ML20024E059
| ML20024E059 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 08/01/1983 |
| From: | Stolz J Office of Nuclear Reactor Regulation |
| To: | Beckham J GEORGIA POWER CO. |
| References | |
| TAC-07939, TAC-07940, TAC-7939, TAC-7940, NUDOCS 8308090105 | |
| Download: ML20024E059 (11) | |
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>:-= m -3 DISTRIBUTION Docket Hos.: 50-321 Docket File and 50-366 NRC PDR ORB #4 Rdg DEisenhut Mr. J. T. Beckham, Jr.
A Uc, 1 19g3 OELD Vice President Engineering NSIC Georgia Power Company EJordan P. O. Box 4545 JTaylor Atlanta, Georgia 30302 AC GR ve ark
Dear Mr. Beckham:
RIngram Gray File
SUBJECT:
MARK I CONTAINMINT LONG TERM PROGPAM - PLANT UNIQUE ANALYSIS REPORT - LOADS AND STRUCTURAL EVALUATIONS Re:
E. I. Hatch Nuclear Plant, Unit Nos. I and 2 The NRC staff and its consultants Brookhaven National Laboratory (BNL) and Franklin Research Center (FRC) are reviewing the loads and structural aspects, respectively, of your plant unique analysis report.
As a result of our review to date we have prepared the enclosed requests for additional information.
To expedite this review it is requested that within three weeks of the date of this letter a meeting betwean the NRC and our consultants, and you and your contractor be held to discuss your response to these requests.
is our intent to resolve these issues at this meeting, it is imperative Since it that you have a representative at this meeting that has the authority to make the detisions necessary to accomplish this goal.
It is suggested that this meeting be held at your contractors office; however we are amenable to having it wherever it is most convenient.
your project manager within seven days of receipt of this letter w!th aPlease notify proposed meeting date.
an alternative one.
If you cannot meet the three week schedule, propose This request for information was approved by the Office of Management and Budget under clearance number 3150-0091 which expires October 31, 1985.
Sincerely, e308090105 esoeo3 M ufnar 4,neg w DR ADOCK 05000321 p
PDR John F. Stolz, Chief Operating Reactors Branch #4 Division of Licensing rnetne,wa.
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Patch 1/2 50-321/366
, Georgia Power Company cc w/ enclosure (s):
Mr. James P. O'Reilly, Regional Adminis trator G. F. Trowbridge, Esq.
U. S. Nuclear Regulatory Commission Shaw, Pittman, Potts and Trowbridge Region II 1800 M Street, N.W.
101 Marietta Street, Suite 3100 Washington, D. C.
20036 Atlanta, Georgia 30303 Ruble A. Thomas Vice President P. O. Box 2625 Southern Company Services Inc.
Birmingham, Alabama 35202 0 zen Batum Charles H. Badger
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Southern Company Services, Inc.
Office of Planning and Budget Post Office Box 2625 Room 610 Birmingham, Alabama 35202 270 Washington Street, S.W.
Atlanta, Georgia 30334 Appling County Commissioners County Courthouse Baxley, Georgia 31513 J. Leonard Ledbetter, Commissioner.
Department of Natural Resources Mr. L. T. Gucwa 270 Washington: Street, N.W.
Georgia Power Company
' Atlanta, Georgia 30334 Engineering Department P. O. Box 4545 Atlanta, Georgia 30302 Mr. H. C. Nix, Jr. General Manager Edwin I. Hatch Nuclear Plant Georgia Power Company P. O. Box 442 Baxley, Georgia 31513 Regional Radiation Representative EPA Region IV 345 Courtland Street, N.E.
Atlanta, Georgia 30308 Resident Irispector U. S. Nuclear Regulatory Commission Route 1, P. O. Box 279 Baxley, Georgia 31513
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E. I. HATCH NUCLEAR PLANT IINITS 1 & 2 REQUEST FOR INFOR4ATION RELATED TO LOADS EVALUATION 0
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l ITEM 1:
PUAR section 1.2.4, AC section 2.13 Provide more detailed informatIion concerning the T-quencher utilized in the Hatch plants.
Specify any differences such as hole spacing, hole diameter, etc. between the T-quencher tested at Monticello and the T-quencher s installed in the Hatch units.
ITEM 2:
PUAR section 4.1.3.4, AC section 2.2 The factors used to account for the non-uniform downcomer flow as shown in Figures 4.1.3-2 and 4.1.3-3 of the PUAR differ from the factors given in the LDR (Rev. 2).
In addition, the statement that "as flow rate decreases, the above factors approach 1.0" seems to imply that the factors were varied, whereas the LDR (Rev. 2) assumes that the factors remain constant during the portion of the vent header thrust load transient after vent clearing.
Provide sufficient information to justify the use of the PUAR factors and clarify how the factors were used during the vent header thrust load transient.
ITEM 3:
PUAR section 4.1.4.4, AC section 2.7.1 Indicate what drag coefficient was used to calculate the pressure due to drag following impact for internal cylindrical structures located above the suppression pool.
ITEM 4:
PUAR section 4.1.7, AC section 2.13.7 The AC required that an asymmetric SRV discharge load case be consid-ered for both first and subsequent actuations with the degree of asym-metric discharge for each event combination being determined from a plant-specific primary system analysis designed to maximize the asym-metric condition.
No mention of an ~ asymmetric SRV discharge load case is made in the PUAR load case discussion.
Provide sufficient informa-tion to satisfy the AC requirements concerning this matter.
ITEM 5: PUAR section 4.1.7.8, AC section 2.13.6 Clarify the ' discussion in the PUAR concerning the maximum S/RVDL and discharge device pipe wall temperature.
It is stated in the PUAR that the temperature distribution is applied from the S/RV exit to the dis-charge device entrance which is consistent with the AC.
However, in the next paragraph it is also stated that an average wall temperature of 350*F was used in the evaluation of the S/RVDL.
Does this average wall temperature correspond to the discharge device temperature which was generically defined in the LDR (Rev. 2) to be 370*F for the T-quencher?
ITEM 6:
PUAR section 8.2.3, AC section 2.3 Describe in more detail what is meant by the statement that the analy-ses were performed assuming an undisturbed pool (i.e.,100 percent water mass) and that no credit was taken for the percent water mass "i n fl ight".
Indicate why this assumption is a conservatism in the an-alyses of the pool swell loads during the upload phase of the transi-ent.
ITEM 7:
Indicate whether all loads covered by the LDR and the AC have been con-sidered during the plant unique analysis and provide justification if any load has been neglected.
ITEM 8:
Hatch 1 PUAR section 4.1.4.1, AC section 2.3 The normal operating co.nditions for Hatch 1 appear to have changed from those reported in the LDR (Rev. 2) and tested in the QSTF tests.
Con-sequently, since no Hatch 1 supplemental tests were performed in the QSTF, the LOCA pool swell loads must now be based on or derived from only one OSTF test (i.e. Tests 5).
The LDR and NUREG-0651 specify a minimum of four tests as a data base for obtaining net torus vertical loads, therefore, the use of one test is an exception to the AC.
Further justification and clarification is required for this AC ex-ception. Have any additional margins been added to account for the larger statistical variance associated with the smaller number of tests at the zero A p condition? Describe in detail how the use of only one test affects other loads such as impact and drag loads, etc.
ITEM 9:
PDAR section 4.1.4.3 & 4.1.4.9, AC sections 2.6 & 2.10 Both Hatch 1 & 2 utilize a vent header deflector in the non-vent bays and no deflector in the vent bays.
Provide a detailed description of how the vent header and vent header deflector impact loads were deter-mined in each bay for both units.
Refer to specific QSTF test data or analytical methods used to help clarify.the issue and where test data is referenced, specify the number of tests used to arrive at the speci-fication.
Include a discussion of what was done for the Hatch 1 vent bay vent header impact load since all the QSTF tests were performed with vent header deflectors.
In a likewise manner for Hatch 2, specify what method was used for the non-vent bay since no QSTF tests were performed with vent header deflectors.
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APPF.NDIX B ADDITIONAL INFORMATION REQUIRED RELATED TO STRUCTURAL EVALUATION ah
. Franklin Research Center A Division of The Franklin institute The Benpman Frankhn Parkway. Phita. Pa 19103(21S)448-1000
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TER-C5506-325 REQUEST FOR INFORMATION 3
Item 1: Provide a summary of the analysis with regard to the vacuum breaker l
piping systems for E. I. Hatch Nuclear Plant Units 1 and 2.
Item 2: With reference to Section 6.4.4.5 of the Hatch Unit 1 PUA report [2],
provide the operability and functionality evaluations for the remaining 10 valves.
If the evaluation has not been completed, provide the proposed schedule of completion.
Item 3:
Indicate whether the fatigue usage factors for the SRV piping and the torus-attached piping are sufficiently small that a plant-unique fatigue analysis is not warranted for piping. The NBC is evaluating the conclusions of a generic study to determine whether it is sufficient for each plant-unique analysis to establish that the expected ust.ge factors for piping are small enough to obviate a plant-unique fatigue analysis of the piping.
Item 4: With reference to Section 6.1.1.6 of the Hatch Unit 1 PDA report [2],
elements in the bottom of the shell near the miter joint were overstressed by 254.
Indicate any conservatisms in the analysis which can offset the overstress and reduce the stresses to the code
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allowables; otherwise, modification is required.
Item 5: Provide and justify the reasons for not considering a 180* beam model of the torus including columns, saddles, and seismic restraints in order to determine the effects of nonsymmetric loads such as SRV and chugging for E. I. Hatch Nuclear Plant Units 1 and 2.
Item 61 Provide and justify the reasons for not considering a 180' beam rodel i
of the vent system in order to determine the effects of seismic and other nonsymmetric loads for E. I. Hatch Nuclear Plant Units 1 and 2.
Item 7: With reference to Section 6.2.2.6 of the Hatch Unit 1 PUA report [2],
there are two columns that are 12% over 'the code allowables, and the top clevis pin bearing stresses are 3% over the code allowables.
Indicate any conservatisms in the analysis which can offset the overstress and reduce the stresses to the allowable.s; otherwise, modification is required.
Item 8: With reference to Section 6.1.1.6 of the Hatch Unit 2 PUA report [3],
stresses in the bottom of the shell exceeded the code allowables by 25%, and the top portion of the shell exceeded the limiting buckling str ess.
Indicate any conservatisms in the analysis which can offset the overstress and reduce the stresses to the Code allowables; otherwise, modification is required,
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TER-C5506-325 Item 9: With reference to Section 6.1.1.6 of the Hatch Unit 2 PUA report [3],
stresses in the ring girder exceeded the allowable by 25%. Indicate any conservatisms in the analysis which can offset the overstress and reduce the stresses to the code allowables; otherwise modification is required.
Item 10: With reference to Section 6.1.1.6 of the Hatch Unit 2 PUA report [3],
the weld stress at one saddle stiffener under the suppression chamber exceeded the allowable by 234. Indicate any conservatisms in the analysis which can offset the overstress and reduce the stresses to the code allowables; otherwise, modification is required.
Item 11: With reference to Section 6.4.4.5 of the Hatch Unit 2 PUA report [3],
provide the operability and functionality evaluations for the remaining 14 valves and two pumps. If the evaluations have not been completed, provide the analytical approach and the proposed schedule of completion.
Item 12: With reference to Table 1 of Appendix B, indicate whether all loads have been considered in the analysis and/or provide justification if any load has been neglected for Hatch Nuclear Plant Units 1 and 2.
Item 13: Provide justification for determining the load combinations indicated in Section 6.1.1.2 [7, 8] to be the limiting load combinations for Hatch Nuclear Plant Units 1 and 2.
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20th and Raca Streets. PMa.. Pa. 19103 (215) 448 1000 gg j g 4 Table 1. Structural Loading (from Referenced)
OtherWetwell Interior Structures Structures
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- 1. Containment Pressure and Temperature X
X X
X X
X X
X X
- 2. Vent System Thrust Loads X
X X
- 3. PoolSwell 3.1 Torus Net Vertical Lcads X
X 3.2 Torus Shell Pressure Histories X
X 3.3 VentSystemimpactand Drag X
X X
3.4 Impact and Drag on Other Structures X
X X
3.5 Frothlmpingement X
X X
X X
3.6 Pool Fallback X
X X
3.7 LOCAJet X
X 3.8 LOCA Bubble Drag X
X X
- 4. Condensation Osci!!ation 4.1 Torus Shell Loads X
X 4.2 Load on Submerged Structures X
X X
e 4.3 Lateral Loads on Downcomers X
X 4.4 Vent System Loads X
X
- 5. Chugging 5.1 Torus Shell Loads X
X 5.2 Loads on Submerged Structures X
X X
l 5.3 Lateral Loads on Downcomers X
X l
5.4 Vent System Lo' ads X
X l
- 6. T-Ouencher Loads l
6.1 Discharge Line Clearing X
l 6.2 Torus Shell Pressures X
X 6.4 Jet Loads on Submerged Structures X
X X
X l
6.5 Air Bubble Drag X
X X
X 6.6 Thrust Loads on T-Ouencher Arms X
l 6.7 S/RVDL EnvironmentalTemperature X
- 7. Ramshead Loads 7.1 Discharge Line Clearing 7.2 Torus Shell Pressures 7.4 Jet Loads on Submerged Structures El 7.5 Air Bubble Drag C
D 7.6 S/RVDL EnvironmentalTemperature l
Loads required by NUREG-0681C+ 3 I
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Not applicable.
TER-C5506-325 REFERENCES FOR APPENDIX B 1.
" Mark I Containment Program Structural Acceptance Criteria Plant Unique Analysis Application Guide" J
General Electric Co., San Jose, CA October 1979 2.
E. I. Hatch Nuclear Plant Unit 1 Plant Unique Analysis Report - Mark I Containment Long-Term Program, Revision 0 Georgia Power Company Southern Company Services, Inc.
January 1983 3.
E. I. Hatch Nuclear Plant Unit 2 Plant Unique Analysis Report - Mark I Containment Long-Term Program, Revision 0 Georgia Power Company Sourthern Company Services, Inc.
February 1983 4.
" Safety Evaluation Report, Mark I Containment Long-Term Program Resolution of Generic Technical Activity A-7" Office of Nuclear Reactor Regulation July 1980 5.
NEDO-21888 Revision 2
" Mark I Containment Program Load Definition Report" General Electric Co., San Jose, CA November 1981 A Dumaan of The Franhen insomme
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