ML20246F148
| ML20246F148 | |
| Person / Time | |
|---|---|
| Issue date: | 06/26/1989 |
| From: | Kenyon T Office of Nuclear Reactor Regulation |
| To: | Scherer A ABB COMBUSTION ENGINEERING NUCLEAR FUEL (FORMERLY |
| References | |
| PROJECT-675A, PROJECT-675F NUDOCS 8907130132 | |
| Download: ML20246F148 (15) | |
Text
{{#Wiki_filter:- ----_ g A. June 26, 1989 g LProject No. 675.* 1 Mr. A. E. Scherer,. Director. Nuclear Licensing Combustion Engineering 1000 Prospect Hill Road Post Office Box 500: Windsor, Connecticut. 06095-0500-
Dear Mr. Scherer:
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION ON CESSAR-DC, SYSTEM 80+ As a result of our review of Chapters 1, 3, 6, 7, 9, 12 and 13 of the System 80+ Standard Design, Amendment E of CESSAR-DC, we require additional information in order to complete our review of the design. The additional information is needed in the areas of emergency preparedness, plant systems, reactor systems,; chemistry, radiation protection, and reactor safeguards and is covered in the enclosed questions. Please respond to this request within 90 days of the date of this letter. If-you have any questions regarding this matter, call me at (301) 492-1120.- Sincerely, q + Thomas J. Keny n, Actin roject Manager Standardization and Non-Power Reactor Project Directorate Division of Reactor P.rojects - III, IV, Y and Special Projects' Office of Nuclear Reactor Regulation As stated cc: See next page . DISTRIBUTION: eDocketxt Hee NRC & Local PDRs PDSNP R/F Ellylton TKenyon -{ EJordan 1' BGrines OGC-Rockville ON3 ACRS (10) . LTR TO MR. SCHERER) II 4[' ( af n i t w ~fd L P4 , NP D:PDSNP bon TK on:cw Chiller 06/}/89 06/p/89 06/lf/89 l 8907130132 090629 [ QPh i PDR PROJ 1 675A PNV
e i %} ' UNITED STATES NUCLEAR REGULATORY COMMISSION '~ [ g, D j-WASHINGTON, D C. 20555 e-June 26, 1989 ...../ Project No. 675 Mr. A. E. Scherer, Director Nuclear Licensing Combustion Engineering 1000 Prospect Hill Road Post Office Box 500 Windsor, Connecticut 06095-0500
Dear Mr. Scherer:
SUBJECT:
REQUEST-FOR ADDITIONAL INFORMATION ON CESSAR-DC, SYSTEM 80+ As a result of our review of Chapters 1, 3, 6, 7, 9,12 and 13 of the System 80+ Standard Design, Amendment E of CESSAR-DC, we require additional information in order to complete our review of the design. 'The additional information is needed in the areas of emergency preparedness, plant systems, reactor systems, chemistry, radiation protection, and reactor safeguards and is covered in the enclosed questions. Please respond to this request within 90 days of the date of this letter. If you have any questions regarding this matter, call me at (301) 492-1120. Sincerely, Thomas J. K [nyon, Acting Project Manager ,<W \\ e i Standardization and Non-Power Reactor Project Directorate 1 Division of Reactor Projects - III, IV, V and Special Projects Office of Nuclear Reactor Regulation j Enclosure. As stated cc: See next page
l .t 1 'b \\ ; *. ~* 1 1 .Cos6ustion Engineering, Inc. Project No. 675 j-Advanced CESSAR l cc: Mr. C. B. Brinkman, Manager l-Washington Nuclear Operations Combustion Engineering, Inc. 12300 Twinbrook Parkway Suite 330 l Rockville, Maryland 20852 Mr. Ernest Kennedy Manager of Licensing. Combustion Engineering 1000 Prospect Hill Road Post Office Box 500 Windsor, Connecticut 06095-0500 t' l I
ENCLOSURE Request For Additional Information Chapter 13 of CESSAR-DC provides extensive description of design 8102 requirements for the Technical Support Center (TSC) and Emergency Opera-tionsFacility(EOF). Emergency preparedness regulations and related guidelines contain requirements and guidance for facilities and functions in addition to the TSC and EOF. For the additional facilities listed below, (1) provide a de crkption of the pertinent design requirements which would enable these f acilities to meet the referenced requirements or guidance, or (2) cite the location of these descriptions in current or projected design requirements, or (3) describe or identify how the equivalent function is contained in might be reflected in Control Room design)y (e.g., many OSC func the design requirements of another facilit Operations Support Center (OSC) (
Reference:
NUREG-0654,611.H.9; a. NUREG-0696,11.3.2,13.0) b. Laboratory Facilities (fixed or mobile) (
Reference:
NUREG-0654, 11.H.6.c;NUREG-0737.II.B.3) Post Accident Sampling System (
Reference:
NUREG-0737,II.B.3) c. d. Onsite Decentaminaticn Facility (
Reference:
10 CFR 54, Appendix E. IV.E.3; 10 CFR 50.47(b)(8)) m
REQUEST FOR ADDITIONAL INFORMATION m 281.57 Provide a technical analysis and evaluation of the containment (6.5) spray system's effectiveness in reducing containment pressure and temperature and lowering radioisotope relea u during postulated dominant severe accident sequences. Discust ;,pecific system design features for enhancing the mitigation of severe accident consequences. 281.58 Provide specific results of a failure modes and effects analysis (6.5) (FMEA) of the containment spray system showing that the system is capable of withstanding a single failure without loss of function. 281.59 Provide a detailed description and evaluation of all systems that "- (6. 5) interface with or support the containment spray system. This should include the potential for support / interfacing system single failures rendering the containment spray system inoperable and all potential systems interactions which could degrade plant safety. 281.60 Provide the following information in order to permit the staff to (6.5) perform an integrated review of the containment spray system (CSS): a. Legible copies of the CSS Piping and Instrumentation Diagrams (P& ids), b. CSS heat exchanger fouling factors (design and expected values), c. Design capacity of each CSS train, and d. Technical data for CSS backup water source (outside containment) including source, transfer capability, pressure andflowdata(seeCESSAR-DCSection6.3.2.2.1). 281.61 Provide the sprayed and unsprayed containment volumes and post-(6.5) accident containment mixing features to ensure acceptable spray 8' coverage of the entire containment per the guidance of SRP 6.5.2, Sections II.I.b and II.1.c. 410.47 Provide a table listing the following parameters which are used to 'i (3.6.1) evaluate postulated piping faf 7ures in fluid systems: ]
- actual pipe dimensions, system locations, (c
piping drawings, (d) design temperatures, and i (e) design pressures ) -_________o
410.48 Providethefollowinginformation(nowshownas(LATER")in (3.6.1.2) order to permit the staff to perform an integrated review of the postulated piping failures in fluid systems: a) Completed CESSAR-DC Tables 3.6-3 concerning high energy lines within containment), b) CESSAR-DC Section 3.5.4.D concerning cross-reference sections for interface requirements on missile protection, c). CompletedTable3.2-1(sheet 4of6),classificationof structures, systems, and components concerning the component cooling water system, spent fuel pool cooling and cleanup system, and station service water system, and d) Completed Table 3.2-4, summary of criteria - structures. 410.49-Clarify the criteria used for protection against the dynamic .(3.6.1.1) effects associated with pos o lated piping failures. Discuss how these criteria meet the guivance of BTP ASB 3-1, and GDC 4 which require the following: (a adequate physical separation and remote location, (b suitably designed protective enclosure, and (c restraints and protective measures. 410.50 Identify the " potential hazards and highlighted susceptibilities" (3.6.1.2)whicharebeingdeveloped,asstated,inSARSection3.6.1.2.C. Identify the design changes which have resulted from this ongoing review. 410.51 For the spent fuel storage cooling analysis in SAR Section 9.1.2.3: (9.1.2.3)(a) Discuss the spent fuel pool storage rack design features which enhance natural convection water circulation within the pool and adequate flow to all rack locations in the pool, and (b) Provide an evaluation of the thermal performance and hydraulic stability of the spent fuel storage racks for all postulated 8, normal and accident conditions. Include analysis for a dropped fuel assembly which is reducing the flow area above fuel storage locations in the pool. 410.52 Provide an evaluation of the containment design features which (9.1.4) 1.reclude any postulated leak or failure of the reactor cavity rsfteling pool seal or mitigate / preclude any level reduction in the spent fuel and refueling pools. i
^ Section 9.1.3.3.2 discusses the possibility of an accidental 410.53 (9.1.3.3.2) opening of the gate between the spent fuel pool and a dry transfer canal and the resulting decrease in spent fuel pool level. Provide an evaluation on the effect of this reduced pool level on spent fuel pool pump operation in light of the elevated location of suction and discharge lines and NPSH requirements. The safety evaluation of both the new and spent fuel storage areas 410.54 (9.I.2, includes an evaluation of the effects of dropping a fuel assembly 9.1.4) and its handling tool from a height of two feet above the storage rack. Provide the following additional information in accordance with SRP 9.1.2, Item III.2.e guidance: Verify that the drop of any allowed lighter loads at a greater height does not result in a higher potential energy than a fuel assembly and its handling tool dropped Perform an evaluation of this from its normal operating elevation. in accordance with SRP 9.1.4 guidance. Provide the following information in order to permit the staff to I 410.55 (9.1.3) perform an integrated review of the spent fuel pool cooling and cleanupsystem(SFPCCS): (a) DesignparametersformajorSFPCCScomponents(e.g., pumps, heat exchangers, tank, filters, demineralizers). Include the following minimum information on SFPCCS heat exchangers: Heat exchanger tube surface area (sq'uarte feet). Heatexchangerconductance(Btu /ftr*F), Spentfuelpoolwaterflowrate,perpump(Lb/Hr), (4 Component cooling water flowrate, per heat exchanger (Lb/Hr),and (5) Design component cooling water inlet temperature to the heat exchanger (*F), (b) System interface requirements, SFPCCS design provisions which permit appropriate inservice (c) inspection and function *' testing as stated in SRP 9.1.3, Section III.1.g guideli.x, and SFPCCS design provisions to maintain acceptable pool water (d) conditions per SRP 9.1.3, Section III.7 guidance in the t i following areas: (1 pool mixing, (2 adequate system capacity,
- (3 acceptable instrumentation and sampling capability, (4) refueling canal coolant processing ability, and features to prevent the inadvertent transfer of spent (5) filter and demineralized media to any place other than the radwaste facility.
e i 4 l Provide the heat generation rate calculations using NUREG-0800, ion Standard Review Plan, Branch Technical Position ASB 9-2 and Sect 410.56 (9.1.3) 9.1.3 guidance for the following cases: Normal refueling until the spent fuel pool is full, (a) in the spent fuel pool are filled by a core offload. (b) Explain the apparent discrepancy in the quantity of spent fuel stored as stated in SAR Section 9.1.2 and that in SAR Sec 410.57 9.1.3.1.4 in accordance with SRP 9.1.3, Section III.1.c guidance. (9.1.2, Also, explain how the design spent fuel storage capacity relates to 9.1.3) the design life of the power plant and the expec during this lifetime. Provide the design information necessary to ensure that in the event failure of drains, inlets, outlets, or piping will not result in 410.58 the spent fuel pool level inadvertently dropping below a point -(9.1.3) approximately ten feet above the top of the activ.e fuel in accordance with SRP 9.1.3, Section III.1.e guidance. Provide SFPCCS information which assures that leakage detection, component / header isolation capability, and inter-system leakage 410.59 provisions are incorporated in this design per guidance of SRP (9.1.3) 9.1.3, Secti,on III.3. Explain the discrepancy in stating that the maximum pool tempe is 150'F in SAR Section 9.1.2.3.5 and that the naximum pool 410.60 temperature is 140*F in SAR Section 9.1.3.1.4. (9.1.2, 9.1.3) Provide an evaluation that assures that any failures in the nonsa related spent fuel pool cleanup and associated systems cannot a 410.61 the functicasi performance of any safety-related components or (9.1.3) systems in accordance with SRP 9.1.3, Section III.5 guidance. You have stated under SAR Section 9.1.3.2.1 that "The spent f In receives normal borated water makeup from a wate 410.62 (9.1.3) i and/or hoses from an alternate water source.information co source" makeup system including related technical data and cross Also, update Figure 9.1-3, Spent Fuel Pool Cooling and Ciennup P&ID concerning the above makeup information. references. " Containment Systems (To Be Revised in Submittal Sec' tion 6.2 states: Group F)" Provide the revised Section 6.2. 480.7 (6.2.1, 6.2.6) ,/
w-REQUEST FOR ADDITIONAL INFORMATION SECTION 9,1 440.5 In addition to assurance that k,ff is less than 0.98 with optimum moderation, the new fuel storage design bases should also include assurance that k,ff is less than 0.95 in the event the fuel area becomes fully flooded with full density unborated, pure water. 440.6 The acceptability of the calculational methods (DOT-4 and KENO-IV) and the qualification of CE in their use should be documented either by inr,luding benchmark calculations performed by CE with these - methods or by referencing previpus NRC approval of CE use of these methods. 440.7 Include a discussion of the method bias and uncertainty as well as other uncertainties considered such as those due to variations in the mechanical and material specifications from their nominal values. Verify that these uncertainties are combined with k,ff to provide a one-sided, upper tolerance limit on k,ff equivalent to a 95/95 probability confidence level for fuel storage calculations. 440.8 Explain why Paragraph C of Section 9.1.2.3.1.2 refers to an assumed boron concentration of at least 2000 pro in the spent fuel pool in evaluatirp a dropped fuel asserbly accident whereas Par 6 graph A of Section 9.1.2.3.1.3 implies that less than one-half of normal (about 1000 ppm)isassumed. 440.9 Explain what is meant in Section 9.3.2.3.3.3 by " borated" or " mixed" modes and which neutron absorption ef fects is credit taken for. This paragraph also seems to imply a two-region pool with burnup i credit allowed. However, this is not described in the spent fuci pool storage rack description in Section 9.1.2.2.2. L
~ s REQUEST FOR ADDITIONAL INFORMATION 281-32 Section 6.5.1.3.K., Chemistry and Sampling, indicates that the containment spray system is designed for 2.5 w/o boric acid at a pH of 77.0. Discuss the spray additive or pH control system and describe how it meets Standard Review Plan Section 6.5.2, Containment Spray As A Fission Product Cleanup System and meets the requirements of GDC 41, 42 and 43. 281-33 Section 9.1.2.2.2, Spent Fuel Pool Stora';e Racks, indicates that the structural design of the spent fuel rack and poo includes provisions for neutron poison inserts to meet future expansion potential. Since this is a likely situation based on current experience, the spent fuel racks with neutron poison inserts should be considered in the reference design. For a spent fuel rack design that includes neutron poison inserts, a coupon surveillance program should be included to monitor the performance of the neutron poison material in the spent fuel pool environment. 281-34 Describe the instrumentation and sampling to monitor the water purity and need for demineralized resin replacement including the chemical and radio-chemical limits and demineralized differential pressure used to initiate corrective action (Section 9.1.3.3.3). O e l q
REQUEST FOR ADDITIONAL INFORMATION 471.1 Chapter 12.0, Radiation Protection, states that: this chapter describes the radiation protection neasures incorporated in the station design and in the operating procedures to ensure that... Section 12.1.3, Operational Considerations, states that: this section to be provided hy site operator. These two statements are contradictory. Please correct this discrepan cy. 471.2 Section 12.1.3, Policy Considerations refers to Regulatory Guide 8.2, " Guide for Administrative Prectices in Radiation Monitoring". It appears that the intent was to reference to Regulatory Guide 8.8. "Information Relevant to Ensuring that Occupational Radiation Expo-sures at #cclear Power Stations Will Be As Low As is Reasonably L Achievable,* rather than R.G. 8.27 Please clarify or correct. I' l 471.3 Ir. section 12.3.2, Design Considerations, paragraph B, states that: equipnEnt is designed to minimite crud buildup and facilitate deCon-tamination; paragraph C, states that speces are provided where I { _ = _ _ - _ _ _ _ _ - - - J
~~ + appropriate to place shielding for the purpose of reducing neutron activation.
- r Please provide examples for statements in paragraphs e and C.
471.4 Sane section as in 471.3, paragraph D, sta'tes that: activated corrosion product buildt,p over the plant life of the design is minimized in the design stage through appropriate selection of corrosion resistant materials... Please clarify the statement: ...over the plant life of the design.... 471.8 In Table 12.2-1, Maximum Heutron Spectra Outside Reactor Vessel, Column: Average Neutron Energy (Mev), it appears that the neutron energy, 3.3 x 10, is incorrect - please verify. ~471.9 In Ttble 12.2-6, CVCS Heat Exchanger Soluble Inventories, Column: Luclide,' I-33 should be revised to 1-133. 471.11 Section 12.3.1.2, Equipment and System Design Features for Control of Onsite Exposure, paragraph G Piping, states that: " design recommendations and information for keeping in-plant personnel exposure ALARA are provided. The following information and recommen-dations are provided. 1. Interface criteria and radiation source terms are provided...". Please specify where this information is provided, or revise the above sentences.
~ .c 2,. I . 471.12 Section 12.3.1.2, paragraph M. Refueling Equipment, first sentence states "1.. All spent fuel transfer and storage operations are desiered to be conducted underwater to insure adequate shielding and to limit the maximum continuous radiation levels.in working areas." ~ Please identify all accessible areas where personnel could receive a radiation dose of 100 rem in one hour, and describe design consider-ations that will ensure that a potentially lethal overexposure of l personnel would not occur. Particular attention should be focused on transient very high radiation areas (e.g., area adjacent to spent fuel transfer tubes), as well as areas with intense sources of radiation continuously present. l' lease addrest the compliance with 10 CFR 20, " Standards for 471.13 Protect. ion Against Radiation," and if exemptions are contemplated, acceptable alternative rethods should be proposed. 1 _____.__m._:___m_._-____-._____
l.. 471.14 Section 12.2.1.1.5, ChemicalandVolumeControlSystem(CVCS),- paragraph B.1 states, in part, that "All nuclides except Xe, Kr, Rb, and Cs have a decontamination factor (DF) of 10 and efficiency of 901, Xe and.Cs have (DT) of 1.0 and efficiency of 0%, Rb and Cs have a DF of 2.0, and efficiency of 501." Please review the accuracy of DF and efficiency for Cs, and specify DF and efficiency for Kr. 471.15 Section 12.2.1.1.5, CVCS, please justify the large difference e between the data quoted in B.I. and b.2. for Rb and Cs. i 1 m
l.. o' REQUEST FOR ADDITIONAL INFORMATION 500.13 The Comission's Severe Accident Policy Statement included the policy ~ that: "The issues of both insider and outsider sabotage threats.. will be emphasized in the design and in the operating procedures developed for newplants."(emphasisadded) Also, NUREG/CR-2643, "A Review of Selected Methods for Protecting Against S.abotage by an Insider", concluded that effective insider protection will ~ requireanintegratedapproachthatincludesthebestfeaturesof(1) physical protection measures, (2) damage control measures, and (3) plant design measures. The physical protection measures studied in that report all involved some impact on site work rules and procedures. However, CESSAR-DC Revision E Section 13.5, Plant Procedures, states that "the site operator's plant procedures is within the site operator's scope and shall be provided in the site-specific SAR." Such a blanket statement seems to remove the possibility of including procedural constraints as a part of the standaro design sabotage protection design philosophy. 500.14 Section 1. s.13, Physical Plant Security and Protection from Sabotage", states that these design features are described in Chapters 2, 3, 7, 8, and 9. To assist in our review, please specify where in those chapters we should find such features. f I 1 l}}