ML13302A362
| ML13302A362 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 06/16/1980 |
| From: | Schwencer A Office of Nuclear Reactor Regulation |
| To: | Dietch R, Gilman B San Diego Gas & Electric Co, Southern California Edison Co |
| References | |
| NUDOCS 8006250392 | |
| Download: ML13302A362 (11) | |
Text
matlutlon JUN 6 1980 11tiO!
JKnight Local PDR VMoore Docket File WKreger LB #3 Files MErnst DEisenhut RPDenise RPurple ELD Docket Nos.5.O1 RTedesco I13)
And 50-362>-
ASchwencer NSIC HRood TIC JLee ACRS(16)
RMattson Mr. Robert Dietch SHanauejr. B. W. Gilman Vice President Senior Vice President - Operations Southern California Edison Company
'San Diego Gas and Electric Company 2244 Walnut Grove Avenue 101 Ash Street P. 0. Box 800 P. 0. Box 1831 Rosemead, California.91770 San Diego, California 92112 Gentlemen:
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION RELATING TO THE STAFF REVIEW OF THE SAN ONOFRE NUCLEAR GENERATING STATION, UNITS 2.AND 3 As a result of our review-of the Final Safety Analysis Report for the San Onofre Nuclear Generating Statidn, Units 2.and 3, we find that we need the additional information listed In the Enclosure. Please contact us if you have any questions about the information requested.
Sincerely, A.
wncer ing Ch Licensing Branch No. 3 Division of Licensing
Enclosure:
Request for Additional Information ccs w/enclosure:
See next page OFFICE D:B#
SURNAME.
HRood:aB......
AS hwencer................
DATE...
6 80.......................
NRC FORM 318 (9-76) NRCM 0240 1,U.S. GOVERNMENT PRINTING OFFICE: 1979-289-369
o AUNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 AUN 1 6 1iou Docket Nos. 50-361 and 50-362 Mr. Robert Dietch Mr. B. W. Gilman Vice President Senior Vice President Operations Southern California Edison Company San Diego Gas and Electric Company 2244 Walnut Grove Avenue 101 Ash Street P. 0. Box 800 P. 0. Box 1831 Rosemead, California 91770 San Dieoo, California 92112 Gentlemen:
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION RELATING TO THE STAFF REVIEW OF THE SAN ONOFRE NUCLEAR GENERATING STATION, UNITS 2 AND 3 As a result of our review of the Final Safety Analysis Report for the San Onofre Nuclear Generatinq Station, Units 2 and 3, we find that we need the additional information listed in the Enclosure. Please contact us if you have any questions about the information requested.
Sincerely, A. chwencer, Acting Chief Licensing Branch No. 3 Division of Licensing
Enclosure:
Request for Additional Information ccs w/enclosure:
See next page
cc:
Charles R. Kocher, Esq.
James A. Beoletto, Esq.
Southern California Edison Company 2244 Walnut Grove Avenue P. 0. Box 800 Rosemead, California 91770 Chickering & Gregory ATTN:
David R. Pigott, Esq.
Counsel for San Diego Gas & Electric Company &
Southern California Edison Company 3 Embarcadero Center - 23rd Floor San Francisco, California 94112 Mr. Kenneth E. Carr City Manager City of San Clemente 100 Avenido Presidio San Clemente, California 92672 Alan R. Watts, Esq.
Rourke & Woodruff Suite 1020 1055 North Main Street Santa Ana, California 92701 Lawrence Q. Garcia, Esq.
California Public Utilities Commission 5066 State Building San Francisco, California 94102 Mr. R. W. DeVane, Jr.
Combustion Engineering, Incorporated 1000 Prospect Hill Road Windsor, Connecticut 06095
Mr. Robert Dietch 2
Mr. B. W. Gilman cc:
Mr. P. Dragolovich Bechtel Power Corporation P. 0. Box 60860, Terminal Annex Los Angeles, California 90060 Mr. Mark Medford Southern California Edison Company 2244 Walnut Grove Avenue P. 0. Box 800 Rosemead, California 91770 Henry Peters San Diego Gas & Electric Company P. 0. Box 1831 San Diego, California 92112 Ms. Lyn Harris Hicks Advocate for GUARD 3908 Calle Ariana San Clemente, California 92672 Richard J. Wharton, Esq.
Wharton & Pogalies Suite 106 2667 Camino Del Rio South San Diego, California 92108 Phyllis M. Gallagher, Esq.
Suite 222 1695 West Crescent Avenue Anaheim, California 92701 Mr. A. S. Carstens, 2071 Caminito Circulo Norte Mt. La Jolla, California 92037 Resident Inspector, San Onofre/NPS c/o U. S. Nuclear Regulatory Commission P. 0. Box AA Oceanside, California 92054
ENCLOSURE 022.0 CONTAINMENT SYSTEMS BRANCH 022.63 The response to 02?.53 does not demonstrate that adequate provisions are made to ensure that any debris entrained in the vented containment's atmosphere in the event of a LOCA will not prevent closure of the contain ment mini-purge system isolation valve. It is our position that the ducting which houses the registers must be capable of remaining intact under acci dent conditions and that the registers in the ducts must be of sufficiently small mesh size to preclude the passage of debris which could inhibit valve closure. Therefore, either demonstrate that the currently proposed system design meets the above requirements or provide an alternative design which assures that blockage of the purge isolation valves will not occur.
022.64 In a letter dated September 10, 1979, the NRC was informed by Virginia Electric and Power Company that overpressurization of the containment at North Anna 3 and 4 could occur as a result of a main steam line break inside containment. This overpressurization resulted when auxiliary feedwater flow was included in the analysis. NRC is currently assessing the generic implications of this letter.
To assist us in determining if a similar circumstance could occur at your facility, you should take the following actions.
- 1) Review your original analysis of this event, and provide NRC with the assumptions used during this analysis. Particular emphasis should be placed on describing how auxiliary feedwater flow (AFF) was accounted for in your original analysis. (Reference to previously submitted information is acceptable if identified as to page number and date.)
Any changes in your design which would impact the conclusions of your original
022-2 analysis should be discussed. We are particularly concerned with design changes that could lead to an underestimation of,the containment pressure following a MSLB inside containment.
- 2) Specifically, provide the following information for the analyses performed to determine the maximum containment pressure for a spectrum of postulated main steam line breaks for various reactor power levels:
- a. Specify the auxiliary feedwater flow rate that was used in your original containment pressurization analyses. Provide the basis for this assumed flow rate.
- b. Provide the auxiliary feedwater rated flow rate, the run out flow rate, and the pump head capacity curve of your current design.
- c. Provide schematic drawings to show the auxiliary feedwater 'ystem arrangement in your current design.
- d. Provide the time span over which it was assumed in your original analysis that AFF was added to the affected steam generator following a MSLB inside containment.
- e. Discuss the design provisions in the auxiliary feedwater system used to terminate the auxiliary feedwater flow to the affected steam generator. If operator action is required to perform this function, discuss the information that will be available to the operator to alert him of the need to isolate the auxiliary feedwater to the affected steam generator, the time when this information would become available, and the time it would take the operator to complete this action. If termination of auxiliary feedwater flow is dependent
022-3 on automatic action, describe the basic operation of the auto-isolation system. Describe the failure modes of the system. Describe any annunciation devices associated with the system.
- f. Provide the single active failure analyses which specifically identifies those safety grade systems and components relied upon to limit the mass and energy release and the containment pressure response.
The single failure analysis should include, but not necessarily be limited to:
partial loss of containment cooling systems and failure of the auxiliary feedwater isolation valve to close.
- g. For the single active failure case which results in the maximum containment atmosphere pressure, provide a chronology of events.
Graphically, show the containment atmosphere pressure as a function of time for at least 30 minutes following the accident. For this case, assume the auxiliary feedwater flow to the broken loop steam generator to be at the pump run out flow (if a run out control system is not part of the current design) for the entire transient if no automatic isolation to auxiliary feedwater is part of the current design.
- h. For the case identified in (g) above, provide the mass and energy release data in tabular form. Discuss and justify the assumptions made regarding the time at which active containment heat removal systems become effective.
022.4 022.65 The response tr 022.59 is not complete.
Discuss the acceptability of the subcomoartment nodalization used to determine the transient loads on the major components due to asymmetric subcompartment pressurization following a LOCA. Provide the transient loads [fx(t), f (t),
z(t)] and moments [M (t), My (t),
M z(t)]
for the component support design basis model and the expanded subcompartment nodalization model [refer to Figures 02221-1 and 022.36-1], resolved at the appropriate elevations for the component supports design evaluation. In doing so provide the transient loads resulting from subcompartment pressurization for two cases, namely, the design basis case from which transient loading data were obtained for use in the component supports evaluation and the case based on the expanded subcompartment nodalization model.
121.0 MATERIALS ENGINEERING BRANCH COMPONENT INTEGRITY SECTION 121.1-8 In Tables 5.3-16 through 5.3-20, it is stated that SA 540 Grade 324 steel is used for reactor vessel bolting material. Paragraph I.C, Appendix G, 10 CFR Part 50, requires that bolting and other types of fasteners not have a specified minimum yield strength greater than 130 ksi.
Therefore, to establish the actual material yield strength, indicate the class(es) of SA 540 Grade B24 material used.
121.19 Paragraph III.B.3, Appendix G, 10 CFR Part 50, specifies that calibration of temperature instruments and Charpy V-notch imoact test machines comply with Paragraph NB-2360 of the ASME Code. Calibration of test equipment for San Onofre Unit Nos. 2 and 3 was conducted in accordance with Paragraoh NA-4600 of the 1971 ASME Code through 1971 Summer Addenda. Paragraph NA-4600 requires that a procedure be in effect to ensure that measuring and testing equipment is calibrated and properly adjusted at specific periods, and that calibration is against certified measurement standards.
Provide details of this required procedure and measurement standards used.
121.20 Provide the qualifications of individuals who performed fracture toughness tests as required by Paragraph III.B.4, Appendix G, 10 CFR Part 50.
Include training and experience.
121.21 Data presented in Tables 5.2-5, 5.2-5A, 121.11-1 through 121.11-27, 121.12-1, and 121.12-2, either do not meet, or are not adequate to determine if the requirements of Appendix G, 10 CFR Part 50, are met.
ThErefore, to help demonstrate compliance with Appendix G, 10 CFR Part 50, supply the following:
(1) for the reactor vessel beltline materials, provide full Charpy V-notch curves, including data points, reported in impact energy and lateral expansion, both as a funct*on of temperature; (2) for welds and weld heat-affected zones in the beltline reaion, provide fracture toughness data from either available data or additional tests.
Include transition ternperature data, upper shelf energy data, and the significant variables that affect fracture toughness properties, e.g., weld wire, flux, base metal combinations, and heat treatment. Correlate this information with data already presented in Tables 121.11-1 through 121.11-27, and provide analyses of the additional data to demonstrate compliance with all the fracture toughness requirements of Appendix G; (3) for all reactor vessel beltline materials, define an initial reference temperature, RTN,, and the most limiting RTNY.
Provide details of the me od used to establish both va ues.
121-2 121.22 Paragraph III.C.2, Appendix G, 10 CFR Part 50, specifies that every fracture toughness 'test specimen from the reactor vessel beltline must be subjected to a heat treatment that produces metallurgical effects equivalent to those produced in the vessel material throughout its fabrication process.
Identify all specimens that do not meet this requirements and provide technical justification for use of such a specimen(s) in establishing fracture toughness properties of the reactor vessel beltline.
121.23 Revise the pressure-temperature limits, presented in Figures 16.3-7A and 16.3-7B of the Technical Specifications, to reflect data requested in Question 121.19.
121.24 The materials surveillance program uses six specimen capsules, contain ing reactor vessel steel specimens of the limiting base material, weld metal material, and heat-affected zone material.
To help demonstrate compliance with Appendix H, 10 CFR Part 50, provide a table that includes the following information for each of the 342 specimens:
(1) actual surveillance material; (2) beltline material from which the specimen was obtained; (3) test specimen type and orientation; (4) fabrication history of each test specimen; (5) chemical composition of each test specimen; and (6) heat of filler material, production welding conditions, and base metal combinations for weld specimens.
Provide the lead factor for each specimen capsule calculated with respect to the vessel inner wall.
121.25 According to 5 16.3.9.2.1 of the Technical Specifications, the ores surizer is limited to a maximum heatup and cooldown of 200 0 F in any one hour period. Paragraph IV.A.2.a, Appendix G, 10 CFR Part 50, requires that the thermal stress intensity factor produced by a heatup and cooldown rate of 200aF/hr plus the membrane stress intensity factor be lower than the reference stress intensity factor by the margins specified in the following equation of Appendix G of the ASME Code:
2 KIM + Kit < KIR To demonstrate compliance with the fracture toughness requirements of Appendix G, 10 CFR Part 50, orovide the calculations and analyses used to determine the critical stress intensity factors produced by the membrane tensile stresses and the radial thermal gradient result ing from a heatup and cooldown rate of 200aF/hr. Calculate the reference stress intensity factors, and demonstrate that the margins required by Appendix G, 10 CFR Part 50, are met.
121.26 To help demonstrate the integrity of the reactor coolant pumo flywheels, supply the Charpy V-notch impact and tensile data for each flywheel, explicitly stating the material used for each flywheel.
121-3 Also, confirm that welding, includinq repair welding, was not performed on any finished flywheel.
If welding was performed, identify the flywheel(s) and location of the welds.
121.27 Information supplied in FSAR Section 16.3/4.4.5 concerning steam generator tube inspection is either incomplete or inadequate.
In order to demonstrate ccmpliance with NRC requirements, revise the following areas in this FSAR section to be consistent with NUREG-0212, Revision 1, "Standard Technical Specifications for Combustion Engineering Pressurized Water Reactors:"
(1) 5 16.4.4.5.1.2.8 and.C with regard to first, second, and third sample of tubes at each inspection; (2) include the additional requirements and acceptance criteria listed in NUREG-0212 regarding eddy current testing in 5 16.4.4.5.1.2.8 and 5 16.4.4.5.1.4.A.1; (3) in 5 16.4.4.5.1.3 add a requirement to increase the inspection frequency if the test results fall into Category C-3; (4) add a requirement in 5 16.4.4.5.1.4.A and Table 16.4-7 for a preservice inspection; and (5) include the details of the reporting reauirements to 9 16.4.4.5.1.5 as listed in NUREG-0212.