ML20153A847

Forwards Request for Addl Info Re GE Application for Certification of Advanced BWR Design.Responses Requested by 880915

ML20153A847
Person / Time
Site:	05000605
Issue date:	07/07/1988
From:	Scaletti D Office of Nuclear Reactor Regulation
To:	Gay J GENERAL ELECTRIC CO.
References
NUDOCS 8807120634
Download: ML20153A847 (28)

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July 7.-1988 a

Docket No. STN 50-605 -

J. S. Gay, Acting Manager

-Licensing & Consulting Services General Electric Company Nuclear. Energy Business Operations Mail Code 682 175 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 is contained in Enclosure 1. Our request for information addresses the areas of SRP Chapters 4, 5, 6, and 15 reviewed by the Plant Systems and Reactor Systems Branches. Additional questions related to the review by the Reactor Systems Branch will be provided in the near future.

In order for us to maintain the ABWR review schedule, we request that you provide your responses to this request by September 15, 1988. If you have any questions on these natters, , sase call me at (301) 492-1104.

Sincerely, Original Signed By:

Dino C. Scaletti, Project Manager Standardization and Non-Power Reactor Project Directorate Division of Reactor Projects III, l

IV, V and Special Projects Office of Nuclear Reactor Regulation

Enclosure:

As stated Distribution:

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July 7, 1988 Docket No. STN 50-605 J. S. Gay, Acting Manager Licensing & Consulting Services-General Electric Company Nuclear Energy Business Operations Mail Code 682 175 Curtner Avenue San Jose, California 95125

SUBJECT:

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

Our request is contained in Enclosure 1. Our request for information addresses the areas of SRP Chapters 4, 5, 6, and 15 reviewed by the Plant Systems and Reactor Systems Branches. Additional questions related to the review by the Reactor Systems Branch will be provided in the near future.

In order for us to maintain the ABWR review schedule, we request that you provide your responses to this request by September 15, 1988. If you have any questions on these matters, please call me at (301) 492-1104.

Sincerely, Dino C. Scaletti, Project Manager Standardization and Non-Power Reactor Project Directorate Division of Reactor Projects III, IV, Y and Special Projects Office of Nuclear Reactor Regulation

Enclosure:

As stated

s.

ENCLOSURE 1 PE0 VEST FOR ADDITIONAL INFORMATION B) THE OFFICE OF NUCLEAR REACTOR REGULATION DOCKET NO. 50-605 430.1 Provide a failure modes and effects analysis of the control rod'~ drive (4.6) system (CRDS) in tabular form with supporting discussion to delitiente the logic employed. The failure analysis should demonstrate that the CROS can perfonn the intended "unctions with the loss of any active single component. These evaluations and assessments should establish that all essential elements of the CROS are identified and provisions It should be are made for isolation from nonessential CRDS elements.

established that all essential equipment is protected from comon mode failures such as failure of moderate-and high-energy lines. The failure mode and effects analysis of the control rod drives should include water, air and electrical failures to CRDs and how the CR0 system op2 ration is affected due to air contamination or water contamination. Before finali-zing the scope of the analysis, refer to ACRS subcomittee meeting pro-ceedings on the ABWR dated June 1, 1988. It is noted that the above information is to be included in Appendix ISB of the SSAR which will be '

submitted at. a later date. However, the evaluation of the functional design of the reactivity control systems cannot be corapleted until this information is provided.

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' 430.2 Regarding Reactor Coolant Pressure Boundary (RCPB) leakage (5.2.5) detection systems, provide information on the following:

a) Describe how the leakage through both the inner and outer reactor vessel head flange seals will be detected and quantified.

b) List the sources that may contribute to the identified leakage collected in the Reactor Building Equipment Drain Sumps.

c) Describe how potential intersystem leakages will be monitored for the

1) Low Pressure Coolant Injection System, 2) High Pressure Core Spray System, 3) Reactor Core Isolation Cooling System (RCIC) - Water side and 4) Residual Heat Removal System-Inlet and discharge sides. Your response should include all the applicable (for tha ABWR design) systems and components connected to the Reactor Coolant System that are listed in Table 1 of SRP Section 5.2.5 and other systems that are unique to ABilR (except those that you have already discussed i SSAR Subsection 5.2.5.2.2, Item 11).

430.3 Discuss compliance of reactor coolant leak detection systems (5.2.5) with Regulatory Guide (RG) 1.45, "Reactor Coolant Pressure Boundtry Leskage Detection Systems," Positions C4, C5, C6, C8 and C9 with respect to the following itens:

l a) Indicators for abnormal water levels or flows in all the affected

. areas in the event of intersystem leakages.

l b) Sensitivity and response time of leak detection systems used for unidentified leakages outside the drywell.

c) Qualification relating to seismic events for drywell equipment drain sump monitoring system and leak detection systems outside the drywell.

l d) Testing Procedures - Monitoring sump levels and comparing them with

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applicable flow rates of fluids in the sumps.

h e)- Inclusion of reactor building and other areas floor and equipment drain sumps in ABWR Technical Specifications for leak detection systems.

Note that a few of the questions above arise because in Subsection 5.2.3.4.1 you state that the total leakage rate includes leakages collected in drywell.

reactor building and other area floor drain and equipment drain sumps.

430.'4 Clarify whether the RCIC makeup capacity is sufficient to provide (5.2.5) also for main steam line leakage through to the mein turbine stop valves. Also, citrify whether this leakage is included in the total leakage mentioned in Subsection 5.2.5.4.1.

430.5 Clarify how Position C.2 of RG 1.29, "Seismic Design (5.2.5) Classification" is met for all applicable leak detection systems (also include the leak detection systems outside the drywell).

430.6 Identify'all the interface requirements relating to RCPB leakage (5.2.5) detection systers.

(Containment SSAR Se:tions 6.2.1, 6.2.1.1.c 6.2.1.2, 6.2.I.3, 6.2.2, 6.2.3, 6.2.4 and 6.2.5) 430.7 In the SSAR section devoted to containment functional design, identify cletrly those areas that are not part of the, scope and provide relevant interf3ce requirements. ABV A f

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430.8 With respect to the design bases for the containment:

a) Discuss the bases for establishing the margin between the maximum calculated accident pressure or pressure difference and the corresponding design pressure or

. pressure difference. This includes the design external pressure, internal pressure, and pressure between subcompartment walls, b) Discuss the ccpability for energy removal from the contain-ment under various single-failure conditions. State and justify the design basis single failure that affects con-tainment heat removal.

430.9 The Standard Safety Analysis Report ($$AR) states that the analytical models used to evaluate the containment and drywell responses to postulated accidents and transients are included in the General Electric Co. report NE00-20533 and its supplement 1, entitled "The G.E. Mark III Pressure Suppression Containment Analytical Model". Provide justifi-cat on that these references are appropriate to use for the ABWR Con-tainment design which is not specified as Mark III. Discuss the simi-larties and differences of the AEWR design to previously approved Mark II and Mark III designs es they relate tn the containment and drp ell respn-ses to the postulated accidents and the analytical model used for the analyses. Include in the discussion the conservatisn. used in the model and assumptions, the applicable test data that support the analytical models, and the sensitivity of the analyses to key parameters.

430.10 With regard to the design features of the containment:

a) Provide general arrangement drawing; for the containment structure.

b) Provide appropriate references to Section 3 of the SSAR which includes the information on the codes, standards, and guides J

applied in the design of the containment and containment internal structures, c) Discuss the possibilities of wati.r entrapment inside containment and its effec't on the accident analysis, d) Provide infomation on qualification tests that are intended 1.0 demonstrate the functional capability of the containment structures, systems and components. Discuss the status of any developmental tests that may not'have been completed.

430.11 Provide a detailed discussion of the likelihood and sensitivity to steam bypass of the suppression pool for a spectrum of accidents.

Include in your discussion the fallowing infomation:

a) A compa-ison of the ABWR pool bypass capability with that for Mark II and Mark III designs b) The measures for minimizing the potential for steam bypass and the systems provided to mitigate the consequences of pool bypass. Discuss and demonstrate the conservatism of assumptions made in the analysis of steam bypass, c) Identify all lines from which leakage (or rupture) could contribute to pool bypass and wetwell air space pressurization.

d) Identify all fluid lines which traverse the wetwell air space and identify those lines which are protected by guard pipe, e) Discuss the rationale and basis for the wetwell spray flow cepacity.

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430.12 With regard to containment response to external pres'sure:

a) Describe the wetwell-to-drywell vacuum breaker system and show the extent to which the requirements of sub-section'NE of section'III of the ASME B&PV Code

- are satisfied. Discuss the functional capability of the system. Provide the design and perfonnance para-meters for the vacuum relief devices, b) Discuss the basis for selecting a low design capability for external pressure acting across the drywell to wetw' ell boundary. It is not apparent that the drywell negative ,

design pressure of 2.0 psid is desirable or sufficient, c) The margin between the calculated wetwell-to-reactor building negative differential pressure (-1.8 psid) and the design differential pressure (-2.0 psid) is not considered adequate. A higher margin of 15% should be provided at this stage of the design. Further, given the reliance of the BWR pressure suppression design on containment venting to control pressure, di:, cuss the basis for not providing wetwell to reactor building vacuum breakers, d) In the analysis of wetwell-to-reactor building negative differential pressure calculation, a 500 gpm wetwell spray flow rate was used. Provide the basis for the assumption and the design basis for the wetwell spray capacity.

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4 430.13 Section 6.2.1.1.3 of the SSAR states that the containment functional evaluation is based upon the consideraticn of several postulated accident conditions including small break accidents. Provide the assumptions, analysis and results of the small break accidents con-sidered, and demonstrate that the identified (in the SSAR) feedwater line and steam line breaks are the limiting accidents.

430.14 Provide analyses of the suppression pool temperature for transients involving the actuation of safety / relief valves. Provide th6 assump-tions and conservatism employed in the analyses so that er, assessment could be made for conformance to the acceptance criteria set forth in NUREG-0783, "Suppression Pool Temperature Limits for BWR Containments."

430.15 Provide the pressure at which the maximum allowable leak rate of 0.5%/ day is quoted.

430.16 Provide engineered safety systems information for containment response analysis (full capacity operation and capability used in the contain-ment analysis), as indicated in Table 6-7 rf Regulatory Guide 1.70, Revision 3.

430.17 In the design evaluation section for containment subcompartments (Section 6.2.1.2.3), provide the information necessary to substan-tiate your assessment that the peak differential pressures do not exceed the design differential pressures. Guidance for th2 infor-rnation required is provided in Regulatory Guide 1.70, Revision 3, Section 6.2.1.2, "Containment Subcompartments", Design Evaluation.

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430.18 Describe the manner in which suppression pool dynamic loads resulting from postulated ~1oss-of-coolant accidents, transients (e.g., relief valve actuation), and seismic events have been integrated into the affected containment structures. Provide plan ar.d section drawings of the containment illustrating all equipment and structural surfaces that could be subjected to pool dynamic loads. For each structure or group of structures, specify the dynamic loads as a function of time, and specify the relative magnitude of the pool dynamic load compared to the design basis load for each structure. Provide justification for each of the dynamic load histories by the use of appropriate >

experimental data and/or analyses.

Describe the manner by which potential asymmetric loads were considered in the containment design. Characterize the type and magnitude of possible asymmetric loads and the capabilities of the affected structures to withstand such a loading profile.

430.19 Provide information to demonstrate that the ABWR design is not vulnerable to a safety relief valve discharge line break within the air space of the wetwell, coupled with a stuck open relief valve after its actuation as a result of the transient.

430.20 Discuss suppression ' pool water makeup under normal and accident conditions.

430.21 With respect to mass and energy release analyses for postulated loss-of-coolant accidents identify the sources of generated and stored energy in the reactor coolant system that are considered in the analyses of loss-of-coolant accidents. Describe the methods used and assumptions made in calculations of the energy available for release from these sources. Address the conservatism in the calculation of the available energy from each source. Tabulate the stored energy sources and the amounts of stored energy. For

_y the sources of generated energy, provide curves showing the energy release rates and integrated energy release.

430.22 In the SSAR sections devoted to containment heat removal systems, identify clearly those areas that may not be part of the GE scope and provide relevant interface requirements.

430.23 The SSAR states that the containment heat removal system is designed to limit the long-tenn temperature of the suppression pool to 207'F.

The calculated peak pool temperature is 206.46'F for the feedwater line break. With respect to this analysis provide the following infor-mation:

a) The justification that this is the limiting accident with respect to the naximum temperature in the suppression pool, b) The bases for the design margin between the design and calculated temperatures.

c) All assumptions used in the analysis and conservatism associated with each. Include the effects of potential temperature stratification in the suppression pool and its effects on heat removal capability of the system.

d) The identification of the decay heat curve used in the analysis.

430.24 Provide the design bases for the spray features of the containment heat removal system. Provide the safety classification of the compo-nents associated with the spray feature of the system, 430.25 Discuss the rationale for continued reliance on sprays as the sole active engineered safety feature for drywell atmosphere pressure and temperature. Discuss the merits of upgrading the design of drywell fan coolers to previde some capacity for pressure, temperature, and humidity control following an accident.

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430.z6 The time period assumed for initiation of the containment heat removal system after a LOCA is 10 minutes requiring operator action. It is the staff's position that this time period is too restrictive. In fact previous BWR designs (Grand Gulf's Mark III) use 30 minutes actuation time. Provide the reasons why the ABWR does not provide more flexibility with respect to the time required for actuation.

430.27 Describe the design features of the suppression pool suction strainers.

Specify the mesh size of the screens and the maximum particle size that could be drawn into the piping. Of the systems that re:eive water through the suppression pool suction strainers under posta:-ident con-ditions, identify the system component that places the limiting require-ments on the maximum size of debris that may be allowed to pass through the strainers and specify the limiting particle size that the component can circulate without impairing system perfomance. Discuss the poten-tial for tha strainers to become clogged with debris. Identify and discuss the kinds of debris that might be developed following a loss-of

-coolant accident. Discuss the types of instilation used in the contain-ment and describe the behavior of the insulation during and after a LOCA. Include in your discussion information regarding con'pliance with the acceptance criteria associated with USI A-43 es documented in NUREG-0897.

430.28 Provide analyses of the net positive suction head (NPSH) available to the RHR pumps in accordance with the recommendations of Regulatory Guide 1.1. Compare the calculated values of available NPSH to the required NPSH of the pumps.

430.?9 In SSAR Section 6.2.3, identify clearly those areas that may not be part of the scope and provide relevant interface requirements.

Ab 430.30 Provide a tabulation of the design and performance data for the secondary containment structure. Provide the types of information indicated in Table 6-17 of Regulatory Guide 1.70, Revision 3.

430.31 Describe the valve isolation features used in support of the secondary containment. Specify the plant protection system signals that isolate the secondary containment and activate the standby gas treatment system.

430.32 Identify and tabulate by size, piping which is not provided with isolation features. Provide an analysis to demonstrate the capability of the Standby Gas Treatment System to maintain the design negative pressure following a design basis accident with all non isolated lines open and the event of the worst single failure of a secondary containment isolation valve to close. .

4'10.33 Discuss the design provisions that prevent primary containment leakage from bypassing the secondary containment standby gas treatment system and escaping directly to the environment. Include a tabulation of potential bypass leakage paths, including the types of infomation indicated in Table 6-18 of Regulatory Guide 1.70, Revision 3. Provide an evaluation of potential bypass leakage paths considering equipment design limitations and test sensitivities. Specify and justify the maximum allowable frac-tion of primary containment leakage that may bypass the secondary con-tainment structure. The guidelines of BTP 6-3 should be addressed in considering potential bypass leakage paths, i

l 430.34 Provide a list of the secondary containment openings and the instrutaentation means by which each is assured to be closed during a

' postulated design basis accident.

i 430.35 Provide a table of design information regarding the containment isolation provisions for fluid system lines and fluid instrument lines l

penetrating the containment which are within the GE scope of the ABWR design. Include as a ninimum the following information:

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1) General design criteria or regulatory guide reeminendations that have been met or other defined bases for acceptability;

2) System name;-

3) Fluid contained;

4) Line sire;

5) ESF system (yes or no);

6) Through-line leakaoe classification;

7) Reference to figure in SSAR showing arrangement of containment isolation barriers;

8) location of valve (inside/outside containment);

9) Type C leakage test (yes or no);

10) Valve type and operator;

11) Primary mode of valve actuation;

12) Secondary mode of valve actuation;

13) Nomal valve position;

14) Shutdowr. valve position;

15) Postaccident valve position;

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16) Power failura valve position;

17) Containment isolation signals;

18) Valve closure time; and

19) Power source.

ABWA 430.36 For isolation valve design'in systems not within the3 scope, identify the systems and the relevant intorface requirements. , Include a discussion on essential and non-essential systems per Pegulatory Guide 1.141 and the means or criteria provided to automatically isolate the nonessential systems by a containment isolation signal. Also, in-clude a discussion on the requirement that the setpoint pressure which initiates containment isolation for nonessential penetrations be reduced to the minimum value compatible with normal operations.

430.37 Specify all plant protection signals that initiate closure of the containment isolation valves.

430.38 Describe the leakage detection means provided to identify leakage for the outside-containment remote-manual isolation valves on the following influent lines: Feedwater, RHR injection, HPCS, standby liquid control, RWCU connecting to feedwater line, RWCV reactor vessel head spray.

430.39 The containment isolation design provisions for the recirculation pump seal water purge line do not meet the explicit requirements of GDC 55 nor does the design satisfy the GDC on some other defined b6 sis as out-lined in SRP Section 6.2.t. It is our position that the isolation design in the instance is inadequate and should be modified to satisfy GDC 55 either explicitly or on some other defined basis, with the appro-priate justification.

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• :. : l 430.40 With respect to Figure 6.2-38a:

a) Include the isolation valve arrangement of the standby liquid control system line, b) Identify the line labeled in the figure as "WDCS-A" (it joins the RWCU line prior to its c(nnection to the feedwater line), and discuss the isolation provisions for that line.

430.41 Provide a diagram or reference to figure (s) showing the isolation valve arrangement for the lines identified below. For the isolation valve design of each of these lines, provide justification for not meeting the explicit requirements of GDC 56, and demonstrate that the guidelines for acceptable alternate containment isolation provisions contained in SRP 6.2.4 are satisfied. The lines in question are:

' HPCS and RHR test and pump miniflow bypass lines

• RCIC pump miniflow bypass line

• RCIC turbine exhaust and pump miniflow bypass lines

• SPCU suction and discharge lines 430.42 Describe the isolation provisions for the containment purge supply and exhaust lines and discuss design confomance with Branch Technical Posi-tion CSB 6-4, "Containment Purge During Normal Operations."

430.43 Discuss the closure times of isolation valves in system lines that can provide an open path from the primary containment to the environment (e.g., containment purge system). Also discuss provisions of radiation monitors in these lines having the capability of actuating containment isolation.

430.44 Identify the system lines whose containment isolation reoutrements are covered by GDC 57 and discuss confomance OT the design to the GDC requirements.

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430.45 For the combustible gas control systems design, identify clearly those areas that may not be part of the scope and provide relevant interface requirements. WA 430.46 According to SRP 6.2.5 specific acceptance criteria related to the concentration of hydrogen or oxygen in the containment atmosphere among others are the following:

a) The analysis of hydrogen and oxygen production should be based on the

- parameters listed in Table 1 of Regulatory Guide 1.7 for the purpose of establishing the design basis for combustible control systems.

b) The fission product decay energy used in the calculation of hydrogen and oxygen production from radiolysis should be equal to or more conservative than the decay energy model given in Branch Technical Position ASB9-2 in SRP 9.2.5.

Provide justification that the assumptions used in the ABWR in establishing the design basis for the combustible gas control systems are conservative with respect to the criteria a. and b. above.

430.47 Provide an analysis of the production and accumulation of combustible gases within the containment following a postulated loss-of-coolant accident including all applicable information specified in Section 6.2.5.3 of Regulatory Guide 1.70, Revision 3.

430.48 Regarding Containnent Type A leakage testing, (6,2.6) a) Provide the values for P, and Pt '

b) Include the acceptance criterion for Ltduring preoperational j leakags rate tests, i.e., L t *L a Iltm/ lam), for the case when l L, (Ltm/bam) = 0.7.

c) Your acceptance criterion for Ltm (SSAR Subsection 6.2.6.1.2.2, Item 1) is at variance with the staff's current practi.ce for acceptance of (

,Lg. Also, it does not comply with the 10 CFR Part 50, Appendix J. l Section III, Item A.1.(a) requirement. Therefore, either provide l sufficient supporting justification for the exemption from i compliance with the above requirement or correct the criterion as appropriate to comply with the requirement. Also, correct the stated acceptance criterion (SSAR Subsection 6.2.6.1.2.2 Item 3) as appropriate to comply with Appendix J Section III, Item A.6.(b) requirement.

d) Regarding ILRT, identify the systems that will not be vented or drained and provide reasons for the same, e) Provide P& ids and process flow drawings for systems that will be vented or drained.

430.49 Regarding Type B tests.

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(6.2.6)

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a) Clarify how air locks opened during periods when containment integrity is required by plant's Technical Specifications will be tested to comply with Appendix J Section III, Item D.2.(b).(iii).

b) Provide the frequency for periodic tests of air locks and associated inflatable seals, c) Provide the acceptance criteria for air lock testing and the associated inflatable seal testing.

c) List all coatainment penetrations subject to Type B tests, d) List all those penetrations to be excluded from Type B testing and the rationale for excluding 3 hem, 430.50 Regarding Type C tests.

(6.2.6) ,

a) Correct the statement (Subsectiqn 0.2.6.3.1, Paragraph 1) as appro-priate to ensure that the hydraulic Type C tests are performed only on those isolation valves that are qualified for such tests per Appendix J. The current statement implies that these tests are not necessarily restricted tc the valves that qualify for such tests, b) List all the primary containment isolation valves subject to Type C tests and provide the necessc y P& ids.

c) Provide the list of valves that you propose to test in the reverse direction and justification for such testing for each of these valves.

d) Identify the valves that you propose to test hydrostatically based on their ability to maintain a 30-day water leg seal. Also, identify

, other valves which you propose to test hydrostatically and provide the basis for such tests. Provide the test pressure for all the valves mentioned above.

e) Indicate test pressures for MSIVs (with justification if it is less than P,) and isolation valves sealed from a sealing system.

f) Indicate how you will perform Type C leak tests for ECCS systems and RCIC system isolation valves.

g) Confirm that the interval between two consecutive periodic Type C tests will not exceed 2 years as required by Appendix J.

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h) State what testing procedures you will follow regarding the valves that are not covered by Appendix J requirements.

430.51 Identify the reporting requirements for the tests. Note that (6.2.6) p ur response should address compliance with the requirements in this regard as stated in Appendix J. Sections III.A.(a), IV.A and V.

(For example, regarding follow up tests after containment modification, you have nM included Type C testing for affected areas).

430.52 Regarding Secondary Containment. .

(6.2.6) a) Identify the special testing procedures you will follow to assure a maximum allowable inleakage of 50 percent of the secondary containment free volume per day at a differential pressure of -0.25" water guage with respect to the outdoor atmosphere (See Section 6.5.1.3.2).

b) Identify all potential leak paths which bypass the secondary contain-ment. (For such identification, see (BTP) CSB 6-3, "Determination of Bypass Leakt.ge Paths in Dual Containment Plants")

c) Identify the total rate of secondary containment bypass leakage to the enviro 1 ment 430.53 Identify all the interface requirements relating to containment (6.2.6) leak testing.

430.54 Regarding Control Room Habitability systems, (6.4) a) Provide the minimum positive pressure at which the control building envelope (which includes the mechanical equipment room) will be maintained with respect to the surrounding air spaces when makeup air is supplied to the envelope at t!.e design basis rate (295 CFM).

b) Provide the periodicity for verification of control room pressurization with design flow rate of makeup air, c) Clarify whether all the potential leak paths (to be provided in Section 9.4.1) include dampers or valves upstream of recirculation fans, d) Identify the action to be taken when there is no flow of the equipment roon return fan and consequently the equipment room is over pressurized (Table 6.4-1 contains no information on the above).

e) Provide the actual minimum distances (lateral and vertical) of the control room ventilation inlets from nafor potential plant release points that have been used in ycur control room dose analysis.

Also, provide a schematic of the location of control room intake vents, f) Provide Figure 6.4-5 (plan view) which you state shows the release points (SGTS vent).

g) Section 6.4.2.4 and Figure 6.4-1 indicate only one air inlet for supplying makeup air to the emergency zone. However, Tables 6.4-2 and 15.6-8 and Section 15.6,0.5.2 indicate that there are two automatic air inlets for the emergency zone. Correct the above dit.crepancy as appropriate. Also describe the characteristics of inese inlets with respect to their relative locations and automatic selection control features. 5 tate how both flow and isolation in each inlet assuming single active component failure will be ensured, h) Describe the design features for protecting against confined area releases (e.g., multiple barriers, air flow patterns in ventilation zones adjacent to the emergency zone).

1) Describe the specific features for protecting the control room operator frou airborne radioactivity outside the control room and direct shine from all radiation sources (e.g., shielding thickness for control room structure boundary, two-door vestibules).

j) Clarify what you mean by "sustained occupancy" (See SSAR Section 6.4.1.1, Item 3) for 12 persons.

k) Provide justification for not specifying any unfiltered infiltration of contaminated air into the control room in SSAR Table 15.6-8,

1) Provide Subsection 6.3.1.1.6 which you state (SSAR Section 6.4.6) contains a complete description of the required instrumentation for ensuring control room habitability at all times.

m) Give schematics for control room emergency mode of operation during a postulated LOCA (this is required for calculating control room LOCA doses).

n) The source terms and control room atmospheric dispersion factors (X/Q values) used in the control room dose analysis (See SSAR Tables 15.6-8 and 15.6-12) to demonstrate ABWR control room compliance with GDC 19 are non-conservative. Therefore, reevaluate control room doses during a postulated LOCA using RG 1.3 source terms and assumptions and the methodology given in Reference 4 of SSAR Section 15.6.7.

Include possible dose contributions from containment shine, ESF filters and airborne radioactivity outside the control room. Also check and correct as appropriate the recirculation rate in the control room (22.4 m3 /sec) given in Table 15.6-8.

I o) Section 6.4.7.1, "External Temperature," provides design maximum external temperatures of 100 F and -10 F. How are these values used in the design and assessments related to the ABWR? What factors, such as insulation, heat generation from control room personnel and I

l equipmeni and heat losses, are taken into account? Do these values represent "instantaneous" values or are they temporal and/or spatial averages?

p) Clarify your position on potential hazardous or toxic gas tources onsite of an ABWR. If applicable, indicate the special features provided in the ABWR design in this regard, to ensure control room habitability.

q) Identify all the irterface requirements for control room habitability systems (e.g., instrumentation for protection against toxic gases in

. general and chlorine in particular; potential toxic gas release points in the environs).

430.55 Regarding ESF Atmosphere Cleanup Systems, (6.5.1) ,

a) Provide a table listing the compliance status of the Standby Gas Treatment System (SGTS) with each of the regulatory positions specified under C of P.G 1.52. Provide justifications for each of those items that do not fully comply with the corresponding requirements. In this context, you may note that the lack of redundency of the SGTS filter train (the staff considers that filter trains are also active components-See SRP 6.4, Acceptance Criterion II.2.b) is not acceptable.

Further, the described sizing of the charcoal adsorbers based on assumed decontamination factors for various chemical forms of iodine in the suppression pool is not acceptable (RG 1.3 assumes a decontamination factor of 1 for all forms of iodine end RG 1.52 requires compliance with the above guide for the design of the adsorber section). There-fore, revise charcoal weight and charcoal iodine leading given in SSAR Table 6.5-1 as appropriate. -

b) Specify the laboratory test criteria for methyl iodide penetration that will be identified as an interface requirement to be qualified for the adsorber efficiencies for iodine given in SSAR Table 15.6-8.

Also, provide the depth of the charcoal beds for the control room emergency system.

c) Provide a table listing the compliance status of the instrumentation provided for the SGTS for read out, recording and alarm provisions in the control room with each of the instrumentation items identified in Table 6.5.1-1 of SRP 6.5.1. For partial or non-com11ance items, provide justifications.

d) Clarify whether primary containment purging during normal plant operation when required to limit the discharge of contaminants to the environment will alwan Le through the SGTS (See SSAR Section 6.5.1.2.3.3).

Clarify whether such a release prior to the purge system isolation has been considered in the LOCA dose analysis, e) Provide the cortpliance status tables referred to in Items (a) and (c) above for the control room ESF filter trains. (The staff notes that you have comitted to discuss control room ESF filter system under SSAR Section 9.4.1. However, since evaluation of the control room habitability system cannot be completed until the information identified above is provided, the above information is requested now.)

f) Identify the applicable interface requirements for the SGTS and the control room ESF atmosphere cleanup system.

430.56 Regarding Fission Product Control Systems and Structures, (6.5.3) a) Provide the drawdown time for achieving a negative pressure of 0.25 inch water gauge for the secondary containment with respect to the environs during SGTS operation. Clarify whether the unfiltered release of radioactivity to the environs during this time for a

i

c. . .

,. I I

postulated LOCA has been considered in the LOCA dose analysis. (Note that the unfiltered release need not be considered provided the required negative pressure differential is achieved within 60 seconds from the time of the accident.)

b) Provide justification (See SRP Section 6.5.3. II.4) for the decontamination factor assumed in SSAR Tables 6.5-2 and 15.6-8 for iodine in the suppression pool, correct the elemental, particulate and organic iodine fractions given in the tables to be consistent with RG 1.3, and incorporate the correction in the LOCA analysis tables.

Alternatively, taking no credit for iodine retention in the suppression pool, revise the LOCA analysis tables. Note that the revision of the LOCA analysis tables (this also includes the control room doses) mentioned above is strictly in relation to the iodine retention factor in the suppression pool (also, there may be need for revision of other parameter (s) given in the tables and these will be identified under the relevant SRP Sections questions),

c) Identify the applicable interface requirements.

~

430.57 Regarding SSAR Section 6.7, the staff notes that the Nitrogen (6.7) Supply System has been discussed under this section, in' stead of the itain Steam Isolation Yalve Leakage Control System (HSIV-LCS) as required by the Standard Format for SARs. The staff will review the material presented in SSAR Section 6.7 along with the material that will be presented in SSAR Section 9.3.1.

Regarding MSIV-LCS, the staff notes that you are committed to provide a non-safety related MSIV leakage processing pathway consistent with those evaluated in NUREG-1169, "Resolution of Generic Issue C-8,"

August 1986. Since the staff has not finalized its position so far on the acceptability of the NUREG findings with regard to the design of the HSIV-LC:s, provide pertinent information on the system design including interface requirements to evaluate the to-be-proposed design against the acceptance criteria of SRP 6.7.

~ 430.58 The accident analyzed under this section considers only the airborne (15.7.3) radioactivity that may be released due to potential failure of a concentrated waste tank in the radwaste enclosure. The'5RP acceptance criteria, however, requires demonstration that the liquid radwaste concentration at the nearest potable water supply in an unrestricted area resulting from transport of the liquid radwaste to the enrestricted area does not exceed the radionuclide concentration limits specifiad in 10 CFR Part 20, Appendix B Table U Column 2. Such a demonstration will require information on possible dilution and/or decay during transit which, in turn, will depend upon site specific data such as surface and ground water hydrology and the parameters governing liquid waste movement through the soil. Additionally, special design features (e.g., steel liners or walls in th%.radwaste enclosure) may be provided as part of the liquid radweste treatment systems et certain sites. The staff will, therefore, review the site specific characteristics mentioned above individually for each plant refer-

, encing the ABWR and confint its review of ABWR, only to the choice of the liquid radwaste tank. Thersfore, provide information on the following:

a) Basis for determining the concentrated waste tank as the worst tank (this may very well be the case, but in the absence of information on the capacities of major tanks, particularly the waste holdup tanks, it is hard to conclude that the above tank both in terms of radionuclide concentrations and inventories will turn out to be the worst tank).

b) Radionuclide source terms, particularly for tha long-lived radionuclides such as Cs-137 and Sr-90 (these m y be the critical isotopes for sites that can claim only decay credit during transit) in the major liquid radwaste tanks.

o.

(SSARSection4.6) 440.1 SPP d.6 identifies the fo11cwino GDCs 23, 25, 76, 77, 28 and 29 in the acceptance criteria. Confinn'c that the reactivity system,

' meet the requirements of the described in Section 4.6 of the above GOCs.

440.2 In Section 4.6.2.3.2.2 Analysis of ma. vr , lon relating to rod ingle malfunctions that withdrawal, it is stated "There are kr ,inale control rod."

cause the unplanned withdrawal of eve Confirn that this is a editorial mistake and correct it if so.

Otherwise, explain in detail the ba'.is for this statement and why this is acceptable.

440.3 In Section 4.6.1.2 it is stated that CRD system in con,iunction with CRCAIS and RPS systems provides selected control rod run in (SCRRI) for reactor stability control. Describe in detail how SCRRI works.

440.4 In Figure 4.6-8a, CRD system P&ID, sheet 1, piping ouality classes AA-D, FC-D, FD-D, FD-B, etc. are shown. Submit the document which explains these classes and relsces them to ASME code classes.

A40.5 In Figure 4.6-8b, the leak receiver tank is shown. What is the function of this tank? How big is this tank? Will a high level in the tank impact the operation of the control rod drive?

440.6 Identify the essential portions of the CR0 system which are safety related. Confirm that the safety related portions are isolable from non-essential portions.

440.7 In the old CRD system, the major function of the coolina water was to cool the drive mechanism and its seals to preclude damaae resulting from long term exposure to reactor temperatures. What is the function of purge water flow to the drives?

440.8 We ur.derstand that the LaSalle Unit ? Submit fit? motion control the test rod results asdrive demonstration test is still in progress.

soon as it is available. '

440.9 In the present CR0 system design, the ball check valve ensures rod insertion in the event the accumulator is not charged or the inlet scram value fails to open if the reactor pressure is above 600 psip.

Confirm that this capability still exists in the APWR design.

440.10 in section 4.6.2.3.1, it is stated the scram time is adequate as shown by the transient analyses of Chapter 15. Specify the scram time.

e i I

1 l

l b) Specify the laboratory test criteria for methyl iodide penetration l that will be identified as an interface requirement to be qualified for the adsorber efficiencies for iodine given in SSAR Table 15.6-8.

Also, provide the depth of the charcoal beds for the control room emergency system.

c) Provide a. table listing the compliance status of the instrumentation provided for the SGTS for read out, recording and alarm provisions in the controi *oom with each of the instrumentation 3tems identified in Table 6.5.1-1 of SRP 6.5.1. For partial or non-cogliance items, provide. justifications.

d ') Clarify whether primary contsinment purging during normal plant operation when required to limit the discharge of contaminants to the environment will always be through the SGTS (See SSAR Section 6.5.1.2.3.3).

Clarify whether such a release prior to the purge system isclation has been considered in the LOCA dose analysis, e) Provide the corrpliance status tables referred to in Items (a) and (c) above for the control room ESF filter trains. (The staff notes that you have comitted to discuss control room ESF filter system under SSAR Section 9.4.1. However, since evaluation of the contral room habitability system cannot be completed until the information identified above is provided, the above information is requested now.)

f) Identify the applicable interface requirements for the SGTS and the control room ESF atmosphere cleanup system.

~

430.56 Regarding Fission Product Control Systems and Structures, (6.5.3) a) Provide th,e drawdown time for achieving a negative pressure of 0.25 inch water gauge for the secondary containment with respect to the environs during SGTS operation. Clarify whether the unfiltered release of radioactivity to the environs during this time for a

e postulated LOCA has been r:onsidered in the LOCA dose analysis. (Note that the unfiltered release need not be considered provided the required negative pressure differential is achieved within 60 seconds from the time of the accident.)

b) Provide justification (See SRP Section 6.5.3, II.4) for the decontamination factor assumed in SSAR Tables 6.5-2 and 15.6-8 for iodine in the suppression pool, correct the elemental, particulate and organic iodine fractions given in the tables to be consistent with RG 1.3, and. incorporate the correction in the LOCA analysis tables.

Alternatively, taking no credit for iocine retention in the suppression pool, ravise the LOCA analysis tables. Note that the revision of the LOCA analysis tables (this also includes the control room doses) mentioned above is strictly in rSlation to the iodine retention factor in the suppression pool (also, there may be need for revisian of other parameter (s) given in the tables and these will be identified under the relevant SRP Sections questions),

c) Identify the applicable interface requirements.

l 430.57 Regarding SSAR Section 6.7, the staff notes that the Nitrogen (6.7) Supply System has been discussed under this section, in' stead of the liain Steam Isolation Valve Leakage Control System (HSIV-LCS) as required by the Standard Format fe 'ARs. The staff will review the material

. presented in SSAR Section o., along with the material that will be presented in SSAR Section 9.3.1.

Regarding MSIV-LCS, the staff notes that you are committed to provide a non-safety related MSIV leakage processing pathway consistent with those evaluated in NUREG-ll69, "Resolution of Generic Issue C-8,"

August 1986. Since the staff has not finalized its position so far on the acceptability of the NUREG findings with regard te the design l of the MSIV-LCS, provide pertinent information on the system design including interface requirements to evaluate the to-be-proposed l

design against the acceptance criteria of SRP 6.7.

. =. -- - - - . --. -- . . - . - ._. .

j I

' 430.58 The accident analyzed under this secti)n considers only the airborne (15.7.3) rdioactivity that may be released due to potential failure of a concentrated waste tank in the radwaste enclosure. The SRP acceptance criteria,'however, requires demonstration thet the liquid radwaste concentration at the nearest potable water supply in an unrestricted area resulting from transport of the liquid raawaste to the unrestricted area does not exceed the radionuclide concentration limits specified in 10 CFR Part 20, Appendix B Table II, Column 2. Such a demonstration will require information on possibic dilution and/or decay during transit which, in turn, will depend upon site specific data such as surface and ground water hydrology and the para:neters governing liquid waste movement through the soil. Additionally, specici design features (e.g., steel liners or walls in the radwaste enclosure) may be provided as part of the liquid radwaste treatment systems at certain sites. The staff will, therefore, review the site specific characteristics mentioned above individually for each plant refer-encing the ABWR and confint its review of ABWR, only to the choice of the liquid radwaste tank. Therefore, provide information on the following:

a) Basis for determining the concentrated waste tank as the worst tank (this may very well be the case, but in the absence of information on the capacities of major tanks, particularly the waste holdep tanks, it is hard to conclude that the above tank both in terms of radionuclide concentrations nnd inventories will turn out to be the worst tank).

l t

i b) Radionuclide source terms, particularly for the long-lived radionuclides such as Cs-137 and Sr-90 (these may be the critical isotopes for sites that can claim only decay credit during transit) in the major liquid radwaste tanks.

l i

l l

l

(SSARSection4.6)

A40.1 SPP A.6 identifies the following GDCs 23, 25, 76, 27, 28 and 29 in the acceptance criteria. Confirm that the reactivity syef en, described in Scction 4.6 of the SSAR, meet the requirements of the above GDCs.

440.2 In Section 4.6.2.3.2.2 Analysis of malfunction relating to rod withdrawai, it is stated "There are known single malfunctio:s that cause the unplanned withdrawal of even a sinole control rod.-

Confirm that this is a editorial mistake and correct it if so.

Otherwise, explain in detail the basis for this stctement and why this i_ acceptable.

440.3 In Section 4.6.1.2 it is stated that CRD system in conjunction with CRCAIS and RPS systems provides selected control rod run in (SCRPI) for reactor stability control. Describe in detail how SCRRI works.

440.4 In Figure 4.6-Ba, CRD system P&ID, sheet 1, piping ouality classes AA-D, FC-D, FD-D, FD-B, etc. are shown. Submit the document which explains these classes and relates them to ASME code classes.

A40.5 In Figure 4.6-8b, the leak receiver tank is shown. What is the function of this tank? How big is this tank? Will a high level in the tank impact the operation of the control rod drive?

440.6 Identify the essential portions of the CRD system which are safety related. Confirm that the safety related portions are isolable from non-essential portions.

440.7 In the old CRD systen, the major function of the coolina water was to cool the drive mechanism and its seals to preclude damaae resulting from long term exposure to reactor temperatures. What is the function of purge water flow to the drives?

A40.8 We understand that the LaSalle Unit 2 fine motion control rod drive demonstration test is still in progress. Submit the test results as soon as it is available. '

440.9 In the present CRD system design, the ball check valve ensures rod insertion in the event the accumulator is not charged or the iniet scram value fails to open if the reactor pressure is above 600 psig.

Confirm that this capability still exists in the APWR design.

440.10 In section 4.6.7.3.1, it is stated the scram time is adequate as shown by the transient analyses of Chapter 15. Specify the serem tine.

.e L _ - - - - . .

4 l

1 440.11 For both the low ("zero") power and operating power region describe l

the patterns of the control rod groups that are expected to be withdrawn simultaneously with the new rod system, and estimate the maximum for the total and differential reactivity wnrth of these groups '. What sort of margin to period scram will exist in the low power range.

440.12 Describe the relative core locatiot of control rods sharing a scram accumulator. Can a failure of the scram accumulator fail to insert adjacent rods? If so, discuss the consequences of that failure.

I

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