ML19345C178
| ML19345C178 | |
| Person / Time | |
|---|---|
| Site: | Shoreham File:Long Island Lighting Company icon.png |
| Issue date: | 11/21/1980 |
| From: | Tedesco R Office of Nuclear Reactor Regulation |
| To: | Pollock M LONG ISLAND LIGHTING CO. |
| References | |
| NUDOCS 8012040075 | |
| Download: ML19345C178 (11) | |
Text
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wq g NUCLEAR REGULATORY COMMISSION J. W ASHINGTON, D. C. 20555 " 's ~)**'*/ NG'l 2 !.1330 m %~V 'R
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.p i .m ,,.li ' Docket No.: 50-322 l,.,_ E Lgj , as; 'P : rra .o.
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- r 2E';m Long Island Lighting Company
" drl.1o A" ATTN: Mr. M. S. Pollock ?! h5G R C 7 Vice President - Nuclear 4 55 S ca 175 East Old Country Road i Hicksville, New York 11801 i
Dear Mr. Pollock:
Subject:
Requests for Additional Information - Shoreham Nuclear Pcwer Station As a result of our review of your application for an operating license for the Shoreham Nuclear Power Station, we find-that we need additional information in the areas of core performance, chemical engineering, mechanical engineering, hydrologic engineering, power systems, and materials engineering. The specific requests for information are enclosed. We have scheduled our review of the Shoreham Control Room Design for. the week of March 30,-1981. In order to maintain this schedule, you must provide a preliminary control room assessment, as discussed in 1.D.1 of NUREG-0737, to the NRC a minimum of two weeks prior to March 30, 1981. 1 If you desire any discussion or clarification of the information requested, please contact J. N. Wilson, Licensing Project Manager, (301) 492-8408. Sincerely, I 2h'c W Robert L. Tedesco, Assistant Director for Licensing Division of Licensing
Enclosure:
Request.for Additional Information ecs w/encls.: See next page 80220d o 075
, 6,, 'O 8 Mr. M. S. Pollock Vice President-Nuclear i Long Island Lighting Company 175 East Old Country Road 4 Hicksville, New York 11801 ccs: Howard L.'Blau, Esq. Honorable Peter Cohalan Blau and Cohn, P.C. Suffolk County Executive. 217~ Newbridge Road County Executive / Legislative Building Hicksville, New York 11801 Veteran's Memorial Highway Hauppauge, New York 11788 Jeffrey Cohen. Esq. 1 i Dept.ty Comissioner and Counsel David Gilmartin, Esq. 1 New York State Energy Office Suffolk County Attorney Agency Building 2 County-Executive / Legislative Building Empire State Plaza Veteran's Memorial Highway Albany, New York -12223 Hauppauge, New York 11788 Energy'Research Group, Inc. MHB Technical Associates 400-1 Totten Pond Road 1723 Hamilton Avenue - Suite K Waltham, Massachusetts.02154 San Jose, California 95125 1 Irving Like, Esq. Stephen Latham, Esq. Reilly, Like and Schnieder Twomey, Latham & Schmitt 200 West Main Street P. O. Box 398 Babylon, New York 11702 33 West Second Street i Riverhead, New York 11901 ) J. P. Novarro Project Manager Joel Blau, Esquire Shoreham Nuclear Power Station New York Public Service Commission P. O. Box 618 The Governor Nelson A. Rockefeller Bldg. Wading River, New York 11792 Empire State Plaza Albany, New York 12223 W. Taylor Reveley, III, Esq. Hunton & Williams Ezra I. Bialik P. O. Box 1535 Assistant Attorney General Richmond, Virginia 23212 Environmental. Protection Bureau New York State Department of Law Ralph Shapiro, Esq. 2 World Trade Center l Camer & Shapiro New York, New York 10047 9 East 40th Street. New York, New-York 10016 Edward J. Walsh, Esq. General Attorney Long Island Lighting Company 250 Old Country Road Mineola, New York 11501 Resident Inspector /Shoreham NPS c/o U.S. Nuclear Regulatory Comission P. O. Box B Rocky Point, New York 11778 ^
112.24 Due.to a long history of problems dealing with inoperable and incorrectly installed snubbers, and due to the potential safety significance of failed snubbers in safety related systems and components, it is requested that maintenance records for snubbers be documented as follows: Pre-service Examination A pre-service examination should be made on all snubbers listed in tables 3.7 4a and 3.7-4b of Standard Technical Specifications 3/4.7.9 This exami-nation should be made after snubber installation but not more than six months prior to initial system pre-operational testing, and should as a mimimum verify the following: (1) There are no visible signs of damage or impaired operability as a result of storage, handling, or installation. (2) The snubber location, orientation, position setting, and configuration (attachments, extensions,-etc.) are accordiny to design drawings and specifictions. _(3) Snubbers are not seized, frozen or jammed. (4) Adequate swing clearance is provided to allow snubber movement. (5) If applicable, fluid is to the recommended level and is not leaking from the snubber system. (6) Structural connections such as pins, fasteners and other conrecting hardware such as lock nuts, tabs, wire, cotter pins are installed correctly. If the period between the initial pre-service examination and initial system pre-operational test exceeds six months due to unexpected situations, re-examination of items 1,4, and 5 shall be performed, Snubbers which are installed incorrectly or otherwise fail to meet the above requirements must be repaired or replaced and re-examined in accordance with the above criteria. , Pre-Operational Testing During pre-operational testing, snubber thermal movements for systems whose operating temperature exceeds 250* F should be verified as follows: (a) During initial system heatup and cooldown, at specified temperature intervals for any system which attains operating temperature, verify the snubber expected thermal movement. (b) For those systems which do not attein operating temperature, verify via observation and/or calculation that the snubber will accommodate the projected thermal movement. (c) Verify the snubber swing clearance at specified heatup and cooldown intervals. Any discrepencies or inconsistencies shall be evaluated for cause and corrected prior to proceeding to the next specified interval.
4 The above described operability program for snubbers should be included and documented by the pre-service inspection and pre-operational test 1 programs. The pre-service inspection must be a prerequisite for the pre-operational testing of snubber thermal motion. This test program should be specified .in Chapter 14 of the FSAR. 4 W I + 4 i j -e i I i J i i j .w-e e, --m. ,r.- ,e.., a g
5 121.0 Materials Engineering Branch - Ccmponent Integrity Section j - 121.34 ' Additional information is needed to justify that the value of -50*F used to estimate the nil ductility temperature for the beltline welds was obtained from test samples that represent the beltline welds in the Shoreham, Unit I reactor pressure vessel. This information should include a comparison of the.significant weld parameters (e.g. weld wire, flux,. thermal treatment) and mechanical properties from both i the sample and beltline welds. 121.35 No data has been provided to demonstrate that the impact properties for the ferritic valve.and bolting materials in the reactor coolant pressure boundary meet the requirements of Paragraph IV.A.3 of Appendix G. The actual test data must be supplied so that compliance witn the paragraph can be demonstrated. If no data exist for the actual materials at Shoreham Unit 1, data from the literature for similar materials and/or f analyses can be used to demons *. rate compliance with Paragrapn IV.A.3 of Appendix G. 121.36-Insufficient data has been supplied to demonstrate that the weld metals 20291/1092/3854, 21935/1092/3889, IP-2815/1092/3869, and 90099/0091/3458 have a minimum upper shelf energy of 75 ft-lb as required by paragraph IV.B of Appendix G. The applicant must provide additional data, infomation frcm the licerature, and/or analyses to demonstrate that an acceptable margin of. safety is assured for nomal operation in the upper shelf temperature region. The additional _ data should be from tests of similar welds; that Lis, those having the same weld wire, flux and themal treatments as the four identified beltline welds. l .121.37-Based on our evaluation we have detemined that the predicted shift in RT c rresponding to the end of. life fluence for the limiting weld NDT in the beltline. region of the reactor vessel exceeds 100*F. For shifts
, in RT th't "I' 9' eater than 100*F but less than 200*F, NDT Paragraph II.C.3.b of Appendix H requires that four surveillance l capsules be included in the surveillance program. Provide technical. ' justification for the fact that the Shoreham Unit I surveillance program has three rather than four surveillance capsules. 121.38 In response to Question 121.16 the applicant provided information concerning the location and properties of the. beltline plates and welds. However, additional information and clarification are needed to ensure an accurate representation of the reactor vessel beltline. The applicant must provide the' exact weld materials (wire, flux), process and heat treatment used for each separately identified weld seam in Figure 121.18. The applicant cust also provide the calculated j value of initial RT for each individual plate and weld seam NDT identified in Figure 121.18; a listing of only the data obtained in accordance with the construction code requirements is not sufficient. For. clarity, the information requested here should be placed in the text of the response to Question 121.18. i 1i.
223.92 In order to assure long term reliability of the diesel generator installations we require that the following procedural modifications be implemented prior to the first refueling: Test Loading: The operating procedures shall be developed to require loading the engine up to a minimum of 25 percent of full load for one hour after eight hours of continuous no load operation, or as recommended by the diesel engine manufacturer.
231.9 The NRC staff has been generically evaluating three materials models-nat are used in ECCS evaluations. Those models predict cladding rupture temperature, cladding burst strain, and fuel assembly ficw blockage. We have (a) discussed our evaluation with venders and other industry repre-sentatives (Reference 1), (b) published NUREG-0630, " Cladding Swelling and Rupture Models for LOCA Analysis" (Reference 2), and (c) required licensees to confirm that their operating reactors would continue to be in conformance with 10 CFR 50.46 if the NUREG-0630 models were substituted for the present materials models in their ECCS evaluations and certain other compensatory modelchangeswereallcwed(References 3anc4). Until we have ecmpleted our generic review and implemented new acceotance criteria for cladding models, we will require that the ECCS analyses in your FSAR be accompanied by supplemental calculations to be performed with tne materials models of NUREG-C630. For these supplemental calculations only, we will accept cther compensatory model changes that may not yet be approvec by the NRC, but are censistent with the changes allowed for the confirmatory operating reactor calculations mentioned above. REFERENCES 1 1. Memorancum from R. P. Cenise, NRC, to R. J. Mattson, " Summary Minutes of Meeting on Cladding Rupture Temperature, Cladding Strain, ano Assembly Ficw Blockage," Novemcer 20, 1979. 2. D. A. Powers and R. O. Meyer, " Cladding Swelling and Rupture Mcdels for LOCA Analy' sis," NRC Report NUREG-0630, April 1980. 3. Letter frcm D. G. Eisenhut, NRC, to all Operating Lign: Water Reactors, dated November 9,1979. 4. Memorandum from H. R. Centen, NRC, to Ccmmissioners, " Potential Ceficiencies in ECCS Evaluation Mcdels," November 26, 1979. e e t I
- ~- O. i . ADDITIONAL INFORMATION REQUIRED BY 1 CHEMICAL ENGINEERING BRANCH i FROM SHOREHAM NUCLEAR POWER STATION, UNIT NO. 1 i. L 281.1 Describe how resin transfers will be monitored in the reactor (5.5.8) water cleanup system (acceptance criterion 1.d of SRP 5.4.8). 281.2 Table 5.2.3-2 sp(cifies the conductivity and chloride concen-(5.5.8) tration limits for~the reactor water to be 2 umho/cm and 0.1 ppm, t respectively, during reactor operation up to.10 percent of rated ( i power. Table 1 of Regulatory Guide 1.56, revision 1, specifies the same limits, but for power. operation at steaming rate less than one percent of rated steam flow. Verify that steaming i rates will be less than one percent of the rated steam ficw at i power levels up to 10 percent of the rated power. } 281.3 Summarize the procedures for determining the pH, chloride con-(5.5.8) centrations, and conductivity in the reactor vessel water (regulatory position C.6 of Regualtory Guide 1.56, revisicn 1). 281.4 Your FSAR does not indicate that chemical analysis for suspended (10.4.6) impurities will be performed in accordance with regulatory position c.1 of. Regulatory Guide 1.56, revision 1. Verify that such i analyses are to be performed and state the sampling and analysis frequency and established-limits and the basis for such limits. 281.5 Describe the water chemistry control program to assure maintaining (10.4.6) the condensate conductivity within the limits of Table 2 of Regulatory Guide 1.56, revision 1. Include conductivity meter alarm set points and the corrective actions to be taken when the limits of Table 2 are exceeded. T 4 1 -) 4 1
HYDROLOGIC ENGINEERING GENERIC QUESTIONS RELATING TO E.O.11988 FLOODPLAIN IGNAGEMENT FOR PLANTS NEAR COMPLETION Definition (from Executive Order 11988 Floodplain Management) Floodplain: The lowland and relatively flat areas adjoining inland and coastal waters including floodprone areas of offshore islands, including at a minimum that area subject to a one percent or greater chance of flooding in any given year. 371.16 Provide descriptions of the floodplains of all water bodies, including inter-mittant water courses; within or adjacent to the site. On a suitable scale map provide delineations of those areas $ hat will be flooded during the one-percent chance flood.in the absence of plant effects (i.e., pre-construction ficodplain). 371.17 Provide details of the methods used to determine the ficodplains in response to 1. above. Include your assumptions of and bases for the pertinent parameters used in the computation of the one-percent flood ficw and water elevation. If studies approved by Flood Insurance Administration (FIA), Housing and Urban Development (HUD) or the Corps of Engineers are available for the site or adjoining area, the details of analyses need not be supplied. You can instead provide the reports from which you obtained the floodplain infor-mation. 371.18 Identify, locate on a map, and describe all structures and topographic alterations in the floodplains. t l l l e
F~ 2-371.19 Discuss the hydrologic effects of all items identified in 3. above. Discuss the potential for altered flood flows and levels, both upstream and down-stream. Include the potential effect of debris accumulating on the plant structures. Additionally, discuss the effects of debris generated from - the site on downstream facilities. 371.20 Provide the details of your analysis used in response to 4. above. The level of detail is similar to that identified in item 2. above. '{ f 4 .4 ( l =}}