ET-NRC-93-4023, Forwards Responses to NRC 930827 RAIs on AP600

Forwards Responses to NRC 930827 RAIs on AP600

ML20058F685
Person / Time
Site:	05200003
Issue date:	11/30/1993
From:	Liparulo N WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:	Borchardt R NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
ET-NRC-93-4023, NUDOCS 9312080227
Download: ML20058F685 (12)

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Westinghouse Energy Systems Bm 355

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Electric Corporation ET-NRC-93-4023 l'

NSRA-APSI 93-0467 Docket No.: STN-52-003 November 30,1993 Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20355 ATTENTION:

R. W. BORCliARDT

SUBJECT:

WESTINGHOUSE RESPONSES TO NRC REQUESTS FOR ADDITIONAL INFORMATION ON THE AP600 r

Dear Mr. Borchardt:

Enclosed are three copics of the Westinghouse respmses to NRC requests for additional information on the AP600 from your letter of August 27,1993. This transmittal completes the responses to that

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letter. Revisions to responses previously tras,Jmitted are also included. A listing of the NRC requests

- l for additional informatian tr'txmded to in this letter is contained in Attachment A. Attachment B is a complete listing of the quer ons associated with the August 26,1993 letter and the corresponding Westinghouse letters that provided our response.

These respmses are also provided as dcctronic files in Wordperfect 5.1 format with Mr. Ilasselberg's copy.

If you have any questions on this material, please contact Mr. Brian A. McIntyre at 412-374-4334.

7l On //*'>'1% &

i Nicholas J. Liparulo, Manager Nuclear Safety & Regulatory Activitics

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B. A. McIntyre - Westinghouse E Ilasselberg - NRR O b O I3.$,

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ET-NRC-93-4023 ATTACIIMENT A AP600 RAI RESPONSES SUBMITTED NOVEMBER 30,1993 l

f RAI No.

Issue 420.110 Design Basis for " Monitor Bus" t

420.116 EMI Test on Protection Cabinet 471.004R01; Rad piping area maps f

471.014R01:

Very high radiation areas j

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471.019R01:

Event / condition requiring additional shield walls 952.005 Primary pump shutoff time 952.010 Steam generator level control i

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I Printed; 11/3093 ATTACHMENT B CROSS REFERENCE OF WESTINGHOUSE RAI RESPONSE TRANSMITTALS TO NRC LETTER OF AUGUST 27,1993 Queston issue NRC Westeghouse No.

Letter Transmittal Date 952.002 ADS sparger draMngs O &27/93

'10/0493 952.003 Insulaton/ heat-tracing 08/27/93 10/0493 952.004 Starting conditons for depressurtraton transient 08/27/93 10/0493 952.005 Primary pump shutoff time 08/27/93 11/3a93 952.006 Test conddons 08/27/93 10/0493 952.007 Collecton of fluids leaving the pnmary 08/27/93 10/0493 952.008 Pressuruer heater controls CW27/93 1004/93 952 009 Safety valve pressure settn9 08/27/93 10/0493 952.010 Steam generator level control 08/27/93 11/3G93 952.011 Core power during a transient 08/27/93 10/0493 i

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4 NRC REQUEST FOR ADDITIONAL INFORMATION HH ug{

H Ouestion 420.110 The staff has concluded that Westinghouse's April 29,1993 response to Q420.23 is not acceptable. The monitor bus is a major component in the I&C system. The functional requirements should he defimed during design certification rather than wait until the combined license stage. The design requirements documents should be made available for staff review.

Response

The monitor bus does not perform a safety-related function. The function of the monitor bus is to transport processed information to the operator to support normal plant operation. The monitor bus is not required for safe shutdown, nor does it play an active role in the automatic control of plant functions or the manual control of plant components.

In the unlikely event of a complete failure of the monitor bus, the automatic functions of the nonsafety-related plant control system will continue to operate. The operator will continue to be able to monitor selected functions by using indications provided by the protection and safety monitoring system and plant control system. The operator will be able to perform orderly shutdown of the plent using the nonsafety-related plant control system.

The automatic functions of the safety-related protection and safety enonitoring system will continue to be available in the unlikely event of a complete failure of the monitor bus. The operator will continue to be provided with sul6cient information to safely shutdown the plant and will remain able to trip the reactor and actuate safety-related plant components in the aormal manner using the safety-related protection and safety monitoring system.

I'unctional requiremeo Y the monitor bus are available for NRC saff review, SSAR Revision: NONE 420.110-1 W westingtmuse

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NRC REQUEST FOR ADDITIONAL INFORMATION

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Ouestion 420.116 1he staff has wncluded that Westinghouse's April 29,1993 response to Q420.56 (EMI test on protection cabinet) is not acceptable. The electronuignetic interference qualification test performed for the South Texas Project (rt ported in WCAP-11341, dated November,1986) does not address many concerns of EMI protection of the safety-related digital systems. The NRC is in the process of issuing a regulatory guide that describes the design, installation, and testing practices that are acceptable to the NRC staff for addressing the susceptibility of digitall&C systems to EMI/H1'l in a nuclear power plant environment (The draft guide includes Military Standards Mll STD-461C and -462 and IEC standard 801). IEEE Std. 1050-1989, "lEEE Guide for Instrumentation and Control Equipment Grounding in Generating Stations," provides guidance for grounding and shield practices for digital systems. It also provides guidance to control the susceptibility of digital I&C systems when these systems are exposed te interfenence sources. Discuss conformance of the AP600 design to IEEE Std. 1050-1989, and provide the interface requirements for liMI/RFI protection.

Response

IEEli Std. 1050 1989 deals mainly with instrumentation and control equipment installation and field wiring grounding practices. The standard is not a licensing requirement or endorsed by a regulatory guide. However, the AP600 will follow the design practices addressed by this standard.

Present EMI/RFI interface requirement levels are described in the response to RAI 420.59 as level 3 of IEC 801-3 (10V/m; 27MHi-500MHz).

SSAR Revision: NONii W Westint'410use

NRC REQUEST FOR ADDITIONAL INFORMATION mi i

Response Revision 1 Question 471.4 Section 12.1.1.1 of the SSAR states that pipes containing radioactive fluids or radioactive sources are adequately shielded and properly routed to minimize exposure to personnel. Indicate and describe on area maps all radioactive horizontal pipe chases and areas that contain radioactive piping that personnel could come in contact with.

Resporise:

The radioactive pipe chases that contain radioactive piping in the Nuclear Island and the Radwaste building and areas that contain radioactive piping that personnel could come in conact with (er.cluding radioactive equipment) have been identified in the area maps provided in Westinghouse letter ET-NRC-93-4022/NSRA-APSle93-0464 dated November 29,19?

Piping areas that are normally inaccessible are distinguished from those that are accessible by a different pattern on these maps. Accessible piping areas are those where personnel may come in contact with the radioactive piping after proceeding through the plant permanent access control provisions such as a locked entrance or an entrance barricade (e.g., rope, chain, et.~.).

Normally inaccessible piping areas are areas confined within shielded compartments where there is no normal access. Access to those areas requires removal of shielded barriers such as hatches or steel plates.

'Ihere are no radioactive pipe chases in the turbine building.

Refer to the response to RAI 471.14 for access control provisions for these areas.

SSAR Revision: NONE W Westinghouse i

1 NRC REQUEST FOR ADDITIONA1. INFORMATION m

!ii Response Revision 1

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n Question 471.14 Section 12.3.1.2 of the SSAR states that Radiation Areas and High Radiation Areas will be posted and controlled.

No mention is made of Very High Radiation Areas as defined in 10 CFR 20.1602. Provide the location of any areas where personnel could be exposed to radiation levels greater than 100 Rads in one hour as stated in the Standard Review Plan, and describe any special controls to prevent personnel entry in to these areas.

Also, Section 12.5.4 of the SSAR states the entrances to High Radiation Areas are equipped with audible and/or visible alarms. Clarify this statement. Will these areas also be locked? Discuss how the requirements of 10 Ci R 20.1601, " Control of Access to High Radiation Areas,' and 10 CFR 20.1602, " Control of Access to Very High Radiation Areas." and the guidance of Regulatory Guide 8.38 (issued in May 1993) will be implemented.

Response

A set of the Radiological Access Control drawings for normal operation / shutdown for the nuclear island, annex building, radwaste building, turbine building, and the site plot plan is provided in Westinghouse letter IIT-NRC 4022/NSRA-APSle93-0464 dated November 29,1993. These drawings have been developed to incorporate the requirements from the revised 10 CI'R 20 and Regulatory Guide 8.38 on the control of access to high and very high radiation areas. These drawings will also be added into Section 12.3 as Figure 12.3-3.

SSAR Revision:

SSAR Sections 12.3 and 12.5 will be revised as follows:

12.3.1.2 Radiation Zoning and Access Control Accem to areas inside the plant structures and plant yard area is regulated and controlled by mdiathm m ing-w*Le**ms*4 posting of radiation signs, control of per6cnnel, and use 'of alarms and locks (Section 12.5). b4t-mdieth 1 - der ***uhe-m!h '

' ? =ge t - hk4*-the gg=g 'e ef-m*4dbuting-*mmeris as4 cow *iat-bvw.hieldine, During plant operation, personnel gain access to radi:!!r

msn bl radiologically restricted areas through the access control areas in th' annex II and radwaste buildings.

Plant areas are categorized into radiation wnes according to eyw4chlesign basis radiation levels and anticipated personnel occupancy with consideration gisen toward maintaining pers(mnel exposures ALARA and within the standards of 10 CFR 20. Rooms, corridors, and pipeways of plant buildings are evaluated for potential radiation sources during normal, shutdown, spent resin transfer, and emergency operations; for maintena;sce occupancy requirements; for general access requirements; and for material exposure limi s to determine apprnprimea t

n ming. Each radiation zone defines the radiation level range to which tho' aggregate of contributing sources is attenuated by shielding. The radiation mne categories employed and zoning for each plant area under normal conditions is s,hown in Figure 12.3-1. The mning for each plant area under accident conditions is shown in Figure 12.3-2. Radiation mnes shown in the figures are based upon conservative design data. ': Actual in-plant zones and i

control of personnel access are based upon surveys conducted by the combined license holder. Access control provisions for each plant area under normal expected conditions are shown in Figure 12.3-3. These provisions l

471.14(R1)-1 W Westinghouse 1

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NRC REOUEST FOR ADCITIONAL INFORMATION

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Response Revision 1 e

implement'the requirements of 10' C1'R 20 and utilize the alternative access. control methods outlined in Regulatory Guide 8.38. Radkaien :cr

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Posting of radiation signs, control of personnel access, and use of alarms and locks are Combined Ucense applicant responsibilities and are in compliance with requirements of 10 CFR 20.

12.5.2.1 Access and Exit of Radiologically Controlled 4testricted Areas Access to the radiologically contr+461 restricted area (RCA) encompassing the containment and potentially contaminated areas of the annext and auxiliary, ar

" buildings is normally through the dmble doors located d'"-

in the noeth+outheast-west corridor that leads to the health physics booth where RCA entry and exit are ccrirolled at-4h*4w.rthwul-of-the-anne &bu;ldmg. Exit from the RCA is normally through the same door. The radwaste building has its own RCA access and exit facility.

12.5.4 Controlling Access and Stay Time ne-e a.trk4cd A controlled access area kw4udes is defined as an _ areas within the primary security boundary. This area 44erther4m+Lwedown4ato+on-radi*tioneca*-ami includes radiologically contn461 restricted areas (RCA)(for radiation protection purposes). Centudhi-The re4.tricted areas ar+can be Turther categorized as radiation areas, high radiation areas, vey high radiation areas, airborne radioactivity areas, contamination areas, and radioactive materials areas, to comply with 10 CFR 20, ami plant procedures and instructions, and Regulatory Guide 8.38.

Entrances to the RCA areas i*-are normally throur;h the access control areas (previously discussed in Subsection 12.5.2).

11ased on conservative design bases, high radiation :treas and very high radiation' areas are segegn:cd and identified in accordance with 10 CFR 20 and Regulatory Gt:ide 8.38 and are shown in Figure 1234. newaranee

. =;uipped-w*h4mdil4e+r:d'c - "r-eianew Access control pmvisions to expected high to4dgh4adin' -

and very high radiation areas are shown in Figure 12.3-3.

471.14(R1 )-2 W Westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION a

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Response Revision 1 e

Oucstion 471.19 Figures 1.2-30 and 1.2-31 of the SSAR show areas around the steam generator blowdown heat exchangers, CCS pumps and heat exchangers, and the condensate polishing unit that would add future shield walls, if required.

Describe what future events or conditions would require the addition of additional shielding in these areas. Also, provide the expected radiation levels in these areas during these events or conditions.

Response

Based on the revised system designs. a set of the Radiological Access Control drawings for normal operation / shutdown for the building is provided in Westinghouse letter ET-NRC-93-4022/NSRA-APSle93-0464 dated November 29,1993.

The addition of shielding around the steam generator blowdown heat exchangers, the component cooling water pumps and heat exchangers, and the condensate polishing system is addressed as follows:

Steam Generator Blowdown Heat Exchancers Permanent shielding for plant personnel protection is provided for the blowdown demineralizers. Radiation monitors and sampling provisions are provided for early detection of radioactive primary-to-secondary leakage on the steam generator blowdown system (BDS). The design basis radiation zones for the blowdown system are zone V inside the demineralizer shielded area, zone I inside the heat exchanger shielded area, and zone I outside the blowdown system shielded area (See Note).

The area around heat exchangers could become Zone Il should the fuel defects and primary to secondary leakage approach the design basis values. However access to this area would not be restricted in that the dose wculd not exceed two millirem in any hour and the dose rate would not be expected to exist for an extended period of time such that the annual dose limit of 50 millirem per year to an individual member of the public would be exceeded.

Canponent Cooline Watei Pumns and Hut Exchancers Radiation monitoring and sampling provisions, continuous and grab, are provided for early detection of radioactive leakage within the component cooling water system (CCS). Radiation mc,nitoring and sampling provisions are addressed in the response to RAls 252.144, 460.14, and 460.15. Early detection of radioactive leakage within CCS permits isolation and repair of the leak; therefore, no shielding is provided around the CCS components. The design basis radiation levels correspond to the designated radiation zones for the CCS components; zone i for the area around CCS (See Note).

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NRC REQUEST FOR ADDITIONAL INFORMATION

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Response Revision 1 Condensate Polishine tJnit lhe space reserved for shielding around the condensate polishing unit may be required in the event of a primary-to-secondary tube leak in the steam generator. Radiation monitoring and sampling provisions in blowdown and condenser air removal systems allow for early detection of leaks; these provisions are addressed in the response to RAl's 252.144, 460.14, and 460.15. Although no radioactive buildup is expected and no shielding is provided, a slow buildup of radioactive material can occur in the polishing system resin. The design basis radiation levels for the polishing unit correspond to the designated radiation zones for this area: zone 11 immediately adjacent to condensate polisher and zone i outside the temporary shielded area.

Note:

Reference SSAR Section 12.3. Plant radiation innes are defined in Figure 12.3-1 (sheet 1). Radiation zone designations for the areas regarding the steam generator blowdown heat exchangers, the CCS pumps and heat exchangers, and the condensate polishing unit are referenced in Figure 12.3-1 (sheets 15,16)

SSAR Revision: N O Nil 471.19(R1)-2 3 Westinghouse

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i NRC REQUEST FOR ADDITIONAL INFORMATION

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Question 952.5 j

When will the primary pumps be shut off relative to the time that the break is opened?

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Response

i lhe current plan is to use logic for the OSU test facility similar to that for the Al%00, in which an 'S' sigt.al is detected by pressurizer pressure or level, or both. The "S" signal could also be a time delay from when the break is initiated. After the "S" signal, the reactor coolant pumps will be tripped following a 8 second delay, which is one-half the time for the Al%00 and is consistent with the OSU scaling logic.

SSAR Revision: None PRA Revision: None i

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NRC REQUEST FOR ADDITIONAL INFORMATION I

Question 952.10 llow will the steam generator leveh be controlled at steady state initial conditions and during a transient? What will the feedwater temperature be? liow will its flow rate be controlled? 110w will the steam flow be controlled?

Response

For steady-state operation, feedwater flow and steam flow will be monitored ann antrolled by a process control program to maintain a constant preset level in the steam generators. Steam generator feed flow and steam flow will be isolated following receipt of an *S* signal. The initial feedwater temperature will be room temperature (~ 70*F) and the flow will be from a feedwafer tank located inside the test facility.

SSAR Revision: None PRA Revision: None 952.10-1 W westingneuse