ML20217K628

Summary of 980416 Meeting W/Westinghouse in Rockville,Md to Discuss AP600 TS Issues.List of Participants & Handouts & Markups Provided by Licensee During Meeting,Encl

ML20217K628
Person / Time
Site:	05200003
Issue date:	04/28/1998
From:	Huffman W NRC (Affiliation Not Assigned)
To:	NRC (Affiliation Not Assigned)
References
NUDOCS 9805040045
Download: ML20217K628 (20)

Text

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f*%e d S ~ 0 0.3 g O UNITED STATES j ,j NUCLEAR REGULATORY COMMISSION l g WASHINGTON, D.C. 20666-0001

%*****#'g April 28, 1998 l I l

APPLICANT: Westinghouse Electric Company

! PROJECT: AP600 i

SUMMARY

OF MEETING TO DISCUSS AP600 TECHNICAL SPECIFICATION (TS) ISSUES The subject meeting as held on April 16,1998, at the Nuclear Regulatory Commission's office in Rockville, Maryland. Attachment 1 is a list of participants. Attachment 2 contains handouts and markups provided by Westinghouse during the meeting.

The purpose of the meeting was to discuso questions that the staff had raised on the most recent update of the AP600 technical specifications (Revision 22 of the AP600 SSAR).

Specifically, in the process of completing its review of the updated technical specifications, the staff had a general concern about a number of inconsistencies. These concems were documented in a letter to Westinghouse on April 9,1998. In addition, the staff also had an issue on the need to have technical specifications for containment closure and the auxiliary building ventilation system during movement of irradiated fuels.

During this meeting, the staff and Westinghouse reached agreement on wording changes that would resolve most of the concems identified in the April 9,1998, letter. Westinghouse also provided a proposed revision to Section 5.2.5.2.3 of the SSAR to justify the technical bases for calling PRHR leakage into the IRWST " Identified Leakage" as it is currently limited in TS LCO 3.4.8 and defined in TS 1.1. The steff, subsequent to the meeting, reviewed the proposed changes to the SSAR and found that 'he revised words would acceptably address the staff concems regarding the classification of PRHR tube leakage as identified leakage.

In response to item (37) of the April 9,1998 letter, Westinghouse noted that a design change to the instrumentation and controllogic was being processed to permit automatic actuation of the containment recirculation valves on an IRWST Low-3 level coincident with the actuation of ADS stage 4 valves. The proposed actuation logic change was acceptable to the staff.  !

Westinghouse noted that this would result in a change to sheet 16 of the SSAR Figure 7.2-1 logic diagrams. //

The staff also had discussions with Westinghouse on the format of TS Table 3.3.2-1. It is the staff's belief that this Table will require a licensee to maintain operability of l&C actuation circuitry 0 Fb3 even though the equipment that the l&C actuation circuitry control was not required to be operable. Westinghouse understood the staff's questions on the table's arrangement but noted that this was not a safety issue but instead, only an operability constraint that Westinghouse did not perceive as burdensome. The staff stated that the TS Table 3.3.2-1 format related to item (38) of the April 9,1998 letter was acceptable to the staff if Westinghouse did not find the constraints to be an operational concem.

I Westinghouse also committed during the April 16,1998 meeting to verify that the 275 'F LTOP enable temperature in TS 3.4.15 was consistent with the plant design basis analysis.

Westinghouse also committed to determine if the information in the TS Bases on ADS valve operability, as noted in item (21) of the April 9,1998, letter was accurate or even necessary.

9805040045 980428 3 PDR ADOCK 052 3 hr mhyy I

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, i 2 April 28, 1998 Although the staff believes that all the above items are technically reso!ved, the staff cannot close these items in its Technical Specification safety evaluation until revised and corrected documentation is submitted by Westinghouse.

Westinghouse also provided the results of its technical specification self assessment in response to the NRC's quality assurance concems (see handout in attachment 2, I Westinghouse concluded that all SSCs relied on to mitigate the AP600 SSAR Chag.ter 6 and Chapter 15 DBA's are covered in TSs. Several backup functions are not covered directly in the TSs but are covered in the same manner that NUREG 1431 is used to cover these systems in STS. Westinghouse stated that it did not plan to make any changes unless directed by the staff.

The staff and Westinghouse also had extensive discussions on the need for technical specifications for containment closure during movement of irradiated fuel in containment and the need to establish an auxiliary building ventilation technical specification for movement of irradiated fuel in the spent fuel pool. The Westinghouse AP600 design does not use safety-related ventilation systems. In addition, Westinghouse analyses of the radiological consequences of fuel handling accident are well within the dose acceptance criteria of 10 CFR 50.34. Westinghouse contends that based on the first three technical specification inclusion criterion of 10 CFR 50.36, no technical specification LCOs are required for containment closure or for auxiliary building ventilation while moving irradiated fuel. The staff's position is that although the analyzed dose from a fuel handling accident may not exceed the dose acceptance criteria, the principle of defense-in-depth makes it prudent to establish some type of containment barrier to a postulated release from a fuel handling accident. The staff has noted to Westinghouse that operating plants have requested relaxation of fuel handling technical specifications to permit opening equipment hatches and personnel air locks while moving fuel.

The staff has granted licensing changes to permit maintaining personnel air locks open as long as specific compensatory measures were taken to assure the staff that the personnel air locks  ;

could be closed quickly. l Westinghouse states that for the AP600, all the direct containment penetrations (including the equipment hatches) open to the radiologically controlled auxiliary building rather than directly to atmosphere. Westinghouse states that if the major equipment hatches were open, the AP600 design has roll doors in the auxiliary building mezzanine areas which can, in conjunction with an l operating containment air filtration system (VFS), function as an attemate barrier, equivalent, from a defense-in-depth perspective, to an equipment hatch on containment with four bolts in- '

place.

Upon further consideration of the AP600 design and the low radiological consequences of a fuel handling accident, the staff has agreed in principle that the use of an a' ternate barrier and the establishment of an operating filtered ventilation system when moving irradiated fuel within the containment represents an equivalent defense-in-depth concept for containment. The staff suggested, however, that additional changes be made to the proposed technical specification by Westinghouse. Final wording, agreeable to all the NRC technical staff, for a containment closure technical specification applicable to the AP600 design, was not reached during the April 16,1998 meeting. In addition, the staff had requested Westinghouse to propose an equivalent defense-in-depth technical specification for the auxiliary building ventilation system for movement of irradiated fuel in the spent fuel pool. The staff discussed how the currently proposed spent fuel pool fuel handling technical specification for the AP600 could be made i acceptable to the staff. During the meeting, Westinghouse questioned the need to have a spent fuel pool ventilation TS for fuel handling. The staff exchanged views among themselves and with Westinghouse regarding precedents for such technical specifications. Subsequent to the

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3 April 28,1998 meeting the staff discussed the STS LCO 3.7.13 " Fuel Building Air Cleanup System (FBACS)" as such a precedent. The staff feels that the extent to which Westinghouse cities the auxiliary building and the VFS as a barrier to the atmosphere for a fuel handling accident inside l

containment emphasizes the need for similar requirements for such an accident within the auxiliary building proper. In this regard, the staff discussed how the currently proposed spent fuel pool fuel handling technical specification could be mada acceptable. Based on the lack of a definitive agreement between the staff and Westinghouse on this issue during the meeting, the establishment of final AP600 related technical specifications for movement of irradiated fuel in the containment and the spent fuel pool area is still unresolved.

A draft of this meeting summary was provided to Westinghouse to allow them the opportunity to comment on the summary prior to issuance.

original signed by:

William C. Huffman, Project Manager Standardization Project Directorate Division of Reactor Program Management Office of Nuclear Reactor Regulation Docket No.52-003 Attachments: As stated cc w/atts: See next page DISTRIBUTION w/ attachments:

. Docket File PDST R/F TKenyon BHuffman PUBLIC JSebrosky DScaletti JNWilson SMagruder JHWilson MDunsaniwskyj DJackson DISTRIBUTION w/o attachments:

SCollins/FMiraglia,0-5 E7 BSheron,0-5 E7 BBoger,0-5 E7 JRoe DMatthews TQuay ACRS (11) JMoore,0-15 B18 WBeckner,0-13 H15 MReinhart,0-13 H15 AChu,0-13 H15 CMiller,0-10 D4 REmch,0-10 D4 GHolahan,0-8 E2 JLyons,0-8 D1 JKudrick,0-8 H7 HLi,0-8 H3 HWalker,0-8 D1 JRaval,0-8 D1 JMonninger,0-8 H7 JPeralta,0-9 A1 JSegata,0-8 D1 GHsii,0-8 E23 RYoung,0-8 D1 DOCUMENT NAME:A:TS4-16MT. SUM *See previous concurrence To re:eive a copy of this document, indicate in the box: "C" = Copy without attachment / enclosure "E" = Copy with attachment / enclosure "N" = No copy OFFICE PM:PDST:DRPM l SC:SPLB:DSSA l SCSB:DSSA l BC:SCSB:DSSA l NAME WCHuffmankgr/L JELyons* JMonninger* CBerlinger*

DATE 04 /98 04/21/98 04/21/98 04/21/98

ONC TSB

ADPR SC:hSB:ADPR, SC:PERB:DRP_M D$i) DST:DRPM f NAME AChu* MReinhart* REmch/ d TRQuay $N DATE 04/ /98 04/23/98 04K498 04/;)f98 OFFICIAL RECORD COPY l

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Westinghouse Electric Corporction Docket No.52-003 cc: Mr. Nicholas J. Liparulo, Manager Mr. Frank A. Ross Nuclear Safety and Regulatory Analysis U.S. Department of Energy, NE-42 Nuclear and Advanced Technology Division Office of LWR Safety and Technology Westinghouse Electric Corporation 19901 Germantown Road P.O. Box 355 Germantown, MD 20874 Pittsburgh, PA 15230 Mr. Russ Bell Mr. B. A. McIntyre Senior Project Manager, Programs l Advanced Plant Safety & Licensing Nuclear Energy Institute l Westinghouse Electric Corporation 1776 l Street, NW  !

l- Energy Systems Business Unit Suite 300 Box 355 ]

Washington, DC 20006-3706 i Pittsburgh, PA 15230 I Ms. Lynn Connor Ms. Cindy L. Haag Doc-Search Associates Advanced Plant Safety & Licensing Post Office Box 34 Westinghouse Electric Corporation Cabin John, MD 20818 Energy Systems Business Unit Box 355 Dr. Craig D. Sawyer, Manager i

! Pittsburgh, PA 15230 Advanced Reactor Programs GE Nuclear Energy Mr. Jack Bastin 175 Curtner Avenue, MC-754 Westinghouse Electric Company San Jose, CA 95125 11921 Rockville Pike 1 Suite 107 Mr. Robert H. Buchholz j l Rockville, MD 20852 GE Nuclear Energy )

175 Curtner Avenue, MC-781 Mr. Sterling Franks San Jose, CA 95125 U.S. Department of Energy NE-50 Barton Z. Cowan, Esq.

19901 Germantown Road Eckert Seamans Cherin & Mellott Germantown, MD 20874 600 Grant Street 42nd Floor Pittsburgh, PA 15219 '

Mr. Charles Thompson, Nuclear Engineer AP600 Certification Mr. Ed Rodwell, Manager NE-50 PWR Design Certification 19901 Germantown Road Electric Power Research Institute Germantown, MD 20874 3412 Hillview Avenue Palo Alto, CA 94303

~ Mr. Robert Maiers, P.E.

Pennsylvania Department of Environmental Protection Bureau of Radiation Protection l Rachel Carson State Office Building -

P.O. Box 8469 Harrisburg, PA ' 17105-8469 i

i 2

f 1

AP600 MEETING TO DISCUSS AP600 TECHNICAL SPECIFICATIONS MEETING ATTENDEES APRIL 16,1998 NAME ORGANIZATION I

MCORLETTI WESTINGHOUSE  !

CSUGGS WESTINGHOUSE SFANTO WESTINGHOUSE BMCINTYRE WESTINGHOUSE TSCHULZ (PART TIME) WESTINGHOUSE j JMONNINGER NRC '

ACHU NRC I BHUFFMAN NRC l JLYONS (PART TIME) NRC MREINHART (PART TIME) NRC RYOUNG (PART TIME) NRC GHSil (PART TIME) NRC HLI (PART TIME) NRC REMCH (PART TIME) NRC TED QUAY (PART TIME) NRC  ;

CARL SCHULTEN (PART TIME) NRC 1

i

Hand outs and Markups from the April 16,1998, AP600 Technical Specification Meeting 1

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l l

Attachment 2 i u___ -

vj f _ 5 ESFAS Instrumentation B 3.3.2

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BASES -

APPLICABLE 5.c. Safeguards Actuation SAFETY ANALYSES, LCOs. and Turbine Trip is also initiated by all Functions that APPLICABILITY initiate the Safeguards Actuation signal. The (continued) isolation requirements for these Functions are the same as the requirements for their Safeguards Actuation Function. Therefore, the recuirements are not repeated in Table 3.3.2 1. Insteac Function 1 is referenced for all initiating Functions and requirements. The Safeguards Actuation signal closes all main feedwater control, isolation and crossover valves, trips all main feedwater pumps, and trips the turbine.

5.d. Reactor Trip Turbine Trip is also initiated by all functions that initiate Reactor Trip. The isolation requirements for these Functions are the same as the requirements for their Reactor Trip Function. Therefore, the requirements are not re mated in Table 3.3.2 1. .

Instead Function 18.a. 8 4 (Reactor Trip), is referenced for all initiating Functions and requirements.

i 6. Main Feedwater Control Valve Isolation The primary Function of Main Feedwater Control Valve Isolation is to prevent damage to the turbine due to water in the steam lines and to stop the excessive flow of feedwater into the SGs. This Function is actuated by Steam Generator Narrow Range Water Level - High 2, by a Safeguards Actuation signal, or manually. The Reactor Trip Signal also initiates closure of the main feedwater control velves coincident with a low-RCS average temperature (Tavg) signal whenever a reactor trip (P 4) is generated.

Closing the Main Feedwater Control Valves on Manual Main Feedwater Isolation, SG Narrow Range Water Level High 2, er-Sefcgucrds-d ir * *C 3 i xActuation-4s. necessary

? 11-to mitigate in MODES the effects of 1,anM a y3-large SLB or a large FLB. This Function is also required (continued)

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@ AP600 en B 3.3 74 04/98 Amendment 0 o .

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/ ESFAS Instrumentation

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BASES APPLICABLE SAFETY ANALYSES, 18.d. Pressurizer Level P 12 (continued)

LCOs, and IRWST actuation, reactor coolant ) ump trip, ter APPLICABILITY purification line isolation. Witt pressurit.er (continued) level channels less than the P 12 setpoint, the operator can manually block low pressurize" level signal used for these actuations. Concur;ent with

. -9"; t" blockelpMNT actuation on low

>ressur zer level, 'IRWST actuation on Low 2 RCS g

d' s .7 d) 1ot leg level is Tenemed. When the pressurizer level is above the P 12 sh int the pressurizer V[

r ei/1  ; level si confirmabal issiautomaticalfy ena, bled and a (g ry open isolation valves on <gnal is issued

he CMT cold legtobalance the 9

I lines. This Function is required to be OPERABLE- Rf inMODES1,2,and3/,f,* 6 18.e. RCS Pressure, P 19 The P 19 interlock is provided to permit water solid conditions (i.e., when the pressurizer water level is [>92*]) in lower MODES without automatic isolation of the CVS makeup pumps. With RCS

• pressure below the P 19 setpoint. the operator can manually block CVS isolation on High 2 pressurizer water level. When RCS pressure is above the P 19 setpoint, this Function is automatically C unblocked. This Function is required to be OPERABLE IN MODES 1, 2, 3, and 4 with the RCS not being cooled by the RNS. When the RNS is cooled by the RNS, the RNS suction relief valve provides the required overpressure protection (LCO 3.4.15).

l

19. Containment Air Filtration System Isolation Some DBAs such as a LOCA may release radioactivity into the containment where the potential would exist for the radioactivity to be released to the atmosphere and exceed the acceptable site dose limits. Isolation of the Containment Air Filtration System provides protection to prevent radioactivity inside contairulent from being released to the atmosphere.

19.a. Containment Radioactivity - High 1 Three channels of Containment Radioactivity -

High 1 are required to be OPERABLE in MODES 1, 2, ,

3. and 4 with the RCS not being cooled by the RNS, ,

when the potential exists for a LOCA, to protect i against radioactivity inside containment j

, (continued) ,

h AP600 B 3.3 97 04/98 Amendment 0

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Tabla 2 - Tech Spec on Equipmsnt Ussd to Mitigato Accidents Tsble 1 Function / Equipment Technical Specification l

Note LCO Survelliance I  ! - RCP trip; CMT act. 3.3.2 Table 3.3.2-1, item 11 j '

- PCS act; cont pres, high 2 3.3.2 Table 3.3.21, item 12 i i - PRHR HX act; SG narrow range lev, low & SFW 3.3.2 Table 3.3.21, item 13 i )

i I l' flow low

SG wide range lev, low 3.3.2 Table 3.3.21, item 13 1 -  ; CMT act 3.3.2 Table 3.3.21, item 13

Pzr level, high 3 3.3.2 Table 3.3.2-1, item 13 I l

- SG blowdown isol; PRHR HX act 3.3.2 Table 3.3.2-1, item 14

SG narrow range level, low 3.3.2 Table 3.3.21, item 14 .

- Boron dil block; source range flux, high 3.3.2 Table 3.3.2-1, item 15 i

' -  ;RT 3.3.2 Table 3.3.2-1, item 15 l - CVS makeup isol; SG narrow range lev, high 2 3.3.2 Table 3.3.2-1, item 16 i j

Pzr level, high 1 & Saf Act 3.3.2 Table 3.3.2-1, item 16 '

Pzr level, high 2 3.3.2 Table 3.3.2-1, item 16

- Cont Purge Isol; cont isol act. 3.3.2 Table 3.3.2-1, item 19  :

- MCR isol / air supply act; MCR rad, high 2 3.3.2 Table 3.3.2-1, iterg 20

- IRWST injection act; ADS 4 act 3.3.2 Table 3.3.2-1, item 22 i

-  ; RC HL level, low 2 3.3.2 Table 3.3.2-1, item 22 l

- Cont recirc act; IRWST level, low 3 & Saf Act 3.3.2 Table 3.3.2-1, item 23 {

, - Pzr heater trip; CMT act 3.3.2 Table 3.3.2-1, item 27 4

Pzr level, high 3 3.3.2 Table 3.3.2-1, item 27 j

! i i

1E de power 3.8.1 3.8.1.1  ;

b Reactor trip j trip RT breakers !3.3.1 Table 3.3.1-1, item 17 ,

c Pzr pressure relief

, - open/close Pzr SV 3.4.7 3.4.7.1 '

d SG pressure relief

- open/close SG SV l 3.7.1 3.7.1.1 l  !

o, e1 Isolation of SG steam lines l close MSIVs I 3 7.2 1701 G31ose SG PORVs I note 1,3 close SG PORV block valves 3.6.3^3.6.3.4

- close Turb stop valves  ; 3.7.2 3.7.2.2

- close Turb control valves l 3.7.2 3.7.2.2 .

- close Turb bypass valves  ! 3.7.2 3.7.2.2  :

- close main steam to aux steam valve i 3.7.2 3.7.2.2 Page 11

, . Icbla 2 - Tech Spec on Equipment Used to Mitigate Accidents l

!' Tcble 1 Function / Equipment Technical Specification Note LCO Surveillance i 5 - close moisture sep reheat valve 3.7.2 3.7.2.2 i f, f t, f2 ' Isolation of MFW to SG I l - close MFW isol valves ' 3.7.3 3.7.3.1

- close MFW control valves 3.7.3 3.7.3.1

- trip MFW pumps 3.3.2 Table 3.3.2-1, item 7 l

g1,g2 Isolation of SFW to SG  !

- close SFW isol valves 3.7.7 3.7.7.1

- close SFW control valves _ _

3.7.7 3.7.7.1 Q trip SFW pumps 3.3.2 Tab I

h l Containmentisolation I - close containment isol valves 3.6.3 3.6.3.4 i PCS actuation -

l - open PCS iso! valves 3.6.6 3.6.6.4  !

PRHR HX actuation l J ,

l - open PRHR HX control valves 3.5.4 3.5.4.4

- close IRWST gutter isol valves i 3.5.4 3.5.4.4

- inle d ormally open _

! 3.5.4 3.5.4.2

- close SG blowdown isol valves] } not k CMT actuation I

- open CMT discharge isol valves. 3.5.2 3.5.2.6

- trip RCPs operable 3.3.2 Table 3.3.21, item 11

- inlet MOV normally open 3.5.2 3.5.2.3 I Accum actuation  :

- confirm discharge MOV normally open and ' 3.5.1 3.5.1.1 power removed m ADS actuation .

- open ADS stage 1/2/3 MOVs  ; 3.4.12 3.4.12.2 open ADS stage 4 squib valves i 3.4.12 3.4.12.3 i

i l n IRWST injection actuation j

- open IRWST squib valves 1 3.5.6 3.5.6.8

- injection line MOV normally open  ; 3.5.6 3.5.6.4 Page 12