ML20045H643

From kanterella
Jump to navigation Jump to search
Forwards Responses to Request for Addl Info on AP600,per NRC 930312 & 0413 Ltrs
ML20045H643
Person / Time
Site: 05200003
Issue date: 07/16/1993
From: Liparulo N
WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To: Borchardt R
NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
ET-NRC-93-3924, NUDOCS 9307210079
Download: ML20045H643 (48)
v • d • e


Text

@

Westinghouse Energy Systems Bm 355 .

Eleetric Corporation Pittsburgh Pennsylvania 15230-0355 ET-NRC-93-3924 NSRA-APSI 93-0249 Docket No.: STN-52-003 July 16,1993 Document Control Desk  ;

U.S. Nuclear Regulatory Commission Washington, D.C. 20555 ATTENTION: R.W.BORCHARDT

SUBJECT:

WESTINGHOUSE RESPONSES TO NRC REQUESTS FOR ADDITIONAL INFORMATION ON THE AP600

Dear Mr. Borchardt:

Enclosed are three copies of the Westinghouse responses to NRC requests for additional information on the AP600 from your letters of March 12,1993 and April 13,1993. This transmittal completes the responses to the March 12,1993 letter. A listing of the NRC nequests for additional information i responded to in this letter is contained in Attachment A. Attachment B is a complete listing of the questions associated with the March 12,1993 letter and the corresponding Westinghouse letters that provided our response, in addition to responses to NRC requests for additional information with identifying numbers, the enclosure includes the response to an unnumbered request for additional information contained in your letter of April 2,1993.

These responses are also provided as electronic files in Wordperfect 5.1 format with Mr. Hasselberg's copy.

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

Nicholas J. Liparuto, Manager Nuclear Safety & Regulatory Activities

/nja Enclosure cc: B. A. McIntyre - Westinghouse F. Hasselberg - NRR

  • ^

190 s 9307210079 930716

.M

,M 1 .

l PDR IU  ;

ADOCK 05200003 E(g A PDR 3 J  !

ET-NRC-93-3924 ATTACHMENT A AP600 RAI RESPONSES SUBMITTED JULY 16,1993 RAINo. Issue 420.024  : Level measurement NC gas concerns 420.043 i bypass and inoperable status display 1

~

420.055  : FMEA, DAS 420.070  ;- Software ITAAC-420.071  : Remote shutdown with single active failure 420.072  : Remote shutdown workstation I&C 420.074 i Remote shutdown workstation j 420.077  : Type F variables 420.078  : Steam generator water level 420.079 i Neutron flux instrument range 420.080  ; RCS boric acid concentration 420.081 1 Condenser air removal radiation instrument range 420.092 DWS isolation 420.103 i LCO, containment purge & exhaust isolation instr  ;

420.104  : LCO, fuel bldg air cleanup sys actuation instru 471.010  ; Lighting in high radiation areas 471.018 l BAS /CAS connections with other systems 471.019- l Event / condition requiring additional shield walls 620.051  : HFE request for matrix i

b I

1

- f

,. , , - - r

1 l

Printcio7/16/93 -

ATTACHMENT B CROSS REFERENCE OF WESTINGHOUSE RAI RESPONSE TRANSMITTALS TO NRC LETTER OF MARCH 12,1993 Question - Issue NRC Westinghouse No. Letter Transmittal Date 100.009 ITAAC rationale and selection criteria 03/12/93 04/29/93 420.009 Conformance to software standards 03/12/93 05/28/93 420.010 Conformance to EMI/RFI standards 03/12/93 05/28/93 420.011 Operation of monitoring system for LOFW 03/12/93 05/1493 420.012 On/off control of plant loads 03/12/93 04'29/93 420.013 IPS/ICS differences 03/12/93 04/29/93 420.014 Interface arrangement between workstations 03/12/93 05/14'93 420,015 Workstation graphic display design features 03/12/93 06/17/93 1 420.016 Alarm system design features 03/12/93 06/17/93 420.017 Special monitoring system 03/12/93 04/29/93 420.018 Reference to IEEE Std 769-1983 03/12/93 0429/93 '

420.019 LLNL report on IEEE Std. 796-1083 03/12/93 05/2&S3 420.020 EMI protection for digital 1&C 03/12/93 04'29/93 ,

420.021 RT group 1/ group 2 inconsistencies 03r12/93 04'29/93 420.022 IEEE Std 796 bus function 03/1293 04/29/93 420.023 Monitor bus design basis 03/12/93 04'2W93 '

420 024 Level measurement NC gas concerns 03/12/93 07/16/93 420.025 RC hot leg / cold leg temperature measurement 03/12/93 06/17/93 ,

420.026 Reactor coolant pump speed rnonitoring 03/12/93 05/1493 420.027 Inter-cabinet communications 03/12/93 05/1493 ,

420.028 ESFAC interfaces 03/12/93 0429/93 420.029 Status indication criteria 03/12/93 05/1493 420.030 Remote VO analog cabinet function 03/12/93 04'29/03 420.031 Reactor trip channel bypass logic 03/12/93 04/29/93 420.032 Fault tolerance design methods 03/12/93 04/29/93 420.033 ESFAC diagram clarification 03/12/93 04/29 S 3 420.034 Commu >ication diagram clarification 03/12/93 04'29/93 420.035 Isolation devices 03/12/93 04'29/93 420.036 Testing of protection system actuated equipment 03/12/93 05/28/93 420 037 Interlocks to prevent simultaneous testing 03/12/93 04/29/93 420.038 Number of control muttiplexer cabimets 03/12/93 05/14/93 420.039 IPS/ICS Interface 03/12/93 05/28/93 420.040 Technical support center ISC design 03/12/93 06/17/93 420 041 GDC 21 with two channels bypassed 03/12/93 04/29/93 420.042 Reactor trip breaker bypass design 03/12/93 0479/93 420.043 bypass and inoperable status display 03/12/93 07/16/93-420.044 Transmission of post accident monitoring info 03/12/93 05/2&S3 420 045 Division synchronization 03/12/93 05/14/93 420.046 P-18 03/12/93 05/14'93 420 047 Manual block control switch location 03/12S3 06/17/93 420.048 Time constants 03/12/93 05/1493 420.049 Reset reactor trip (not redundant) 03/12/93 05/1493 420.050 Video display unit qualification nethods 03/12/93 04/29/93  !

420 051 Portable testers 03/12/93 04/29/93 420 052 Extreme environmental & energy supply conditions 03/12/93 05/26/93 420.053 FMEA, protection cabinet power supply arrangement 03/12/93 05/28/93 420.054 DAC design V&V process 03/12/93 04/29/93 420.055 FMEA, DAS 03/12/93 07/16/93 420.056 Standard EMI seismic cabinet 03/12/93 04'29/93 420 057 Different bits of resolution 03/12/93 04'29/93 420 058 Cnterion for use of EEPROMs 03/12/93 04/29/93 420 E9 Max field parameter for EMI/RFI protection 03/12/93 04'29/93 420 060 Isolation transformer and EMl/RFI effects 03/12/93 04/29/93 420 061 Limits for shipping environment 03/12/93 04'29/93 420 062 Cabinet cooling assembly power supply 03/12/93 0479/93 420 063 Automatic calibtation feature 03/12/93 04/29/93 420 E4 Communication subsystem single failure analysis 03/12/93 04/29/93 420E5 Automatic tester anaytsis 03/12/93 04'29/93 420 066 Scope of automatic surveillance testing 03/12/93 04/29/93 420 067 Portable tester software developmen; 03/12/93 04'29/93 420 068 Data highway for protection & monitoring system 03/12/93 04/29/93 420.069 Number of microprocessor based subsystems 03/12/93 04'29/93 420 070 Software lTAAC 03/12/93 07/16/93 l

Page1 1

Question lasue NRC Westinghouse No. Letter Transmittal Date ,

420.071 Remote shutdown with single active failure 03/12/93 ' 07/16/93 420 072 Remote shutdown workstaten l&C 03/12/93 07/16/93 420.073 Remote shutdown workstation 03/12/93 06/17 S3 420.074 Remote shutdown workstation 03/12/93 07/16/93 420.075 Remote snutdown workstation 03/12/93 0479/93 420 076 Remote shutdown workstation securtty 03/12S3 0479/93 ,

420.077 Type F variables 03/12/93 07/16/93 1 420.078 Steam generator water level 03/12/93 07/1693 420.079 Neutron flux instrument range 03/12/93 07/16S 3 420.080 RCS boric acd concentraton 03/12S3 07/16.93 420.081 Condenser air removal radiation instrument range 03/12/93 07/16S3 420.082 Safety panel displays 03/12/93 05/1493 420.083 Procedure for loss of alarm system 03/12S 3 04/29/93 420.084 impact of environment on alarm system 03/12/93 04/29/93 420.085 Nuisance starms 03/12/93 '05/1493 420.086 Time response dunng upset condition - 03/12/93 OSr28/93 420.087 Muttiplexer connection to each workstaten 03/12S3 0429/93 420 088 First stage ADS vafve 03/12 S3 05/1493 420.089 Trip-Normal-Bypass switch descripton 03/12/93 0429S3 .!

420.090 Units involved in generating S signal 03/12S3 05/14/93  ;

420.091 Request for clanication 03/12S3 06/17/93 420.092 DWS isolatkm 03/12/93 07/1693 420.093 WCAP-12648 03/12/93 06/17/93 620.051 HFE request for matnx 03/12/93 07/16/93 Records printed. 87 m.

i I

1

. we s .

NRC REQUEST FOR ADDITIONAL INFORMATION y

2

_ e Question 420.24 Describe the level measurement arrangements for the pressurizer water level, the steam generator water level (narrow and wide range), and hot leg water level. Address concerns related to potential problems of non-condensable gases in the reference leg that have been raised in NRC Information Notice No. 92-54 " Level Instrumentation inaccuracies Caused by Rapid Depressurization," and in Generic letter 92-04, " Resolution of the Issues Related to Reactor Vessel Water Level Instrumentation in BWRs Pursuant to 10 CFR 50.54(f)."

(Section 7.1.2.8).

Response

Pressurizer level Instrumentation Four pressurizer liquid level transmitters provide signals to the protection and safety monitoring system (PMS).

These transmitters also provide isolated input to the diverse actuation system (DAS) and to the plant control system (PLS).

He concerns raised in the NRC Information Notice and the Generic letter deal with level errors resulting from a rapid depressurization of the reactor coolant system, which causes the dissolution of noncondensable gases from the liquid in the pressurizer reference legs. His phenomenon can result in a higher pressurizer level indication than is actually present in the pressurizer following large, rapid depressurizations in the primary system.

The design and layout of the AP600 pressurizer level instrumentation minimize the potential for the dissolution of noncondensable gases. This design includes a downward sloping impulse line from the pressurizer and no condensate pot. Steam condensation to maintain and replenish a reference level occurs on the exposed inner surface of the impulse line, on the liquid surface, and on the root valve inner surface when they are exposed (uninsulated).

Noncondensable gases will not concentrate (as they do in a condensate pot) because they have direct access to the pressurizer steam space and a buoyancy force to diffuse them in that direction.

Steam Generator Level Instmmentation Four wide-range and four narrow-range liquid level transmitters per steam generator provide signals to the j protection and safety monitoring system and isolated input signals to the plant control system. The wide-range steam j generator level transmitters also provide isolated input signals to the diverse actuation system.

]

The level instrumentation systems of the pressurizer and steam generators of the AP600 are functionally similar.

However, the steam generators do not contain a significant concentration of noncondensable gases such as hydrogen j on the secondary side. Gases that are present, such as ammonia and boric acid vapor, are highly soluble in water j and will not build up during normal operations to levels significantly higher than those that exist in either the steam l

generator secondary side or the condenser. At the concentration level of these highly soluble gases in both the fluid j and steam volumes of the steam generator, off-gassing during depressurization would not be significant. l 420,2M W Westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION

~

iii it!r!

IQ In summary, the steam generator water level instrumentation is not subject to significant depressurization-induced error associated with the presence of noncondensable gases in the reference leg. l Hot leg level Instrumentation The hot leg level instrumentation is provided primarily for shutdown operations. The design and layout of this instrumentation are similar to that of the AP600 pressurizer level instmmentation. Therefore, because of the lower levels of dissolved gases and layout requirments, off-gassing should not significantly affect the accuracy of the hot leg level measurement.

SSAR Revision: NONE l

420.24-2 T westinghouse l

l 1

1 NRC REQUEST FOR ADDITIONAL INFORMATION u!! In U IN

_ t Question 420.43 Describe the design of the bypass and inoperable status display in the main control room. Are the displays located on operator's workstation? Are they continuously indicated or do they need to be retrieved by the operator?

Discuss the conformance of the design to Regulatory Guide 1.47, " Bypassed and Inoperable Status Indication for Nuclear Power Plant Safety Systems." (Section 7.1.4.2.13)

Response

Regulatory Guide 1.47 states that the operator needs to know of the operating status of the safety-related systents and needs to know the extent to which the bypassing activity will affect those systems and whether that effect is permissible within the provisions of the license. He AP600 information system is designed to provide the operator with this information. His required information is incorporated into the alarm system and the information system of the main control room. At the operators' workstations, physical and functional displays are available. Alarms are available on the operators' workstations and on the wall panel information system. High level plant status during any plant state is continuously available on the wall panel information system.

The physical displays show how a component's availability or unavailability impacts the alignment and availability of the system. This is indicated on the display, including bypasr,ed or deliberately induced inoperability of the protection system and the systems actuated or controlled by that protection system, as required by Regulatory Guide 1.47.

De functional displays show the functional relationships between the pieces of equipment, like sources of water available for cooling, alternate paths available to transport the water, and selectable injection points. These displays also indicate the status and availability of the support systems necessary to satisfy the functional purposes. In this way, the indications aid the operator in recognizing the effects on plant safety of seemingly unrelated or insignificant events and show the status of systems relative to their supporting or auxiliary systems.

The alarm system presents abnorrnal conditions as alarms. Improper safety system alignments, safety-related component unavailability, and bypassed protective functions are considered in the alarm logic. His information is continuously monitored by the alarm system. The alarm system establishes priorities and indicates the highest-priority alarms on the wall panel information system. Abnormalities identified through the alarm system are used to direct the operator to the appropriate place in the information system on the operators' workstations, where additional information can be found. If the wall panel information system fails for any reason, the displays are available at the operational display system of each operator workstation. Any abnormalities not of the highest priority are built into the information system in the location where the operator has to operate controls or monitor status.

In addition to the preceding features integrated into the display and alarm systems, indication will be provided for the following conditions, in accordance with Regulatory Guide 1.47:

420.43-1 l W W85tirigh0llSB l I

i I

NRC REQUEST FOR ADDITIONAL INFORMATION l

'1 E l

  • If it is expected to occur more frequently than once a year
  • If it is expected to occur when the affected system is normally required to be operable Th: capability to manually indicate that something is inoperable is provided in the system.

SSAR Revision: NONE 420.43 2 W Westinghouse e

NRC REQUEST FOR ADDITIONAL INFORMATION

't rW +:

it!

n Ouestion 420.55 Provide a failure modes and effects analysis to demonstrate that inadvertent actuation of the DAS does not result in an additional plant transient, plant trip, or ESF actuation. (Section 2.11 of WCAP-13382)

Response

Credible single failures of the diverse actuation system (DAS) could be postulated to result in a plant transient, plant trip, or ESF actuation. The DAS is designed to minimize these potential failures. An evaluation of the potential transients initiated by inadvertent DAS actuation is underway.

The results of the failure modes and effects analysis (FMEA) including the effects of an inadvertent actuation of the DAS caused by credible single failures will be provided by September 15, 1993.

SSAR Revision: NONE 420.55-1 W Westinghouse

l 4

NRC REQUEST FOR ADDITIONAL INFORMATION Ouestion 420.70 WCAP 13383, "AP600 Instrumentation and Control Hardware and Software Design Verification and Validation Process Report," describes an engineering approach to aid the development of hardware, software, and system design. Provide a formal design implementation process with a phased inspection, test, analysis and acceptance criteria (ITAAC) for design development. The detailed process description should be non-proprietary. The information presented in WCAP-13383 should be included in a submittal of the design description and ITAAC under

" Hardware and Software Development." The description of the development plan should include details of the hardware and software management plan, the configuration management plan, and the verification and validation plan.

Response

The AP600 Tier 1 information submitted on December 15, 1993 includes system ITAAC for the following instrumentation and control systems:

  • Protection and Safety Monitoring System
  • Diverse Actuation System
  • Data Display and Processing System
  • Incore Instrumentation System
  • Plant Control System
  • Radiation Monitoring System ne Tier 1 information and associated ITA ACs provided for these systems show that the instrumentation and control systems are built and installed in accordance with the functional design described in the SSAR for design certification. The Tier 1 system function and hardware information is complete.

The process to design, manufacture, install, operate, maintain, and modify the instrumentation and control syste ns is described in SSAR Chapter 7 and WCAP-13383, " AP600 Instrumentation and Control Hardware and Software Design Verification and Validation Process Report." The process summarized in WCAP-13383 is proprietary because the information reveals the distinguishing aspects where prevention of the use of the process by any of Westinghouse's competitors without license from Westinghouse constitutes a competitive economic advantage over other companies and use of the process by a competitor would reduce his expenditure of resources or improve his competitive position in the design, manufacture, or licensing of similar instrumentation and control systems. The design of the AP600 instrumentation and control systems will be parformed consistent with existing Westinghouse procedures that have been utilized in the design of other instrumentation and control systems licensed by the NRC.

Rese procedures are available for NRC review. The detailed process and procedures are not Tier 1 information, but are available for NRC review. Review points will be established in the design process but not included as Tier 1 information. See the response to RAI 420.7 for additional information.

SSAR Revision: NONE 42oaoa W wes1inghouse 4

NRC REQUEST FOR . ADDITIONAL INFORMATION 1m ..g

. e Question 420.71 Section 7.4.3.1.1 of the SSAR states that the remote shutdown workstation 's designed to allow safe shutdown of the plant following an eva:uation of the control room coincident with the loss of offsite power and a single active failure. Describe the method to achieve safe shutdown if the single active failure is the remote shutdown workstation itself.

Response

In a main control room evacuation event, safe shutdown is automatically achieved and er itained by the protection uhaonitoring system actuating the passive safety-related fluid systems. He remote sht.t . awn workstation provides the operator with the capability of performing these same actions manually. The workstation has safety-related soft cuatrols that provide manual actuation of the safety-related features. There are also nonsafety-related soft controls through which nonsafety-related features can be controlled.

The complete failure of the remote shutdown workstation is not a credible single failure due to the redundancy within the workstation and its support features. The redundancy includes soft controls, multiplexers, data links, and instrument displays. The operator's ability to control the safety-related features is maintained with the failure of one soft control, multiplexer, data link, or display.

SSAR Revision: NONE I

420.71-1 W Westinghouse 1

7 3

NRC REQUEST FOR ADDITIONAL INFORMATION Question 420.72 Provide a list of controls and instrumentation to be located on the remote shutdown workstation. If the controls include both the safety and the noa-safety circuits, describe how the design satisfies the separation criteria of Regulatory Guide 1.75. (Section 7.4.3)

Response

h Al*R remote shutdown workstation contro s and monitors the same equipment and instrumentation as do the main control room workstations. The remote shutdown workstation controls both safety-related and nonsafety-related equipment, as do the main control room workstations. The separation features in the remote shutdown workstation are equivalent to tL,se in the main control room workstations. See SSAR Subsection 7.1.4.2.6 for a discussion of these separation features. Appendix 1 A provides a discussion of AP600 conformance with Regulatory i

. Guid31.75 for the nuun control room and the remote shutdown u orkstations.

SSAR Revision: NONE s

420.72-1

NRC REQUEST FOR ADDITIONAL INFORMATION

!ai! 1Eu 19 2 Ouestion 420.74 Describe the design that allows the plant to be brought to cold shutdown from the remote shutdown workstation and other local controls. (

Reference:

Section 4.9.3 of the EPRI ALWR Utilities Requirements Document)(Section 7.4.3 of the SSAR)

Response

SSAR Section 7.4.3 states that the shutdown basis for the AP600 remote shutdown workstation is a safe shutdown.

Safe shutdown is automatically achieved by the protection and monitoring system and the passive safety-related systems (see SSAR Section 7.4.1). This SSAR Section describes the AP600 automatic safe shutdown operation, including the equipment used. The operator is provided with manual control of these passive safety-related systems from both the main control room workstations and from the remote shutdown workstation. These workstations have safety-related soft controls that provide manual control capability to achieve safe shutdown.

The remote shutdown workstation can also control nonsafety-related equipment. When the nonsafety-related systems are available, this capability allows the operator to put the plant in an appropriate shutdown condition including cold shutdown.

SSAR revision: NONE l

l l

l 420,74-1 W Westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION E

~

y.

Ouestion 420.77 The AP600 SSAR defines a " Type F" variable which is not defined in Regulatory Guide 1.97. Section 7.5.2.1.6 of the SSAR defines " Type F" variables as those that provide the information to allow the operator to take manual actions using non-safety-related systems to prevent the unnecessary action of safety-related systems, and to monitor the performance of the non-safety system. Discuss the features that minimize human errors due to any misleading information from " Type F" variables, including any problems or concerns with the instrument qualification. The staff considers that the draft international standard IEC 1226, "The Classification of Instrumentation and Control Systems important to Safety for Nuclear Power Plants," has addressed the appropriate requirements for this type of instruments. How will the design comply with the requirements and guidance given in both IEC 1226 and IAEA 50-SG-D8, " Safety-Related Instrumentation and Control Systems for NPP's." (The latter document provided the safety principles for the IEC standard.)

ne selection process for Type F variables appears to be incomplete. For example, there are no spent fuel pit cooling system variables included in Table 7.15-9 of the SSAR; however, the spent fuel pit cooling system is listed as a defense-in-depth system in Section 2.3.1.3 of Chapter 3 of Volume IH of the EPRI ALWR Utilities Requirement Document. This would indicate that some of the spent fuel pit instrumentation given in Figure 9.1-8 of the SSAR should be Type F. Clarify your reason for not including any of them in the design.

Response

Type F variables are included in SSAR Section 7.5 to address monitoring of the nonsafety-related systems that provide defense-in-depth functions. He instrumentation associated with these systems, including the post-accident monitoring instrumentation, is designed to the same classification as the system unless it is used to perform a higher-level function. For example, an instrument in a nonsafety-related system used to perform a safety-related function is designed to Class IE requirements.

Design features intended to minimize human errors due to any misleading information from nonsafety-related instrumentation are discussed in SSAR S .ction 18.9 Consistent with the NRC/ industry agreement regarding regulatory treatment of nonsafety systems, the AP600 nonsafety-related systems are being evaluated for potential risk significance. Draft international standard IEC 1226, "The Classification of Instrumentation and Control Systems Important to Safety for Nuclear Power Plants," and I AEA 50-SG-D8, " Safety-Related Instrumentation and Control Systems for NPP's" are not regulatory requirements.

The applicability of the guidance provided by these documents is dependent on the outcome of the evaluation nonsafety-related systems and will be determined upon completion of that evaluation.

The Type F variables identified in SSAR Tables 7.5-1 and 7.5-9 address the AP600 nonsafety-related system I defense-in-depth ftmetions including spent fuel pit cooling. SSAR Tables 7.5-1 and 7.5-9 are revised to identify spent fuel pit pump flow, spent fuel pit temperature, and spent fuel pit water level as Type F, Category 2 variables.

420 m W westinghouse

NRC REQUEST FOR ADDITIONAL. INFORMATION m..t n Although startup feedwater flow is identified in SSAR Table 7.5-1, Sheet 3, as a Type F, Category 2 variable, it was inadvertently omitted from Table 7.5-9. 7he revised Table 7.5-9 includes startup feedwater flow.

SSAR Revision: Revise Table 7.5-1, sheet 7 as follows:

Table 7.5-1 (Sheet 7 of 12)

Post-Accident Monitoring System Rangel Typel Qualification Number of ODPS Variable Status Category . Instruments Power Indication Remarks Environmental Required Supply (Note 2)

Scismic (Note 1)

MCR chilled On/Off F2 None None 1/ pump Non- 1 E No water pump status MCR chilled Open/ F2 None None 1/ valve Non-1 E No water valve Closed status Spent fuel pit 0-1000 D2lF2 None None 1/ pump Non-1E No pump flow gpm Spent fuel pit 50- D24F2 None None 1 Non-1 E No temperature 250*F Spent fuel pit 0-100 % D2,F2 Harsh Yes 2 lh Yes water level PXS cold leg to Open/ D2 flarsh Yes 1/ valve IE Yes core makeup Cloned tank valve status Core makeup Open/ D2 liarsh Yes 1/ valve lE Yes tank to reactor Closed vessel valve status Core makeup 0-100 % D2,F2 liarsh Yes 3/ tank IE Yes Trending required tank level (Note 4)

IRWST to Open/ D2 Harsh Yes 1/ valve IE Yes reactor vennel Closed valve status ADS to RCDT Open/ D2 Harsh Yes 1/ valve !E Yes valve status Cloned 4203L2 W-Westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION

jili uHi M  !!!

Revise Table 7.5-9, sheets 2 and 4 as follows:

Table 7.5-9 (Sheet 2 of 4)

Summary of Type F Variables Variable Type / Category Startup feedwater control valve status F2 Startup feedwater flow F2 Main feedwater flow F2 Steam generator level (WR) F2 Reference leg temperatures F2 Steam flow F2 Deaerator storage tank level F2 Deaerator storage tank to SFW pump valve status F2 Condensate storage tank to SFW pumps valve status F2 Main feedwater control valve status F2 Main feedwater isolation valve status F2 Main steam isolation valve status F2 Main feedwater pump status F3 Startup feedwater pump status F3 Condenser steam dump valve status F2 Condensate storage tank level F2 Fressurizer spray - cold leg to pressurizer valve status F2 Auxiliary spray - regenerative heat exchanger to pressurizer valve F2 status Regenerative heat exchanger to SG/CL valve status P2 Makeup pump to regenerative heat exchanger control valve status F2 Makeup flow F2 Makeup pump status F3 420.77-3 l W_

Westinghouse 1

i

NRC REQUEST FOR ADDITIONAL INFORMATION

!!!F u!s Table 7.5-9 (Sheet 4 of 4)

Summary of Type F Variables Variable Type / Category Service water pressure F2 Service water pump status F3 Service water valve status F3 Instnlment air compressor status P2 Instrument air valve status F2 i spent fud pit pumji flos F2

~ Spent fuel pit tem;ierators F2

lSpentTuel pit lwateilevel F2 420.77-4 W Westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION ti!N  :!!i

=

!![

Question 420.78 Regulatory Guide 1.97 specifies that the steam generator water level should be a Category i variable. Sheet 1 of Table 7.5-1 of the SSAR indicates that this indication is a Category 2 variable (1/SG). Justify this deviation from the Regulatory Guide.

Response

For current Westinghouse PWRs. steam generator water level is utilized by the operator to determine safety injection termination and to maintain adequate heat sink capability. The steam generators in the AP600 provide a nonsafety related heat sink capability, not required for mitigating SSAR Chapter 15 events. The safety-related heat sink capability is provided by the passive residual heat removal heat exchangers and the passive containment cooling system. The variables provided to monitor this capability are in-containmert refueling water storage tank level, passive residual heat removal heat exchanger flow, passive residual hest removal heat exchanger outlet temperature, passive containment cooling system storage tank water level and passive containment cooling water flow. As indicated in SSAR Table 'i 5-1, these indications are Category 1 variables.

Steam generator water le,el is provided as a Type D, Category 2 variable, since it provides the operator with information to verify the safety-related functions of the passive residual heat removal heat exchangers and passive containment cooling system to control secondary pressure and level. Steam generator water level is also provide <1 as a Type F, Category 2 variable, since it provides the operator with information to perform preplanned, manual, nonsafety-related system actions and to verify nonsafety-related system performance. For additional information regarding the use of Type F variables, see the response to RAI 420.77.

SSAR Revision: NONE E Westinghouse i

NRC REQUEST FOR ADDITIONAL INFORMATION ND liii i;

Ouestion 420.79 Regulatorj Cuide 1.97 specifies that the neutron flux instrument range should be 10-6% to 100%. Sheet I of Table 7.5-1 of the SSAR indicates the range of the AP600 design is 10-5% to 200%. Justify this deviation.

Response

The post-accident monitoring neu.rnn flux instrument range is 10-6 percent to 200 percent power. SSAR Table 7.5-

1. Sheet I will be revised accor$ ny.

1 l

420. m 1 E westinghouse I

NRC REQUEST FOR ADDITIONAL INFORMATION

!!ili 'IIIil N

!l.l.

. t SSAR Revision: SSAR Table 7,5-1, Sheet 1, to be revised as follows:

l Table 7.5-1 (Sheet 1 of 12)

Post-Accident Monitoring System j Rangel Type / Qualirration Number of QDPS l Variable Status Category Instruments Power Indication Remarks Environmental Msmic RmM kpply Note 2)

(Note 1) i l

RCS pressure 0-3300 B1, H2, Harsh Yes 3 IE Yes Located outside psig D2, Cl, containment F2 RCS TH 0-700* BI, B2, Harsh Yes 2 IE Yes Diverse Measure-l (Wide Range) F Cl, D2, ment:

l F2 Incore thermocou-l P*

(

RCS Tc 0-700* B1, B2, Harsh Yes 3 IE Yes

( (Wide Range) T Cl, D2,

! N Steam gen 0-100 % D2,F2 liarsh Yes I/SO IE Yes water level of span (wide range)

Steam gen 0-100 % D2,F2 Harsh Yes 1/S0 lE Yes water level ofspan (narrow range)

Steam gen ref 50- D2, F2 liarsh Yes 1/$0 lE Yes leg temp 420*F (W.R.)

!/S0 (N.R) l Pressurizer 0-100 % B1, D2, Harsh Yes 3 IE Yes I level F2 (Note 4)

B1,D2 Pressurizer 50- Harsh Yes 3 IE Yes reference leg 420*F (Note 4) temperature Neutron flux 1046 B1 Harsh Yes 3 IE Yes 200 % (Note 4) power Control rod 0-225 B3, D3 None None 1/ control Non-1E No position steps rod Containment 0 100 % B1, B2, Harsh Yes 3 IE Yes 72 A to 108 ft.

water level C2,F2 (Note 4) 420.79-2 W-Westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION if j Ouestion 420.80 Regulatory Guide 1.97 specifies that the RCS boric acid concentration range should be O to 6000 ppm. Sheet 3 of Table 7.5-1 of the SSAR indicates that this range is N/A. Provide ranges for this variable.

Response

Table 2 (PWR Variables) of Regulatory Guide 1.97 (Revision 3. May 1983) identifies RCS soluble boron concentration as a Type B, Category 3 backup variable. It is used to verify that a specific plant safety function, reactivity control, is being accomplished.

For the AP600, RCS soluble boron concentration measurement is specified as a Type B, Category 3 variable in conformance with P.egulatory Guide 1.97. This information is provided by manual sampling of the reactor coolant system. Therefore, direct indication of RCS boric acid is not required, and the specification cf an indication range is not applicable for this variable.

SSAR Revision: NONE

[ W85tingh00S8 1

i i

NRC REQUEST FOR ADDITIONAL INFORMATION Ouestion 420.81 Regulatory Guide 1.97 specifies that the condenser air removal radiation instrument range be 10-6 to 10+5 Ci/ce.

Sheet 11 of Table 7.5-1 of the SSAR indicates that the range is 10-6 to 10"I Ci/cc. Justify the deviation. Also, justify the ranges for the control room air radiation and the steam generator blowdown radiation instruments.

Response

As stated in the response to RAI 460.16a, the condenser air removal effluent noble gas range will be modified to detect from 10-6 to 10+5 Ci/ce.

The range specified for the main control room supply air duct radiation monitor,10-7 to 10-2 Ci/ce, is sensitive enough to initiate main control room ventilation isolation in the event of high outside radiation.

The sensitivity of the steam generator blowdown radiation monitor will be changed from 10-8 Ci/cc to 104 pCi/cc to provide a monitor that will detect a steam generator tube rupture in the presence of normal background radiation.

SSAR Tables 7.5-1 and 11.5-1 will be revised as follows:

SSAR Revision:

Table 7.5-1 (Sheet 11 of 12)

Post-Accident Monitoring System Rangel Typel Qualification Number Of QDPS Variable Status Category Instruments Power Indication Environmental Seismic Required Supply (Note 2)

(Note 1)

Steam generator 104- E2 None None 1 Non-1E No blowdown 10'I discharge radia- Ci/cc tion 4 20.8, .,

W weeneouse

NRC REQUEST FOR ADDITIONAL INFORMATION Table 11.5-1 Radiation Monitor Detector Parameters Detector Type Service isotopes Nominal Range BDS-JE-RE007 y Steam Generator Blowdown Discharge Cs-137 1.0E4 to 1.0E-1 pCi/cc l

i 420.81-2 W Westinghouse j l

1 l

l l

NRC REQUEST FOR ADDITIONAL INFORMATION 95 a4

(

Question 420.92 Section 15.4.6.2.5 of the SSAR states that any reactor trip signal will isolate unborated water from the demineralized water system (DWS). However, there is no signal shown on the reactor trip logie diagram. Explain how the reactor trip functions to isolate the unborated water from the DWS. (Sections 7.2 and 15.4.6.2.5)

Response

He chemical and volume control system is designed to operate following a reactor trip. It borates the RCS.

Following any reactor trip or source range flux doubling, the line from the makeup pump suction to the demineralized water system is i. olated by closing two safety-related, motoroperated, series valves. A nonsafety-related, three-way pump suction control valve is also realigned on the same signals so that the makeup pumps take suction from the boric acid tank. Additionalinformation on the chemical and volume control system operation is

! found in SSAR Subsection 9.3.6.4.5.1.

SSAR Sections 7.3 and 16.1 will be revised to include the reactor trip signal isolation of unborated water from the demineralized water system.

SSAR Revision:

SSAR Subsection 7.3.1.1 will be revised as follows:

l 7.3.1.1.4.7 Signal to Block Boron Dilution i l

A signal to block boron dilution et shutdown is derived from source range neutron flux increasing at an I excessive rate (source range flux doubling), indicating an unplanned boron dilution. This is illustrated in Figure 7.2-1, Sheet 3. An average of the source range count rate, sampled at least N times over the most recent time period i Tg, is compared to a similar average take at time period T 2 carlier. If the ratio of the current average count rate l to the earlier average count rate is greater than a value A, a logical "1" is generated in the division. On a l coincidence of two of the four divisions, boron dilution is blocked, nis is actuated as the closure of the chemical volume control system suction valves to demineralized water storage tanks.

The source range flux doubling signal may be blocked manually above the P-6 power level. Itis automatically reinstated below P-6.

~

Baron dilution is also blocked following a reactor trip by closing two safety-related/ mot 6r opardtsdiseries valves that isolate the line from the chemical. and volume' control system makeup pump.. suction to the demineralized water system.

l l

l 1

420.92-1 W-Westinghouse l

NRC REQUEST FOR ADDITIONAL INFORMATION l

!!! " !!!ill l 9 $ i SSAR Subsection 7.3.1.1.5 will be revised as follows:  ;

1 7.3.1.1.5 Blocks, Permissives, and Interlocks for Safeguards Actuation  !

1 ne interlocks used for engineered safety features actuation are designated as "P-xx" permissives and are listed in Table 7.3-2.

The interlocks from reactor trip are as follows:

Manual blocks to engineered safety features actuations are described below:

  • The block of boron dilution source range flux doubling can be manually actuated above the P-6 intermediate range power level.

Safeguards actuation on pressurizer low pressure, low steam line pressure, or low Teold can be manually blocked when pressurizer pressure is below the P-1I setpoint.

Steam line isolation on low compensated steam line pressure and low Teold can be nu ually blocked when pressurizer pressure is below the P-11 setpoint. When steam line isolation is manually blocked below P-11, steam line isolation on high negative steam pressure rate is automatically unblocked.

  • Feedwater isolation on lew-l or lew-2 Tayg may be manually blocked when pressurizer pressure is below the P-1I setpoint.

He steam /feedwater isolation and safeguards block control is provided to allow startup and cooldown.

The manual block control is used to permit blocking of automatic steam line isolation and power-operated relief valve block valve closure, feedwater isolation, and certain safeguards signals. The block control consists of two momentary controls mounted in the main control room. Operating either control blocks the following protection

. functions:

  • low Teold safeguards signal a

low Teold steam line isolation a low Teold startup feedwater isolation a low-l reactor coolant system Tavg feedwater isolation

420.92-2 W-Westinghollse l

1 l

1 NRC REQUEST FOR ADDITIONAL INFORMATION Automatic block of chemical and volume control system isolation and passive residual heat removal system ;

actuation on high pressurizer water level occurs when pressurizer pressure is below the F-11 setpoint.

SSAR Table 7.3-1, Sheet 6 will be revised as follows:

Table 7.3-1 (Sheet 6 of 8)

Engineered Safety Features Actuation Signals Actuation Signal No. of Actuation Permissives and Interlocks Channels / Logic Switches

14. Block of Boron Dilution (Figure 7.2-1, Sheet 3)
a. Flux doubling 4 2/4-BYP1 ---

calculation C

^

b! - Reactor' trip 1/ division $/4 (P-4)

15. Chemical Volume Control System isolation (See Figure 7.2-1, Sheet 6)
a. liigh pressurizer 4 2/4-BYPI Auto unblock above P-11 vcater level 420.92-3 T Westingh00S8

NRC REQUEST FOR ADDITIONAL INFORMATION SSAR Table 7.3 4, Sheet 4 of 5, will be revised as follows:

Table 7.3-4 (Sheet 4 of 5)

Engineered Safeguafds Features Actuations Variables, Limits, Ranges, and Accuracies (Nominal)

Protative Functions Variables to be Conditions of the Range of Protative Systern Restenw Mnnitored Variable or Other Variables Accuracy Tirne Safeguards Actua- (Sw)(II tion Signals That Initiate Protwtive Action

12. Passive ResidualIIcat Sicam generator Steam generator 0 - 100% of i 10.0% of span 1.6 Removal water level narrow range water span (narrov.

level range taps) coincident with: Startup feed

  • ster 0 - 500 gpm 4.0% of span 1.6 Startup feedsster flow - Low flow Steam generator Lnw steam generator 0 - 100% of 15.5% of span 1.6 water level wide range level span (wide range taps)

Reactor coolant See item 6 See item 6 See Item 6 See item 6 system depres-surization actus-tion Steam generator liigh-2 compensated 0 100% of i 10.0% of span 1.6 wster level steam generator nar- span (narrow row range water level rsnge taps)

Pressurizer s ater liigh pressurizer 0 - 100% of i 7.0% of span 1.2 level water level cyhndrical por-tion of pressuriz-er 13, Accumulator lajection Pressurizer pres- liigh pressurizer 1700 to i 4.5% of span 0.6 sure pressure 2500 psig 14, Binck ik>ron Ddution Neutron flux liigh source range I to 106c/sec i 5.0% of equiva, 10.0 neutron flux lent linear fidi acale output P-4 Reactor trip N/A N/A 0.3 1

420.92-4 '

W***

Westinghouse 1

i I

NRC REQUEST FOR ADDITIONAL INFORMATION N

SSAR Section 16.1. Table 3.3.2-1, page 7 of 8, will be revised as follows:

Table 3.3.2 1 (page 7 of 8)

Engineered Safeguards Actuation System Instrumentation FUNCTION APPLICABLE REQUIRED CONDITIONS SURVEILLANCE NOMINAL DAL(a)

MODES CHANNELS / REQUIREMENTS TRIP DIVISIONS SETPOINT

e. Steam 1,2,3,4 4/SG F,1 SR 3.3.2.1 a67% to.25%

Generator SR 3.3.2.4 (SG) Wide SR 3.3.2.5 Range Water Level- Low

f. RCS Refer to Function 7.a, 7.b, 7.c, and 7.d (RCS Depressurization) for all Depressurizat initiating functions and requirements ton
g. Pressurizer 1,2,3(c) 4 F,1 SR 3.3.2.1 s;92% 10.25%

Water Level-- SR 3.3.2.4 High SR 3.3.2.5

11. Demineralized Water System Makeup Isolation
a. ESFAC Logic 2(b),3,4,5 2 ESFAC C,1 SR 3.3.2.2 N/A N/A with 2 SR 3.3.2.5 redundant logic groups
b. PLC 2(b),3,4,5 2 G,1 SR 3.3.2.2 N/A N/A SR 3.3.2.5
c. Source Range 2(b),3,4,5 4 F,1 SR 3.3.2.1 Source 23.0%

Neutron Flux SR 3.3.2.4 Range flux Doubling SR 3.3.2.5 Doubling in 10 minutes d, - . Reactor Trip, Refer to Function' 14.a (ESFAS Interlocks, Reactor Trip, P 4)' f rir[all P-4 ' initiating;funct{onsandrequiremrmtsp

12. Chemical Voltrae and Control System Makeup Isolation
a. ESFAC Logic 1,2,3 2 ESFAC C,1 SR 3.3.2.2 N/A N/A with 2 SR 3.3.2.5 redundant logic groups
b. PLC 1,2,3 2 G,1 SR 3.3.2.2 N/A N/A SR 3.3.2.5 W Westinghouse l

l l

l I

NRC REQUEST FOR ADDITIONAL INFORMATION l

l c. Steam 1,2,3 4 per SG F,1 SR 3.3.2.1 s79% 10.25%

l Generator SR 3.3.2.4 l 4

(SG) Narrow SR 3.3.2.5 l Range Water l Level *-High 2 I

d. Pressurizer 1,2,3(c) 4 F,1 SR 3.3.2.1 s 92% $0.25%

Water Levet-- SR 3.3.2.4 High SR 3.3.2.5 (continued) I (a) Deviation f rom "as lef t" (DAL) is the maximin acceptable deviation in percent of channel span where deviation equals the absolute value of the "as found" setpoint from the "as left" setpoint.

(b) Below the P-6 (Intermediate Range Neutron Flux) interlocks.

(c) Above the P-11 (Pressuriger Pressure) interlock.

l l

l l

420.92-6 ,

W

~~

Westinghouse j l

i l

NRC REQUEST FOR ADDITIONAL INFORMATION  :

1 The Bases for Section 3.3.2, page 16.1-2cl will be revised as follows:

BASES APPLICABLE 11.a. E.JFAC Looic SAFETY ANALYSES, LCOs, and ESFAC Logic consist of the same features and APPLICABILITY operate in the same manner as described for ESFAS ,

(continued) Function 1.b.

11.b. ELQ ,

PLC provide the logic and power interface between the actuated components. These cabinets must be OPERABLE to support both automatic actuation and manual actuation. Only two divisions are provided for CVS Makeup Isolation. 1 1 1.c. Source Ranoe Neutron Flux Four channels of Source Range Neutron Flux are provided.

Two-out-of-four channels indicating flux doubling willisolate i demineralized water makeup to preclude a boron dilution event.

Midh[B56tyhTypfP3 j Demineralized water' makeup is 'also lsolated by~all fundtlon's' i that initiate Reactor Trip via the P-4 signal. .The isolation 'I requirements ,for these functions are the 'same' as the requirements for their Reactor Trip function. ' Therefore, the i requirements are not repeated in Table 3.3.21. Instead I Function 14.a, P-4 (Reactor Trip Breakers), is referenced for j allhit!.ating functions;and,requi(ementsj  :

)

i i

420.92a W Westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION

[ *iR

. . e e, Revise the Bases for Section 3.3.2, page 16.1-293, as follows:

BAGES APPLICABLE interlock functions backup manual actions to ensure SAFETY ANALYSES, bypassable functions are in operation under the LCOs, and conditions assumed in the safety analyses.

APPLICABILITY (continued) 14.a. Reaciqr Trio. P-4 The P-4 interlock is enabled when either reactor trip breaker on 2-out-of-4 divisions are open. The functions of the P-4 interlock are:

  • Trip the main turbine Permit the block of automatic S actuation after a predetermined time interval following automatic S actuation.
  • 1 llsofstsTdernineralizediw~atsDnsksUri The reactor trip breaker position switches that provide input to the P-4 interlock only function to energize or de-energize or open or close contacts. Therefore, this function has no adjustable Trip Setpoint with which to associate a DAL.

This function must be OPERABLE in MODES 1,2, and 3 when the reactor may be critical or approaching criticality. This function does not have to be OPERABLE in MODES 4,5, or 6, because the main turbine, the Main Feedwater System, and 1 the Steam Dump System are not in operation.

14.b. Pressurizer Pressure. P-11 The P-11 interlock permits a normal unit cooldown and ,

depressurization without actuation of S or main steamline and )

feedwater isolation. With pressurizer pressure channels less l than the P-11 setpoint, the operator can manually block the  ;

Pressurizer pressure -Low, Steam Line Pressure--Low, and l Tcold--Low S signals and the Steam Line Pressure -Low and  ;

1 420.92-8 3 Westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION 8

Question 420.103 Provid- the LCO for the Containment Purge and Exhaust isolation Instrumentation, as dermed in the Westinghouse Plant Standard Technical Specifications (NUREG-1431) orjustify why it is not included in Chapter 16 of the SSAR for the AP600.

Response

Limiting Conditions for Operation (LCO) for instrumentation, that isolates the containment air ventilation system, is found in the LCO for engineered safety features actuation system (ESFAS), Functions 3 and 15. Function 3 is a general containment isolation that actuates on an S signal and a high radiation signal. Function 15 is a containment air filtration system containment isolation and actuates on a Radioactivity-High 1 signal.

SSAR Revision: NONE 0,103-1 W Westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION

}.

Question 420.104 Provide the LCO for the Fuel Building Air Cleanup System Actuation Instrumentation, as defined in the Westinghouse Plant Standard Technical Specifications (NUREG-1431) orjustify w hy it is not included in Chapter 16 of the SSAR for the AP600.

Response

Actuation instrumentation for the radiologically controlled area ventilation system (VAS) has not been included in the AP600 technical specifications because it does not meet the NRC technical specification selection criteria identified in Subsection 16.1.1 of the AP600 Technical Specifications.

The VAS serves the radiologically controlled portions of the annex 11 building, the fuel building, the fuel handling area, and auxiliary building. 'lhere is no safety design basis for the VAS. However, the failure of nonsafety, nonseismie HVAC equipment /ductwork will not compromise any safety-related systems, structures, or components.

As indicated in SSAR Subsection 15.7.4.2, "For the fuel handling accident postulated to occur in the spent fuel pit area, there is assumed to be no filtration in the release pathway. Activity released from the pool is assumed to pass directly to the environment with no credit for holdup or delay in release in the building." Thus, no LCO is required for the VAS or the instrumentation that actuates it.

SSAR Revision: NONE

[ WC5tiligh0LISS

i i

NRC REQUEST FOR ADDITIONAL INFORMATION Question 471.10 l

Section 123.1.1.1 of the SSAR states that multiple electric lights are provided for each cell or room containing highly radioactive components. Is suf0cient lighting provided in high radiation areas in the auxiliary building and containment to reduce the need to rig temporary lighting / power during maintenance periods or outages? Justify your answer.

Response

As stated in SSAR Subsection 123.1.1.1, wherever practical, multiple circuited incandescent lights are provided in high-radiation areas. The circuit design in these areas is such that alternate incandescent lighting fixtures are fed from different lighting panels powered from diesel-backed buses. less of one panel or circuit will not affect all lights in an area. Sufficient lighting will still be available so that replxement of burned-out lamps or circuit j components can be made when the system is secured. Temp)rary lighting may be needed in congested areas where l

it is not practical to provide multiple lights. Convenience outlets are provided to facilitate rigging of the temporary l

lights. This lighting design reduces the need to rig temporary lights during maintenance periods or outages.

SSAR Revision: NONE

)

l

  1. 'I O W Westinghouse 1 1

l l

_______________________________________.______-.__________________________a

NRC REQUEST FOR ADDITIONALINFORMATION Ouestion 471.18 Section 93.1 of the SSAR describes the breathing air system (IlAS) as a subsystem of the compressed and instrument air system (CAS). Describe the location of the suction of the CAS air mmpressors and describe how the air quality requirements of Paragraphs 20.1001-20.2401(2)(d) of Appendix A to 10 CITt 20 are met. Describe what precautk)ns are taken to prevent the CAS/I3AS from bemming cross <onnected with other systems (e.g., waste gas or low pressure nitrogen systems). Also, describe the locations of these 13AS service connections inside the mntainment relative to operations that will necessitate personnel use.

Response

De CAS air mmpressors are in the turbine building. De inlet source is from the turbine building environment (see SSAR Figure 1.2-32 Proprietary). As a subsystem of the CAS, BAS air is filtered through the compressor's inlet air filter, compressed, cooled through the air system's aftercxx)lers, and filtered again through the compressed air prefilters and breathing air subsystem filters (see SSAR Subsection 93.1.2.2).13AS air quality will be sampled and monitored periodically to meet the requirements of Footnote d.2(d) of Appendix A to 10 Cllt 20, Sections 20.1001-20.24)l.

No piping connections uist between CAS/ BAS and other plant gas systems. Also unique quick disconnects are provided for the breathing air system. De dismnnects will distinguish BAS connections from service air and connections from other gas system's (oxygen, nitrogen, etc.).

De llAS unique quick-disconnect adapters inside the mntainment will be k)cated near areas requiring breathing air supply.

SSAR Revision: NONE l

471.18-1 W

Westinghouse l

l l l

I l

\

I 4

I NRC REQUEST FOR ADDITIONAL INFORMATION N

Ouestion 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 1cvels in these areas during these events or conditions.

Response

1his is an interim response to the referenced question.

Figures 1.2-30 and 1.2-31 will be revised as agreed during our May 4 and 5,1993 meeting. A final response to the question, including figures, will be submitted by October 30. b>93.

SSAR Revision: NONE 1

3 Westinghouse

NRC REQUEST FOR ADDITIONALINFORMATION

?

Question 620.51 Provide a matrix to identify or map the information mntained in Chapter 18 of the SSAR and other applicable SSAR chapters with the 8 elements and specific components of each element described in the document "IIFE Program i Review Model and Acaptance Criteria for Evolutionary Reactors," that was transmitted to Westinghouse by letter dated September 16,1992.

Response

At the June 9 and 10,1993 meeting held with the NRC to discuss the AP600 man-machine interface design, the scope of this RAI was expanded by the NRC. The NRC indicated that the response to this RAI should contain identification of those elements and subelements of the proposed IIFE model that Westinghouse intends to (xrtify.

Efforts are under way to provide a response containing the additional information. The response including the matrix, along with the additional scope requested, will be provided to the NRC by August 31,1993.

SSAR Revision: NONE 3

620m T Westinghouse i

1 I

I

NRC REQUEST FOR ADDITIONAL INFORMATION dh tiii

~

ii!

Request for Additional Information related to AP600 Containment Buckling Isstes The following Request for Additional Information was received in a letter from Frederick W. Hasselberg to Nicholas J. Liparuto, dated April 2,1993.

At a meeting on February 10,1993 Westinghouse presented infonnation on asymmetric temperature stresses in the AP600 during passive containment cooling. These stresses were for the case where n= 100 (100 water streams on the shell). Additionally, Westinghouse provided the Nuclear Regulatory Commission (NRC) with utn sses for the case where n=50. Following these discussions the NRC staff has determined that he worst case for compressive stresses in the shell may not have been investigated.

Accordingly, Westinghouse is requested to evaluate and report to the NRC the cases where n=200 and n=400.

These cases should combine stresses due to temperature gradients and pressure and compare the results with the allowable buckling stresses in ASME Code Case N-284. This information is necessary in order to determine whether or not additional research will be required in this area.

Response

The following discussion provides a summary of the asymmetric thermal stress evaluation and includes both material presented at the meeting noted in the request for additional information and subsequent work in response to the NRC request for additional information. He material presented at the meeting is documented in Reference 1.

AP600 Containment Vessel Evaluation for Asymmetric Temnerature Distribution Introduction The AP600 includes a passive containment cooling system in which water flow onto the top of the containment dome is used to cool the containment. Tests of a preliminary water distribution system, reported in WCAP-13296, have shown that there may be wet and dry sections of the containment vessel. Typically the tests showed a number of vertical dry strips surrounded by wet areas. Review of the test data indicated that the wet areas covered about 70 percent of the surface and that the dry areas could have a maximum width of about 15 inches. In the safety analyses temperatures of the vessel were calculated separately for the wet and dry regions with no consideration of heat conduction from one to the other. These analyses showed a maximum difference in temperature between wet and dry regions of 68'F.

Structural analyses were performed to investigate the effect of these temperature variations on the vessel stresses.

He temperature difference was conservatively specified as 80*F, providing margin above the maximum calculated  ;

difference reported in the preceeding paragraph. Below elevation 132'-3", it was assumed that the metal temperature was at the vessel design temperature of 280'F. At all elevations above elevation 132'-3", it was j assumed that the metal temperature was constant on a given azimuth and that the circumferential temperature was '

l def'med '_y 15-inch-wide dry strips at a temperature of 280*F, altemating with 34-inch-wide wet strips at a 1

Response to 4/2/93 Letter-1

NRC REQUEST FOR ADDITIONAL INFORMATION M 'un temperature of 200*F. This case was represented by the zero harmonic plus four higher harmonics of 100, 200, 300, anc 9 in the Fourier series, with the corresponding emplitudes of 224.4*,41,7*,23.9*,4.4*, and -8.2*F, respectiven. He zero harmonic is an axisymmetric part of temperature distribution and does not contribute to the asymmetric distribution, and therefore to the asymmetric stress pattern. This case does not produce any stresses in the region away from the discontinuities at Elevations 100'.0" and 132'-3", it is considered independently, since the subject of the present discussion is to investigate the effects of asymmetric temperature distribution only.

An additional case was considered with 50 waves around the circumference. This has dry regions larger than the maximum observed in the water distribution tests. It was chosen based on the containment structural analyses that show that critical clastic buckling of the AP600 head occurred for n = 65 and plastic buckling occurred for n =

33. The purpose of analyzing this case was to investigate the sensitivity of the stress results to the number of waves.

Stresses due to Asymmetric Temperature Distribution Analyses for the 50 and 200 harmonic loads with an amplitude oN0*F were performed using the axisymmetric shell of revolution model of the overall containment vessel as used in the vessel analyses described in SSAR Subsection 3.8.2.4.1. Analyses for the 100,200, and 400 harmonic loads were performed using local models of the top head.

Comparison of the results in the knuckle region for the 100 harmonic showed that the local model gave results that were consistent with those of the overall model, he analyt.es were performed for asymmetric temperatures specifiec' by a cosine distribution with an amplitude of 40*F. A single Fourier harmonic, n, of 50, IM 200, and 400, one term at a time, was used. Table I shows stress results for n = 50 and 100 for a typical locaion in the cylinder away from the hoop stiffeners. Table 2 shows results for n = 50,100, 200, and 400 at the knuckle location where the maximum compression occurs under internrd pressure. The maximum meridional tensile / compressive membrane stress is essentially the same for each harmonic, n, grea.sr than 50, and approaches a maximum value of 7.0 ksi. This is found in both the cylindrical shell and the top head knuckle. This value is equal to the stress in a plate restrained in the meridional direction, free in the circumferential direction, and subjected to a uniform temperature difference of 40*F. A small membrane hoop stress of 0.6 ksi occurs for n = 50 in the knuckle at the location waere the maximum compression occurs under internal pressure. Hoop and in-plane shear stresses around the circu.nference are near zero for the higher hanaonics in the knuckle and for all harmonies in the cylinder.

The preceeding results for Fourier harmonics of 100,200,300, and 400 with the amplitudes of 41.7*,23.9*,4.4*,

and -8.2*F, respectively, were combined to simulate the response due to the asymmetric portion of the temperatures defined by the case of 15-inch-dry and 34-inch-wet strips (Figures 1 and 2). Maximum combined meridional compressive membrane stresses in the cylindrical shell and the top head knuckle (c = 105 degrees, tangent line 4

= 90 degrees) are 10.8 ksi and 9.7 ksi, respectively. These meridional stresses are very local around the circumference of the vessel. There are no other thermal stress components present at these locations. He stress distribution patterns repeat every 49 inches for the cylindrical shell and every 48.5 inches for the top head knuckle at & = 105 degrees. These stress plots are shown in Figures I and 2, respectively. There are 100 repetitions around the circumference.

i I

Response to 4/2/93 Letter.2 I W

Westinghouse 1 I

l l

NRC REQUEST FOR ADDITIONAL INFORMATION iii p l iM e

Buckling Evaluation for Asymmetric Temperatures Cylindrical Shell The alternate compression / tension meridional loading was investigated considering simply supported edges r.t the azimuths of zero meridional stress (compression changing to tension and vice versa). This assumption permits the buckling investigation of a cylindrical shell model to be represented by a cylindrical panel model. The circumferential width of a panel is equal to the distance over which meridional compxssion is acting and is Joout If. inches (Figure 1). The width is conservatively used as 24.5 inches (one-half of 49 inches), which makes an angle of 1.8 degrees = ir/100 radian subtended at the axis of revolution (Figure 3).

From page 486 of Reference 2, the theoretical clastic buckling stress, act, is calculated as 477 ksi for a uniform meridional compression. Applying a capacity reduction factor of 0.252 (due to imperfection), a plasticity reduction factor of 0.358, and a factor of safety of 2, based on ASME Code Case N-284 (Reference 3), the allowable buckling meridional stress, a43, is 21.5 ksi. %e applied maximum meridional compressive stress is about 10.8 ksi (Figure 1). (Note tt r the 10.8 ksi is the maximum stress and is not the average stress over the width of the panel.)

Therefore, it is concluded that there is no potential buckling of the cylindrical shell due to asymmetric temperatures.

An alternate approach of distributiny the meridional load uniformly over the half-wave length, calculated based on a cylindrical shell subjected to exterul pressure and then comparing with the N-284 allowable, leads to the same conclusion. This approach is considered applicable because a cylindrical shell with closed ends subjected to external pressure develops a stress condition of meridional and circumferential compression. His stress pattern is more severe than one with meridional compression alone. De average meridional compressive stress over the half-wave length of 106.5 inches is 0.8 ksi(Tigure 1). The corresponding ASME Code Case N-284 allowable is 4.5 ksi using the theoretical elastic buckling stress of cr a = 35.8 ksi, a capacity reduction factor of 0.252, a plasticity reduction factor of 1.0, and a factor of safety of 2.

Top llead Knuckle When subjected to external pressure, an i -

shell (top head) develops a stress condition of meridional compression and circumferential tension in t,.. a Q . region, whereas a spherical shell develops a stress condition of equal meridional and circumferential compression. The external pressure buckling evaluation of the top head knuckle with meridional compression alone can be treated as an equivalent spherical shell of a radius equal to the conical radius of the top head knuckle. (Note that the effect of tension on buckling is ignored in ASME Code Case N-284.)It is therefore conservative to use the minimum value of the circumferential buckling wave length of an equivalent spherical shell when calculating the average stress.

He meridional load is uniformly distributed over the half-wave length calculated based on equivalent spherical shell subjected to external pressure. His is demonstrated in Figure 2. He average meridional compressive stress over the half. wave length of 63 inches is 1.8 ksi. The corresponding ASME Code case N-284 allowable is 3.6 ksi using the theoretical clastic buckling stress of cr a = 34.9 ksi, a capacity reduction factor of 0.207 (this is based on proposed revision I of the Code Case, since this paragraph was in course of preparation when the original code case gg Response to 4/2/93 Letter-3

NRC REQUEST FOR ADDITIONAL INFORMATION d VR l![

was issued), a plasticity reduction factor of 1.0, and a factor of safety of 2. Herefore, it is concluded that there is no potential buckling of the top head due to asymmetric temperatures.

Buckling Evaluation for Combined Asymmetric Temperatures and Internal Pressures Cylindrical Shell The asymmetric temperatures produce meridional stresses only (alternating compression and tension around the circumference), whereas internal pressu e develops meridional tension and circumferential tension. When both loadings are combined, the meridional compression will be reduced, resulting in less severe buckling potential than that with asymmetric temperatures, as discussed earlier. Herefore, there is no need to explore this further.

Top Head Knuckle ne circumferential distribution of membrane stresses due to asymmetric temperatures plus internal pressure is shown in Figure 4. As described before, circumferential and meridional compressive membrane stresses are, respectively, zero and 1.8 ksi (using the average over the half-wave length) due to asymmetric temperatures. The circumferential and meridional membrane stresses due to 45 psig are 10.3 ksi compression and 11 ksi tension, respectively. Combining these stresses with the stresses due to the internal pressures of 5 and 45 psig, the resulting membrane stresses and their N-284 buckling evaluation are as follows:

CASE 1: Asymmetric temperatures + 5 psig in' mal pressure Stress Component Stress Circumferential compression 1.1 ksi Meridional compression 0.6 ksi N-284 allowable compression for circurnferential stresses 10.3 ksi N-284 allowable compression for meridional stresses 2.2 ksi nerefore, there is no potential buckling for this load case.

CASE 2: Asymmetric temperatures + 45 psig internal pressure Stress Component Stress Circumferential compression 10.3 ksi Meridional tension 9.2 ksi N-284 allowable compression for circumferential stresses 10.3 ksi N-284 allowable compression for meridional Stresses 2.2 ksi nerefore, there is no potential buckling for this load case. (Also, the meridional tension will have a positive effect on preventing any possible buckling.)

Response to 4/2/93 Letter-4 W-Westinghouse I .

NRC REQUEST FOR ADDITIONAL INFORMATION

.?H ns W U Considering the magnitude of the internal pressure buckling capacity of the top head, as determined by BOSORS analysis (Reference 4), and the margin available in the buckling evaluation of the top head subjected to asymmetric temperatures alone, the combined loadings of maximum 45 psig design internal pressure and asymmetric temperatures will not lead to any potential buckling.

Conclusions ne containment vessel was evaluated for temperature variations around the vessel that were conservatively postulated based on a review of the water distribution tests and other safety analyses. Shell stresses due to the thermal loads were conservatively evaluated and demonstrated large margin against buckling. The evaluation demonstrates that such temperature variations are not significant to the design of the containment vessel.

Refercoces

1. letter from N. J. Liparulo to Document Control Desk, USNRC, dated February 8,1993.
2. Timoshenko, S. P. and Gere, J. M., " Theory of Elastic Stability,' Second Edition,1961, McGraw-Hill Book Co.
3. ASME Code Case N-284.
4. AP600 SSAR Subsection 3.8.2.4.2.2 - Buckling Evaluation of Top Head.

SSAR Revision: NONE Response to 4/2/93 Letter-5 g,g

NRC REQUEST FOR ADDITIONAL INFORMATION TABLE 1 Summary of Stresses in Cylindrical Shell (Elevation 150'-0")

due to n th Harmonic Temperature Loading above Elevation 132'-3" (T = 40 x Cos(n x 0))

Harmonic, n 6 Ixcation og og 7 4g Inside -7.0 0.0 0.0 50 0.0 hiembrane -7. ) 0.0 0.0 Outside -7.0 0.0 0.0 1.8 Membrane 0.0 0.0 -0.1 Inside -6.8 -0.1 0.0 0.0 Membrane -7.0 -0.1 0.0 IN Outside -7.2 0.0 00 0.9 Membrane 0.0 0.0 -0.1 ap, o0' 769: Meridional stress, circumferential stress, and shear stress, respectively. Stresses in ksi.

6 is the distance from the center of the dry strip n.casured in degrees (in circumferential direction).

I Response to 4/2/93 Letter-G

[ WB5tingh0USD 1

1 l

NRC REQUEST FOR ADDITIONAL INFORMATION

!!iit uHi TABLE 2 Summary of Stresses in Top Head Knuckle (Elevation 224'-G", o = 105 degrees) due to n th Harmonic Temperature Loading above Elevation 132'-3" (T = 40 x Cos(n x 6))

llarmonic, n 6 12) cation a6 rg a4 Inside -3.3 -5.6 0.0 50 0.0 Membrane -1.7 0.1 0.0 Outside -0.1 5.7 0.0 1.8 Membrane 0.0 0.0 0.1 Inside -7.4 -4.7 0.0 100 0.0 Membrane -5.9 -0.0 0.0 Outside -4.5 4.6 0.0 0.9 Membrane 0.0 0.0 0.0 Inside -7.4 -1.4 0.0 200 0.0 Membrane -7.0 0.0 0.0 i Outside -6.5 1.4 0.0 i l

0.45 Membrane N.C. N.C. N.C. j Inside -7.1 -0.3 0.0 400 0.0 Membrane ~7.0 0.0 0.0 i

Outside -6.9 0.3 0.0 0.225 Membrane N.C. N.C. N.C.

1 1

og, og, rg: Meridional stress, circumferential stress, and shear stress, respectively. Stresses in ksi.

6 is the distance from the center of the dry strip measured in degrees (in circumferential direction).

N.C. Not calculated as the magnitudes are expected to be zero or near zero. (See the corresponding results for n = 50 and 100.)

Response to 4/2/93 Letter-7

r p NRC REQUEST R R ADDITIONAL INFORMATION r

( (SYMMETRY) i - 34.0" s ,

(WET STRIP)

6. 0 - -

--190 SC .

i

_ 7.. .. .. .y. .. . ...

g.. ,

-200 ,

i

~

^ T  :

c- i i  :

0  ! i 24.p.  ; j --210 7 o_.

g i l i 49.0" i

  • C 7!5 15.0 22.5! 30.0 37.5 !45.0 !52.5 ! .0 67.5 --220 0 C. ,[
J. i .d' j; . ; j. . [ i ;;[. ;;'3, {

.:.a.ui;u, . :.G;iu;;u:.w..si;;;;u;;2:.G;iu;ia:.s...;. :;;s:a:2.La*.u;ig

! l 224 p l

ARC, S {\N) --230 g 6
.

- s3.2r'

\

h! ~20 -- -800 PSI, AVERAGED OVER i  ! -Q y THE HALF-WAVE LENGTH, l j (W /2) -- 240 g m Wo- 106 5"(THREE TIMES i l n.

4 0-- 3 3 THE SQUAREROOT OF 'Rt') i  !

I BASED ON BUCKLING DUE l
--250 W g z i

TO EXTERNAL PRESSURE i i

i j

l o i i  !.l i --260 5 i  !  ! l I 6 -8.0 "-

3 l CYLINDRICAL SHELL i OUTSIDE RADIUS, '

l- !llj j i e

--270 i R - 781.625"  ! i

-10. 0 - .. ....: OUTSIDE PE RIMETE R - 4911" i jl '

--280 (APPROX.100 WAVES OF 49") ,

THICKNESS, t - 1.625" , 15.0" m

-12.0 --

DRY STRIPT --290 SC: COMBINED STRE SS USING n - 100,200,300 AND 400 FOURIER HARMONIC RESULTS TENSION IS POSITIVE AND COMPRESSION NEGATIVE.

FIGURE 1 CIRCUMFERENTIAL DISTRIBUTION OF STRESSES AND TEMPERATURES AT EL.150'-0 (CYL. SHELL) i Response to 4/2/93 Letter-8 3 W85tirigh00S8 c -- --

W j

i NRC REQUEST FOR ADDITIONAL INFORMATION Q (SYMMETRY)

I

, - 33.6" s

6. 0 '- (WET STRIP) --180 SC , __190
4. 0 --  ;

g ................p. ..... ..... ..; 7..

-200 U  !  ! .  !< T  !

82.0--

! l l

48.4" i

i

--210 m u.

m  :  ! 24.2" '!

!  :  : o w 7l 14.8 22.2 d 37.1 !44.5 l51.9 i 59.3 66.7 --220 [

0.0-- .I  : .l. .

i' . 29.h

' J j *l

  • l l l l -224 ui

. ' ii!! 31.\ 5,,  ! l I ARC, S (IN) --230 $ ,

E

.i

g

. L;;;;;;s.2;;;;;ce.e..:.;;;;;;.:;.' (Wt /2) i g o -2.0. _

-- 24 0 m .

m  ; o.

mi w -1800 PSI, AVERAGED OVER j

x 3 THE HALF-WAVE LENGTH,

--250 W F i

d 40-- W - 63.0" (1.75 TlMES THE .

--260

$  ! ShUAREROOT OFg 'R t) BASED i l j 5 i ON EQUlVALENT SPHERICAL i .!  !

if -6.0 -- i SHELL BUCKLING DUE TO i l  : --270

$  ! EXTERNAL PRESSURE, R -  !  !  !

. ' . .....j..... '

.......! 799"(CONICAL AT KNUCK[E) .

--280

-8. 0 --

TOP HEADKNUCKLE OUTSIDE l RADIUS, AT PHI - 105* (T. L. 90* ) --290 l-R - 772" (HORIZONTAL) 3

-10. 0 -- OUTSIDE PERlMETER- 4851" ,, --300 (APPROX.100 WAVES OF 4R 5")

THICKNESS, t - 1.625" iD STRik SC: COMBINED STRESS USING n - 100,200,300 AND 400 FOURIER HARMONIC RESULTS TENSION IS POSITIVE AND COMPRESSION NEGATIVE.

ELGURE 2 CIRCUMFERENTIAL DISTRIBUTION OF STRESSES AND TEMPERATURES AT EL. 224'-6 (TOP HEAD)

Response to 4/2/93 Letter-9

=

r

..j.

i.-

NRC REQUEST FOR ADDITIONAL INFORMATION NONUNIFORM LOADING

/ s

,...., f Y,

I

- 9.

[:. .

I l

I'

/

1P Vjl '['s ' '

i

~ s B = n/10p.

f_____ _ _ _ _ .__'_ _s q l

l l

I

, R  :

s .

I

I I

MMU l A -

E'%ss s% i a

Jk.

_ .I-  :

l

[,,...'..,7 l I ,

?

i FIGUREJ BUCKLING OF A CYLINDRICAL PANEL

,f

.- Response to 4/2/93 Letter-10 '

W-Westinghouse- .

i i

i

'm

NRC REQUEST FOR ADDITIONAL INFORMATION

!!1Fu!tiIt in - !g

=

l 8 I 20. 0--  !

O, DUE TO C2 150--

l l

100-- l g 0, DUE TO C1 1 0 50--

0

% 7. 14.0 22.2 29.6 37.1 44. 1.9 59.3 66.7

0. 0 q l l l l z l l 7 l l ARC, S (119

$-5.0-- Ce DUE TO C1 )

y )

l l , DUE TO C2

- 10. 0--

0, : MERIDIONAL STRESS

~ 0, : CIRCUMFERENTI AL STRESS AVERAGED OVER jA t , R3 - 200", t - 1.625" j C1: ASYMMETRIC TEMPERATURES + 5 PSIG INTERNAL PRESSURE i C2: ASYMMETRICTEMPERATURES + 45 PSIG INTERNAL PRESSURE 0, - O DUE TO ASYMMETRICTEMPERATURES l i

R3 - MERIDIONAL RADIUS AT p- 105* i TENSION IS POSITIVE AND COMPRESSION NEGATIVE.

I FIGURE 4 I CIRCUMFERENTIAL DlSTRIBUTION OF STRESSES l DUE TO ASYMMETRIC TEMPERATURES AND l lNTERNAL PRESSURE AT j EL. 224'-G (TOP HEAD)

Response to 4/2/93 Letter-11 i

l Il

_~-

Retrieved from "https://kanterella.com/w/index.php?title=ML20045H643&oldid=2554652"