ML20155K515

Forwards Request for Addl Info Re GE Application for Certification of Advanced BWR Design.Info Includes Suppression Pool Water,Condensate Storage Tank,Discharge Line Fill Pump & Standby Liquid Control Sys

ML20155K515
Person / Time
Site:	05000605
Issue date:	10/26/1988
From:	Scaletti D Office of Nuclear Reactor Regulation
To:	Marriott P GENERAL ELECTRIC CO.
References
NUDOCS 8811010278
Download: ML20155K515 (11)

Text

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OCT 2 6 1988 i

Docket No. STN 50-605 Mr. Patrick W. Marriott, Manager Licensing & Consulting Services GE Nuclear Energy General Electric Company 175 Curtner Avenue San Jose, California 95125

SUBJECT:

REQUEST FOR ADDITIONAL INFORMATION REGARDING THE GENERAL ELECTRIC ,

COMPANY APPLICATION FOR CER11FICATION OF THE ABWR DESIGN  !

In our review of your application for certification of your Advanced Boiling Water Reactor Design, we have identified a need for additional information.

Our request for additional information, contained in the enclosure, addresses the areas of SRP Chapters 6.3, 9.3.5 and 15 reviewed by the Reactor Systems Branch. This completes the initial request for additional information on the ABWR related to SSAR Chapters 4, 5, 6 & 15. However, the need for additional information may occur during the development of the staff's safety evaluation.

If this should occur, the need will be identified in a draft Safety Evaluation Report which will be provided for your consideration.

In order for us to maintain the ABWR review schedule, we request that you provide your responses to this request by December 1,1988. If you have any concerns regarding this request please call r.e on (301) 492-1104.

Sincerely, IS/

Dino C. Scaletti, Project Manager Standardization and Non-Power Reactor Project Directorate

Division of Reactor Projects - I!!, IV, Y and Special Projects Office of Nuclear Reactor Regulation

Enclosure:

As stated cc: See next page DISTRIBUTION:

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• WASHINGTON, D. C. 20655 i j

%,,,,, OCT 2 61988 Docket No. STN 50-605 Mr. Patrick W. Marriott, Manager Licensing & Consulting Services GE Nuclear Energy General Electric Company 175 Curtner Avenue San Jose, California 95125

SUBJECT:

REQUEST FOR ADDITIONAL INFORMATION REGARDING THE GENERAL ELECTRIC COMPANY APPLICATION FOR CERTIFICATION OF THE ABWR DESIGN In our review of your application for certification of your Advanced Boiling Water Reactor Design, we have identified a need for additional information.

Our request for additional inforr.ation, contained in the enclosure, addresses the areas of SRP Chapters 6.3, 9.3.5 and 15 reviewed by the Reactor Systems Branch. This completes the initial request for additional information on the ABWR related to SSAR Chapters 4, 5, 6 & 15. However, the need for additional information may occur during the development of the staff's safety evaluation.

If this should occur, the need will be identified in a draf t Safety Evaluation Report which will be provided for your consideration.

In order for us to maintain the ABWR review schedule, we request that you provide your responses to this request by December 1,1988. If you have any concerns regarding this request please call me on (301) 492-1104.

Sincerely, (htb .

C. Scalett , Project Manager Standardizatlon and Non-Power Reactor Project Directorate Division of Reactor Projects - III, IV, Y and Special Projects Office of Nuclear Reactor Regulation

Enclosure:

As stated cc: See next page i

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, GE-ABWR Docket No. STN 50-605 cc: Mr. Robert Mitchell General Electric Company 175 Curtner Avenue San Jose, California 95114 f

Mr. L. Gifford, Program Manager Regulatory Programs GE Nuclear Energy 12300 Twinbrook Parkway Suite 315 Rockville, Maryland 20852 Director Criteria & Standards Division OfficeofRadiationPrograms U. S. Environmental Protection Agency 401 M Street, S.W.

Washington, D.C. 20460 ,

Mr. Daniel F. Giessing L Division of Nuclear Regulation and Safety Office of Converter Reactor Deployment, NE-12 ,

Office of Nuclear Energy Washington, D.C. 20545 i i

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ll ADVANCED BOILING WATER REACTOR (ABWR)

SSAR CHAPTERS 6.3. 9.3.5 AND 15 440.75 In the ABWR design, the HPCF is tested by taking suction from and (6.3) returning water to the suppression pool. Norwelly the suppression pool water is a lower quality than that of the CST; therefore, draining, flushing and refilling the system is required prior to returning the system to standby after testing. Please discuss the pros and cons of using the CST for testing the HPCF system.

440.74 Address the following THI-2 action items related to ECCS.

II.K.1.5  !

(6.3) (a)

(b) II.K.1.10 (c) II.K.3.17 (d) II.K.3.18 (e) II.K.3.21 (f) II.K.3.25 (g) II.K.3.30 '

(h) II.K.3.31 440.77 Confirm that the HPCF system meets the guidelines of Regulatory (6.3)- Guide 1.1regardingpumpNetPositiveSuctionHead(NPSH).

440.78 SRP 6.3 identifies GDCs 35, 36 and 37 in the acceptance criteria.

(6.3) Confirm that the HPCF system, described in Chapter 6.3 of the SSAR, j meets the requirements of the above GDCs.

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440.79 Normally, the HPCF pump takes suction from the Condensate Storage (6.3) Tank (CST). But, the CST is not seismically qualified or safety related. Confirm that the system piping and level transmitters, which interface with CST, will be designed and installed such that the automatic switchover to the suppression pool takes place without failure.

440.80 What is the minimum quantity of water required in the condensate (6.3) storage tank (CST) for HPCF operation? Give the basis for the required quantity of water in the CST.

440.81 What is the closing time of test return valves F009.01C, F011.01C, (6.3) F015B and F01687 They should close earlier than 36 seconds to l L

prevent any flow diversion to the suppression pool during a LOCA.

440.82 Since HPCF is part of the ECCS network, the HPCF pump minimum flow (6.3) line should be designed to operate for a reasonable length of time.

How long can HPCF run in minimum flow sede?

440.83 In the resolution of TMI-2 Action Item II.K.3.13, the BWR Owners' (6.3) Group decided and the staff agreed to keep the initiating RPV level setpoint L2 for starting RCIC and HFCI systems. In ABWR design RCIC is still started at RPV level L2, but the HPCF is started at level 1.5. What is the basis for the initiating level 1.5 for HPCF7 440.84 In Section 7.3.1.1.1.2(3)(9) it is stated that HPCF pump discharge (6.3) pressure is used as a permissive to start ADS automatically. If HPCF is available ADS may not be requ' d. In the current BWR designs, only low pressure pumps discharge pressures, not HPCS, are used as permissive to start ADS. What is the basis for this change in ADS logic?

440.85 In SSAR Section 7.3.1.1.1.2, it is stated that ADS timer wili be set (6.3)- at 29 seconds. Submit the analysis to support the 29 seconds time i delay.

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40.86 SSAR Section 6.3.1.1.4, "ECCS Environmental Design Basis" refers to (6.3) SSAR Section 3.11 for Qualification of ECCS Equipment. However, Section 3.11 does not provide the required information. Provide the necessary equipment qualification information.

440.87 Confirm that there are provisions for equipment maintenance during (6.3) long-term coolant recirculation in the post LOCA environment for ECCS equipment.

440.88 Confim that long-term cooling capacity is adequate in the event of (6.3) failure of any single active or passive component of the ECCS.

440.89 in SSAR Section 6.3.6, References 1 and 4, latest approved revisions (6.3) of NEDE-29011-P-A are given as references. Identify the latest revisions which are used for ABWR.

440.90 In Table 6.3-6 "Plant variables with nominal and sensitivity study (6.3) values," Item #5 inetal water reaction rate, nominal value is given as "EPRI coefficients." Confirm that the EPRI coefficients are the same as used in the model already approved by the staff or identify the EPRI report which discusses these EPRI coefficients.

440.91 SSAR Sections 6.3.3.5 and 6.3.3.6 refer to Reference 4 instead of (6.3) eddressing the subjects "use of dual function components for ECCS" and "Limits on ECCS system paraineters." Briefly describe the above subjects in the SSAR.

440.92 List all computer codes used in the LOCA analysis and give a brief (6.3) description of each code.

440.93 Section 6.3.3.7.2 accident description refers to Reference 4. Provide (6.3) a brief description of the accident. For details Reference 4 can be

, used. ,

440.94 Why is there no discharge line fill pump provided for the HPCF (6.3) system? How does the system design reduce water hausner during the pump start-up?

440.95 List the capacity and settings of all relief valves provided for the (6.3) ECCS to satisfy system overpressure.

440.96 Revise SSAR Section 6.3.2.2.1 HPCF to include a description of relief (6.3) valves provided in the suction and discharge of the HPCF pump.

440.97 SSAR Table 5.4-2 gives the design parameters for RCIC system (6.3) components. Provide similar information for RHR and HPCF systems.

440.98 Confirm that 0.099ft2 is the lower limit of pipe break size for which (6.3) ECCS operation is required.

440.99 In the RCIC system description (Ref. 5.4.6.1.1.1) it is stated that (6.3) the mixture of the cool RCIC water and the hot steam quenches the steam. Since RCIC is injected to the reactor through the feedwater system, this statement may not be true.

440.100 In the Remote Shutdown System RCIC controls are replaced by HPCF (6.3) controls. Traditionally, RCIC was used for remote shutdown because the system will be available during station blackout. Describe the basis for replacing RCIC controls with HPCF controls in the Remote Shutdown Panel.

440.101 SSAR Table 9.3-1 is not complete. Include pump flow and other (9.3.5) parameters for all modes of operation. The existing Table 9.3-1 gives only test modes.

440.102 In current BWRs explosive valves are used at SLCS pump discharge.

(9.3.5) Why are they deleted? How is boron leakage into the reactor vessel prevented during testing? s

440.103 The ATWS rule states that "Each Boiling Water Reactor must have a (9.3.5) Standby Liquid Control System (SLCS) with a minimum flow capacity and baron content equivalent in control capacity to 86 gpm of 13 weight percent sodium pentaborate solution." (251 vessel,Ref: NEDE 31096-P-A) How does the ABWR design with 278 diameter vessel meet the requirement of the ATWS rule, 10 CFR 50.627 440.104 In the ABWR design, SLCS pump is started manually. But the ATWS (9.3.5) rule 10 CFR S0.62 states that "The SLCS initiation must be automatic and must be designed to perform its function in a reliable manner for plants granted construction permits after July 26, 1984." How does the ABWR design satisfies the ATWS rule?

440.105 We understand that boron mixing tests were performed for optimizing (9.3.5) the location of boron injection. Describe the test criteria and the test results.

440.106 In SSAR Section 9.3.5.3, under criterion 26, it is stated that "The (9.3.5) requirements of this criterion do not apply within the SLCS itself." Elaborate on this assumption.

440.107 In SSAR Section 9.3.5.3, under criterion 27, it is stated that "this (9.3.5) criterion applies no specific requirements onto the SLCS and therefore is not applicable." Describe in detail the justification for the above statement.

440.108 Provide further justification for the fact that the input parameters (15) and initial conditions for analyzed events are conservative.

Provide a list of what parameters will be checked at startup and which will be in the Technical Specifications. You should define the range of operating conditions and fuel types for which your input parameters will remain valid. For example, would these parameters remain valid for 9x9 or 7x7 fuel or for a similar large change in the fuel lattice.

i 440.109 Provide an analysis of the loss of instrument air (nitrogen).

(15) 440.110 In SSAR Table 15.0-2, the following transients are not categorized (15) as moderate frequency event [(Category (a)]

(a) Runout of two feedwater pumps (Cat. c)

(b) Opening of all Control and Bypass Valves (Cat. c)

(c) Pressure Regulator Downscale failure (Cat. c)

(d) Generator Load Rejection, Failure of One Bypass Valve (Cat, b)

(e) Generator Load Rejection with Bypass Off (Cat. c)

(f) TurbineTripwithFailureofOneBypassValve(Cat.b)

(g) Turbine trip, Bypass Off (Cat. c)

(h) Loss of Aux. Power Transformer and one S/up transformer (Cat c)

(1) Trip of all Reactor Internal Pumps (Cat. c)

(j) Fast Runback of all Reactor Internal Pumps (Cat. c)

(k) InadvertentHPCFpumpstart-up(Cat.b)

Category b refers to Infrequent event and Category c refer to limiting faults.

The above categorization of transients is a significant deviation from the SRP and hence sufficient justification must be subnitted to support the change in the categorization.

440.111 Provide a table similar to 15.0-2 showing your evaluation of (15) anticipated transients with single failure. List the single failure chosen for each event and provide a justification for why the chosen failure is the most limiting.

440.112 Provide the following:

(15)

(1) A listing of all equipment which is not classified as safety-related but is assumed in FSAR analyses to mitigate the consequences of transients or accidents.

-7 (2) Justification for the assumption of operability of this equipment based upon equipment quality, reliability, and proposed surveillance requirements.

(3)-Discusstheconsequencesofthoseeventsconcerning(1) number of fuel failures, (ii) delta CPR and (iii) delta peak pressure that would l result if only safety grade systems or components were considered in the specific transients analyses taking credit of (

non-safety grade systems or components. >

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440.113 You have classified the trip of all reactor internal pumps as a l (15) limiting fault. This is based on ywr assumption that the loss of f greater than three reactor internal purps is 10'0 per year. Provide I operating experience data to justify this failure rate.

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r 440.114 The ABWR feedwater control system and the steam bypass and pressure (15) control s,wstem use a triplicated digital system. Your claim is that

, no single failure in these systems will cause a minimum demand to all (

4 turbine control valves and bypass valves or the runout of two

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feedwater pumps.  !

a) What is the reliability of the systems?

b) What design feature of these systems prevent comon mode l

failure to more than one channel? t

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! c) What protection is provided in these systems against a  ;

technician disabling a second channel while performing l maintenance on the first? [

d) What are the most limiting events for the case where two l channels are lost in these systems?  ;

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l 440.115 Provide further analysis and numerical ,fustification for your (15) assertion that FMCR0 design is equivalent to an ARI system and that the SLCS is not required to respond to an ATWS event.

440.116 For each transient and accient, identify the computer code used fri (15) the analysis in the respective section of Chapter 15.

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