ML19347E156
| ML19347E156 | |
| Person / Time | |
|---|---|
| Site: | Waterford  |
| Issue date: | 04/16/1981 |
| From: | Tedesco R Office of Nuclear Reactor Regulation |
| To: | Aswell D LOUISIANA POWER & LIGHT CO. |
| References | |
| NUDOCS 8104240138 | |
| Download: ML19347E156 (12) | |
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UNITED STATES
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APR 1 6 1981 Docket No.: 50-382 Mr. D. L. Aswell Vice President, Power Production Louisiana Power & Light Company 142 Delarende Street New Orleans, Louisiana 70174
Dear Mr. Aswell:
REQUEST FOR ADDITIONAL INFORMATION - WATERFORD STEAM E
SUBJECT:
STATION, UNIT 3 We have determined that certain additional information is required in order to permit us to complete our review of your application for an operating The enclosure to this license for Waterford Steam Electric Station, Unit 3.
letter is couposed of requests for additional information in several different review areas.
Please advise us of the date you expect to provide responses to the enclosed If you require any clarification, please contact the staff's request.
assigned project manager.
I Sincerely, t
i W *C44u Robert L. Tedesco, Assistant Director for Licensing Division of Licensing
Enclosures:
Request for Additional Infonnation cc w/ enclosures:
See next page.
81042.4013%
Mr. D. L. Aswell Vice President, Power Production Louisiana Power & Light Company 142 Delaronde Street New Orleans, Louisiana 70174 cc:
W. Malcolm Stevenson, Esq.
Monroe & Lemann 1424 Whitney Building New Orleans, Louisiana 70130 Mr. E. Blake Shaw, Pittman, Potts and Trowbridge 1800 M Street, N. W.
Washington, D. C.
20036 Mr. D. B. Lester Production Engineer Louisiana Power & Light Company 142 Delaronde Street New Orleans, Louisiana 70174 Lyman L. Jones, Jr., Esq.
Gillespie & Jones P. O. Box 9216 Metairje Louisiana __70005 Luke Fontana, Esq.
Gillespie & Jones 824 Esplanade Avenue New Orleans, Laufstana 70116 Stephen M. Irving, Esq.
One American Place, Suite 1601 Baton Rouge, Louisiana 70825 Resident Inspector /Waterford NPS P. O. Box 822 Killona, Louisiana 70066 e
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APR 16 1981 DISTRIBUTION:
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321.13 Table 10.4-11 of the FSAR indicates that the valves of the SGBS will (10.4.8) meet design code ANSI B 16.11 criteria instead of ANSI B 31.1; yet your response to Question 321.3 states that the SGBS will confonn to Quality Group D which corresponds to quality standard ANSI B 31.1 for valves. Justify why ANSI B 16.11, which addresses " Forged Steel f
Fittings, Socket Welding and Threaded," is more appropriate for valves then ANSI B 31.1.
321.14 The Reactor Auxiliary Building (RAB) ventilation system does not (9.4, 11.3) include a preheater prior to the medium efficiency filter, the HEPA filter, and the charcoal adsorber. With the high relative humidity (i.e., > 70%) in the area of the Waterford Plant (Table 2.3-7, Waterford ER) absence of a preheater will result in deleterious affects on the charcoal adsorbers. Describe the design features that have been incorporated in the RAB ventilation system to maintain c
the relative humidity below 70% so as not to impact on the efficiency of the charcoal adsorbers? A filter efficiency of 505 will be credf-ted to the charcoal adsorbers if the system has not been designed to the approximate expected relative humidity and the relative hum l4f ty of the inlet air is not maintained below 705.
321.15 Your response to Question 321.4 did not address either in-place (9.4, 11.3) testing criteria or laboratory testing criteria for activated carbon.
Our position is that such testing will be required to be perfonned in accordance with ANSI N509-1976 and ANSI N510-1976.
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. 321.16 Position C.4.a of Regulatory Guide 1.140 specifies Section 2.3.8 (9.4, 11.3) of ERDA 76-21 for accessibility provisions. Address the provisions of Section 2.3.8 as they apply to the design of your RAB ventila-tion system.
321.17 To our knowledge, there are no 00P generators available which will generate enough 00P for flow rates of 92,190 cfm and 151,330 cfm to ensure adequate testing capability of the RAB ventilation system for normal operation and purge operating conditions. Therefore, it is our position that a shroud will be required in order to test the HEPA filters and that the shroud be used such that the entire face of the HEPA filter is ultimately tested.
321.18 Inspection and testing of the vaelous components of the liquid waste (10.4, 11.2) management system has not been addressed in the FSAR or the response to Question 321.8. This must be addressed for SER approval of the liquid waste management system.
321.19 Design and fabrication codes for piping and valves of the liquid waste (11.2) l management system have not been indicated in either the response to Question 321.8 or in Section 11.2 of the FSAR. Both must be addressed l
i for SER approval of the liquid waste management system.
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321.20 The response to Question 321.8 did not indicate the design code (11.2) specifications for material of the various components of the
. 321.20 liquid waste management system. The design codes for these (11.2) materials must be indicated prior to SER approva1 of the liquid waste management.
321.21 The response to Question 321.8 did not contain information as to (11.3) the code specifications for materials which will comprise the components of the gaseous waste management system. In addition, inspection and testing was not addressed. These items must be addressed prior to SER approval of the gaseous waste management sys tem.
321.22 In Section 11.4.1.6 of the FSAR, it is indicated that the design (11.4) amounts of waste input and off-site shipment of solid waste were based upon the recommendations of the Environmental Survey of I
Transportation of Radioactive Materials from Nuclear Power Plants I
l (AEC 12/72). Review of the semi-annual operating reports from nuclear power plants through 1979 indicates that approximately 3
13,000 ft /yr of solidified wet waste would be expected fram the 3
primary system wet waste, approximately 300 ft /yr wculd be expected 3
from the SGB desineralizers resin, and apprcximately 10,000 ft /yr of dry waste would be expected from a 3400 MWt PWR. Discuss the capability of your radwaste system to handle this quantity of waste.
It is our position that the Waterford 3 solid radwaste system should be capable of handling such volumes of waste.
. 321.23 Provide a discussion of operating experience encountered by the designers and vendors of radwaste systems similar to those installed at Waterford 3.
In particular,1) address modifications that have been necessary because of any earlier operational problems and,
- 2) provide a summary of any presently inoperable equipment in operat-ing reactors of which you are aware that is similar to equipment you have installed. Address modifications that have been incogorated in your design to preclude earlier identified operational problems.
321.24 The response to Question 321.11 did not address the sampling and (11.5) monitoring provisions of Table 1A of SRP 11.5, Rev. 2.
In addition, 1
l many monitcrs are not discussed in Section 11.5 of the FSAR. These items must be addressed prior to approval of the process and effluent radiological monitoring and sampling system for the SER.
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321.25 Isolation of the containment purge upon high radiation signal is l
(11.5, l
9.4, only addressed for the refueling ventilation mode. Our position is 11.3) that this same purge isolation capability, upon receipt of a high radiation signal in the plant stack, is also required for the purge mode during non-refueling situations.
w ATTACHMENT 2 LOUISIANA POWER & LIGHT COMPANY WATERFORD STEAM ELECTRIC STATION UNIT NUMBER 3 ADDITIONAL GUIDANCE FOR PRESERVICE INSPECTION PROGRAM MATERIALS ENGINEERING BRANCH DIVISION OF ENGINEERING 121.1 Our review of the preservice inspection program includes evaluation of requests for relief fron. ASME Section A Code examination requirements.
Since the preservice inspection is usually completed late in the construction phase and all relief requests may not be identifiable until an examination is performed, indicate the latest date you anticipate for submission of relief requests.
121.2 The examination requirements in later editions and addenda of Section XI which are incorporated by reference in 10 CFR 50.55a(b) may be utilized for your program, subject to staff review. A request for relief is not required.
The use of Appendix III and IWA-2200 in the 1977 Edition through Sununer 1978 Addenda is acceptable, with the following additional conditions for the inservice inspection program:
a) Indications of 50% of DAC or greater shall be recorded.
b) Any indication 100% of DAC or greater shall be evaluated by a Level II or Level III examiner to the extent necessary to determine the size, shape, identity, and location of the reflector, c) Any non-geometric indication 20% of DAC or greater discovered during l
the ultrasonic examination shall be recorded and investigated by l
a Level II or Level III examiner to the extent necessary to determine the shape, identity, and location of the reflector.
121.3 Paragraph IWC-1220(c) in the 1974 Edition of Section XI pennits ECCS components to be exempted from volumetric and surface examinations provided the control of fluid chemistry is verified through periodic sampling. The control of fluid chemistry is intended to minimize corrosive effects. The " chemistry control" provision was deleted from the 1977 Edition of Section XI because practical evaluation, review, and acceptance standards were not defined.
This exemption is not an acceptable basis for eliminating volume'.'ic or surface examinations of ECCS components.
121.4 Your FSAR states in Section 6.6.8 that high energy fluid system piping penetrating containment which is four inches in diameter or smaller is exempt from inservice examination, in accordance with IWC-1220 of Section XI.
The Section XI exemption criteria do not apply to the augmented examination requirements of SRP 3.6.1; hence, these welds must be examined volumetrically as part of the augmented inservice inspection program.
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121.5 Paragraph B.2.d in BTP APESS 3-1 of Standard Rtview Plan requires that fluid system piping which connects to the reactor ce slant pressure boundary on a continuous or intermittent basis be subjected to augmented ISI. Describe the technical basis for not including this piping in the augmented ISI program.
In addition, justify the. elimination of augmented ISI on mainsteam and feedwater piping, inside contaiament, that is enclosed in guard pipes and thus is ou*. side the break exclusion area.
- -i ENCLOStPE Waterford Steam Electric Station, JJnit 3 Louisiana Power and Light Company Docket Number 50-382 Chemical Engineering Branch Corrosi_on Engineering Section
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282.0 Corrosion Engineering Section 282.1 Provide the following references cited in your (5.4.2.1) secondary water chemistry control program:
OP-901-006 Condenser Tube Leakage
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OP-901-007 Abnormal Condensate /Feedwater/ Steam Generator Chemistry 282.2 Provide specifics of actions
- you will take to (5.4. 2.1 )
correct off-control point chemistry conditions.
282.3 Identify and provide a list of the materials of (9.1.2 )
construction of the spent fuel pool racks.
Provide detail drawings or sketch of rack design showing how poison materials are contained. Verify if there is venting of the poison material compartment.
.282.4 Provide specifics of your spent fuel pool material (9.1.2) coupon surveillance procram.
Provide schedule of sample removal, location of samples in pool, description of sample construction, and sample parameters to be tronitored.
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- Sranch Technical Position MTEB 5-3 describes the acceptable means for monitoring secondary side water chemistry in PWR steam generators, including corrective actions for off-control point chemistry conditions. However, the staff is amenable to alternatives, particularly to Branch Technical Position B.3.b(9) of MTEB 5-3 (96-hour time limit to repair or plug confirmed condenser tube leaks).
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1 REQUESTS FOR ADDITIONAL GEOLOGIC I'NFORMATION WATERFORD STEAM ELECTRIC STATION, UNIT NO. 3 LOUISIANA POWER AND LIGHT COMPANY DOCKET NO. 50-382
~361.8 In order to determine the possible presence or absence of structure please utilize all relevant, available geophysical data within an approximate 40 sector extending eastward from the Waterford plant at least 5 miles. Present the results of your interpretation in the FSAR, revising text and figures as necessary. As a part of your response provide a new FSAR figure showing the seismic coverage within 5 miles of the Waterford 3 site.
361.9 Conflicting opinicas are presented in a 1975 Corps of Engineers reference by Kolb et al (Pleistocene Sediments of the New Orleans-Lake Pontchartrain Area) regardirj the age of the Pleistocene sediments irmtediately under-i lying the '..aterford 3 plant. One alternative considers the uppermost Pleistocene as the Deweyville Terrace (17,000-30,000 ybp) while the other alternative suggests the material is the Prairie Terrace whose estimated age ranges from 80,000 to 110,000 ybp. The FSAR (p. 2.5-24) l states that the Prairie Terrace underlies the plant area. Considering the age conflict discussed in this paper, e.g., the uppermost Pleistocene unit as cited in Kolb, provide a discussion of the reason for selecting the Prairie instead of the Deweyville. Please revise the FSAR accordingly.
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l LOUISIANA POWER & LIGHT COMPANY WATERFORD STEAM ELECTRIC STATION UNIT NO. 3 DOCKET NUMBER 50-382 10.2.3 Turbire Disc Integrity We have reviewed Section 10.2.3 of the Final-Safety Analysis Report (FSAR) for the Waterford Steam Electric Station Unit No. 3.
With the exception of in-complete In-Service Inspection requirements, the FSAR assures that the integrity of the turbine will be adequate and that reasonable assurance is provided that the applicable parts of General Design Criterion of 10 CFR Part 50 will be met.
The turbine discs and rotors are forged from vacuum degasses steel by processes that minimize flaws and provide adequate fracture toughness. These materials t
have the lowest fracture appearance transition temperatures and highest Charpy V-notch energies obtainable on a consistent basis. The maximum tangential stress in discs and rotors resulting from centrifugal forces, interference fit and thernal gradients does not exceed 0.75 of the yield strength of the materials of 115% of the rated speed.
Since 1979 the staff has known of the stress corrosion problems in low pressure rotor discs in Westinghouse turbines. Appropriately conservative inspection intervals have been effective in monitoring crack growth to permit repair or replacement of discs well in advance of failure. The method used to predict crack growth rates is based on evaluating all of the cracks found to date in i
Westinghouse turbines, past histcry of similar tur;ine disc cracking, and results of laboratory tests. This prediction metaod takes into account two main parameters; the yield strength of the disc, and the temperature of the disc at the bore area where the cracks of concern are occuring. The higher the yield strength of the material and the higher the temperature, the faster the l
crack growth will be.
The preservice inspection program calls for 1005 ultrasonic inspection (UT) of each rotor and disc forging before finish machining and magnetic particle (MT) after finish machining.
In-service inspection requirements provide for visual and MT inspection when the turbine is disassembled. The FSAR does not provide requirements for volu-metric inspection of the turbine discs nor calculation methods of critical crack sizes. When this information is recefved and reviewed by the staff, approval action will be completed and inspection intervals will be established.
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