Long Term Program In your February 26, 1981 Letter (EF2-51,714) to Robert L. Tedesco, of NRC on the subject of Mark I Containment Long Term Program - Schedule for Submittal of the Plant Unique Analysis, you made a number of statements which can be construed as follows:
a.
The interim PUA which has been provided to the staff should provide enough information for the staff to make an acceptance evaluation cf the FERMI 2 Containment Program in the SER.
b.
The applicant committed to perfom a confirmatory review af ter the staff's acceptance of the Mart I Containment LDR.
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c.
The applicant will use the alternate criteria in NUREG-0661, App'endix A, Article 2.13.9 for assessment of the safety relief valve loads through a series of in-plant tests. The results of which will be used to calibrate the structure analytical models as a means of confirmatory review. The results of this review l
will be available at the plant site, which means, they will not i
be submitted for staff evaluation.
Your approach to the resolution as summarized above for the Mark I containment issues is unacceptable and it is the staff's position i
that a plant unique ana13 sis incorporating the load definition and criteria as established by the Mark I Containment Long Tenn Program l
l should be submitted for staff revin. < evaluation. The need for such a submittal is justified as
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i 130-11
1.
On page 1.3 of your interim structural evaluation report it is stated that once the final load definition and acceptance criteria are established by the Mark I Owners Group and approved by the staff, a confimatory review and fatigue evaluation will be conducted to verify that the modifications which have bed. made to the containment are indeed adequate for the approved loads.
2.
In your responses (Amendment 22) to the staff's questions on various areas of concerns it is indicated that they will be considered ia your confirmatory review.
3.
There are a number of structures which according to your evaluation, are overstressed and it is necessary to reassess
these structures on the basis of the loads and criteria as established by the Long Term Program.
Therefore it is required that a reasses'sment of FERMI 2 Mart I Containment be made on the basis of the most up-to-data infomation generated in the Long Tem Program as approved by the staff in NUREG-0661.
l 130.6A Mark"I Containment Interim Structural Evaluation 1.
In your response to question 5 it is indicated that with the exception of downcomer lateral loads the cogdensation oscillation (CO) loads have not been considered in the e
l interim structural evaluation, and will be included in i
130-12 l
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l your confirmatory review.
It is requested that the torus shell, the vent system, and the supports should be evaluated for the C0 loads on the bases as described in the summary of the meetings held on March 4,1981 with the Mark I Owner's Group, issued on March 16, 1981.
2.
In your response to question 9 it is indicated that the single post supports are used to link the water mass to the torus beam elements. Provide a discussion on how the fluid-structure interaction is taken into consideration in your analysis by such an idealization.
3.
In your response to question 18 it is stated that sample computations will be provided only for specific areas of the structures. The following are such areas:
a.' In Table 6.1.1-6 the computed upward load is 498 kips but the allowable is only 410 kips.
Indicate the contributions considered, and for the dynamic loads how the responses are combined, SRSS or ABS.
If, by using the loads and criteria established from the Long Term Program, there is still no reduction in i
such high level of overstress, it is the staff's position f
that a modification of the. design of the tie down base plate should be made so that the' Allowable will l
not be exceeded.
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t 130-13 l
b.
In Tables 6.2.1-3 and 6.2.1-4 indicate the contributions of each load in the load combi lation to the computed total stress intensities of 18 ksi and 22 ksi respec-tively. Indicate how the dynamic responses are established and combined. The computed stress intensity of 22 ksi is greater than the allowable value of 19.3 and you justify it on the basis that the magnitude of impact pressure in the computation is conservative. A reissessment should be made on the basis of more realistic load magnitude and the stress criteria as established by the Long Term Program.
'c.
In Table 6.2.1-5, indicate the contributions of each load in the load combination to the coc?uted total column compression load of 72 kips and total column tension loac of 104 kips, and specify the allowable for each.
d.
For the torus, its internal structures and its supports, fatigue should be included in the evaluation.
4.
As mentioned before you plan to use the alternate criteria in NUREG-0661 Appendix A, Article 2.13.9 for asressment of the
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Safety Relief Valve Load. This approach involves in-plant l
tests and the establishment of a coapled load-structure analytical model. Provide a description of such a model 130-14
l which you are going to calibrate together with the basis i
for the analytical model adopted.
5.
Since there is a change in the seismic input response spectra, the reassessment of the Mark I containment should take the effect of this change into consideration.
130.9 HIGH DENSITY SPENT FUEL RACKS 1.
Indicate whether the proposed fuel racks modifications conform with the NRC position on " Fuel Pool Storage and Handling Application" dated' April,1978 and amanded January 1979.
If any deviations exist identify and justify these deviations.
2.
The responses to questions 130.7.2 and 130.7.3 are not satisfactory, provide the following:
a.
Justify why the spent fuel pool liner was designed in accordance with ASME Boiler and Pressure Vessel Code,Section IIIV, Division I instead of Section III, Subsection NE and ACI 359.
b.
Indicate in detail the methooology used to demonstrate the leaktight integrity of the fuel pool liner when subjected to the postulated fuel assembly drop over the spent fuel racks or directly falling over the fuel pool liner. The fuel assembly drop should be analyzed for the titled position and straight drop.
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130-15 l
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3.
With regard to the fuel assembly drop on top of the rack, provide the following:
a.
The acceptance criteria used for this case.
b.
Detailed descriptions of the method used to satisfy the acceptance criteria.
Comparison between the drops in the titled position, straight c.
drop on the middle of the rack and on the edge of the rack.
d.
Indicate whether other modes of failure of the racks exist beside crushing.
e.
Discuss these same concerns for the stainless steel racks and for the aluminum racks.
4 4.
Indicate whether material, fabrication, installation and quality control of the racks conform with Subsection NE of the ASME Code.
5.
In Section 6.2.4 of the Joseph Oat Corporation Report, it is stated that only fluid damping is included in the analysis and is simulated by inclusion of appropriate equivalent linear damping.
1 Indicate what this damping value is and justify the domping value used in the analysis.
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6.
Indicate why the empty or nearly empty rack case was not used in e
the analysis to predict the maximum rack displacement.
t 130-16
7.
Indicate how the increase in the value of the plant design earthquake and cons
- ently the change in the floor response spectrum could affect the seismic analysis done so far for the fuel pool racks.
8.
Because different type (169,108 and 35 cells) modules were used in the proposed modification with different sizes and weights, indicate which type was used in the seismic and sliding analysis.
Indicate also how other types were qualified for the postulated loadings.
t 130-17
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