ML19326B795
| ML19326B795 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 12/13/1971 |
| From: | Morris P US ATOMIC ENERGY COMMISSION (AEC) |
| To: | Holmes H ARKANSAS POWER & LIGHT CO. |
| References | |
| NUDOCS 8004180549 | |
| Download: ML19326B795 (7) | |
Text
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DR. Reading VHWilson, DRL (2)
PWR Branch 3 Reading ACRS (16)
AEC PDR Seismic Consultant Local PDR EJBloch, DR
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CKBeck, DR erson, m Docket No. 50-313 SHanauer, DR FSchroeder, DRL Mr. J. D. Phillips TRWilson, DRL Vice President and Gief Engineer RSBoyd, DRL Arkansas Power & Light company DJSkovholt, DRL Sixth and Pine Streets HRDenton, DRL Fine Bluff, Arkansas 71601 RCDeYoung, DRL RWKlecker, DRL DRL/DRS Branch Giefs
Dear Mr. Phillips:
I indicated in my letter of November 1,1971, that further requests for additional information would be transmitted to you as our review of your application for an operating license for the Arkansas Nuclear One - Unit
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No. 1 facility progressed. Additional information required before we can complete our review is described in the enclosure to this letter.
Please contact us if you have any questions regarJing the additional information required.
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Sincerely, 1
R. C. DeYoung, Assistant Director for Pressurized Water Reactors Division of Reactor Licensing
Enclosure:
Additional Information Required cc:
Mr. Harlan T. Holmes Nuclear Project Manager Arkansas Power & Light Company Sixth '& Fine Streets Pine'31uff, Arkansas 71601 Mr. Horace Jewell House, Holas, & Jewell 1550 Tower 9milding Little Rock, Arkansas 72201 Mr. Roy B. Snapp 1725 K Street, N. W.
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f ADDITIONAL INFORMATION REQUIRED
- ARKANSAS PO((ER & LIGHT COMPAhT ARKANSAS NUCLEAR ONE - UNIT 1 1
DOCKET No. 50-313 1.0 CENE RAL 1.8 Discuss the considerations given to accidents involving commer'cial traffic on nearby roads, railroads and navigable waterways in the design and in the preparation of operating procedures for the plant.
Consider the potential for explosions, fires, and release of toxic gases due to such accidents and their potential effect on the safety of the nuclear plant.
2.0 SITE AND ENVIRONMENT 2.6 Provide maps of the plant site and surrounding areas of suitable scale, that* clearly indicate the following:
2.6.1 1 hat portion of the plant site co be established as a
" restricted area", as defined in 10 CFR Part 20.3.
2.6.2 The boundary that you propose be used to establish technical specification limits for radioactive gaseous effluents.
2.6.3 Each point within the facility from which gaseous effluents containing or potentially containing radioactivity may be released and the distance of each from the nearest boundary line in 2.6.2 above.
2.6.4 The location and nature of any non-plant related activities (e.g., farming, picnic areas, camps) that will be conducted on the plant site.
2.6.5 The Low Population Zone and location of all schools, hospi-tals, institutions and other facilities that may require special consideration in the implemestation of the emergency plan.
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,s 5.0 STRUCTURES 5.61 State your criteria for selection of protective coatings and paints for use within the containment to withstand accident conditions, including consideration of radiation, boric acid spray washdown, steam environdent, and jet impingement effects. Include an evaluation of the potential impairment of the performance capa-
- bilities of engineered safety features due to flo'w blockage, fouling of heat transfer surfaces, or other events that might result from failure of the protective coatings and paints.
5.62 Identify any aspects of your program for testing reinforcing bars used in Class I concrete structures that do not conform to Safety Guide 15 and indicate your bases for believing that these are,
acceptable.
5.63 Identi.fy any aspects of your containment structural acceptance test that will not conform to Safety Guide 18 and indicate your bases for believing that these are acceptable.
5.64 Provide the missing information in Table 5.1 pertaining to con-tainment penetrations 25 and 45.
5.65 Does the -isolation valving for the containment penetrations satisfy General Design Criteria Nos. 54, 55, 56 and 57? For each penetration for which the valving does not satisfy these criteria, indicate your bases for believing that the design is acceptable.
7.0 INSTRUMENTATION AND CONTROL 7.21 We note that all the nuclear power range channels required for reactor protection are also used for control functions. Discuss how this design meets the requirements of Section 4.7 of IEEE-279 (1971) with particular emphasis on (1) results of tests and analyses 3
that verify the isolation devices can withstand all credible faults without preventing the protection, system from meeting 11ts minimum
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performance requirements, (2) measures proposed to verify that degradation of the isolation devices during the plant operating life will not violate the requirements of IEEE-279 and (3) the methods used to isolate the averaging networks from the protection system.
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l' 9.0 AUXILIARY' AND EBtERGENCY SYSTEMS 9.4 Identify any aspects of the fuel storage facility that do not conform to the provisions of Safety Guide 13 and provide your basis
-for believing that these are acceptable.
10.0 STEAM AND POWER CON"ERSION SYSTEMS 10.1 The turbine stop valves serve as containment isolation valves and for isolation of the unaffected steam generator in the event of a steamline rupture accident. Provide information on the design, operation, inspection and testing of these valves relative to these functions.
Indicate whether the control system that closes these valves meets IEEE-279 criteria.
10.2 Provide information describing the locations of both emergency feedwater pumps, the location of the atmospheric exhaust associated with the steam turbine driven pump and a escription of how the d
electrically driven pump can be manually connected to the diesel generator busses. Estimate the time associated with such a transfer of power sources.
11.0 RADIOACTIVE WASTE AND RADIATION PROTECTION 11.6-The Commission, on June 9,1971, published for comment a proposed Appendix I " Numerical Guides for Design Objectives and Limiting Conditions 'for Operation to Meet the criterion ' As Low As Practicable' for Radioactive Material in Light-Water-Cooled Nuclear Power Peactor Effluents" to 10 CFR Part 50.
Review your waste-processing system design in lig'ht of the new Appendix I and evaluate whether each of the radioactive waste processes meet the "As Low As Practicable" criterion. Identify each effluent stream from the plant and justify why further processing i
is not practicable.
Calculate the~ total radiological impact of operation of the ARK-1 plant on the environs in view of the proposed numerical guides stated in Appendix I.
Consider (1) operation with expected levels of radioactivity in the primary and secondary systems, and (2) operation with technical specification limits of radioactivity and primary to secondary system leakage.
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12.0 CONDUCT OF OPERATIONS 12.10 Provide information to demonstrate compliance with the' provisions contained in Safety Guide 17, Protection Against Industrial Sabotage.
' 14.0 SAETY ANALYSIS 14.1 Discuss the probable consecuences associated with the following types of fuel loading errors.
Indicate by what means and with what degree of confidence these loading errors could be detected after the fuel assembly was placed in the reactor.
14.1.1 Erroneous enrichment loading of fuel pellets or fuel pinc
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during the fabrication process, 14.1.2 Erroneous location or orientation of fuel assemblies during initial core loading or during subsequent refueling operations.
14.2 Relative to the steam line failure accident analysis :
14.2.1 Justify the use of a moderator temperature coefficient i of -3.0 x 10-4 6 k/k/0F for this analysis.
s 14,.2.2 Provide the results of -
snalysis indicating the effects on the primary system
..g this accident.
Include the time-history tracas for all significant parameters such as DNBR, pressure, temperature, pressurizer level, and reactivity.
Indicate the time sequences for significant events such'as reactor trip, turbine valve closure, feedwater valve: closure, steam dump and bypass valve operation, relief valve operation, and high pressure injection actuation.
14.2.3 We note that you have analyzed this decident assuming certain
. operator actions in one case and no operator action in an-other case. ' The results obtained appear to be identical.
Provide a' clarification of what actions the operator is expected-to perforr during this accident and discuss whether these activns can significantly ef fect the potential consequences of the accident.
y 14.3 Submit a, summary of a detailed analysis to show that the plant can
't-be shutdown safely in the event of a loss of all AC power. (offsite.
and onsite), including:
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S-14.3.1 A list of all necessary components, control systems, instrumentation and other necessary electrical systems that must be supplied from the station batteries and discuss how power will be provided from the batteries. Estimate the power consumption of each component and indicate the maximum time the plant could maintain this condition safely.
14.3.2 Section 14.1.2.8.4 of the FSAR indtcates that in the event the condensate scorage tank is unavailable as a source of
~ emergency feedwa:er, an alternate supply is availabic from the service water system. Provide an analysis showing that NPSH requirements for both the steam driven and motor driven emergency feed pumps can be met in the event suction must be taken through the idle service water pumps.
14.4 Relative to your analysis of the potential consequences of a loss of load incident:
14.4.1 Justify the use of 10 CFR Part 100 limits as acceptable criteria for radiological doses associated with this event.
14.4.2 Describe the testing that will be performed during the startup test program to verify the assumptions and results of your analyses of the loss of load incident.'
15.0 TECHNICAL SPECIFICATIONS 15.1 The proposed technical specifications submitted in Amendment 20 are not complete. Submit a complete set of, proposed : technical specifications, preferably based on the set recently developed for Unit 1 of the Oconee Station and identify any specifications that are different and the bases for these differences. Note that the reporting requirements of your proposed technical specifications should be based on Safety Guide 16, Reporting of Operating Informa-tion.
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